ML20154J187

Amend 91 to License DPR-6,changing Tech Specs to Reflect Features & Terminology Used W/New out-of-core Power Range Instrumentation

ML20154J187
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	05/17/1988
From:	Muller D Office of Nuclear Reactor Regulation
To:
Shared Package
ML20154J170	List: ML20154J167 ML20154J187 ML20154J198 ML20154J201
References
NUDOCS 8805260214
Download: ML20154J187 (11)

Text

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• UNITED STATES 3N  % NUCLEAR REGULATORY COMMISSION j I g WASHING TON, D. C. 20555 y e

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CONSUMERS POWER COMPANY BIG ROCK POINT PLANT DOCKET N0. 50-155 AMENDMENT TO FACILIY OPERATING LICENSE Amendment No. 91 License No. DPR-6

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

A. The application for amendment by Consumers Power Company (the licensee) dated November 9, 1987, 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. Tne 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 I

E. The issuance of thir, amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have i been satisfied,

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and Paragraph 2.C.(2) of Facility Operating License No. OPR-6 is hereby amended to read as follows:

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2-L Techn Eal Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 91, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Daniel R. Muller, Acting Director Project Directorate III-1 Division of Reactor Projects - III, IV, V

& Special Projects

Attachment:

Changes to the Technical Specifications Date of Issuance: May 17, 1988

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ATTACHMENT TO LICENSE AMENDMENT NO. 91 FACILITY OPERATING LICENSE N0. OPR-6 DOCKET NO. 50-155 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.

REMOVE INSERT 5-2 5-2 6-4 6-4 6-5 6-5 6-6 6-6 6-7 6-7 6-8 6-8 6-9 6-9 7-6 7-6

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,.S PUEL SUNCLt AND CH ANNEL POSITION COCROINATES h ouT-CP-CORE INSTRuutNTATION 3-2 A'rendment no. 91

Qg8 .2 (Continued) .

Trip Contacts Coincidence Scram- Warning Sensor and in Each in Each Setting and Instrument Annunciation l

Trip Device Channel Channel Tolerance Special Features Ranges Trip Set Point Loss of auxiliary 1 1 out of 1 52I 20 volts Closes all automat-l power supply ically actuated con-(voltage relay) cnntainment sphere isolation valves.

l Nigh Neutron Flux 3a 2 out of 3 120 1 SI Interlocka 0 to 150I 105 1 SI (Prevents prevent control (log readout l

(each of 3 Power from 0 to II, Range Monitors rod withdrawal outward rod

has a flux trip as described in linear motion) contact) Section 6.2.1 readout from 1 to 150%)

Power Range Monitor 3 1 downscale One Monitor N/ One Monitor cn Circuit Failure 1 upscale <1 x 10 I7 <1 x 10 7I In (downscale) (downscale) and either (Prevents of other two outward rod monitcre >120I notion) or <10 seconds (upscale)

Short Feriod (each 3* 2 out of 3 10 1 2 seconds Automatically -100 seconds 15 1 2 seconds of 3 Fower Range bypassed at to infinity (Prevents out-Monitors has a approximately II to +10 seconds ward rod motion) period trip (Also 20 1 2 second alarm f{,g,r contact) operated by

,j(jg Source Range jg e Honitor)

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, . Manual Scram 1 1 out of 1

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• goth High N=stron Flux and Short Period operate the same set of 3 trip contacts.

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6.1.2.1 Source Range Monitor Channel - Channels 6 and 7 shall provide logarithmic ntutron flux level and period information from source level to seven decades above source level, wich'out moving detectors (approximately 10 10 to 10~3% of rated power). The principal components in each channel shall be a neutron detector, current pulse amplifier, source range monitor instrument, log count rate metet, log count rate recorder and period meter. Gas-filled Boron-10 lined proportional counters with a sensitivity of approximately 12 counts /nv shall be used as detectors. Provisions shall be made for remotely positioning the detectors. By moving the detectors away from the midplane of the core, their effective range may be extended.

A short period on either channel shall be annunciated in the control room.

6.1.2.2 Deleted 6.1.2.3 Power Range Monitor Channel - Channels 1, 2 and 3 shall provide logarithmic neutron flux level information and period scram protection from approximately 1 x 10~7% to 1% rated power and linear neutron flux level information from approximately 1% to 150% rated power f or the 84 fuel bundle' core. The principal components in each channel shall be a neutron detector, DC-vide range monitor, power level and period recorders and operator display assembly which indicates power level and period. The detectors shall be gamma compensated ion chambers with a design sensitivity cf at least 2.2 x 10 1" amperes /nv. The channel output shall be connected to the reactor saf ety system to provide high neutron flux and short period scram protection.

6.1.2.4 In-Core Flux Monitors - In-core flux monitors shall be used to evaluate predicted power distributions and detect power oscillations or deviations from expected power distributions in time for the operator to take corrective action to avoid exceeding local heat flux limits.

t 6-5 Change No. 3, 7.

Amendment No. 45, 23. 27,91

E 6.1.2.5 Deleted 6.1.3 Reactor Safety System Bypass The following tabulation gives the permissive functional con-ditions during which certain react safety system sensors are bypassed by the reactor safety system mode selector switch.

A key-lock reactor mode ; witch shall b provided, having "Shut-down", "Refuel", "Bypass Dump Tank" and "Run" positions, Tnese positions shall have the following functions:

Mode Selector Switch Position Trip functions 3yoassed Run None (e)

.'." ss Dump Tank (a) Low Steam Drum Water level Recirculation Waterline Valves Closed Steam Line Backup Isolation Valve Closed High Water Level in Scram Dump Tank (b)

High Condenser Pressure Refuel (d) Low Steam Drum Water Level Recirculation Waterline Valves Closed Steam Line Backuo Isolation Valve Closed High Condenser Pressure Shutdown None (c)

(a) Cont,ol rod withdrawal is prevented by interlock while switch is in this mode position.

(b) Bypass of this trip function is necessary to enable emptying the dump tank af ter a scram.

Change No. 7, 6-6 Amendment No.;i

6.1.3 (Contd.)

(c) Wit 5themodeswitchinthe"shutdown"position,brththe scram circuit and the control rod withdrawal circuit are i o pe.n . The ventilating duct circuit power supply-is transferred to a point which provides penetration closure protection through signals from "high containment sphere pressure" and low water level in reactor vessel." This permits rernel ventilation in the containment sphere during shutdown when the control rods are ,

held in the full-in prsition, hone of tha reactor safety system signals are bypassed since there is no need to withdraw control rods.

(d) With the mode switch in the refuel position ard the crane positioned over the reactor vessel, crane operation is preverted if any one rod is withdrawn from full-in oosition.

(e) High condenser pressure reactor trip is automatically bypassed any time steam drun pressure is below a set poi t naxirum of 500 psig.

6.1.4 Related Systems (a) Emergency Condenser Control A pressure switch shall initiate automatic operation of the emergency condenser if the reactor pressure reaches 100:10 psi abnve the reactor operating pressure.

6.1.5 Operatinn Requirements (a) Except as otbe vise provided in these Technical Specifications, the reactor safety systen shall be OPERABLE during POWER OPERATION as indicated in Section 6.1.2. This systen shall be functionally tested during each MMOR RETUELING shutdown, but not less frequently than once every 18 months and in addition shall be tested not less frequently than once a month using the switches provided to simulate sensor trips.

(b) The emergency condenser system control initiation sensors shall be functionally tested at each KUOR REFUELING shutdovu but not l less frequently than once every 18 months.

6-7 /

AmendmentNo.26.J.%,91

6.1.5 (Centd)

(c) Both source range monitor channels shali be OPERABLE and measuring flux from the core during reactor start-ups. prior to the power level at which t.be power range monitor channels become operative. However, if one of the two source range monitor l channels become inoperative during the course of start-up and -

prior to the power level at which the power range monitor channels become operative, then it will be permissible to hold l the control rod pattern and power level attained until both channels are again OPERABLE.

(d) Deleted.

(e) All power range monitor channels shall be OPERABLE and able to provides (1) Logarithmic neutron flux level information and period scram protection during reactor start-up from approximately 1 x 10 7% to approximately 1% of rated power. For reactor operation above approximately 1% of rated power, logarithmic neutron flux level information and period scram protection are not required.

(2) Linear neutron flux level infor1sation and high neutron flux scram protection from approximately 1% of rated power to rated power.

Any one of the three power range monitor channels may be taken out of service for surveillance testing or maintenance during reactor operation. If one channel is out of service, a trip on either of the two remaining channels shall scram the reactor.

When maintenance is necessary, no major changes in power level, flux distribution or control rod pattern shall be made.

(f) During POWER OPERATION in-core flux monitors shall be operating to insure that local heat flux limits specified in Section 5.2.1 are not exceeded.

(g) Protection against a "cold-water accident" is provided by recirculation pu=ps and valve interlocking. The valves on either side of the recirculating pumps are interlocked with pu=p power such that each valve must be in its proper position before the pump motor can be started.

is closed, the mote vill be tripped.If the suction valve to the If the discharge pump valve e

and bypass valve are closed, the motor will be tripped.

(h) Minimum nuclear instrumsntation in operation during SHUTDOWN operation shall be the same as that required for REFUELING OPERATION, except that only one source range monitor channel shc11 be required. '

Change 3.7 6-8 Amendment No. M 91

6.2 CONTROL ROD WITHDRAWAL PERMISSIVE SYSTEM 6.2.1 Interlocks Interlocks shall prevent control rod withdrawal when any of the following conditions exists (a) When any two of the thirty-two scram accumulators are at pressure below 700 psig.

(b) When any one of the three power range monitor channels reads:

(1) Less than 1 x 10 %7 power, when reactor power is above the operating range of the source range monitoring channels, (2) Greater than 105% power, or (3) A reactor period less than 15, seconds.

(c) When the scran dump tank is bypassed.

(d) When the mode selector switch is in the shutdown position.

6.2.2 Operatina Requirements The. control rod withdrawal permissive interlocks shall always be OP ERABLE. No further withdrawal of control rods will be permitted if one of these circuits is found to be inoperable.

Permissive circuits shall be functionally tested prior to 'each KAJOR l REFUELING but no less frequently than every 18 months. However, the refueling interlocks will be functionally tested prior to each MAJOR RETUELING. .

6.3 REFUELING OPERATION INTERLOCK SYSTEM 6.3.1 Reactor Refueling System All of the trip devices not bypassed by the mode selector switch in the refuel position shall be OPERA 3LE during all REFUELING OPERATIONS. This shall include the sensors and trip devices of the reactor safety system as specified for POWER OPERATION as follows:

High Reactor Pressure Low Reactor Water Level High Containment Sphere Pressure High Scram Dump Tank Level Loss of .arx111ary Powet Supply High Neutron Flux Short Period Manual Scram 6-9 Amendment No. 70, 79,91

9 7.3.4 NORMAL POWER OPERATION During normal POWER OPERATION, the initial pressure regulator shall maintain the reactor pressure at its rated value by operating the turbine admission valves. The turbine-generator load shall be established by the control rod positions. The principal function of the operating personnel during this period shall be as follows:

(a) The maintenance of a continuous watch in the control room for prompt attention to any annunciated alarms.

(b) The adjustment of the control rod pattern to accommodate changes in reactivity and to maintain the desired power distribution.

(c) The evaluation of abnormal conditions and the initiation of corrective action as required.

7.3.5 EXTENDED SHUTDOWN An extended SHUTDOWN shall be accomplished as follows:

(a) Reactor power shall be reduced by manipulation of the control rods, and the main generator load shall be decreased simultaneously. The turbine-generator shall be separated from the system.

(b) All control rods shall be inserted.

(c) The removal of reactor decay heat and the reduction of reactor pressure shall be accomplished by controlling reactor steam flow. The rate of cooling cf the reactor vessel shall not be allowed to exceed 100*F per hour.

l (d) The reactor shutdown cooling system shall be placed in operation i whenever reactor pressure drops below a pressure sufficient to maintain turbine seals. This system will complete the cooling l

of the reactor water to 125'F.

(e) A =ini="- of one source range monitor channel and one power range monitor channel shall be lef t in operation. All l instru'm antation pertaining to control of activity release shall be 1tle in operation.

7.3.6 SHORT DURATION SHUTDOWN l A SHUTDOWN of short duration may be accomplished while maintaining  ;

system pressure. The turbine-generator shall be unloaded and separated from the system. Reactor heat shall be accommodated by system losses or bypassing steam to the main condenser.

7-6 Amendment No. 87, 91

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