ML20209H048
| ML20209H048 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 01/28/1987 |
| From: | Zwolinski J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20209H046 | List: |
| References | |
| NUDOCS 8702050519 | |
| Download: ML20209H048 (10) | |
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, UNITED STATES
! g NUCLEAR REGULATORY COMMISSION D <j WASHINGTON, D. C. 2%66
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CONSUMERS POWER COMPANY DOCKET NO. 50-155 BIG POCK POINT PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 87 License No. DPR-6
- 1. The Nuclear Regulatory Comission (the Comission) has found that:
A. TheapplicationforamendmentbyConsumersPowerCompany(the licensee) dated August 15, 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 Comission; C. There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endancering 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 Comission's regulations and all applicable requirements have 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:
'0702050519 070120 PDR ADOCK 05000155 P PDR
a
-2 (2) Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 87, 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 k 1.,-
John A. Zwolinski, Director BWR roject Directorate #1 Division of BWR Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: January 28, 1987 4
O f
ATTACHMENT TO LICENSE AMENDMENT tl0. 87 FACILITY OPERATING LICENSE NO. DPR-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 chanoe.
REMOVE INSERT 6-5 6-5 6-8 6-R 7-3h 7-4 7-4 7-5 7-5 7-6 7-6 7-7 7-7 7-8*
7-8
- Pagination change only I
6.1.2.1 Source Ranae Monitor Channel - Channels 6 and 7 shall provide logarithmic neutron flux level and period information from source level to seven decades above source level, without moving detectors (spproximately 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 meter, 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 Intermediate Range Channels - Channels 4 and 5 provide logarithmic neutron flux level and period information from approximately 10-5g g, rated power for the 84-bundle core. The principal components in each channel shall be a neutron detector, dual high-voltage power supply, Log-N and period amplifier, Log-N indicator, period indicator, and Log-N recorder. The detectors shall be gamma compensated ion chsabers with a design sensitivity of at least 2.2 x 10-14 amperes /nv.
6.1.2.3 Power Range Channels'- Channels 1, 2 and 3 shall provide linear neutron flux level information from approximately 10-I% to 125% rated power for the 84 fuel bundle core. The principal components in each channel shall be a neutron detector, dual power supply, picoammeter with console range switch and power level indicator, and power level recorder. The detectors shall be gamma compensated ion chambers with a design sensitivity of at least 2.2 x 10-14 amperes /nv. The amplifier output shall be connected to the reactor safety system.
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.
6-5 Change No. 3, 7 Amendment No. ,43,g7,,87
6.1.5 (Centd)
(c) Both source range monitor channels shall be OPERABLE and measuring flux from the core during reactor stsrt-ups, prior to the power level at which the intermediate or power range channels become operative. However, if one of the two source range monitor channels become inoperative during the course of start-up and prior to the power level at which the intermediate or power range channels become operative, then it will be permissible to hold the control rod pattern and power level attained until both channels are again OPERABLE.
(d) There shall be a minimum of two intermediate range channe3s providing logarithmic neutron f3ux level information and period sergm protection during reactor start-up from approximately 10- % to approximately SI of rated power.
For reacter operation above approximately 5% of rated power, the logarithmic neutron flux level information and period scram protection are not required (see Section 6.1.2).
(e) Any one of the three power range flux monitors may be taken out of service for maintenance during reactor operation. If one monitor is cet of service, a trip on either of the two remaining monitors shall scram the reactor. When maintenance is necessary , no major changes in power level, flux distribution or the 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 aFainst a " cold-water accident" is provided by recirculaticn pumps and valve interlocking. The valves on either side of the recirculating pumps are interlocked with pump power such that each valve must be in its proper position before the pump motor can be started. If the suction valve to the pump is closed, the motor will be tripped. If the discharge valve and bypass valve are closed, the motor will be tripped.
(h) Minimum nuclear instrumentation in operation during SHUTDOWN operation shall be the same as that required for REFUELING OPERATION, except that only one source range monitor channel shall be required.
Change 3,7 6-8 Amendment No. 87
. s . ..
7.3.2 (Contd)
, (d) The source range monitor shall indicate a minimum of three counts per second with a signal-to-noise ratio of 3 to 1. This will be accomplished by withdrawing the proportional counter to a region of lower flux and observing the reduction in count rate.
In the event that neutron source strength is insufficient to produce the required count rate, an approach to criticality for reactor start-up shall be allowed provided that the following conditions are met:
(i) Prior to the first start-up after development of the low source strength condition. a critical approach (es) with the reactor vessel head off shall be performed to evaluate the source range monitor response and the control rod withdrawal sequence. Two additional low-level detectors shall be temporarily inserted in the vessel to monitor this head-off critical approach (es). These additional detectors shall indicate a minimum of 3 counts per second with a signal-to-noise ratio of 3 to 1.
(ii) Evaluation of source range monitor responses during the head-off critical approach (es) shall demonstrate that, by 4
the time the estimated k of the core reaches 0.995, i eachsourcerangemonitoIksreadingatleast3cpswitha I
3 to 1 signal-to-noise ratio.
(iii) Critical approaches with the head on and without the temporary in-vessel low-level detectors in service shall be permitted provided that the instrument response
! requirements of (ii) above have been demonstrated and provided that these same requirements are also met during each critical approach with the head on. Normal start-up may continue any time source range monitor count rate is at least 3 counts per second with a minimum signal-to-noise ratio of 3 to 1.
(iv) The procedure of (1) and (ii) above shall be repeated by conducting an additional head-off critical approach (es) in the event of either a core configuration change or a significant change in that part of the control rod with-drawal sequence for which the minimum count rate and signal-to-noise ratio specified above are not attainable.
1
! (v) The site reactor engineer or his alternate will be on site during all head-off critical approaches and instrument response evaluations and during the initial head-on start-up following head-off evaluations.
1 Change No. 6 7-4 Amendment No. 29,gg, 87 l
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7.3.2 (Contd)
(e) Critical approaches shall be monitored using source range monitors. The start-up rate shall be restricted to demonstrate that the CIC picoammaters overlap the source range monitors at about 10-5% indicated full power on the picos m tera is normal and satisfactory for control and safety purposes before continuing further into the power range. Control rod withdrawal sequence shall be specified and limited to those sequences shown by previous analysis or tests to preserve fuel integrity in the event of accidental reactivity insertion either while starting up or at power.
(f) The power shall be adjusted once criticality is reached to maintain a reactor vessel temperature rise rate not to exceed 100*F per hour.
(g) The turbine shaft sealing system shall be placed in service as soon as sufficient steam pressure is available (approximately 150 psig).
(h) The condenser shall be evacuated with the mechanical vacuum pump and the air ejector will be placed in service.
(1) Turbine heating shall be started during this operation sequence.
After turbine heating is completed, the turbine shall be gradually brought up to speed.
(j) The mode of turbine control shall be transferred to the initial pressure regulator.
(k) The control rods shall be adjusted to provide the desired power distribution within the core.
7.3.3 Hot Start-Up Whenever the plant has been shut down for a period of time with the reactor vessel and auxiliaries remaining pressurized, a hot start-up procedure shall be followed to return the plant to service. This procedure will be essentially independent of the cause of shutdown assuming that the cause is recognized and any nonstandard conditions have been corrected. The reactor instrumentation shall be reset and downscaled and a hot start-up checklist shall be completed prior to the withdrawal of control rods.
A coupling integrity check shall be unde in accordance with Section 5.2.2(d). The start-up shall then proceed in accordance with Paragraphs (d) through (k) of Section 7.3.2 of the normal cold start-up procedure outlined above.
I Change 6. 7 Amendment No. 87 l 7-5
. . . . . . . . . _ . ._ .. s. .
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 of the reactor vessel shall not be allowed to exceed 100*F per hour.
(d) The reactor shutdown cooling system shall be placed in operation whenever reactor pressure drops below a pressure sufficient to maintain turbine seals. This system will complete the cooling of the reactor water to 125'F.
.i (e) A minimum of one source range monitor channel and one power range channel shall be left in operation. All instrumentation pertaining to control of activity release shall be left in operation.
7.3.6 Short Duration Shutdown 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.
l 76 Amendment No. 87
7 7.4 REFUELING CPERATION The refueling operation shall be conducted in accordance with the following basic principles:
(a) Detailed written procedures shall be available prior to each refueling outage.
(b) The insertion and removal of fuel bundles and channels shall be done through the top of the reactor vessel aftor opening reactor vessel head closures as appropriate. Water shielding shall be provided by flooding the reactor vessel and the refueling extension tank. Fuel bundles and channels shall be handled by means of a grapple, transfer cask, and crane.
Fuel shall be replaced according to the following sequences (i) Removal of selected bundles from core and transfer to spent fuel storage.
(ii) Reshuffling of remaining bundles in core as desired.
(iii) Insertion of new bundles in vacant positions as desired.
Shutdown margin verifications and suberiticality checks shall be made as required by Section 5.2.5. Assembly replacement shall proceed as described above until the desired number of fuel assemblies have been changed.
(c) The trip devices specified in Section 6.3.1 shall be in service and connected to the reactor safety system during all refueling operations.
No additional instrumentation need be placed within the core lattice if the out-of-core instrumentation produces a significant response to the suberiticality check in the region to be altered. If this criterion cannot be met, a low-level neutron detector, measuring neutron flux, shall be located near the region to be altered.
In addition, both source range monitor channels shall be in service and measuring neutron flux during all REFUELING OPERATIONS.
(d) The procedure which shall be used for cere alterations which increase reactivity shall be as outlined in Section 5.2.5(a).
Communications between the control room and the loading area shall exist during all core alterations.
Change No. 7 7-7 Amendment No. 70,87 i
i 7.4 (Contd)
(e) The liquid poison system shall be available and ready for use.
4 (f)) Containment sphere integrity provisions shall be in effect during refueling operations.
(g) Unirradiated fuel shall nomally be stored in air in a new i fuel storage area within the containment sphere.
i (h) Irradiated fuel and irradiated channels shall be stored i
in the spent fuel storage pool.
(i) The minimum refueling crew during refueling operations shall be four men. There shall be a licensed operator in the con- i trol room at all times, and the Shift Supervisor shall be l in charge. I l
(j) Functional testing of the trip mechanism of the fuel l transfer cask safety catch device shall be performed prior !
, to commencing refueling activities.
1 7.5. 7.5.1 - 7.5.6 Deleted (Change 46,12/19/75)
I 7.5.7 It shall be permissible to remove a control rod drive from the i
reactor vessel when the reactor is in the shutdown condition and the mode selector switch is locked in the " Shutdown" position.
The core shutdown margin of 0.3% keff /keff with the strongest y i control rod out of the core shall have been met prior to the
! control rod drive removal; and in addition, the equipment shall bo l properly tagged. The control rod drive that was removed shall '
l without delay be replaced by a spare control rod drive or the i original control rod drive shall be reinstalled. One control rod i drive pump shall be operating daring removal and reinsertion and
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during the time the control rod drive is outside the reactor vessel.
I 7.6 OPERATIONAL TESTING OF NUCLEAR SAFEGUARD SYSTEMS i Procedures for testing of plant components and safety systems ,
l Which have a potential safeguards function are prescribed in I i Sections 3.0 through 6.0. These tests and frequency of testing I i
shall be as tabulated.
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Change 8, 46 '
78 Amendment No. 70.87
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