ML20212G586
| ML20212G586 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 10/27/1997 |
| From: | Mckee P NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20212G570 | List: |
| References | |
| NUDOCS 9711060235 | |
| Download: ML20212G586 (9) | |
Text
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@ ist p
t UNITED STATES g
j NUCLEAR REGULATORY COMMISSION l
WASHINGTON, D.C. 20065 4 001 o
49.....,d M BTl1 EAST NUCLEAR ENERGY COMPANY DOCKET NO. 50-245 MILLSTONE NUCLEAR POWER STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 103 License No. DPR-21 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Northeast Nuclear Energy Company (the licensee) dated May 15, 1997, 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 13 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.
9711060235 971027 PDR ADOCK 05000245 P
PDR l
' 2.
Accoruingly, 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. DPR-21 is hereby amended to read as follows-(2)
Iechnical-Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 101 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 issuance, to be implemented within 60 days of issuance.
FOR T UCLEAR REGULATORY COMMISSION Phillip F~.
McKee Deputy Director for Licensing Special Projects Office 3
Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
October 27, 1997
ATTACHMENT TO LICENSE AMENDMENT NO.103 FACILITY OPERATIN3 LICENSE NO. DPR-21 DOCKET NO. 50-131 Replace the following pages of the Appendix A, Technical Specifications, with the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
Remove Insert 3/4 1-3 3/4 1-3 3/4 1-5 3/4 1-5 3/4 1-8 3/4 1-8 B 3/4 1-2 B 3/4 1-2 B 3/4 1-9 B 3/4 1-9 B ~,/4 1-10 s
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TABLE 3.1.1 (Continued)
I3 Reactor Protection System (Scram) Instrumentation Requirements a
Notes:
e h
5.
NOT USED l
~
6.
The design permits closure of any one valve without a scram being initiated.
7.
May be bypassed when necessary by closing the manual instrument isolation valve for scram of PS-1621 A through D during purging for containment inerting or deinerting.
-8.
When the reactor is subcritical 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.
High Flux IRM d.
Scram Discharge Volume High Level w1 e.
APRM Reduced High Flux 9'.
Not required to be operable when primary containment integrity is not required.
- 10. NOT USED l
- 11. Trip functions are not required to be operable if all control rods are fully inserted, and either g
electrically or hydraulically disarmed in accordance with Specification 4.1.D.
oo.
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TABLE 4.1.2 jf-SCRAM INSTRUMENTATION CALIBRATION 5
3 MINIMUM CALIBRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUMENT CHANNELS I
Calibration Minimum Calibration f
Instrument Channel SComLE1 Method Freauency (2)
APRM Output Signal (4)
B Heat Balance Once every 7 days APRM Flow Blas Trip B
Standard Current Source Once every 3 months APRM Reduced High Flux Trip (5)
B Standard Current Source once every 3 months IRM (5)
B Standard Current Source Refueling High Reactor Pressure A
Pressure Standard Every 3 months g
High Drywell Pressure A
Pressure Standard Every 3 months low Reactor Water A
Delta Pressure Standard Every 3 months
{
Condenser Low Vacuum A
Vacuum Pump Every refueling Generator Load Rejection A
Pressure Standard Every refueling High Water Level ir Scram Discharge A
Water Level Every 3 months Notes:
1.
A description of the 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 g
or are tripped.
If tests are missed, they shall be performed prior to returning i
f, the systems to an operable status.
3.
Maximum calibration frequency required is once per week.
4.
The heat balance method serves the calibration of the normal APRM high flux trip and the reduced APRM high flux trip.
y 5.
The IRM and SRM channels shall be determined to overlap during each startup after entering the STARTUP/ HOT STAND 6Y MODE and the IRM and APRM channels shall be determined to overlap 4
during each controlled shutdown, if not performed within the previous 7 days.
ow
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3.1 ' REACTOR PROTECTION SYSTEM BASES The outputs of the subchannels are combined in a 1 out of 2 logic; i.e.,
an input signal on either one or both of the subchannels will cause a trip system trip. The outputs of the trip systems are arranged so that a trip on both channels is required-to produce a reactor scram.
.9 This system meets the requirements of the proposed IEEE Standard for Nuclear-Power Plant Protection Systems issued September 13, 1966.
The system has a reliability greater than that of a 2 out of 3 system and somewhat less than that of a 1 out of 2 system.
With the exception of the Average Power Range Monitor (APRM),
Intermediate Range Monitor (IRM) channels, Main Steam Isolation Valve Closure, and Turbine Stop Valve Closure, each subchannel has one instrument channel.
When the minimum condition for operation on the number of operable instrument channels per untripped protection logic channel is met or if it cannot be met and the affected protection trip system is placed in a tripped condition, the effectiveness of the protection system is preserved; i.e.,
the system can tolerate a single failure and still perform its intended function of scramming the reactor.
Three APRM instrument channels are provided for each protection trip system.
APRMs #1 and #3 operate contacts in a logic subchannel and APRMs #2 and
- 3 operate contacts in the other logic subchannel.
APRMs #5, and f4 together with 86 and #4 are arranged similarly in the'l other protection trip system.
Each protection trip system has one more APRM than is necessary to _ meet the minimum number required per channel.
This allows the bypassing of. one APRM per protection-trip system for maintenance, testing or calibration.
MILLSTONE UNIT 1 B 3/4 1-2a Amendment No. pp, Jp/ 103 0312
4.1 REACTOR PROTECTION tiV5 TEM -
BASES Group (C) devices are active only during a given portion of the operational cycle. For example the IRM is active during startup and inactive during full-power opera,cion. Thus, the only test that is meaningful is the one performed just prior to shutdown or startup; i.e.,
the tests that are performed just prior to use of the instrument.
While included in Group (C), the f.ondenser Low Vacuum trip is treated differently. This is because the condenser low vacuum trip sensor can only be tested during shetdown. The primary function of this trip is to protect the turbine and condenser, although it is connected into the reactor protection system; thus testing the sensor at each refueling outage is adequate.
Calibration frequency of the instrument channels is divided into two groups. These are as follows:
a.
Passive type indicating devices that can be compared with like units on a continuous basic.
b.
Vacuum tabe or semiconductor devices and detectors that drift or lose sensitivity.
Experience with passive type instruments in generating stations and substations indicates that the specified calibrations are adequate.
For those devices which employ amplifiers, etc., drift specifications call for drift to be less than 0.4%/ month; i.e., in the period of a month a drlft of 0.4% would occur thus providing for adequate margin.
For the APRM system, drift of electronic apparatus is not the only consideration in determining a calibration frequency. Change in power distribution and loss of chamber sensitivity dictate-a calibration every seven days.
Calibration on this frequency assures plant operation at or below thermal limits.
SRM/IRM/APRM overlap surveillances are established to ensure that no gaps in neutron flux indication exist from subcritical to power operation for monitoring core reactivity status.
The overlap between SRMs and IRMs is required to be demonstrated to ensure that reactor power will not be increased into a neutron flux region without adequate indication. This is required prior to withdrawing SRMs from the fully inserted position since indication is being transitioned from the SRMs to the IRMs.
Tht. overlap between IRMs and APRMs is of concern when reducing power into the.IRM range. On power increases, the system design will prevent further increases (by initiating a rod block) if adequate overlap is not maintained. Overlap between IRMs and APRMs exists when sufficient IRMs MILL 5 TONE UNIT 1 B 3/4 1-9 Amendment No. 192, 103 0247
3.1 REACTOR '"!0TECTION SYSTEM BASES (continued) and APRMs concurrently have onscale readings such that the transition between the RUN and STARTUP/ HOT STANDBY Modes can be made without either APRM downscale rod block, or IRM upscale rod block. Overlap between SRMs i,
and IRMs similarly exists when, prior to withdrawing the SRMs from the fully inserted position, IRMs are above mid-scale on range 1 before SRMs have reached the upscale rod block.
As noted, IRM/APRM overlap is only required to be met during entry into STARTUP/ HOT STANDBY Mode from the Run Mode. That is, after the overlap requirement has been met and indication hat transitioned to the IRMs, maintaining overlap is not required (APRMs may be reading downscale once in the STARTUP/ HOT STANDBY Mode).
If overlap for a group of channels is not demonstrated (e.g., IRM/APRM overlap), the reason for the failure of the surveillance should be determined and the appropriate channel (s) declared inoperable. Only those appropriate channels that are required in the current condition should be declared inoperable.
The RPS LOGIC RESPONSE TIME surveillance ensures that the logic response time is less than or equal to the value assumed in the safety analyses.
The logic response time is measured from the opening of the sensor contact up to and including the opening of the trip actuator contacts (de-energization of the SCRAM pilot valve solenoids). The RPS LOGIC RESPONSE TIME acceptance criterion is 50 milliseconds.
RPS LOGIC RESPONSE TIME tests are conducted at least every OPERATING CYCLE. This frequency is consistent with the Millstone refueling cycle and is based upon plant operating experience, which shows that random failures of instrumentation components causing serious response time degradation, but not channel failure, are extremely unlikely.
B.
The peak heat flux shall be checked once per day to determine if the APRM scram requires adjusWent.
This will normally be done by checking the LPRM readings. Only a small number of control rods are moved daily, thus the peaking factors are not expected to change significantly and a daily check of the peak heat flux is adequate.
MILLSTONE UNIT 1 B 3/4 1-10 Amendment No. 103 l
0313