ML20137L516

Amend 37 to License DPR-22,revising Tech Specs to Provide Limiting Conditions for Operations & Surveillance Requirements for Overtime Limitations,Rcic Suction Transfer & Addl Accident Monitoring Instrumentation

ML20137L516
Person / Time
Site:	Monticello
Issue date:	01/22/1986
From:	Zwolinski J Office of Nuclear Reactor Regulation
To:
Shared Package
ML20137L519	List: ML20137L516 ML20137L526
References
NUDOCS 8601280008
Download: ML20137L516 (38)

Text

kmatro UNITED STATES o,,

E.

NUCLEAR REGULATORY COMMISSION o

5

E WASHINGTON, D. C. 20555

%,...../

NORTHERN STATES POWER COMPANY DOCKET NO. 50-263 MONTICELLO NUCLEAR GENERATING PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 37 License No. DPR-22 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment by Northern States Power Company (the licensee) dated June 24, 1983, 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 confo'mity 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 endangering the health and safety of the public and (ii) that such activities will be conducted in compliance with the Comission's regulations; 1

D.

The issuance of this amendment will 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 I

have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, l

and paragraph 2.C.2 of Facility Operating License No. OPR-22 is hereby l

amended to read as follows:

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00CK05000263 08 860122 P

PDR l

2 Technical Specifications The Technical Specifications contained in Appendix A 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 COMM SSION

~

/

w John Zwolinski, Director BWR Pr ect Directorate #1 Division of BWR Licensing

Attachment:

Change 1 to the Technical Spec.fications Date of Issuance:

January 22, 1986.

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ATTACHMENT TO LICENSE AMENOMENT NO. 37 FACILITY OPERATING LICENSE NO. DPR-22 DOCKET N0. 50-263 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised areas.are indicated by marginal lines'.

Pages 1, iv, vi, vii 46a 50 60d 61 69a 71a 111 lila 112 229b 229c 229d 229e - 229t (renumber of pages only) 233 233a 249a 6

f l

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TABLE OF CONTENTS page 1.0 DEFINITIONS 1

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEN 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 OPERATION 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 Instrumentation 45 A.

Primary Containment Isolation Functions 45 I

B.

Emergency Core Cooling Subsystems Actuation 46 C.

Control Rod Block Actuation 46 D.

Other Instrumentation 46a l

E.

Reactor Building Ventilation Isolation and Standby Gas Treatment System Initiation 47 F.

Recirculation Pump Trip Initiation 48 G.

Safeguards Bus Voltage Protection 48 H.

Instrumentation for S/RV Low-Low Set Logic 48

[

3.2 Bases 64 4.2 Bases 72 3.3 and 4.3 Control Rod System 76 A.

Reactivity Limitations 76 B.

Control Rod Withdrawal 77 C.

Scram Insertion Times 81 D.

Control Rod Accumulators 82 E.

Reactivity Anomalies 83 F.

Scram Discharge Volume 83A G.

Required Action 83A 3.3 and 4.3 Bases 84 l

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Amendmeat No. 30, 37 l

1 l

. -. ~.

i P. age 3.13 and 4.13 Fire Detection and Protection Systems 223 A.

Fire Detection Instrumentation 223 B.

Fire Suppression Water System 224 C.

Hose Stations

'226 D.

Yard Hydrant Hose Houses 227 E.

Sprinkler Systems 227a F.

Halon Systems.

227b G.

Penetration Fire Barriers 227b 3.13 Bas'es 228 4.13 Bases 229 3.14 and 4.14 Accident Monitoring Instrumentation 229a 3.14 and 4.14 Bases 229e 3.15 and 4.15 Inservice Inspection and Testing 229f 3.15 and 4.15 Bases 229g 3.16 and 4.16 Radiation Environmental Monitoring Program 229h A.

Sampling and Analysis 229h B.

Land Use Census 229j C.

Interlaboratory Comparison 229k i

3.16 and 4.16 Bases 229t 5.0 DESIGN FEATURES 230 5.1 site 230 5.2 N actor 230 5.3 Reactor Vessel 230 5.4 containment 230 5.5 Fuel Storage 231 5.6 Seismic Designs 231 6.0 ADMINISTRATIVE CONTROLS 232 6.1 Organization 232 6.2 Review snd Audit 237 6.3 Special Inspection 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 l

iv Amendment 7, 15, 37 l

LIST OF TABLES Tr[bleNo.

P_ age 28 3.1.1 Reactor Protection System (Scram) Instrument Requirements i

4.1.1 Scram Instrument Functional Tests - Minimum Functional 32 f

Test Frequencies for Safety Instrumentation and Control Circuits 1

4.1.2 Scram Instrument Calibration - Minimum Calibration 34 Frequencies for Reactor Protection Instrument Channels 3.2.1 Instrumentation that Initiates Primary Containment 49 Isolation Functions 3.2.2 Instrumentation that Initiates Emergency Core Cooling Systems 52 3.2.3 Instrumentation that Initiates Rod Block 57 3.2.4 Instrumentation that Initiates Reactor Building Ventilation 59 Isolation and Standby Gas Treatment System Initiation 3.2.5 Instrumentation that Initiates a Recirculation Pump Trip 60 3.2.6 Instrumentation for Safeguards Bus Degraded Voltage and 60a Loss of Voltage Protection i

3.2.7 Instrumentation for Safety / Relief Valve Low-Low Set Logic 60b l

3.2.8 Other Instrumentation 60d 4.2.1 Minimum Test and Calibration Frequency for Core Cooling, Rod Block and Isolation Instrumentation 61

{

3.6.1 Safety Related Snubbers 131 l

3.7.1 Primary Containment Isolation 172

. Radioactive Liquid Effluent Monitoring Instrumentation 1891 3.8.1 3.8.2 Radioactive Gaseous Effluent Monitoring Instrumentation 198k j

4.8.1 Radioactive Liquid Effluent Monitoring Instrumentation 198m Surveillance Requirements

(

4.8.2 Radioactive Gaseous Effluent Monitoring Instrumentation 198n Surveillance Requirements I

4.8.3 Radioactive Liquid Waste Sampling and Analysis Program 198p 4.8.4 Radioactive Gaseous Waste Sampling and Analysis Program 198s 3.11.1 Maximum Average Planar Linear Heat Generation Rate 215 vs. Exposure 3.13.1 Safety Related Fire Detection Instruments 227c vi Amendment 29, 30, 37

3.14.1 Instrumentation for Accident-Monitoring 229b 4.14.1 Minimum Test and Calibration Frequency for Accident 229d Monitoring Instrumentation 4.16.1 Radiation Environmental Monitoring Program (REMP)

'229-1 Sample Collection and Analysis 4.16.2 REMP - Maximum Values for the Lower Limits of Detection 229q 4.16.3 REMP - Reporting Levels for Radioactivity Cpncentrations 229s in Environmental Samples 6.1.1 Minimum Shift Crew Composition 236 I

i l

vii Amendment No. 15,37

3.0 LIMITING CONDITIONS-FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 1

e 2.

Rod Block Monitor (RBM) (continued) b.

RBM Setpoints for control rod block are given in Table 3.2.3.

The upscale LTSP shall be applied above 30% and up to 65% of rated thermal power. The i

upscale 1 TSP shall be applied at and above 65% and up to 85% of rated thermal power. The upscale HTSP shall be applied 4

at and above 85% of rated thermal power.

The RBM Bypass time delay shall be less than or equal to 2.0 seconds.

i D.

Other Instrumentation

}

Whenever the reactor is in the RUN Mode, the limiting conditions for operation for the instrumentation listed in Table 3.2.8 shall l

l be met.

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3.2/4.2 46a i

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TABLE 3.2.1 - Continued Min. No. of Operable Total No. of Instru-or Operating Instru-ment Channels Per ment Channels Per Trip Required Function Trip Settings Trip System System (1,2).

Conditions b.

High Drywell Pressure (5) 12 psig 2

2' D

3.

Reactor Cleanup System (Group 3) l a.

Low Reactor Water

>10'6" above Level the top of the active fuel 2

2 E

b.

High Drywell Pressure 12 psig 2

2 E

4.

HPCI Steam Lines a.

HPCI High Steam Flow 1150,000 lb/hr 2(4) 2 F

1 0 second 6

with time delay 3

b.

HPCI High Steam Flow 1300,000 lb/hr 2(4) 2 y

i c.

HPCI Steam Line 1200*F 16(4) 16 F

Area High Temp.

i 5.

RCIC Steam Lines a.

RCIC High Steam Flow 145,000 lb/hr 2(4) 2 G

with 5 12

[p see time delay e

l 5

b.

RCIC Steam Line Area 1200*F 16(4) 16 G

z P

6.

Shutdown Cooling Supply Isolation-wa-a.

Reactor Pressure

<75 psig 2(4) 2 C

{*

Interlock at pump

'd suction 1

3.7/4.2 50

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Table 3.2.8 Other Instrumentation Minimum No. of Minimum No. of Oper-Operable or Total No. of Instru-able or Operating Required Function' Trip Setting Operating Trip ment Channels Per Instrument Channels Conditions

• System (1)

Trip System Per Trip System (1) 4 i

A.

RCIC Initiation

1. Low-Low Reactor Level

>6'6" & <6'10" 1

2 2

B above top of active fuel L

t i

B.

HPCI/RCIC Turbine Shutdown

a. High Reactor Level

<14'6" above 1

2 2

A top of active i

fuel

)

C.

HPCI/RCIC Turbine Suction Transfer

{

c. Condensate Storage

>2'0" above 1

2 2

C l

Tank Low Level tank bottom j

NOTE:

1.

Upon discovery that minimum requirements for the number of operable or operating trip systems or instrument channels l

Ere not satisfied, action shall be initiated to:

' g

a. Satisfy the requirements by placing the appropriate channels or systems in the tripped condition

' g.

(Turbine /Feedwater Trip only), or

b. Place the plant under the specified required condition using normal operating procedures.

o.

R

{NA Required conditions when minimum conditions for operation are not satisfied:

l5 1-A. Reactor in Startup, Refuel, or Shutdown Mode.

B. Comply with Specification 3.5.F.2.

j i

C. Align HPCI and RCIC suction to the suppression pool. Restore channels to operable status

' status within 30 days or place the plant in Required Condition A.

j i

3.2/4.2 60d

i Table 4.2.1 Minimum Test and Calibration Frequency For Core Cooling Rod Block and Isolation Instrumentation Instrument Channel Test (3)

Calibration (3)

Sensor Check (3)

=

ECCS INSTRUMENTATION 1.

Reactor Low-Low Water Level (Note 7) once/ month once/3 months once/Shif t 2.

Drywell High Pressure (Note 7) once/ month Once/3 months None 3.

Reactor Low Pressure (Pump Start)

Note 1 Once/3 months None 4.

Reactor Low Pressure (Valve Permissive) Note 1 Once/3 months None 5.

Undervoltage Emergency Bus Refueling Outage Refueling Outage None 6.

Low Pressure Core Cooling Pumps Discharge Pressure Interlock Note 1 Once/3 months None 7.

Loss of Auxiliary Power Refueling Outage Refueling outage None 8.

Condensate Storage Tank Level Refueling Outage Refueling Outage None 9.

Reactor High Water Level Once/ month Once/3 months Once/ day ROD BLOCKS 1.

APRM Downscale Notes (1,5)

Once/3 months None 2.

APRM Flow Variable Notes (1,5)

Once/3 months None 3.

IRM Upscale Notes (2,5)

Note 2 Note 2 4.

IRM Downscale Notes (2,5)

Note 2 Note 2 5.

RBM Upscale Notes (1,5)

Once/3 months None 6.

RBM Downscale Notes (1,5)

Once/3 months None 7.

SRM Upscale Notes (2,5)

Note 2 Note 2 8.

SRM Detector not in Start-up Position Note 2 Note 2 Note 2 i

9.

Scram Discharge Volune-High Level Once/3 months Refueling

• outage None f

MAIN STEAM LINE ISOLATION s

EI 1.

Steam Tunnel High Temperature Refueling Outage Refueling Outage None 2.

Steam Line High Flow Note 1 Cnce/3 months Once/ Shift E.i h

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3.2/4.2 61 i

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Bases Continued:

j open and instrumentation drift has caused the nominal 80-psi blowdown range to be reduced to l

60 psi. Maximum water leg clearing gfme has been calculated to be less than 6 seconds i

for the Monticello design.

Inhibit timers are provided for each valve to prevent j

the valve from eaing manually opened less than 10 seconds following valve closure.

j Valve opening is sensed by pressure switches in the valve discharge line. Each valve is provided with two trip, or actuation, systems. Each system is provided with I

two channels of instrumentation for each of the above described functions. A two-out-of-two-once logic scheme ensures that no single failure will defeat the low-low set 4

function and no single failure will cause spurious operation of a safety / relief valve.

Allowable deviations are provided for each specified instrument setpoint. Setpoints within I

the specified allowable deviations provide assurance that subsequent safety / relief valve actuations are sufficiently spaced to allow for discharge line water les clearing.

t Although the operator w.tll set the set points within the trip settings specified in Tables 3.2.1 through 3.2.8, the actual values of the various set points can differ appreciably from the value l

the operator is attempting to set.

The deviations could be caused by inherent instrument error, operator setting error, drift of the set point, etc.

Therefore, these deviations have been accounted for in the various transient analyses and the actual trip settings may vary by the following amounts:'

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References:

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1.

" Average Power Range Monitor, Rod Block Monitor and Technical Specifications Improvement (ARTS) f tj Program for Monticello Nuclear Generating Plant", NEDC-30492-P, April, 1984, i

i 1

i.

3.2 BASES 69a J

I

Trip Function Deviation

O Instrumentation for. Safety / Relief Valve Reactor Coolant System 220 peig Low Low Set Logic Pressure for Opening / Closing Opening - Closing Pressure

>60 psi Discharge Pipe Pressure

!!O paid Inhibit '

Timer Inhibit

-3 sec

+10 see Other Instrumentation High Reactor Water Level

+6 inches Low-Low Reactor Water Level

-3 inches Low Condensate Storage Level

-6 inches i

i 4

E h

A violation of this specification is assumed to occur only when a device is knowingly set outside of the i

limiting trip settings, or, when a sufficient number.o* devices have been affected by any means such that r+

3 i

g 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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3.2 BASES 71a 1

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j 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 4

shutdownshallbeinitiatedimmedlately and the reactor pressure shall be reduced to 150 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter.

F.

Reactor Core Isolation Cooling System (RCIC)

F.

Surveillance of Reactor Core Isolation Cooling System (RCIC) 1.

Except as specified in 3.5.F.2 below, the Surveillance of the RCIC System shall be RCIC system shall be operable whenever performed as follows:

the reactor pressure is greater than 150 l

psig and irradiated fuel is in the reactor 1.

Testing vessel. To be considered operable, the RCIC l

system shall meet the following conditions:

Item Frequency Pump operability Once/ month a.

The RCIC shall be capable of delivering 400 gym into the reactor vessel at 150 psig.

Motor operated once/ month l

valve operability b.

The controls for automatic transfer of the RCIC pump suction from the condensate Flow rate test After major pump i

storage tank to the suppression chamber maintenance and shall be operable.

every three months i

j c.

The controls for automatic restart on Simulated automatic Once/ Operating Cycle j

subsequent low reactor level after it actuation, transfer j

has been terminated by a high reactor of suction to sup-p level signal shall be operable.

pression pool, and j

g automatic restart cp on subsequent low g

reactor water level a

e, b

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1 1

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3.5/4.5 Ill I

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3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS

{

2.

From and after the date that the RCIC system 2.

When it is determined that the RCIC system 1

is made or found to be inoperable for any is inoperable, the HPCI system shall be reason, except automatic transfer of pump demonstrated to be operable immediately suction, reactor operation'is permissible only and daily thereafter.

during the succeeding 15 days unless such system is sooner made operable, provided that during i

such 15 days all active components of the HPCI system are operable. With the controls for automatic transfer of pump suction inoperable, operation for up to 30 days is permissible if the pump suction is aligned to the suppression i

pool.. If these conditions cannot be met, an orderly shutdown shall be initiated and the reactor pressure reduced to 150 psig within l

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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3.5/4.5 lila

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS e

G.

Minimum Core and Contain~ ment Cooling System G.

Surveillance of Core and Containment Availability Cooling System 1.

During any period when one of the standby 1.

When it is determined that one of diesel generators is inoperable, continued the standby diesel generators is reactor operation is permissible only inoperable, all low pressure core during the succeeding seven days, provided cooling and containment cooling that all of the low pressure core cooling service water systems connected to and containment cooling subsystems connec-the operable diesel generator shall ted to the operable diesel generator shall be demonstrated to be operable in-be operable. If this requirement cannot mediately and daily thereafter. In be met, an orderly shutdown shall be addition, the operable diesel gen-initiated and the reactor water temperature erator shall be demonstrated to be shall be reduced to less than 212*F within operable immediately and daily 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

thereafter.

2.

Any combination of inoperable components in the core and containment cooling systems shall not defeat the capability of the remaining operable components to fulfill g

the core and containment cooling functions.

e 5.

2=r+

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I 3.5/4.5 112 1

i Table 3.14.1 Instrumentation for Accident Monitoring

=

Function Total No. of Minimum No. of Required Instrument Channels Operable Channels Conditions

• Reactor Vessel Fuel Zone Water Level 2

1

'A, B Safety / Relief Valve Position 2

1 A, C (One Channel Pressure Switch and One Channel Thermocouple Position Indication per Valve) l Drywell Wide Range Pressure 2

1 A, B i

Suppression Pool Wide Range Level 2

1 A, B Drywell High Range Radiation 2

1 A, D 1

1 Drywell and Suppression Pool 2

1 A, B Hydrogen and Oxygen Monitor Offgas Stack Wide Range Radiatien 2

1 AD Reactor Bldg Vent Wide Range Radiation 2

1 A, D r

j

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• Required Conditions a

i A. When the number of channels made or found to be inoperable is such that the number of operable channels is i

less than the total number-of channels, either restore the inoperable channels to operable status within

!E seven days, or prepare and submit a special report to the Commission pursuant to. Technical Specification 6.7.B.2 within the next 30 days outlining the action taken, the cause of the inoperability, and the plans and schedule for restoring the system to. operable status.

E$

i 229b 3.14/4.14 i

Table 3.14.1 (continued)

Instrumentation for Accident Monitoring

• Required Conditions (continued)

• B. When the number of channels made or found to be inoperable is such that the number of operable channels is less than the minimum number of operable channels shown, the minimum number of channels shall be restored to operable status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least Hot Shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

C. When the number of' channels made or found to be inoperable is such that the number of operable channels is less than the minimum number of operable channels shown, the torus temperature shall be monitored at least once per shift to observe any unexplained temperature increase which might be indicative of an open SRV; the minimum number of channels shall be restored to operable status within 30 deys or be in at least Hot Shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

D. When the number of channels made or found to be inoperable is such that the number of operable channels is less than the minimum number of operable channels shown, initiate the preplanned alternate method of monitoring the appropriate parameters in addition to submitting the report required in (A) above.

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3.14/4.14 229c

Table 4.14.1 l

Minimum Test and Calibration Frequency for Accident Monitoring Instrumentation e

Instrument Channel Test (Note 1)

Calibration (Note 1)

Sensor Check (Note 1)

Reactor Vessel Fuel Zone Water Level Monitor Once/ Operating Cycle Once/ month (Note 3) l

'fafety/ReliefValvePosition(PressureSwitches)

Once/ Operating Cycle Once/ month (Note 2)

Saf4ty/ Relief Valve Position (Thermocouples)

Once/ Operating Cycle Once/ month (Note 2)

- brywell Wide Range Pressure Monitors Once/ Operating Cycle Once/ month Suppression Pool Wide Range Level Monitors Once/ Operating Cycle Once/ month Drywell High Range Radiation Monitors Once/ Operating Cycle Once/ month Once/ Operating Cycle Once/ month i

Dryvell and Suppression Pool Hydrogen and Oxygen Monitors Offgas Stack Wide Range Radiation Monitors Once/ Operating Cy61e Once/ month Once/ Operating Cycle Once/tonth l

Reactor Bldg Wide Range Radiation Monitors l*

Notes:

o

$I (1) Functional tests, calibrations, and sensor checks are not required when the instruments are not required 4

to be operable.

If tests are missed, they shall be performed prior to returning the instruments to an

((

operable status.

o',

(2) Proper instrument response shall be verified during each safety / relief valve actuation.

gj (3) These instruments are off-scale high during normal plant operation.

i 3.14/4.14 229d l

Bases:

3.14/4.14 The operability of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables during and following an accident. This capability is consistent with the recommendations of NUREG-0578, "TMI-2 Learned Task Force Status Report and Short Tern Recomunendations".

I N

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D 3.14/4.14 229e

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 3.15 INSERVICE INSPECTION AND TESTING

• 4.15 INSERVICE INCFECTION'AND TESTING t

Applicability:

Applicability:

l Appliec to components which are part of Applies to the periodic inspection and i

the reactor ccolant pressure boundary and testing of components which are part of their supports and other safety-related the reactor coolant pressure boundary pressure vessels, piping, pumps, and and their supports and other safety-valves.

related pressure vessels, piping, pumps, and valves.

Objective:

Objective:

To assure the intergrity of the reactor.

To verify the integrity of the reactor coolant pressure boundary and the coolant pressure boundary and the operational readiness of safety-related operational readiness of safety-pressure vessels, piping, pumps, and related pressure vessels, piping, pumps, valves.

and valves.

Specification:

Specification:

A.

Inservice Inspsetion A.

Inservice Inspection d

1.

To be considered operable Quality 1.

Inservice inspection of Quality Group A, B, and C components shall Group A, B, and C components shall g

satisfy the requirements contained be performed in accordance with I

in Section XI of the ASME Boiler the requirements for ASME Code Class e

El and Pressure Vessel Code and appli-1, 2, and 3 components, respectively, j

cable Addenda for continued service contained in Section XI of the ASME 5

of ASME Code Class 1,.2, and 3 compo-Boiler and Pressure Vessel Code and z

nents, respectively, except where applicable Addenda as required by

?

relief has been granted by the 10 CFR 50, Section 50.55a(g), except Commission pursuant to 10 CFR 50, where relief has been granted by the in

[,

Section 50.55a(g)(6)(1).

the Commission pursuant to 10 CFR 50, Section 50.55a(g)(6)(1)

N 4

I 3.15/4.15 229f s

i Bases 3.15 and 4.15 e

The inservice inspection program for the Monticello plant conforms to the requirements of 10 CFR 50 Section 50.55a(g). Where practical, the inspection of components classified into NRC Quality 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 ASME Boiler and Pressure Vessel Code.

If a Code required inspection is impractical for the i

Monticello facility, a request for a deviation from that requirement is submitted t6 the Commission in j

accordance with 10 CFR 50, Section 50.55a(g)(6)(1).

Deviations which are needed from the procedures prescribed in Section XI of the ASME Code and applicable Addenda vill be reported to the Commission 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 inspection will be reported quarterly throughout the inspection period.

A program of inservice testing of Quality Group A, B, and C pumps and valves is also in effect at the Monticello plant. Technical Specifications related to this program will be issued following NRC

. review and approval of the pump and valve testing program.

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yn ti 3.15/4.15 BASES 229g 1

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 4.16 RADIATION ENVIRONMENTAL MONITORING PROGRAM Applicability Applies to the periodic monitoring and recording of radioactive effluents found in the plant environs.

Objective

~

To provide for measurement of radiation levels and radioactivity in the site environs on a continuing basis.

Specification A.

Sample Collection & Analysis I

1.

The Radiation Environmental Monitoring Program given in Table 4.16.1 shall be conducted. Radioanalysis shall be conducted meeting the requirements of Table 4.16.2.

i A map and a table identifying the locations of the sampling points shall be provided in the Offsite Dose Calculation Manual (ODCM).

El 2.

Whenever the Radiation Environmental Monitoring Program is not being conducted

?

5 as specified in Table 4.16.1, the Annual Radiation Environmental, Monitoring Report se P

shall include a description of the reasons for not conducting the program as required 3.

and plans for preventing a recurrence.

O 229h 3.16/4.16 i

i

._u2 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS e

3.

Deviations are permitted from the required sampling schedule if samples are unobtainable due to hazardous conditions, seasonable unavaila-bility, or to malfunction of automatic sampling equipment.

If the latter occurs, every effort shall be made to complete corrective action prior to the end of the next sampling period.

J 4.

With the level'of radioactivity in an environ-mental sampling medium exceeding the reporting levels of Table 4.16.3 when averaged over any calendar quarter, in lieu of any other report, prepare and submit to the Commission within 30 days from the end of the affected' calendar quarter a Report pursuant to Specification 6.7.C.2.a.

When more than one of the radionuclides in Table 3.12-2 are detected in the sampling medium, this report shall be submitted if:

concentration (1) + concentration (2) +...

.0 limit level (1) limit level (2)

When radionuclides other than those in Table 3.12-2 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to an individual is equal to or greater 2f than the calendar year limits of Specifications 3.8.A.2, 3.8.B.2, or 3.8.B.3.

This report is not if required if the measured level of radioactivity was not the result of plant effluents; however, in such l

an event, the condition shall be reported and described if in the Annual Radiation Environmental Monitoring' Report.

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3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 5.

Although deviations from the required sampling schedule are permitted under Item 3, above, whenever milk or leafy green vegetables can no longer be obtained from the designated cample locations required by Table 4.16.1, the Semi-annual Radioactive Effluent Release Report for this period shall explain why the samples can no longer be obtained and will identify the locations which will be added to and deleted from-the monitoring program as soon as practicable.

B.

Land Use Census 1.

A land use census shall be conducted and shall identify the location of the nearest milk animal, the nearest residenge, and the nearest garden of greater than 500 ft producing fresh leafy vegetables, in each of the 16 meteorological sectors within a distance of five miles. The census shall also identify the allmilkanimalsandall500ft{ocationsof or greater gardens producing broad' leaf vegetation in each of the meteorological sectors within a distance of three miles. This census shall be conducted at least once per year between the dates of May I and October 31 by door to door survey, aerial lI survey, or by consulting local agricultura1' s

authorities associations.

8e 3.16/4.16 229j

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3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 2.

With a land use census identifying a location which yields a calculated dose or dose commitment 1

(via the same exposure pcthway) 20 percent greater than at a location from which samples are currently being obtained in accordance with Specification 4.16.A.1, the Semiannual Radioactive Effluent Release Report for this period shall identify the new location. The new location shall be added to the radiological environmental monitoring program within 30 days. The sampling location, excluding the control station location, having the lowest calculated dose or dose commitment (via the same i

exposure pathway) may be deleted from this moni-i toring program after October 31 of the year in which this land use census was conducted.

C.

Interlaboratory Comparison Program 1

1.

Analyses shall be performed on radioactive materials supplied as part of sn NRC approved interlaboratory l

comparison program as described in the ODCM.

2.

The results of analyses performed as a part of the above required program shall be included in the Annual Radiation Environmental Monitoring Report. When i

required analyses are not performed, corrective k

action shall be reported in the Annual Radiation g

Environmental Monitoring Report.

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3.16/4.16 229k

gp Table 4.16.1 g

(Page 1 of 5)

$g MONTICELLO NUCLEAR GENERATING PLANT r+

RADIATION ENVIRONMENTAL MONITORING PROGRAM SAMPLE COLLECTION AND ANALYSIS U

Number of Samples t$

Exposure Pathway and Sampling and Type and Frequency and/or Sample Sample Locations **

Collection Frequency of Analysis 1.

Airborne Radioiodine &

Samples from 5 locations:

Continuous Sampler Radiciodine analysis Particulates 3 samples from offsite operation with sampler Weekly for I-131 locations (in different collection weekly.

sectors) of the highest Particulate:

calculated annual average Cross beta activity ground level D/Q, I sample on each filter weekly *.

from the vicinity of a com-Analyses shall be per-munity having the highest formed more than 24 calculpted annual average hours following filter ground-level D/Q, and change. Perform gamma 1 sample from a control isotopic analysis on location 8-20 miles dis-composite (by location) tance and in the least sample quarterly.

pre alent wind direction 2.

Direct-Radiation 37 TLD stations established Quarterly Gamma Dose quarterly with duplicate dosimeters placed at the following locations:

If gross beta activity in any indication sample exceeds 10 times'the yearly average of the control sample, a gamma u

icotopic analysis is required.

• • S;mple locations are given on the figure and table in the ODCM.

3.16/4.16 229-1

jf Table 4.16.1 (Page 2 of 5) 5 MONTICELLO NUCLEAR GENERATING PLANT RADIATION ENVIRONMENTAL MONITORING PROGRAM ll SAMPLE COLLECTION AND ANALYSIS 1

0:

Number of Samples R$

Exposure Pathway and Sampling and Type and Frequency and/or Sample Sample. Locations **

Collection Frequency of Analysis 2.

Direct Radiation (Con't.)

1. Using the 16 meterological wind sectors as guidelines, I

an inner ring of stations-in the general area of the site boundary is established and an outer ring of stations at 4 to 5 mile distance from the plant site is established.

Because of inacessibility, two sectors in the inner and i

outer rings are not covered.

2. Seven dosimeters are established at special interest areas and a control. station.

3.

Waterborne a.

Surface Upstream & downstream Monthly composite of Gamma Isotopic analysis locations weekly samples (water of each monthly composite

& ice conditions permitting)

Tritium analysis of quarterly composites of monthly composites o* Simple locations are given on the figure and table in the ODCM.

3.16/4.16 229m 1

3I Table 4.16.1

. (Page 3 of 5)

S MONTICELLO NUCLEAR GENERATING PLANT RADIATION ENVIRONMENTAL MONITORING PROGRAM f

SAMPLE COLLECTION AND ANALYSIS U

Number of Samples k$

Exposure Pathway and Sampling and Type and Frequency and/or Sample Sample Locations **

Collection Frequency of Analysis 3.

Waterbourne (con't.)

b.

Ground Three samples from wells Quart' rly Gamma Isotopic and within 5 miles of the tritium analyses of plant site and one sample each sample from a well greater than 10 miles from the plant

site, c

c.

Drinking One sample from the City of Monthly composite of I-131 Analysis and Minneapolis water supply weekly samples.

Gross beta and Gamma isotopic analysis of each monthly composite Tritium analysis of 1

quarterly composites of monthly composites d.

Sediment from One sample upstream Semiannually Gamma isotopic analysis Shoreline of plant, one sample of each sample downstream of plant,

~

and one sample from shoreline of recreational j

area I

oo Sample locations are given on the figure and table'in the ODCM.

3.16/4.16 229n

g Table 4.16.1 g

(Page 4 of 5) o.l MONTICELLO NUCLEAR GENERATING PLANT r+

RADIATION ENVIRONMENTAL MONITORING PROGRAM SAMPLE COLLECTION AND ANALYSIS

~*

Number of Samples 1

Exposure Pathway and Sampling and Type and Frequency and/or Sample Sample Locations **

Collection Frecuency of Analysis 4.

Ingestion a.

Milk One sample from dairy Monthly or biweekly Gamma isotopic and farm having highest D/Q, if animals are on I-131 analysis of one sample from each of pasture each sample three dairy farms cal-culated to have doses from I-131 > 1 ares /yr, and one sample from 10-20 miles i

b.

Fish and One sample of one game Samples collected Gamma isotopic Invertebrates specie of fish located semi-annually analysis on each upstream and downstream sample (edible of the plant site.

portion only on fish).

One sample of Invertebrates upstream and downstream of i

the plant site.

C0 Sample locations are given on the figure and table in the ODCM.

4 3.16/4.16 229 o

_ _ _ _ =

1 g

Table 4.16.1 (Page 5 of 5)

MONTICELLO NUCLEAR GENERATING PLANT RADIATION ENVIRONMENTAL MONITORING PROGRAM SAMPLE COLLECTION AND ANALYSIS U

Number of Samples 1

y Exposure Pathway and Sampling and Type and Frequency and/or Sample Sample Locations **

Collection Frequency of Analysis c.

Food Products One' sample of corn At time of harvest Gamma isotopic r

from highest D/Q farm analysis of edible and one sample from portion of each -

10-20 miles sample One sample of potatoes At time of harvest Gamma isotopic from highest D/Q farm analysis of edible and one sample from portion of each.

10-20 miles sample One sample of broad At time of harvest I-131 analysis j

leaf vegetation from of edible portion highest D/Q garden and of each sample one sample from 10-20 miles i

l cc Sample locations are given on the figure and table in the ODCM.

4 3.16/4.16 229p

..~..-- -... -.

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Table 4.16.2 (Page 1 of 2)

MAXIMUM VALUES FOR THE LOWER LIMITS OF DETECTION (LLD)***

=

4 I

Airborne Particulate Water or Gag Fish Milk Food Products Sediment Analysis (pC1/1)

(pCi/m )

(pCi/kg, wet)

(pCi/1)

(pCi/kg, wet)

(pCi/kg, dry)

D

-2 l

l grecs beta 4

1 x 10 3

(10M)

H 1

54 15 130 h

59,

30 260 7

t 58, 60 15 130 Co 65 30 260 95 15" Zr-Nb

-2 d

131 1'

7 x 10 3

60 7

-2 134,137 15(10 ), la 1 x 10 130 15 60 150 Cs 140 15*

15" h-h l

I a

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TABLE 4.16.2 (Page 2 of 2)

TABLE NOTATION The LLD is the smallest concentratiod of radioactive material in a sample that will be detected with 95% probability a

with 5% probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement system (which may include radiochemical separation):

Y 4.66 s m

E LLD

=

E, V. 2.22. Y exp(-AAt) 5 where-LLD is the a priori lower limit of detection as defined above (as picoeurie per unit mass or volume),

wI" a is the standard deviation of the background counting rate or of the counting rate of a blank sample b

tj as appropriate (as counts per minute). Typical values of E. V, Y and at shall be used in the calculations.

E is the counting efficiency (as counts per tratisformation)

V is the sample size (in units of mass or volume) 2.22 is the number of transformations per minute per picocurie 1

l Y is the fraction radiochemical yield (when applicable) r A is the radioactive decay constant for the particular radionuclide At is the elapsed time between sample collection (or end of the sample collection period) and time of counting b - LLD for drinking water, c - Total for parent and daughter.

d - Applies to specific isotope analysis-not to gamma spectrum analyses.

- Other peaks which are measurable and identifiable, together with the radionuclides in Table 4.16.2, shall l

o be identified and reported.

229r 3.16/4.16

Table 4.16.3 l

REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES Reporting Levels Airborne Particulate Water or Gag Fish Milk Vegetables Analysis (pCi/1)

(pCi/m )

(pCi/kg, wet)

(pCi/1)

(pCi/kg, wet) i H-3 2 x 10 (*)

3 4

Mn-54 1 x 10 3 x 10 2

Fe-59 4 x 10 1 x 10 3

4 Co-58 1 x 10 3 x 10 4

Co-60 3 x 10 1 x 10 4

Zn-65 3 x 10 2 x 10 Zr-Nb-95 4 x 10 (b) 2 2

1-131 2

0.9 3

1 x 10 3

3 Cs-134 30 10 1 x 10 60 1 x 10 3

3 Cs-137 50 20 2 x 10 70 2 x 10 g

Ba-La-140 2 x 10 (b) 3 x 10 (b) 2 2

a z

)

P a - For drinking water samples b - Total for parent and daughter w*

O 3.16/4.16 229s J

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i 3.16 and 4.16 BASES l

A. Simple Collection & Analysis i

1 1

i The Radiation Environmental Monitoring Prggram required by this specification provides measurements of radiation and of rcdioactive materials in those exposure pathways and for those radionuclides which lead to the highest potential rs:diation exposures of individuals resulting from the plant operation. This program thereby supplements the radio-Icgical effluent monitoring by verifying that the measurable concentrations of radioactive materials and levels of I

redi ' ton are not higher than expected on the basis of the effluent measurements and modeling of the environmental cxpo e pathways. After a specific program has been in effect for at least three years of operation, program i

changes may be initiated based on this experience.

1

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t The detection capabilities required by Table 4.16.2 are state-of-the-art for routine environmental measurements in l

l I

industrial laboratories. The LLD's for drinking water meet the requirement of 40CFR 141.

t

.t i

B. Land Use Census l

j This specification is provided to ensure that changes in the use of off site areas are identified and that modifications to the monitoring program are made if required by the results of this census. The best survey i

information from door-to-door, serial or consulting with local agricultural authorities shall be used. Titis c nsus satisfies the requirements of Section IV.B.3 of Appendix I to 10 CFR Part 50.

Restricting the census to gardens of greater than 500 square feet provides assurance that significant exposure pathways via' leafy

{

v2getables will be identified and monitored since a garden of this size is the minimum required to produce j

the quantity (26 kg/ year) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child.

To determine this minimum garden size, the following assumptions were used,1) that 20% of the garden was 3

i used for growing broad leaf vegetation (i.e., similar to lettuce and cabbage), and 2) a vegetation yield of 2 kg/ square meter.

C. Interlaboratory Comparison Program g The requirement for participation in an interlaboratory comparison program is provided to ensure that independent 3

g checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices

g. cre performed as a part of a quality assurance program for environmental monitoring in order to demonstrate that the g

results are reasonable valid.

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.E. A training program for individuals serving in the fire brigade shall be maintained under the direction of a designated member of Northern States Power management. This program shall meet 4

the requirements of Section 27 of the NFPA Code - 1976 with the exception of training scheduling.

Fire brigade training shall be scheduled as set forth in the training program.

F. Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety-related functions; e.g.,

senior reactor operators, reactor operators, health physicists, auxiliary operators, and key maintenance personnel.. Procedures shall include the following provisions:

1.

Adequate shift coverage shall be maintained without routine i

heavy use of overtime. The objective shall be to have operating personnel work a normal 8-hour day, 40-hour week while the plant i

is operating. However, in the event that unforeseen problems j

require substantial amounts of overtime to be used, or during l

extended periods of shutdown for refueling, major maintenance or major plant modifications, on a temporary basis, the following guidelines shall be followed:

I An' individual should not be permitted to work more than a.

16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> straight, excluding shift turnover time.

i I

j b.

Overtime should be limited for all nuclear plant staff l

personnel so that total work time does not exceed 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in any 24-hour period, nor more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any t

48-hour period, not more than 84 hovra in any seven day period, all excluding shift turnover time.

Individuals should not be required to work more than 15 consecutive days without two consecutive days off.

l, li c.

A break'of at least eight hours including shift turnover time 3

should be allowed between work periods.

c s"

d.

Except during extended shutdown periods,'the use of overtime jf should be considered on an individual basis and not for the entire staff on a shift.

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$2 i

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233 6.1 i

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l e.

Shift Technical Advisor (STA) and Shift Emergency Coordinator (SEC)

I on-site rest time periode shall not be considered as hours worked I

when determining the total work time for which the above limitations

)

apply.

r 2.

Any deviation from the above guidelines shall be authorized by the Plant Manager or designee, or higher levels of management, in accordance with established procedures and with documentation of the basis for i

granting the deviation. During plant emergencies the Emergency l

Director shall have this authority. Controls shall be included in

{

the procedures such that individual overtime shall be reviewed monthly 3

i to assure that excessive hours have not been assigned. Routine i

deviation from the above guidelines is not allowed.

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i 6.1 233a

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The radioactive effluent release report to be submitted 60 days after January 1 of each year shall also include _an i k ccsessment of radiation doses to the likely most axposed member of the general public from reactor releases and other

E esarby uranium fuel cycle sources (including doses from primary effluent pathways and direct radiation) for.the previous "5

12 consecutive months to show conformance with 40 CFR 190, Environmental Radiation Protection Standards for Nuclear 45 Power Operation.

e The radioactive effluent release reports shall include the following information for solid waste shipped offsite during

w the report period.

y a.

container volume, b.

total curie quantity (specify whether determined by measurement or estimate).

c.

principal radionuclides (specify whether determined by measurement or estimate),

j d.

type of waste (e.g., spent resin, compacted dry waste, evaporator bottoms),

1 e.

type of container (e.g., LSA, Type A, Type B. Large Quantity), and l

l 1

f.

solidification agent (e.g., cement, urea formaldehyde).

i l

Th2 radioactive effluent release reports shall include unplanned releases from the site of radioactive materials in goceous and liquid effluents on a quarterly basis, changes to the ODCM, a description of changes to'the PCP, a report' cf when milk or vegetable samples can not be obtained as required by Table 4.16.1, and changes in land use resulting

}

in significant' increases in calculated doses.

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5. Annual Summearies of Meteorological Data An annual summary of meteorological data shall be submitted for the previous calendar year in the form of joint frequency distributions of wind speed, wind direction, and atmospheric stability at the request of l

tha Commission.

I

6. R2 cort of Safety / Relief Valve Failures and Challenges. An annual report of safety / relief valve failures and l

challenges shall be submitted prior to March let of each year.

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