ML19291B062
| ML19291B062 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 08/02/1979 |
| From: | Schwencer A Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19291B063 | List: |
| References | |
| NUDOCS 7908290269 | |
| Download: ML19291B062 (2) | |
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o UNITED STATES g
NUCLEAR REGULATORY COMMISSION
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NORTHERN STATES POWER COMPANY DOCKET NO. 50-282 PRAIRIE ISLAND NUCLEAR GENERATING PLANT UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 37 License No. DPR-42 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Northern States Power Company (the licensee) dated January 10, 1978 supplemented March 15, 1979, 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.
le 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.
2013 123 908290Alr9 7
. 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. DPR-42 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 37, 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 (AdWW$b A. Schwencer, Chief Operating Reactors Branch #1 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance: August 2,1979 2013 124
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UNITED STATES NUCLEAR REGULATORY COMMISSION y'
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,o NORTHERN STATES POWER COMPANY DOCKET N0. 50-306 PRAIRIE ISLAND NUCLEAR GENERATING PLANT UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 31 License No. DPR-60 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Northern States Power Company (the licensee) dated January 10, 1978 supplemented March 15, 1979, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Conmission'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 ar.d 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.
20l3 125 i = = c;;
. 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. DPR-60 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 31, 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 WLftddOp A. Schwencer, Chief Operating Reactors Branch #1 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance:
August 2,1979 20i3 126
ATTACHMENT TO LICENSE AMENDMENTS AMENDMENT NO. 37 TO FACILITY OPERATIt!G LICENSE NO. DPR-42 AMENDMENT N0. 31 TO FACILITY OPERATING LICENSE NO. DPR-60 DOCKET NOS. 50-282 AND 50-306 Revise Appendix A as follows:
Remove Pages Insert Pages TS 2.3-1 TS 2.3-1 Table TS 4.1-1 (page 1 of 4)
Table TS 4.1-1 (page 1 of 5)
Table TS 4.1-1 (page 2 of 4)
Table TS 4.1-1 (page 2 of 5)
Table TS 4.1-1 (page 3 of 4)
Table TS 4.1-1 (page 3 of 5)
Table TS 4.1-1 (page 4 of 4)
Table TS 4.1-1 (page 4 of 5)
Table TS 4.1-1 (page 5 of 5) 2013 127 MW
TS.2.3-1 2.3 LIMITING SAFETY SYSTEM SETTINGS, PROTECTIVE INSTRUMENTATION Applicability Applies to trip settings for instruments monitoring reactor power and reactor coolant pressure, temperature, flow, and pressurizer level.
Objective To provide for automatic protective action in the event that the principal process' variables approach a safety limit.
Specification A.
Protective instrumentation settings for reactor trip shall be as follows:
1.
Startup protection a.
High flux, interr,cdiate range (high set ' poin'.)
current equivalent to 1401, of full power.
b.
High flux, power range (low set point) 125% of rated power.
c.
High flux, source range -
neutron flux 1106 counts /second 2.
Core protection a.
High flux, power range (high set point) -
1108% of rated power.
b.
High pressurizer pressure - 12385 psig.
I c.
Low pressurizer pressure - 11815 psig, d.
Overtemperature t.T 1+ris ATf 16T
[K-K2 (T-T ' ) (l+,, g) +K 3(P-P') - f(AI)]
o 3
wherc Indicated AT at rated power 6T
=
O Average temper ature,
F T
=
567.3 F l
T'
=
Pressurizer pressure, psig P
=
psig 2235 P'
=
l.11 Kt 0.0090 K2
=
0.000566 K3
=
20i3 128 ry 30 sec.
=
4 sec.
1
=
2 Amendment No. 37, Unit 1 Amendment No. 31, Unit 2
TABIS 'IS !.1-1 (Page 1 of 5)
MI!'IMUM FREQUENCIES PCR CHECKS, CM.IS RAT.I O:15 AND TFST OF I!!STRUMENT_CHANNEU Channel Functional Responso Cose:ri p t.on Check.
Calibrate T'est Test Renarks
- 1) Cm:?/shiU when in cer/t;e 1..
t!uclear Power S ! '_)
D (2 )
M(!)
R
- 3) Heat balance
?.ange
- Mt_,
O (L )
M(5 )
3, Signal to aT: bistable action (per-Mki) missive, rod step, trips), with the P(7) exception of the items covered in Remark #7.
'4) Upper and lower charc.bers for axial of f-set using in-core detectors
- 5) Simulated signal for testing peri-tive and negative rate bistable action
- 6) Quadrant Power Tilt :-ioni tor
- 7) P8 and P10 permissives and the 257. High Plux Low Setpoint Trip.
l
- 2. Nuclear Inter-
- S i l)
MA T (2)
R
- 1) Once/shif t when ira service
.wediate Range
- 2) Log icvel; bistable action (per-missive, rod stop, trips)
- 3. Nuclear Source
- S(l)
NA T (2)
R
- 1) Once/shif t when in service l
Range
- 2) Bistable ace. ion (alarm, tripa) 4 Reactor Coolant *Sil,2)
It( 1,2,3)
M(1)
R(1)
- 1) Overtemperature AT Temocrature M(2)
R(2)
- 2) Overpower AT g
13)
- 3) Conv.rol Rod Bank Inseztien Lini:
m E
Monitor t!
6-g
- 5. 3-ractor Coolant S
R M
NA pc s
Flo.s
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- 6. ?ressurizer 5
R H
NA Mater Level bd_.
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- 7. Pressurizer S
R H
NA g
Pressure g
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TABLE TS.4.1-1 (Pa e 2 of 5) g MINIMilM FREQUENCIES FOR CHECKS. CALIBRATIONS AND TEST OF INSTRUMENT CHANNELS Channel Functional
Response
Description Check Calibrate Test Test Rema rks S.
4KV Voltage t.
NA R
M NA Reactor protection circuits only Frequency Sa. RCP Breakers NA R
T NA
- 1) with step counters
- 9. Analog Rod S(1)
R IT2)
NA
- 2) Rod Position Deviation Monitor Position M(2)
Tested by updating computer bank count and comparing with analog rod position test sional.
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TABLE TS.4.1-1 l
(Page 3 of 5)
Functional Response Channel Description Check Calibrate Lest Test Remarks 0.
Rod Position Bank S(1,2)
NA T( 3)
NA
- 1) With analog rod position Counters M(3)
- 2) Following rod motion in excess of six inches when the computer is out of service
- 3) Control rod banks insertion limit monitor and control rod position deviation monitors, 11.
Steam Generator Level S
R M
NA 12.
Steam Generator Flow S
R M
NA Mismatch 13.
Charging Flow S
R NA NA 14.
Residual Heat Removal S (l)
R NA NA
- 1) When in operation Pump Flow 15.
Boric Acid Tank Level D
R(1)
M(1) NA
- 1) Transfer logic to Refueling Water Storage Tank 16.
Refueling Water Storage W
R M(1) NA
- 1) Functional test can be performed by bleeding transmitter.
Tank Level 17.
Volume Control Tank S
R N.'.
NA Level 18a. Containment Pressure S
R M(1) NA Wide Range Containment Pressure 1)
Isolation Valve Signal SI Signal S
R M
NA
-Narrow Range Containment Pressure 18b. Containment Pressure Steam Line Isolation p, CD S
R M
NA 18c. Containment Presspre_.
Containment Spray ty NA R
R NA 18d. Annulus Pressure (Vacuum Breaker)
Amendment No. 37, Unit 1 Amendment No. 31, Unit 2
TABLE TS.4.1-1 (Page 4 of 5 )
l Channel Functional Response Description Check Calibrate Test Test Remarks
- 19.
Radiation Monitoring
- D R
M NA Includes all channels used for leak detection per Spec. 3,1 D.
and effluent release monitoring per Spec.
- 3. 9 and 5. 5.
20.
Boric Acid Make-up Flow NA R
NA NA Channel 21.
Containment Sump Level NA R
R NA Includes Sumps A, B,
and C 22.
Accumulator Level S
R R
NA and Pressure 23.
Steam Generator Pressure S
R M
NA 24.
Turbine First Stage S
R M
NA Pressure 25.
- M R
M NA Includes those named in the emergency Radiation Instruments procedure (referenced in Spec.
b.5 A.6.)
26.
Protection Systems NA NA M
NA Includes auto load sequencers Logic Channel Testing 27.
Turbine Overspeed NA R
M NA Protection Trip Channel
@ypagnmggShell M
NA NA NA Includes those used per Spec. 3.6 D.
28.
29.
Containment Air Temperature M NA NA NA Includes.those used per Spec. 3.6 C.
30.
Environmental Monitors M
NA NA NA Includes those used per Spec. 4.10 rN) 31.
Seismic Monitors R
R NA NA Includes those reported in Item 4 of CD Table TS.6.7-1 32.
Coolant Flow - RTD S
R M
NA LN Bypass Flowmeter 33.
CRDM Cooling Shroud S
NA R
NA FSAR page 3.2-56 Exhaust Air Temperature ha 34.
Reactor Gap Exhaust S'
NA R
NA-FSAR page 5.4-2 Air Temperature Amendment No. 37, Unit 1 Amendment No. 31, Unit 2
TABLE TS.4.1-1 (Page 5 of 5) l Channel Functional Response Description Check Calibrate Test Test Remarks 35.
Post-Accident Monitoring M
NA NA NA Includes all those in FSAR Table Instruments 7.7-2 that are not itemized.in Table TS.4.1-1.
36.
Steam Exclusion W
R M
Actuation System See FSAR Appendix 1, Section 1.14.6 S
Each Shift D
Daily W
Weekly M
Monthly Q
Quarterly R
Each refueling shutdown P
Prior to each startup if not done previous week T
Prior to each startup following shutdown in excess of 2 days if not done the previous 30 days.
NA Not, applicable J
C See Spee. h.1. D.
u uu Amendment No. 37, Unit 1 Amendment No. 31, Unit 2
UNITED STATES fi o,,
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,o SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION, SUPPORTING AMENDMENT NO. 37 TO FACILITY OPERATING LICENSE NO. DPR-42 AND AMENDMENT N0. 31 TO FACILITY OPERATING LICENSE NO. DPR-60 NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT NOS. 1 AND 2 DOCKET NOS. 50-282 AND 50-306 Introduction I
3 By letter dated January 10, 1978 supplemented on March 15, 1979, Northern States Power Company (the licensee) submitted a proposed amendment to the Technical Specifications for Prairie Island Nuclear Generating Plant, Units 1 and 2.
The amendment changes the surveillance frequencies of the nuclear instrumentation system low power level reactor trip functions.
The licensee has provided an analysis for our review indicating that the proposed change would not degrade safe operation of the plant.
In addition, we have corrected T' to read 567.3 F in 2.3. A.2.d on page TS 2.3-1. In amendment requests dated November 4,1976 and.lanuary 4, 1977 T' was proposed by the licensee to be 566.1 F.
A January 28, 1977 licensee withdrew that request and requested that T' remain at 567.3 F.
T' was inadvertently changed to 566.1 F in Amendments 28 and 22 issued March 28,1978.
Since the plant nomally operates at around 560*F, that error was of no safety significance.
It is appropriate at this time to revert to the original value of 567.3 F in the Technical Specifi-cations.
The effect of the requested change in surveillance frequencies would be to require functional testing of the nuclear source and intermediate range power level monitoring instrumentation channels prior to each startup following a shutdown which lasts longer than two days, provided the functional testing has not been performed in the previous 30 days.
The present specification requires the functional test prior to each startup if not done the previous week.
'7908290 M
. Background
Functional testing of a channel consists of electrical simulation of a signal as close as practicable to the channel sensor.
The level of the simulated signal is varied such that the operability and setpoints of the various bistables are checked. Thus, " functional testing" should not be ccnfused with " calibration."
Functional testing is a check of the electronic analog and logic sections of the channel.
Source Range Channels The source range channels are used by the operator to monitor r.eutron flux levels and rates of increase during reactor startup.
The two channels also operate an alarm and a trip in a one-out-of-two logic. The analyses of the startup accidents do not take credit for this trip, but instead conservatively assume failure of both source range channels, both inter-mediate range channels, and two of the four power range channels.
With such high redundancy, an extended functional test interval would be acceptable if trip failure were the only concern. However, the functional test also includes a checkout of the analog circuitry, which in addition to outright failure, is subject to drift.
The Technical Specifications require that the source range neutron flux 6
reactor trip setpoint be <10 counts'per second (CPS). This is the limit of the range of allowable values:
if the setpoint were discovered above this value, it would constitute a Technical Specification violation.
Unlike the Standard Technical Specifications 2, the Prairie Island Techni-cal Specifications do not specify a target value for the operator.
It is the responsibility of the licensee go specify values in his procedures whichhavesufficientmargigtothe10 CPS limit to allow for instrument drift.
The licensee stated that current practice is to set the trip set-5 CPS.
The source range channels read out on a 1ogarithmic point at 10 scale, so the margin is not as large as it would first appear.
The licensee has stated 3 that in the past five years, the RMS deviation observed has been equivalent to 4000 CPS in a six month period, with a maximum deviation equivalent to 5000 CPS.
This implies that in an eighteen mgnth fuel cycle, 95% of the time the actual setpoint will be f.20 x 10 CPS, whic.h (on a log scale) is less than 1/3 of the margin to the Technical Specification limit.
This is acceptable.
20i3 135
. The alarm setpoint drift need not be considered here because the only direct safety function of the alarm is to warn plant personnel of positive reactivity insertions during core alterations.
This is a shutdown function which is not addressed by the particular specification under consideration.
Finally, the indicating function to the operator during startup, which is used to calculate periods and other relative rather than absolute observations, should not be impaired by a 20% drift, particularly since the operator has access to both channels. Therefore, we find the longer test interval to be acceptable.
Intemediate Range Channels In addition to indicating neutron flux levels to the operator during startup, the intermediate range channels feed the P-6 pemissive, a rod stop, and a trip through a one-out-of-two logic. The trip is backed up by the source and power range trips, as discussed previously.
Once again, an extended functional test interval would be acceptable based only on trip failure considerations.
However, the analog circuitry testing must also be considered.
The Technical Specifications require that the intemediate range neutron flux reactor trip setroint be less than or equal to the equivalent of 40%
of rated core thermal power (not detector current). This is in contrast to the source range trip setpoint, which is stated directly in tems of detector output (counts per second).
Thus, there are two contributions to intermediate range setpoint drift:
deviations in the channel circuitry and changes in the relationship between neutron flux at the detector and core thermal power. The licensee has stated prevjously that total setpoint drift can be as high as 7% over a fuel cycle.
The functional test will check the channel circuitry contribution only. No check of detector-flux-to-power ratio is required by the Technical Specification, although data taken later in the startup sequence can be used to determine whether a violation has taken place. The licensee uses 30% of rated thermal power as a target value, thus allowing 10% margin to the 40% rated themal power 1imit in the Technical Specifications.
3 The licensee has stated that the RMS instrument deviation in an average seven and one-half month period has been equivalent tc 1.3% power.
For an eighteen month refueling cycle, this would correspond to 3.12% power RMS, or 5.13% power for a one-sided 95% probability.
This would imply that the 7% maximum total drift would be increased to 8.7%.
This is still less than the 10% margin to the Technical Specification limit.
Moreover, the sole purpose of this 10% margin is to allow for this drift.
Other uncertainties are accounted for elsewhere. Therefore, we find a refueling outage interval to be acceptable for functional tests of the intermediate range trip function.
2013 136
. The intemediate range rod stop is not a Technical Specification requirement.
Its sole function is to prevent inadvertent trips during startup, and is not required for safety. Therefore, it will not be considered further.
The P-6 permissive allows manual blocking of the source range trip, and de-energizing of the source range channels to present detector burnout only when the reactor neutron flux as indicated by the intermediate range channels is above the P-6 setpoint.
(Re-instatement of the source range channels is automatic when the intermediate range signal drops below the setpoint.) The purpose of the P-6 permissive is to ensure that the intermediate range channels are on scale before the source range channels are shut off.
The Technical Specifications do not address the P-6 permissive directly, but range detector signal is less than 10,g be active whenever the intermediate instead require the source range trip amperes.
(This assures at least one decade of source range to intermediate range overlap.) Thus, the P-6 permissive is a hard-wired backup to the operator's manual action. Down-ward drifts of the P-6 setpoint would not necessarily lead to a Technical Specification violation.
Because of this, and because it is expected that the instrument drift rate of the P-6 setpoint will be of the same order as the other bistables, we find the longer functional test interval to be acceptable for the P-6 circuitry.
Finally, the indicating functions of the intermediate range channels to the operator during startup could drift ~ up to 8.7% instead of 7% in terms of core thermal power. This is not a sufficient increase to cause a problem with range overlap, nor would it impair the ability of the operator to safely maneuver the reactor.
Therefore, the longer functional test interval is acceptable for the indicating functions.
Summary All safety requirements of the source and intermedi5te range channels have been examined, and in no case will the longer functional test interval significantly affect the reliability of the channels'. cquired safety functions.
Therefore, we find the proposed Technical Specification change to be acceptable.
2013 137
. Environmental Consideration We have determined that the amendments do not authorize a S..nge in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this determination, we have further concluded that the amendments involve an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR 51.5(d)(4), that an environmental impact statement or negative declaration and environ-mental impact appraisal need not be prepared in connection with the issuance of these amendments.
Conclusion We have concluded, based on the considerations discussed above, that:
(1) because the amendments do not involve a significant increase in the probability or consequences of accidents previously considered and do not involve a significant decrease in a safety margin, the amendments do not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of these amendments will not be inimical to the common defense and security or to the health and safety of
~
the public.
Date: August 2,1979 2013 138
f References
/
1.
Letter, L. O. Mayer (NSP) to Director of Nuclear Reactor Regulation (NRC) dated January 10, 1978, enclosing Request for Amendment to Operating License No. DPR-42 and DPR-60, dated January 10, 1978 by L. J. Wachter (NSP).
2.
" Standard Technical Specifications for Westinghouse Pressurized Water' Reactors," NUREG-0452, June 15,1978.
3.
Letter, L. O. Mayer (NSP) to Director of Nuclear Reactor Regulation (NRC) dated March 15, 1979.
i t,
4.
Letter, L. O. Mayer (NSP) to D. K. Davis (NRC) dated August 10, 1977.
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