ML20004F211

Amend 6 to License DPR-22,replacing Current Inservice Insp Tech Specs W/Inservice Insp Program for Period 1978-1981 That Meets 10CFR50.55a Requirements

ML20004F211
Person / Time
Site:	Monticello
Issue date:	06/06/1981
From:	Ippolito T Office of Nuclear Reactor Regulation
To:
Shared Package
ML20004F212	List: ML20004F211 ML20004F216 ML20004F217
References
NUDOCS 8106160619
Download: ML20004F211 (21)

Text

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s C itly UNITED STATES

• E j "c (,

g NUCLEAR REGULATORY COMMISSION

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a E WASHINGTON, D. C. 20555 k'

NORTHERN STATES POWER COMPANY DOCKET NO. 50-263 MONTICELLO NUCLEAR GENERATING PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 6 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 August 30, 1977, as suppleme ted by letter dated March 15, 1978 and revised August 28, 1978, January 5, 1979,.

Februa ry 26,1979, July 27,1979, March 5,1930, and July 16, 1980, Act of 1954, as amended (the Act) quirements of the Atomic Enargy complies with the standards and re and the Ccmission's rules and ngulations 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 (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such aq.tivities will be conducted in ccmpliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to tbs health and safety of the public; and E.

The issuance of this amenament is in accordance with 10 CFR Part 51 of the Comission's regulationr and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifi-caticns as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operatin9 License No. DPR-22 is hereby amended to read as follows:

2.

Technical Specif_ications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 6

, are hereby incorporated in the T

anse. The licensee shall operate the facility in accordance with the Technical Specifications.

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3.

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

FOR THE NUCLEAR REGULATORY COMMISSION J

Chief Operating Reactors Branch #2 Division of Licensing

Attachment:

I Changes to the Technical Specifications Date of Issuance:

June 6, 1981 49 G

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ATTACHMENT TO LICENSE AMENDMENT NO. 6 FACILITY OPERATING LICENSE NO. DPR-22 DOCKET NO. 50-263

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Replace the following pages of the Appe[idix "A" Technical Specifications with the enciqsed pages. The revised pages are identified by Amendment number and contain vertical ifnes indicating the area of change.

Remove Insert ii 11 iv iv vi vi 70 70 71 71 128 128 138 13A -

139 139 140 140.

1 41 141 142 142 143 in 144 144 151 151 152 152 153 153 T

229e 229f

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Page 3.4 and 4.4 Standby Liquid Control System 93 A.

Normal Operat ion 93 3.

Operation with Inoperable Components 94 C.

Volume-Concentration Requirements 95 3.4 and 4.4 Bases 99 3.5 and 4.5 C:re and Centainment cooling Systems 101 A.

Core Spray System 101 3.

L?CI Subsystem 103 C.

RER Service Water System 106 D.

RPCI System 105 E.

Autematic Pressure Ralief System 109 F.

RCIC System 111 C.

Minimum Core and Containment Cooling System Availability 112 H.

Deleted I., Recirculation System 114 3.5 3ases 115 4.5 Bases 12C 3.6 and a.6 Primary System Soundary

- 1[1 A.

Reactor Coolant Heacup and Ccoldown

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121 3.

Reactor Vessel Temperature and Pressure 122 C.

Coolant Chemistry 123 D.

Coolant Leakage 126 E.

Safety / Relief Valves 127 F.

Deleted l

C.

Jet Pumps 123 H.

Shock Suppressors (Snubbers) 129 3.6 and 4.6 Bases 144 3.7 and 4.7 Ccntainment Systems 156 T

A.

Primary Centainment 156 3.

Standby Cas Treat =ent System 166 C.

Secondary Containment 169 D.

Primary Containment Isolation Valves 170 5.7 Bases 175 i.7 Bases 133

. Amendment No. 6 P00R ORIGINAL

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3.13 and 4.13 Fire Detection and Protection Systems 223 A.

Fire Detection Instrumentation 223 B.

Fire Suppression Water Syttem 224 C.

Rose Statioas 226 D.

Fire Barrier Penetration Fire Seals 227 3.13 Bases ~

225 4.13 Bases 229 3.14 and 4.14 Accident Monitoring Instrumentation 229a 3.14 and 4.14 Bases 229d 3.15 and 4.15 Inservice Inspection and Testing 229e 3.15 and 4.15 B.ses 229f 5.0 DESIGN FEATURES 230 5.1 Site 230 230 5.2 Reactor

+

5.3 Reactor Vessel 230 5.4 Containment 230 5.5 Fuel Storage 231 5.6 Seismic Design 231 6.0 ADMINISTRATIVE CONTROLS 232 6.1 Organization 232 6.2 Review and Audit 237 6.3 Special Inspections and Audits 243 6.4 Action to be taken if a Safety Limit is Exceeded 243 6.5 Plant Operating Procedures 244 6.6 Plant Operating Records 246 6.7 Reporting Requirements 248 6.8 Environmental Qualification 254

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P00R ORIGINAL Amendment No. 6

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• 5 4 -

LIST OF TABLES Table No.

Page 3.1.1 Reactor Protec:ien Systes (Scram) Instrumen: Requiremen:s 28 4'.1.1 Scram Ins:ru=ent Functional Tests - Minimum Functional Test Frequencies for Safety Instru=enta:icn and Control Circui:s 32 4.1.2 Scram Instrucen Calibration - Minimum Calibration Frequencies for Reactor Protec: ion Instru=en: Ch annels 34 3.2.1 Int: u=en:atica that Initiates Primary Containment.

1sc '= tion Funct ions 49 3.2.2 Ins:rumentatica : hat Ini:iates F.=crgency Core Cooling 52 Systems 3.2.3 Instrumentatica : hat Initiates Rod 31cek 57 3.2.4 Ins: ucen:ati:n that Ini:iates Reac:or Building Ven:ilation Iselatien and Standby Gas Trea:=en:

59 c -j Eystem Initiation 3.2.5 Ins:ru=entation that Initiates a Recircula: ion 60 Pump Trip 3.2.6 Instrumen:ation for Safeguards sus 2egraded 6Ca Vol: age and Less of voltage Protec:icn 70 3.2.7 Trip Functions and Deviations 4.2.1 Mini =um Test and Calibration Frequency for Core Cooling, Rad Block and Isolatien Instru=entation 61 3.2.6 Trip Functicas and Deviatiens 70 3.6.1 Safety Related Snubbers 131

3. 7.1-Pri=ary Centainment Isola: ion 172 4 3.1 Monticello Nuclear Plan: - Environ = ental Monitoring Program Sa=ple Collec:ica and Analysis 193 3.11.1 Maximum Average Planar Linear Hea Gene rat ien Ra:e 214 vs. Exposure 3.14.1 Instrumenta: ion for Accident Moni:oring 229b 4.14.1 Minimun Test and Calibra:ica Frequency for Accident M nitoring Instrumentation 229e 6.1.1 Mini =um Shif: Crew Composi:icn 236 P00R ORIGINAL l

Amendment No. 6 l

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Table 3.2.7 Trip Functions And Deviations z

s Trip Function Deviation m

Reactor Ih2ilding Ventilation Isolation and Ventilation Plenum

+0.2 Mr/lfr Standby Gas Treatment System Initiation Ihdiation !!onitors Specification 3 2.E.3 and Table 3.2.34 Bertieling Floor Ibdistion !!onitors

+5 Mr/lir Iow Ileuctor W ter IcVel 6 inches liigh Drywell Pressure

+ 1 psi Primary Contairanent Isolation Functions Inv Iov Vater Icvel

-3 inches Table 3 2.1 l

liigh Flow in linin Steam Line

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liigh Temp. in liain Steam 42 F Line Tunnel Iow Pressure in Main Steam

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liigh Dryvell Pressure

+1 pai Iow Reactor Water IcVel

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+2 F 3.'2 1%SES 70

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' Trip Function and Deviations Trip Function Deviation Instrumentation That Initiates Emergency Low-Low Reactor Water Level

-3 Inches CodeCoolingSystems Table 3.2.2 Reactor Low Pressure (Pump

-10 psi-Start) Pe rmis s ive liigh Drywell Pressure

+1 psi Low Reactor Pressure (Valve

-1D psi Pe rmissive Instrumentation Ibat Initiates IRM Downscale

-2/125 of Scale Rod Block IRM Upscale

+2/125 a,f Scale Table 3.2.3 APRM Downscale

-2/125 ef Scale APRM Upscale See Basis 2.3 RB!! Downscale

-2/125 of Scale RB!! Upscale Same as APRM Upscale Instrumentation That Initiates liigh 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 af fected by any c.eans 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.

3.2 BASES 71

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[30 LDtITING CONDITIONS FOR OPERATION I.0 SURVEILIANCE RB2UIREMENTS 4

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G.

Jct Pumps G.

Jet Pumps Whenever the reactor is in the Startup Whenever there is recirculation flow vith the or Run codes, all jet pumps shall be oper-reactor,in the Startup or Run modes, jet pump able. If it is determined that a jet putr.p is operability shall be checked daily by verify-inoperable, the plant shall be placed in a ing that all the following conditions do not cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

occur simul'taneously:

1.

The two recirculation loop flows are unbalanced by 17,6 or score when the i

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recirculation pumps e c operating at the same speed.

l 2.

The indicated value of core flow rate is 10% or more less than the value de-rived from loop flow measurements.

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'1he safety /relier valves have two functions; i.e. pover relier cr self-actuated by high pressure.

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'Ihe solencid actuated function,( Automatic Pressure Relief) in.'hich external instrumentation si nals of E

ccincident hi6h drf ell pressure and Ic.i-low vnter level initiate opening of the valves. Ti.is function is di", cussed in Specificatica 3.,~.E.

In addition, the valves can be operated r.anually.

'lhe safety function is perfomed by the sane safety /relier valve with self-actuated intc6ral tel.lovq and pilot valve ca.: sing nain valve cperstien...rticle 9 of the Af11E Pressure Vessel Code Section III Ilueleaf Vessels requires that these tellous be mcnitorc~. for failure since this vculd defeat the safety itnetion of the safety / relief valve.

4 It is reall: ed that there is no voy to repair or replace the bellous during operation snd the plant must be shut down to do this.

'Ihe thirty-day period to do this allows the operator flexibility to choose his time for shutdown; neanwhile, because of the redundancy present in the design and the continuing monitoring of the integrity of the other valves, the overpressure pressure protection has not been comprrunised.

'Ihe cuto-relief function would not be impaired by a failure of the bellows. llovever, the self-actuated overpressure safety function vould be impaired by such a failure.

Provision also has been made to detect failure of the bellows monitoring system. Testing of this system quarterly provides assurance of bellows integrity.

When the setpcint is being bench checked, it is prudent to disassemble one cf the safety / relief valves to examine for crud buildup, bending of certain actuator members or other signs of possible deterioration.

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30 I.IMITIir, CONDITIONS E'OR OPEfMTION

!*. O SUrWEILIANCE REQUIRENENPS 3.15 INSEH1 ICE INSPECTION AND TESTIlO 4.15 INSEfNICE INSPECPION AND TESTIfn Applicability:

' Applicability:

lAppliestocomponentswhichatupartof Applies to the periodic inspection and the reactor coolant pressure boundary and testind of components wtaich are part of their supports and other parety-related the reactor coolant pressure boululary pressure vessels, piping, pumps, and and their supports and other safety-

valves, related pressum vessels, piping, pumps, and valves.

Objective:

Objective:

To assure the integrity of the reactor

'Ib verify the integrity of the reactor coolant pressure boundary and the coolant pressure boundary and the operational readiness of safety elated operational readiness of safety-pressure, vessels, piping, pumps, and related pressure vessels, piping, pumps,

valves, and valves.

Specification:

Specification:

A.

Inservice Inspection A.

Inservice Inspection 1.

Inservice inspect. ion of Quality 1.

To be considered operable, Quality Group A, B, and C components shall-Group A, B, and C components shall be perfonned in accordance with satisfy the requirements contained the requirements for ASME Code Class in Section XI of the ASME Uo11er 1, 2, and 3 components, respectively, and Pressure Vessel Code and appli-contained ir.Section XI of the ASME cable Addenda for continued service Ik>iler and Pr?asure Vessel Code arx1 of ASME Code Class 1,

, and 3 compo-g applicable Addenda as required by nents, respectively, except where 10 CFR 50; Section 50 55a(g), e:ttept relier has been granted by the whem relief has been granted by the Conunission pursuant to 10 CFR 50, Ccunnission pursuant to 10 CFR Part 50, Section 50.55a(g)(6)(1).

Sect ton 50. 55a(g)(6)(i).

229e 1.15/4.15

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The inservice inspection program for the Monticello plant confonns to the requirements of 10 CFR 50, Section 50.55a(g). Where practical, the inspection of components classified into NRC Quali y Groups A, B, and C conforms to the requirements of ASME Code Class 1, 2, and 3 components, respectively, contained in Section XI of the ASIE Boiler and Pressure Vessel Code.

If a Code required inspection is impractical for the Monticello facility, a request for a deviation from that requirement is submitted to the Cotmuission in accordhncewith10CFR50,Section50.55a(g)(6)(1).

Daviations which are needed from the procedures prescribed in Section XI of the ASFE Code and applicable Addenda will be reported to the Consnission prior to the beginning of each 10-year inspection period if they are known to be required at that time.

Deviations which are identified during the course of inspecgion will be reported quarterly throughout the inspection period.

l A program of inservice testing of Quality Group A, B, and C pumps and va,1ves is also in effect at the J

Monticello plant. Technical Specifications related to this program will be issued following NRC review and approval of the pump and valve testing program.

s.,

3.15/4.1) BASES 229f a