ML20053F099
| ML20053F099 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 05/20/1982 |
| From: | Vassallo D Office of Nuclear Reactor Regulation |
| To: | Northern States Power Co |
| Shared Package | |
| ML20053F100 | List: |
| References | |
| DPR-22-A-010 NUDOCS 8206100440 | |
| Download: ML20053F099 (11) | |
Text
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. UNITED STATES h
NUCLEAR REGULATORY COMMISSION g
,.t WASHINGTON, D. C. 20555
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NORTHERN STATES POWER COMPANY DOCKET NO. 50-263 MONTICELLO NUCLEAR GENERATING PLANT AMENDMENT TO FACILITY OPERATING LICENSE
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Amendment No.10 License No. DPR-22 1.
The Nuclear Regulatory Comission (the Commission) has found that:
A.
The application for amendment by Northern States Power Company (the licensee) dated October 10, 1980 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission'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 Comission's regulations; D.
The issuance of this amendment wi?1 not be inimical to the comon 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 Specifi-cations as indicated in the attachment to this license amendment, and paragraph 2.C.2 of Facility Operating License No. DPR-22 is hereby amended to read as follows:
2.
Technical Specifications The Technical Specifications contained in Appendices A and B as revised through Amendment No.10 are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.,
820610044o 820510 PDR ADOCK 05000263 P
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This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing'
Attachment:
Changes to the Technical Specifications
'Date of Issuance:
May -20, 1982 J
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9 ATTACHMENT TO LICENSE AMENDMENT NO. 10 l
FACILITY OPERATING LICENSE NO. DPR-22 DOCKET NO. 50-263
- . Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
Remove Insert l
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32 32 57 57 61 61 71 71 83 83 83A 92 92 i
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TABLE OF CONTENTS Page 1.0 DEFINITIONS 1
2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS
-6 2.1 and 2.3 Fuel Cladding Integrity 6
2.1 Bases 10 2.3 Bases 14 2.2 and 2.4 Reactor Coolant System 21 2.2 -Bases 22 2.4 Bases 24 3.0 ' LIMITING CONDITIONS FOR'0PERATION AND 4.0 SURVEILLANCE REQUIREMENTS 26 3.1 and 4.1 Reactor Protection System 26 3.1 Bases 35.
4.1 Bases 41 3.2 and 4.2 Protective Instrumentatio'n 45 A.
Primary Containment Isolation Functions 45 B.
Emergency Core Cooling Subsystems Actuation 46 C.
Control Rod Block Actuation 46-D.
Air Ejector Off-Gas System 46 E.
Reactor Building Ventilation Isolation hnd Standby Gas Trea,tment System Initiation 48 F.
Recirculation Pump Trip Initiation 48 64
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3.2' Bases 4.2 Bases 72.
76 3.3 and 4.3 Control Rod System A.
Reactivity Limitations' 76 B.
Control Rod W. thdrawal 77 i
C.
Scram Insertion Times 81 D.
Control Rod Accumulators 82 83 E.
Reactivity Anomalies
.83 F.
Scram Discharge Volu.me 83A G.
Required Action 84 o
3.3 and 4.3 Bases I
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V TABLE h.l.1 SCRAM INSTRUMENT FUNCTIONAL TES*IS MINIlut EUNCTIONAL TEST FREQUENCIES FOR SAFETY INSTRUME! CATION AND COh"ITIOL CIRCUITS IIGTRUMENT CilANNEL
- GROUP" FUNCTIONAL TEST MINIMUM FREQUENCY (h)
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liigh Reactor Pressure A
Trip Channel armi Alarm Note 1 liigh Drywell Pressure A
Trip Channel and Alarm Note 1 Low Reactor Water Level (2)
A Trip Channel and Alarm Note 1 l
liigh Water Level in Scram Discharge A
Trip Channel and Alarm Once Each Month r
Condenser Low Vac.
A Trip Channel and Alarm Note 1-r Main Steam Line Isolation Valve Closure A
Trip Channel and Alarm Note 1 Turbine Stop Valve Closure A
Trip Channel and Alarm Note 1 Manual Scram A
Trip Channel and Alarm Note 1 Turbine Control Valve Fast Closure A
Trip Channel and Alam Note 1 APRN/ Flow Reference (5)
D Trip Output Relays Once each eek IRM(5)
C Trip Channel and Alarm Note 3 liigh Steam Line Rad. (5)
B Trip Channel and Alarm
- Once each week Mode Switch in Shutdown C
Place mode switch in Each refueling outage shutdown 32
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Amendment No.10 l
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l Table 3.2.3 - Continued Instrumentation That Initiates Rod Block Reactor Modes in Whicir Hin. No. of Oper-Function Must Be Operable Total No. of able or Operating or Operat ing and Allow-Instrument Instrument Channels able Bypass Conditions **
Channels per Per Trip System Required Function Trip Settings Refuel Startup Run Trip system (Notes 1,6)
Conditions
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HBM a.
Upscale 165W + 43 X(c) 1 1* *(Note 5)
D or E (flow ref-(Note 2)
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erenced) b.
Downscale
>3/125 fulI X(c) 1 1 (Note 5)
D or E
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Scram Discharge Volume Water Level.<l8 gal k
X 1
1 B and D, or A liigh Notes:
(1) There shall be two operable or operating trip systems for eAch function.
If the. minimum number of operable or operating instrument channels cannot be met for one of the two trip systems, this condition may exist up to seven days provided that during this time the operable systers is functionally tested immediately and daily thereaf ter. 'Ihis note is not applicable to the Scram Discharge Volume Rod Block since it exists in only one trip system.
(2)
"W" is the reactor recirculation driving fl'ow in percent.
-(3) Only one of the four SRM clIannels may be bypassed.
(4) There must be at least one operable or operating IRM channel monitoring each core quadrant.
(5) One of the two RBMs may be bypassed for maintenance and/or t'esting for periods not in excess of
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24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 30 day period.
An REM channel will be considered inoperable if there are less than half the total number of normal inputs from any LP'RM level.
.e, 3.2/4.2
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l Amendment No.10
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'Ibble 1 '.2.1 6
Minimisa Test arx1 Calibration Frequency For Core Cooling Ilod Block and Isolation Instnanentet. ion Inotnament Channel Test (3)
Calibration (3)
Senoor check (3)
ECCS IllSTI5JMENTATION 1.
Ecoctor Iov-Inv Woter tevel (Note 7) once/ month Once/3 monthe Once/Shi f t 2.
Drywell liigh Precoure (Note 7) once/ month Once/3 monthe None 3
Ilcoctor Iow Prenoure (nunp Stort,)
0 :te 1 Ouec/3 monthe None is.
Ecoct.or lov Prcanure (Valve Ihnniocive)
Nnte 1 Onec/3 monthe None 5
Undervoltage linergency Duo Icrueling Outoge Denseling Outage None 6.
Iow Prenoure Core Cooling nsapa Dicchorge Preocure Interlock Note 1 Once/3 montba None
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Ioos of Auxiliary Power Henselin6 Catage Fefbeling Outoge None.
HO'D DIOCKS 1.
APH4 Downscale Notes (1,5)
Once/3 monthe None 2.
APH4 Flow Yoriable Notes.(1,5)
Once/3 monthe None.
Notes 2,5 Note 2 Note 2 3
IIM Upocale h.
IIH Downacale Notes 2,5 Note 2 Note 2 Notes 1,5 Once/3 months None 5
Jur4 Upocale Noten 1,5 Once/3 monthe None 6.
Ult 4 Downocale Notes 2,5 Note 2 Note 2 7
SH4 Upacale 8; Sit 4 Detector.not in Start-up Position Note 2 Note 2 Note 2
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Scram Discharge Volume-Iligh Level Once/3 months Refueling Outage None l
MAIN STFJ\\!4 IINE ISOIATION 1.
Stcom 'Ibnnel liigh Temperature Denicling Outoge Beibeling Outage None 2.
Stcom Line Illgh Flow Note 1 Once/3 months Once/3hif t 9
68 3 2/Is.2 61 Amendmen t No.,2', 10
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)ft Table 3.2.7 - Continued Trip Function and Deviations o
Trip Function Deviation Instrumentation That initiate's Emergency Low-Low Reactor Water Level
-3 Inches Core Cooling Systems Table 3.2.2 Reactor Low Pressure (Ppop
-10 psi' Start) Permissive High Drywell Pressure
+1 psi Low Reactor Pressu're (Valve
-10 psi Pennissive)
IRM Downscale
-2/125 of Scale Instrumentattop That Initiates Red Block IRM Upscale
+2/125 af Scale Table 3.2.3 APRM Bownscale
-2/125 ef Scale APRM Upscale See Basis 2.3 s
RBM Downscale
-2/125 of Scale RBM Upscale Same as APRM Upscale Scram Discharge Vo.ume-High
+ 1 gallon Level Instrumentation That Ini tia tes High Reactor Pressure
+ 12 psi Recirculation Pump Trip Low Reactor Water Level
,3 Inches A violation of this specification is assumed to occur only when a device is knowingly set outside of the limiting trip settings, or,, when a sufficient number of devices have been affected by any means such that the automatic function is incapable of operating within the allowable deviation while in a reactor mode in which the specified function must be operable or when actions specified are not initiated as specified.
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3.0 LIMITITU CDtDITICOS FOR OPERATIQ1 4.0 SURVEIIINJCE IUQUIREMENIS E.
Reactivity An malies E.
Reactivity Anmalies At a specific steady state base condition During the startup test progra and at of the reactor actual control rod inventory each startup following refueling outages, will be periodically cmpared to a normal-the actual rod inventory shall be cnn-ized amputed prediction of the inventory.
pared to a normalized ccuputed prediction If the difference exceeds one per cent, delta k, of the inventory. %ese cmparisons will reactor power operation shall not be per-be used as base data for reactivity non-mitted until the cause has been evaluated itoring during subsequent power operation and appropriate corrective action has been throughout the fuel cycle. At specific ampleted.
power operating conditions, the actual rod configuration will be ampared to the configuration expected based upon appropriately corrected past daca. 'Ihis cmparison will be made at least every equivalent full power month.
P.
'Ihe scram discharge volume drain and vent 1.
The scram discharge volume drain and valves'shall be operable whenever more than vent valves r. hall be verified open at one operable control rod is withdrawn (not least once per month. Each valve shall including rods renoved per Specification be cycled quarterly. These valves 3.10.E or inoperable rods allowed by may be closed intermittently for testing 3.3.A.2).
under administrative control.
During each refueling outage verify the scram discharge volume drain and vent
- valves, Close within 30 seconds af't'er receipt n.
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of a reactor scram signal and b.
Open when the scram is reset.
2.
If the scram discharge volume drain or vent valve is made or found incperable, at leest all but one, operable control rods (not including rods renoved per Specification 3.10.8 or inoperable rods allowed by 3.3.A.2) shall bg fully inserted within ten hours.
83 3.3/4.3 Amendment No.10
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3.0 LIMITING CONDITONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS
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Required Action
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If Specifications 3.3.h through D above are not met, an orderly shutdown shall be initieted and have reactor' in the cold shutdowa condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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Amendment No.- 10
D uez Continuid:3 3 and 4.3:
Deviations beyond.this magnitude vould not be expectzd cod vsuld require th: rough sv21uation..
One per cent reactivity limit is considered safe e,ince an insertion of this reactivity into the core would not 1 cad to transients exceeding design conditions of the reactor system.
As was noted above reactivity ancualies can be found by cmparison of the actual control rod
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inventory to the predicted inventory at a selected base condition. For example, the predicted control rod inventory atI)0% power at a specified point in time can be compared to the actual i
control rod inventory at 100% power and at the specified time to detemine if a reactivity anmaly exists. he Monticello Plant has been designed to increase or decrease power level as the system load danand changes. For this type of plant an equilibrium cdadition of the variables important to making a control rod inventory prediction, specifically the reactivity effects of the xenon, is rarely achieved. Se uncertainties of calculating the control rod inwntory with non-equilibrium xenon conditions can result in eirors which can be misconstrued as reactivity ananalies. herefore, this specification calla for perfomin6 of rod inventory comparisons at a time when xenon will not be a source of error.
s The closure time of 30 seconds was based on a letter dated 7/25/80 to J. G. Keppler F.
(Region III) from D. E. Gilberts (NSP) concerning IE Bulletin No. 80-14.
Ten hours to insert the required rods will allow time to shutdown in a controlled manner without causing an undue rate of change of the discharge channel temperature.
Whenever a specification (or specifications) can not be met for a particular mode of operation,
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the reactor would be placed in a mode for which the specification (or specifications) are not required, his requires insnediate initiation of a reactor shutdown upon discovery that specif-ications 3 3A through 3 3D are not met.
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hacnJment_No..10 g
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