ML20034G444

Amend 158 to License DPR-46,authorizing Removal of MSL Radiation Monitor Scram & Group I Ci Functions & Modifying TS Accordingly

ML20034G444
Person / Time
Site:	Cooper
Issue date:	03/02/1993
From:	Hubbard G Office of Nuclear Reactor Regulation
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ML20034G446	List: ML20034G444 ML20034G452
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NUDOCS 9303090523
Download: ML20034G444 (18)

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UNITED STATES g

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NUCLEAR REGULATORY COMMISSION ap WASHINGTON, D. C. 20555 9,

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NEBRASKA PUBLIC POWER DISTRICT DOCKET NO 50-298 COOPER NUCLEAR STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.158 License No. DPR-46

'i 1.

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

A.

The application for amendment by Nebraska Public Power District (the licensee) dated May 4, 1992, as supplemented by letters dated October 15, 1992, and January 13, February 12, and February 24, 1993, 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 license 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.

9303090523 930302 PDR ADOCK 05000298 P

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Accordingly, the license is amended by changes to the Technical'Specifi-cations as indicated in the attachment to this license' amendment and Paragraph 2.C.(2) of-facility Operating License No. DPR-46 is hereby amended to read as follows:

2.

Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No.15% are hereby incorporated in the license.

The licensee shall operate the facility in accordance-with the Technical Specifications.

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The license amendment is effective as of its date of issuance.

i FOR THE NUCLEAR REGULATORY COMMISSION j

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nW George T. Hubbard, Acting Director Project Directorate IV-I Division of Reactor Projects - III/IV/V f

Office of Nuclear Reactor Regulation l

Attachment:

Changes to the Technical Specifications Date of Issuance:

March 2, 1993 i

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ATTACHMENT TO LICENSE AMENDMENT NO. 158 l

l FACILITY OPERATING LICENSE NO. DPR-46 t

DOCKET NO. 50-298 I

Replace the following pages of the Appendix A Technical Specifications with the enclosed pages.

The revised pages are identified by Amendment number and-i contain vertical lines indicating the area of change.

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t REMOVE PAGES INSERT PAGES 4

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30 30 33 33 34 34 t

35 35 36 36 39 39 48 48 50 50 52 52 63a 63a i

a 68 68 j

78 78 l

81 81 84 84

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g COOPER NUCLEAR STATION g

TABLE 3.1.,1 (Page 2) g REACTOR PROTECTION SYSTEM INSTRUMENTATION REQUIREMENTS rt U

e Minimum Number Action Required Applicability Conditions of Operable When Equipment Reactor Protection Mode Switch Position Trip Level Channels Per Operability is System Trin Functicn Shutdown Startup Refuel Run Settinn Trlo Systems (1) Not Assured (1)

Main Steam Line M

Isolation Valve Closure X(6)(9)

X(6) s 10% of valve 4

A or C m

MS-LRS-86 A,B.C. & D closure 4

A or C MS-IliS-80 A,B,C, & D Turbine Contt i Valve X(4) a 1000 psig turbine 2

A or B Fast Closure control fluid TC F-63/0PC-1,2,3,4 Turbine stop Valve Closure X(4) s 10% of valve 2

A or B SVOS-1(1), SVOS-1(2)

Closure 4

SV05-2(1), SVOS-2(2) e Turbine First Stage Permissive MS-PS-14 X(9)

X s 30% first 2

A or B A,B,C, 6 D stage press.

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NOTES FOR TABLE 3.1.1 r

1.-

There shall be two operable or tripped trip systems for each function.

If the l

minimum number of operable instrument channels for a trip system cannot be met, the l

affected trip system shall be placed in the safe (tripped) condition, or the appropriate actions listed below shall be taken.

A.

Initiate insertion of operable rods and complete insertion of all operable rods within four hours.

t B.

Reduce power to less than 30% of rated.

C.

Reduce power level to IRM range and place mode switch in the Startup position vithin 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and depressurize to less than 1000 psig.

.l 2.

Permissible to bypass, with control rod block, for reactor protection system' reset i

in refuel and shutdown positions of the reactor mode switch.

3.

This note deleted.

4 Permissible to bypass when turbine first stage pressure is less than 30% of full load.

t 5.

IP3's are bypassed when APP 3's are onscale and the reactor mode switch is in the run position.

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6.

The design permits closure of any two lines without a full scram being initiated.

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k' hen the reactor is suberitical, fuel is in the vessel, and the reactor water temperature is less than 212*F, only the following trip functions need to be operable.

a.

Mode switch in shutdown.

b.

Manual scram.

c.

IRM high flux.

120/125 indicated scale, f

d.

APP 3 (15%) high flux scram.

B.

Not required to be operable when primary containment integrity is not required.

9.

Not required while performing low power physics tests at atmospheric pressure during l

or after refueling at power levels not to exceed 5 MW(t).

10. Not required to be operable when the reactor pressure vessel head is not bolted to the vessel.

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Amendment No. 73, M, 158 _

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COOPER NUCLEAR STATION j.

TABLE 4.1.1 (Page 2)

REACTOR PROTECTION SYSTEM (SCRAM INSTRUMENTATION) FUNCTIONAL TESTS MINIMUM FUNCTIONAL TEST FREQUENCIES FOR SAFETY INSTR. AND CONTROL CIRCUITS U

Instrument Channel Group (2)

Functional Test Minimum Frecuency (3) liigh Water Level in Scram Discharge A

Trip Channel and Alarm Once/3 Months m

j Volume CRD-LS-231 A & B a

CRD-LS-234 A & B CRD-LT-231 C & D bo CRD-LT-234 C & D i

Main Steam Line Isolation Valve A

Trip Channel and Alarm Once/ Month (1)

Closure MS-LMS-86 A,B,C, 6 D MS-LMS-80 A,B,C, &D Turbine Control Valve Fast Closure A

Trip Channel and Alarm Once/ Month (1)

TCF-63/0PC -1,2,3,4 Turbine First Stage Pressure A

Trip Channel and Alarm Once/3 Months Permissive MS-PS-14 A,B,C, & D Turbine Stop Valve Closure A

Trip Channel and Alarm Once/ Month (1)

SVOS-1 (1), SVOS-1 (2)

SVOS-2 (1), SVOS-2 (2) m.

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NOTES FOR TABLE 4.1.1 i

5 1.

Initially once per month until exposure (M as defined on Figure 4.1.1) is 2.0 x 10 ;

thereafter, according to Figure 4.1.1 with an interval not 4ess than one month nor more than three months after review and approval of the NRC.

The compilation of instrument failure rate data may include data obtained from other boiling water reactors for which the same design instrument operates in an environment similar to that of CNS.

2.

A description of the three groups is included in the Bases of this Specification.

3.

Functional tests are not required when the systems are not required to be operable or are tripped.

If reactor startups occur more frequently than once per week, the maximum functional test frequency need not exceed once per week.

If tests are missed, they shall be performed prior to returning the systems to an operable status.

4 Deleted.

l 5.

Test RPS channel after maintenance.

6.

The water level in the reactor vessel will be perturbed and the corresponding level indicator changes will be monitored. This perturbation test will be performed every month after completion of the monthly functional test pro 5 ram.

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J Amendment No. So, 158 4 w.

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COOPER NUCLEAR STATION

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TABLE 4.1.2 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT CALIBRATION MINIMUM CALIBRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUMENT CIIANNELS z

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,s Instrument Channel Group (1) Calibration Test (5) Minimum Frequency (2) "w IRM liigh Flux C Comparison to APRM on Controlled Note (4) o) Shutdowns (7) APRM liigh Flux Dutput Signal B llent Balance Once/ Week Flow Bias Signal B Internal Power and Flow Test once/ Refueling Outage with Standard Pressure Source (8) LPRM Signal B TIP System Traverse Note (9) liigh Reactor Pressure A Standard Pressure Source Once/3 Months !!!gh Drywell Pressure A Standard Pressure Source Once/3 Months Reactor Low Water Level A Pressure Standard Once/3 Months liigh Water Level in Scram Discharge A Note (6) Note (6) Volume Main Steam Line Isolation Valve A Note (6) Note (6) Closure Turbine First Stage Pressure Permissive A Standard Pressure Source Once/6 Months Turbine Control Valve Fast Closure A Standard Pressure Source Once/3 Months Turbine Stop Valve Closure A Note (6) Note (6) l i l l l

I NOTES FOR TAEl.ES 4.1.2 l 1. A description of three groups is included in the bases of this Specification. 2. Calibration tests are not required when the systems are not required to be operable i or are tripped but are required prior to return to service, i 3. Deleted. l l I 4 Maximum frequency required is once per week. 5. Response time is not a part of the routine instrument channel test, but will be checked once per operating cycle. The response time measurement will be the time se5 ment from the time the sensor contacts actuate to the time the scram solenoid valves deenergize. l 6. Physical inspection and actuation of these position switches will be performed during the refueling outages. 7. On controlled shutdowns, the IRM reading 120/125 of full scale will be set equal to l or less than 45% of rated Power. All range scales above that scale on which the most recent IRM calibration was performed will be mechanically blocked. [ \\ 8. The Flow Bias Scran Calibration will consist of calibrating the sensors, flow converters and signal offset networks during operation. The instrumentation is an i analog type with redunlent flow signals that can be compared. The flow bias trip and i upscale will be functionally tested according to table 4.3.1 to assure proper operation during the operating cycle. Refer to Bases of 4.1 for further explanation of calibration frequencies. 9. LPRM detectors shall be calibrated every six weeks of reactor power operation above 20% of rated power. [ i l l + i i i i 4 I i 4 Amendment No. M, 158 4 i

I.IMITING CONDYTIONS FOR OPERATION SURVEILIANCE REOUIREMENTS I 3.1 MSES (Cont'd. ) 4.1 BASES (cont'd.) initiate the cere standby cooling 2. The factor M is the exposure equipment. A high dryvell pressure hours and is equal to the num-1 scram is provided at the same set-ber of sensors in a Eroup, n, ting as the core standby cooling times the elapsed time T (M - i systems (CSCS) initiation to mini-nT). mize the energy which must be accom-modated during a loss of coolant 3. The accumulated number of un-accident and to prevent return to safe failures is plotted as an criticality. This instrumentation ordinate against M as an ab-is a backup to the reactor vessel scissa on Figure 4.1.1. water level instrumentation. l 4. After a trend is established, A reactor mode switch is provided the appropriate monthly test. which actuates or bypasses the vari-interval to satisfy the goal cus scram functions appropriate to will be the test interval to the particular plant operating sta-the left of the plotted tus. Ref. paragraph VII.2. 3.7 FSAR. points. 5. A test interval of 1 month The manual scram function is active 1 be used inWally untn a w in all modes, thus providing for a manual means of rapidly inserting trend is esta m shed, W eh is based on system availability control rods during all modes of reactor operation. analysis and good engineering judgement plus operating expe-The APRM (High flux in Start-Up or Refuel) system provides protection Group (B) devices utilize an analog against excessive power levels and sensor followed by an amplifier and { short' reactor periods in the a bi-stable trip circuit. The sen-start-up and intermediate power sor and amplifier are active compo-ranges. nents and a failure is almost always accompanied by an alarm and an indi-The IRM system provides protection cation of the source of trouble. In the event of failure, repair or substitution can start immediately. An "as-is" failure is one that 1 " sticks" mid-scale and is not capa-ble of going either up or down in response to an out-of-limits input. This type of failure for analog devices is a rare occurrence and is detectable by an operator who ob-serves that one signal does not track the other three. For purpose of analysis, it is assumed that this rare failure wi~ti be detected within two hours. The bi-stable trip circuit which is a part of the Group (B) devices can i sustain unsafe failures which are 1 (6) Reliability of Engineered Safety Features as a Function of Testing Frequency, I.M. Jacobs, " Nuclear Safety", Vol. 9, No. 4 July-Aug. 1968, pp. 310-312. i Amendment No.158. -.

LIMITING CONDITTONS FOR OPERATION SURVEILVANCE REOUIREMENTS 3.2 (cont'd.) 4.2 (cont'd.) l D. Radiation Monitorine Systems -Isola-D. Radiation Monitorine Systems -Isola-tion & Initiation Functions tion & Initiation Functions 1. Steam Jet Air Ejector Off-Gas System 1. Steam Jet Air Ejector Off-Gas System a. Operability of the Steam Jet Instrumentation surveillance Air Ej ector Off-Gas System requirements are given on monitor is defined in Table 4.2.D. Table 3.21.A.2. b. The time delay setting for closure of the staan jet air ejector isolation valves shall not exceed 15 minutes, c. Other limiting conditions for operation are given on Table 3.2.D and Specifica-tions 3. 21. A. 2 and 3.21.C.6. 2. Reactor Building Isolation and 2. Reactor Building Isolation and Standby Gas Treatment Initiation Standby Gas Treatment Initiation The limiting conditions for opera-Instrumentation surveillance tion are given on Table 3.2.D. requirements are given on Table 4.2.D. i 3. Liquid,Radwaste Discharge Isolation 3. Liquid Radwaste Discharge Isolation The limiting conditions for opera-Instrumentation surveillance tion are given on Table 3.2.D and requirements are given on Table Specification 3.21.B. 4.2.D. l 4. Main Control Room Ventilation Isola-4. Main Control Room Ventilation tion Isolation The limiting conditions for opera-The instrument surveillance tion are given on Table 3.2.D and requirements are iS ven on Table the Section entitled " Additional 4.2.D. Safety Related Plant Capabilities." 5. Mechanical Vacuum Pump Isolation 5. Mechanical Vacuum Pump Isolation a. The mechanical vacuum pump The instrument surveillance require-shall be capable of being ments are given on Tables 4.2. A, and l automatically isolated and 4.2.D. secured by a signal of high radiation in the main steam line tunnel whenever the main steam isolat.ica C ves are open. b. If the limits of (3.2.D.5.a) are not met, the vacuum pump shall be isolated. Amendment No. 3 5, S O, E 9,125,157. 158 -4B-

N COOPER NUCLEAR STATION h. TABLE 3.2.A (Page 1) PRIMARY CONTAINMENT AND REACTOR VESSEL ISOLATION INSTRUMENTATION + e 5 x ,o Minimum Number Action Required + $ of Operable When Component O Instrument Components Per Operability is Instrument I.D. No. Settinc Limit Trio System (1) Not Assured (2) Main Steam Line High RMP-RM-251, A,B,C,6D s 3 Times Full Power 2 E o l Radiation 8 j) Reactor Low Water Level NBI-LIS-101, A,B,C,6D #1 a+4.5 in. Indicated Level 2(4) A or B I$ Reactor Low Low Low Water NBI-LIS-57 A & B #1 2-145.5 in. Indicated Level 2 A or B ra Level NBI-LIS-58 A & B #1 g Main Steam Line Leak MS-TS-121, A,B,C,6D s 200*F 2(6) B EA Detection 122, 123, 124, 143, 144, 145, 146, 147, 148, 149, 150 Main Steam Line liigh MS-dPIS-116 A,B,C,6D s 150% of Rated Steam 2(3) B Flow 117, 118, 119 Flow Main Steam Line Low MS-PS-134, A,B,C,6D a 825 psig 2(5) B Pressure liigh Drywell Pressure PC-PS-12, A,B,C,6D s 2 psig 2(4) A or B liigh Reactor Pressure RR-PS-128 A & B s 75 psig i D l Main Condenser Low MS-PS-103, A,B,C,6D a 7" l{g (7) 2 A or B Vacuum Reactor Water Cleanup RWCU-dPIS-170 A & B s 200% of System Flow 1 C System liigh Flow 1 l l l I t ,r-,< .,m.. -. -.., - -, - -, - - ~... - - ,.m. .--,,,,,,,c.... v ,--,.w ,m..~--.

NOTES FOR TABLE 3.2.A 1. Whenever Primary Containment integrity is required there shall be two operable or tripped trip systems for each function. 2. If the minimum number of operable instrument channels per trip system requirement ' cannot be met by a trip system, that trip system shall be tripped. If the requirements cannot be met by both trip systems, the appropriate action listed below shall be taken. A. Initiate an orderly shutdown and have the reactor in a cold shutdown condition in 24 hours. B. Initiate an orderly load reduction and have the Main Steam Isolation Valves shut within 8 hours. C. Isolate the Reactor Water Cleanup System. D. Isolate the Shutdown Cooling mode of the RHR System. E. Isolate the Reactor Water Sample Valves. 3. Two required for each steam line. 4 These signals also start the Standby Gas Treatment System and initiate Secondary Containment isolation. 5. Not required in the refuel, shutdown, and startup/ hot standby modes (interlocked with the mode switch). 6. Requires one channel from each physical location for each trip system. I 7. Low vacuum isolation is bypassed when the turbine stop is not full open, manual bypass switches are in bypass and mode switch is not in RUN. 8. The instruments on this table produce primary containment and system isolations. The following listing groups the system signals and the system isolated. Group 1 Isolation Signals: 1. Reactor Low Low Low Water Level (2-145.5 in.) 2. Main Steam Line Low Pressure (2825 psig in the RUN mode) 3. Main Steam Line Leak Detection (s200*F) 4. Condenser Low Vacuum (27" Hg vacuum) 5. Main Steam Line High Flow (s150% of rated flow) Isolations: 1. MSIV's 2. Main Steam Line Drains i Amendment No. 45, S1, S E,215,152,158 -52 I

NOTES FOR TABLE 3.2.D I j, 1. Action required when component operability is not assured. A. (1) If radiation level exceeds 1.0 ci/sec (prior to 30 min. delay line) for ' a period greater than 15 consecutive minutes, the off-gas isolation valve shall close and reactor shutdown shall be initiated immediately and the reactor placed in a cold shutdown condition within 24 hours. A. (2) Refer to Specification 3.21.A.2. B. A minimum of one instrument channel per trip system shall be operable i when handling irradiated fuel inside secondary containment, and when moving loads inside secondary containment which have the potential to damage irradiated fuel. If this requirement cannot be met by a trip system, then that trip system shall be tripped. If this requirement cannot be met by both trip systems, then the following actions shall be taken: (1) Cease handling of irradiated fuel'inside secondary containment and remove the load from over the irradiated fuel via the most direct path, or i (2) Isolate secondary containment and start SBGT. C. During release of radioactive vastes, the effluent control monitor shall be set to alarm and automatically close the waste discharge valve prior to exceeding the limits of Specification 3.21.B.1. i D. Refer to Section entitled " Additional Safety Related Plant Capabilities". E. Refer to Section 3.2.D.5 and the requirements for Primary Containment J Isolation on high main steam line radiation, Table 3.2.A. j 2. Trip settings to correspond to Specification 3.21.B.1. 1 i 3. Trip settings to correspond to Specification 3.21.C.6.a. 4 Minimum nu=ber of channels operable shall be one during mechanical vacuum pump operation. j 1 i I i l Amendment No. 39,167, 158 -63a-I J i

N g COOPER NUCLEAR STATION p TABLE 4.2.A (Page 1) m PRIMARY CONTAINMENT AND REACTOR VESSEL ISOLATION SYSTEM 3 TEST AND CALIBRATION FREQUENCIES m ? Instrument Item Item I.D. No. Function Test Frea. Calibration Freo. Check ra m Instrument Channels Reactor Low Water Level NBI-LIS-101, A,B,C,6D Once/ Month (1) Once/3 Months once/ Day U" Reactor Low Low Water Level NBI-LIS-57, A & B #2 Once/ Month (1) Once/3 Months Once/ Day NBI-LIS-58, A & B #2 Reactor Low Low Low Water Level NBI-LIS-57, A & B #1 Once/ Month (1) Once/3 Months once/ Day NBI-LIS-58, A & B #1 Main Steam Line liigh Radiation RMP-RM-251, A,B,C,6D Once/ Month (1) (13) Once/3 Months Once/ Day (14) s, Main Steam Line Leak MS-TS-121, A,B,C,&D Once/ Month (1) Once/ Operating None = Detection 122, 123, 124, 143, 144, Cycle 145, 146, 147, 148, 149, 150 Main Steam Line liigh Flow MS-dPIS-116, A,B,C,6D Once/ Month (1) Once/3 Months None 117 Once/ Month (1) Once/3 Months None 118 Once/ Month (1) Once/3 Months None 119 Once/ Month (1) Once/3 Months None Main Steam Line Low Press. MS-?S-134, A,B,C,&D Once/ Month (1) Once/3 Months None liigh Reactor Pressure RR-PS-128, A & B Once/ Month (1) Once/3 Months None Condenser Low Vacuum MS-PS-103, A,B,C,&D once/ Month (1) Once/3 Months None Reactor Water C.U. liigh Flow RWCU-dPIS-170, A & B once/ Month (1) Once/3 Months None Reactor Water C.U. liigh Space RUCU-TS-150 A-D, 151, 152, Once/ Month (1) Once/ Operating None Temp. 153, 154, 155, 156, 157, Cycle 158, 159. RUCU-TS-81, A,B,E.F RWCU-TS-81 C,D,G,Il .-,-.e-w- i- + -m -m- -w.,... e--,-%=.ir w-s-+ rw -=m + awr ~ w e m7 1--rze' f----?*t?-t*- v~ p e-21r 4-&- e-'W e-m

0 COOPER NUCLEAR STATION S TABLE 4.2.D f MINIMUM TEST AND CALIBRATION FREQUENCIES FOR RADIATION MONITORING SYSTEMS P 3 Instrument Instrument System I.D. No. Functional Test Freo. Calibration Freo. Check j lustrument Channels "j Steam Jet Air Ejector Off-Gas System RMP-RM-150 A & B '12) (12) (12) Reactor Building Isolation and RMP-RM-452 A,B,C6D (12) (12) (12) 38; Standby Gas Treatment Initiation Liquid Radwaste Discharge Isolation RMP-RM-1 (11) (11) (11) Main Control Room Ventilation RMV-RM-1 Once/ Month (1) Once/3 Months Once/ Day isolation Mechanical Vacuum Pump Isolation RMP-RM-251, A, B, C & D See Table 4.2.A l ' Lonic Systems SJAE Off-Gas Isolation Once/18 Months Standby Cas Treatment initiation Once/18 Months Reactor Building Isolation Once/18 Months Liquid Radwaste Disch.-Isolation Once/6 Months Main Control Room Vent Isolation Once/6 Months Mech-ale Vacuum Pump Isolation Once/ Operating 4 Cycle i e --.-+.e . m se~e e.-*-rr - +- ,-s,r--- w .-er.- ev.- s+-. r ~ - -.=r-

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i NOTES FOR TABLES 4.2. A THROUCH 4.2.F

1. Initially once every month until exposure (M as defined on Figure 4.1.1) is 2.0 X 10 ; thereafter, according to Figure 4.1.1 (after NRC approval).

The compilation of instrument failure rate data may include data obtained from other - boiling water reactors for which the same design instrument operates in an environment similar to that of CNS. 'T 2. Functional tests shall be performed before each startup with a required frequency not to exceed once per week. 3. This instrumentation is excepted from the functional test definition. The functional test will consist of applying simulated inputs. Local alarm lights representing upscale and downscale trips will be verified but no rod block will be produced at this time. The inoperative trip will be initiated to produce a rod block (SRM and 1RM inoperative also bypassed with the mode switch in RUN). The functions that ~ cannot be verified to produce a rod block directly will be verified during the operating cycle. 4 4 Simulated automatic actuation shall be performed once each operating' cycle. Where possible, all logic system functional tests will be performed using the test jacks.

5. Reactor low water level and high drywell pressure are not included on Table 4.2.A l

since they are tested on Table 4.1.2. 6. The logic system functional tests shall include an actuation of time delay relays and timers,necessary for proper functioning of the trip systems.

7. These units are tested as part of the Core Spray System tests.

B. The flow bias comparator will be tested by putting one flow unit in " Test" (producing a rod block) and adjusting the test input to obtain comparator rod block. The flow bias upscale will be verified by observing a local upscale trip light during operation and verifying that it will produce a rod block during the operating cycle. 9. Performed during operating cycle. Portions of the logic is checked more frequently during functional tests of the functions that produce a rod block.

10. The detector will be inserted during each operating cycle and the proper amount of travel into the core verified.

j

11. Surveillance requirements for this system are definee in Table 4.21.A.1.

12. Surveillance requirements for this system are defined in Table 4.21.A.2.

13. This instrumentation is exempted from the instrument channel test definition.

The i instrument channel functional test will consist of injecting a simulated electrical signal into the measurement channels to test the alarm and trip functions. 14 Calibration shall be performed using a standard current source. The current source provides instrument channel alignment. Calibration using a radiation source shall be made each refueling outage. I t ~ Amendment No. 75,f?, 158 .gl. . Q

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3.2 BASES

(Cont'd) and the guidelines of 10CFR100 will not be exceeded. For large breaks up to the complete circumferential break of a 28-inch recirculation line and with the trip I setting given above, CSCS initiation and primary system isolation are initiated in time to meet the above criteria. Reference Paragraph VI.S.3.1 USAR. The high drywell pressure instrumentation is a diverse signal for malfunctions to the 1 water level instrumentation and in addition to initiating CSCS, it causes isolation i of Group 2 and 6 isolation valves. For the breaks discussed above, this instrumenta-tion will generally initiate CSCS operation before the low-low-low water level instrumentation; thus the results given above are applicable here also. The water level instrumentation initiates protection for the full spectrum of loss-of-coolant accidents and causes isolation of all isolation valves except Groups 4 and 5. l Venturis are provided in the main steam lines as a means of measuring steam flow and ) also limiting the loss of mass inventory from the vessel during a steam line break j accident. The primary function of the instrumentation is to detect a break in the main steam line. For the worst case of accident, main steam line break outside the j drywell, a trip setting of 150% of rated steam flow in conjunction with the flow limiters and main steam line valve closure, limits the mass inventory loss such that fuel is not uncovered, fuel clad temperatures peak at approximately 1000*F and q release of radioactivity to the environs is below 10CFR100 guidelinea. Reference = Section XIV.6.5 USAR. Te:rperature monitoring instrumentation is provided in the main steam tunnel and along the steam line in the turbine building to detect leaks in these areas. Trips are s 7 l provided on this ins tru:tentation and when exceeded, cause closure of isolation t valves. See Spec. 3.7 for Valve Group. The setting is 200*F for the main steam leak 9 detection system. For large breaks, the high steam flow instrumentation is a backup [ to the temp. instrumentation. High radiation monitors in the main steam tunnel have been provided to detect gross [ fuel failure as in the control rod drop accident. These monitors alert control room I t operators to potential fuel degradation by means of an alarm set at s1.5 times the normal background, and initiate a Group 7 isolation at 53 times the normal background. i t Pressure instrumentation is provided to close the main steam isolation valves in RUN j Mode when the main steam line pressure drops below Specification 2.1.A.6. The Reactor Pressure Vessel thermal transient due to an inadvertent opening of the j turbine bypass valves when not in the RUN Mode is less severe than th-loss of 1 feedwater analyzed in Section XIV.5 of the USAR, therefore, closure of the Main Steam j Isolation valves for thermal transient protection when not in RUN mode is not j required. { The Reactor Water Cleanup System high flow and temperature instrumentation are t arranged similar to that for the HPCI. The trip settings are such that core uncovery j is prevented and fission product release is within limits. 4 I s j i l l A:tendment No. 75,S3,SS,95, 158 .g4 r . --}}