ML19341D609

Amend 3 to License DPR-22,adding Surveillance Requirements for Reactor Protection Sys Instrumentation & Tech Specs Re Degraded Grid Voltage Protection Mods

ML19341D609
Person / Time
Site:	Monticello
Issue date:	03/27/1981
From:	Ippolito T Office of Nuclear Reactor Regulation
To:
Shared Package
ML19341D610	List: ML19341D609 ML19341D611 ML19341D612
References
NUDOCS 8104080002
Download: ML19341D609 (31)

Text

.

. _..=

. d "IG p

t f,

UNITED STATES

• h NUCLEAR REGULATORY COMMISSION j'

w n

W ASHINGTON. D. C. 20555

.;}

tw e

% l%

/

+....

NORTHERN STATES POWER COMPANY DOCKET NO. 50-263 MONTICELLO NUCLEAR GENERATING PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 3 License No. OPR-22 1.

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

A.

The application for amendment by Northern States Power Company (the licensee) dated May 15, 1980, 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 con-e ducted in compliance with the Comission*s regulations; 1

D.

The issuance of this amendment will not be inimical to the cormon

~

defense and security or to the health and safety of the public; and E.

The issuanc'e of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

~

2.

Accordingly,lthe license is amended by. changes;to the Technical Spec.ifi-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 Specificatio E The. Technical -Specifiedtions; contained i_n' Appendices A and B, as revised

^

through Amendment No.,3

,; are hereby' incorporated in the license.

The. licensee shall-_ operate the facility Lin accordance dth the

~

lechnical Specifications.

8104 08 0 00W

. 3.

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

FOR THE I40 CLEAR REGULATORY C0fdISS10!i

?

/

'.veb>A

%' ;t Thomas M. Ippolito, Chief Operating Reactors Branch #2 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: March 27,1981

?

ATTACHMfNT TO LICENSE AMENDMENT NO.

FA,CIl 11 Y_ Ol'lIJA1, LNG l.1,CI,N5f NO. D)),4-J'?,

DOCKET NO. 50-?63 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 ii ii vi.

v.i 14 14 34 34 48A (add) 54 54 55 55 60A (add) 62 62 63 63 81 81 91 91 122 122 130 130 150-150 155 155 169 169 170 170 202 202 203 203 234 234 235 235 237 237 239 239 240 240 241 241 242 242 243 243

C 3.4 and 4.4 Standby Liquid Control System 93 A.

Normal Operation 93 8.

Operation with Inoperable Components 94 C.

Volume-Concentration Requirements 95 3.4 and 4.4 Bases 99 3.5 and 4.5 Core and Containment Cooling Systems 101 A.

Core Spray System 1 01 B.

LPCI Subsystem 103 C.

RHR Service Water System 106 D.

HPCI System 108 E.

Automatic Pressure Relief System 109 F.

RCIC System 111 G.

Minimum Core and Containment Cooling System Availability 112 H.

Deleted I.

Recirculation System 114 3.5 Bases-115 4.5 ' Bases 120 3.6 and 4.6 Primary System Boundary 121 A.

Reactor Coolant Heatup and Cooldown 121 B.

Reactor _ Vessel Temperature and Pressure 122 C.

Coolant Chemistry 123 D. -Coolant Leakage 126 E. - Safety / Relief Valves 127 F.

Structural Integrity

'128 128 G.

Jet Pumps 129 l

H.

_ Shock Suppressors (Snubbers) i l

3.6 and 4.6 Bases 144 156 3.7 and 4.7 Containment Systems 156 A.

Primary Containment l

B.

Standby Gas Treatment System 166 169-C.

Secondary Containment D.

Primary Containment Isolation Valves' 170 175 3.7 ' Bases 183 4.7 Bases rr 11 Amendment No. 3

LIST OF TABLES Page Ta bl e No._

3.1.1 Reactor Protection System (Scram) Instrument Requirements 28 4.1.1 Scram Instrument Functional Tests - Minimum Functional Test Frequencies for Safety Instrumentation and Control Circuits 32 4.1.2 Scram Instrument Calibration - Minimum Calibration Frequencies for Reactor Protection Instrument Channels 34

3. 2.1 Instrumentation that Initiates Primary Containment Isolation Functions 49 3.2.2 Instrumentation that Initiates Emergency Core Cooling 52 Systems 3.2.3 Instrumentation that Initiates Rod Block 56 3.2.4 Instrumentation that Initiates Reactor Building Ventilation Isolation and Standby Gas Treatement

~

59 System Initiation 3.2.5 Instrumentation that Initiates a Recirculation i=

Pump Trip 60 3.2.6 Instrumentation for Safeguards Bus Degraded Voltage 60A and Loss of Voltage Protection

.4.2.1 Minimum Test and Calibration Frequency for Core Cooling, Rod Block and Isolation Instrumentation 61 3.2.6 Trip Functions and Deviations 70 l

3. 6.1 Safety Related Snubbers 131 I'

4. 6.1 In-Service Inspection Requirements for Monticello -

138

3. 7.1 Primary Containment Isolation 172

4. 8.1 Monticello Nuclear Plant - Environmental Monitoring Program Sample Collection and Analysis 193 3.11.1

-Maximum Average Planar Linear Heat Generation Rate 214 vs. Exposure 3.14.1 Instrumentation fo'r Accident Monitoring 229b 4.14.1 Minimum Test and Calibration Frequency for Accident Monitoring Instrumentation 229c I

6.1.1 Minimum Shift Crew composition.

236 1

t L

Amendment No. 3 -

Amendment No. Z

}

(

(

(^

. gages.:

he abnormal operational transiente applicable to operation of the Monticello Unit have been analysed Ma throughout the spectrtam of planned operating conditions up to the theriimipower level of 1670 Wt.

2.3 analyses were based upon plant operation in accordance with the operating map given in Figure 3-2-3 of I

He licensed insrh power level 1670 We represet :

the FSAR.

-not knowingly be exceeded.

Conservatism is incorporated in'the transient analyses in estimating the controlling factors, such as void reactivity coefficient, control rod scram worth, scram delay time, peaking factors, and axial power nese factors are selected conservatively with respect to their effed on the applicable

shapes, his transient model, evolved over many transient results as determined by the current analysis model.

years, has been substantiated in operation as a conservative tool for evaluating reactor dynamic per-Results obtained froma a General Electric boiling water reactor have been compared with predictions made by the model. He comparisons and results are sumanarized in Reference 1.

formance.

he absolute value of the void reactivity coefficient used in the analysis is conservatively estimated Me to be about 257. greater than the nminal maximum value expected to occur during the core lifetime.

Doppler reactivity feedback coeffi,cient has conservatively been derated to 907 of the expected value.

ne scram worth _used has been derated to be equivalent to approximately 807. of t i

We

-the control rods.

set equal to the. longest delay and slowest insertion rate acceptable by Technical Specifications.all conservatively effect of scram worth, scram delay time and rod insertion rate, h e rapid insertion of greatest significance in the early portion of the negative reactivity insertion.

n e early portion of negative reactivity is assured by the time requirements for 57 and 20% insertion.the s De times for 507. and 907, insertion are given to assure proper completion of the and to establish the ultimate fully shutdoorn transient around.

expected performance in the earlier portion of the transient, steady-state condition, i

l 14 l

2.3 BASES Amendment No. 3

TAELE 4.1.2 SCRAM INSTRUMENT CALIBRATION MINIMUM CALIBRATION FREQUENCES FOR REACTOR PROTECTION INSTRUMENT CIIANNELS INSTRUMENT CilANNEL GROUP CALIBRATION METHOD MINIMUM FREQUENCY (2)

APRM E

Heat Balance Once every 3 days (4)

IRM E

Heat Balance See Note 1 High Reactor Pressure D

Pressure Standard Every 3 months High Drywell Pressure D

Pressure Standard Every 3 months Low leactor Water D

Pressure Standard Every 3 months High Water Level in Scram Discharge D

Water Level Every 3 months Condenser Low Vacuum D

Vacuum Standard Every 3 months High Steam Line Radiation E

See Note 3 See Note 3 Main Steamline Isolation Valve Closure D

Observation Every Operating Cycle Turbine Control Valve Fast Closure

.D Pressure Standard Every 3 months D.

Obse rvation Every Operating Cycle Turbine Stop Valve Closure Pressure Standard Every 3 months Recirculation Flow Meters &

Flow Instrumentation Notes:

1.

Perform calibration test during every startup an. normal shutdown.

c 2.

Calibration tests are not required when the systems are not required to be operable or are tripped.

If tests are missed, they shall be performed prior to retur ning the systems to an operable status.

3.

This instrument will be calibrated every three months by means of a build-in current source, and each refueling outage with a known radioactive source.

4.

This calibration is performed by taking a heat balance and adjusting the APRM to agree with the heat balance. Alarms and trips will be verified and calibrated if necessary during weekly functional test.

• CROUPS:

D.

Passive type devices.

34 E.

Vacuum tube or semiconductor devices and detectors

t. hat drif t or lose sensitivity.

3. 1/4.1 Amendment No. 3

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS

.. G. Safeguards Bus Voltage.Protecti.on 1.

Whenever the safeguards auxiliary electrical

. power system is required to be operable by

~

. Specification 3.9, the Limiting conditions for Operation for the. instrumentation listed in Table.3.2.6.shall be met.

T V

s t

}

e 4

48A

'3.2/4.2 Amendment No. 3 4

4 sa'

=

-ee

Table'3.2.2 - Continued Instrumentation Th_at Initiates Emergency Core Cooling System-Min. No of Oper-Min. No.

able or Operating of Operable Total No. of Instru-Instrument Channels or Operating ment Channels Per Per Trip System lepire.1 (3)

Cood it ion s*

Function Tr iy,,Se t,t ing, Trip Systetas,(3_),_

Trip System D.

Diesel Generator 1.

Degraded or Loss of Voltage Essential Bus (5) 2.

Low Low Reactor

>6'6"<6'10" 2

4(4) 4 D.

Water Level' 3.

High Drywell Pres-

<2 psig 2

4(4) 4 D.

NOTES:

1.

High drywell pressure may be bypassed when necessary only by closing the manual containment isolation valves during purging for containment inerting or de-inerting. Verification of the bypass condition shall be noted in the control room log. Also need not be operable when primary containment integrity is not required.

-2.

Due instrument channei is a circuit breaker contact and the other lo an undervoltage relay.

3.2/4.2 54 Amendment No. 3

Table 3.2.2 - Continued Notes:

~

3.

UPon discovery that minimu:n requir-ments for the number of operable or operating trip systems, or instrument channels are not satisfied action shall be initiated to:

(a) Satisfy the requirements by placing appropriate channels or systems in the tripped condition, or (b) Place.the plant under the specified required conditions using norinal operating procedures.

4.

All instrument channels are shared by both trip systems.

5.

S'ee Table 3.2.6.

Bequired conditions when mini-iw conditions for operation are not satisfied.

A.

Caply with Specification 3 5.A.

B.

Comiply with Specification 3 5.D.

C.

Beactor pressure 4150 peig.

D..Caply with Specification 3 9.B.

a 32/4.2 55 Amendment No.

3..

- 55

. ii;

]ii' "'

I si: '1 '

i r

Table 3.2.6 Instrumentation ' for Safeguards Bus Degraded Voltage: and Loss of Voltage Protection Mintwum No. of-Total No. of-Minimum No. o' per-Required Operable or Instrument Channels able or Operk

.ag Conditions

• Operating Trip Per Trip Syntem Channels Per Function-Trip Setting Systems (1)

Trip System (1) 1.-

Degraded Voltage 3885 + 18V 1/ bus 3

3 A

Protection (3) 10

+ 1 see 2.

Loss of Voltage 2625 7 175V 2 ous 2

2 A

Protection (2)-

T 1 see NOTE:

1.

Upon discovery that minimum requirements for the number of operable or operating trip systems or instrument channels are not satisfied, action shall be initiated to:

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

b.

Place the plant under the specified required conditions using normal operating procedures.

2.

One out of two twice logic.

3.

Two out of three logic.

• Required conditions when minimum conditions for operation are not satisfied:

A.

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

60A 3.2/4.2 Amendment No. 3 4

. Table 4.2.1 - Continued Minimum Test and Calibration Frequency For Core Cooling Rod Block and Isolation Instrumentation

. Instrument Channel

. Test (3)

Calibration G)

Sensor Check (3) 3.

Steam Line Low Pressure-Once/ month Once/3 months None 4.

Steam Line High Radiation Once/ week (5)

Note 6 Once/ shift HPCI ISOLATION s

. 1.

. Steam Line High Flow

~Once/ month Once/3 months None

-- 2.

Steam Line High Temperature 3nce/ month Once/3 months None RCIC ISOLATION 1.

Steam Line High Flow Once/ month Once/3 months None 2.

. Steam Line -High Temperature Note 1 Once/3 months None REACTOR BUILDING VENTIALTION 1.

Rad'stion Nonitors (Plenum)

Note 1 Once/3 months Once/ shift

'2.

Radiation knitors (Refueling Floor)

Note 1 Once/3 months (4) 4 OFF-CAS ISOLATION 1.

Radiation Montiors (Air. Ejectors)

Notes (1,5)

Note 6 Once/ shift RECIRCULATION PUMP TRIP

-1.

Reactor High Pressure Note 1 Once/ Operating Cycle-Once/ Day I

Transmitter Once/3 Nonths-Trip Unit 2.

Reactor Low Water Level (Note 7)

Once/ month Once/ Operating Cycle-Once/ shift Transmitter Once/3 Months-Trip Unit 3.2/4.2 62 Amendment No. 3.

i -

Table 4.2.1 - Cort inued Minimum Test and Calibretion Frmuency for Core Cooling, Rod Block and Isolation Instrumentation Instrument Channel Test (3)

. Calibration (3)

Sensor Check (3)

SAFECUARDS BUS VOLTAGE 1.

Degraded Voltage Note 1 Quarterly Not applicable

- Protection 2.

Loss of Voltage Note 1 Once/ Operating Cycle Not applicable Protection 6

NOTES:

(1)~ Initially. once per month until exposure hours (M as defined on Figures 4.1.1) is 2.0 x 10, thereaf ter according to Figure 4.1.1 with an interval not greater than three months.

~

(2) Calibrate prior to normal shutdown and start-up and therea f ter check once per shif t and test once per veck until no longer requ' red. Calibration of this instr.mment prior to normal shutdown means i

adjustment ' channel trips so that they correspond, within acceptable range and accuracy, to a simulated signal injected into the instrument (not primary sensor). In addition, IPM gain adjustment will be performed, as necessary, in the ArRM/ IRK overlap region.

(3) ninctional tests, calibrations and sensor checks are not required when the systems are not. required to be operable or are tripped. If tests are missed, they shall be performed prior to returnin6 the systerns to on operable status.

(h) Whenever fuel handling is in proccas, a sensor check shall be perfonned once per shift.

(5) A Mmetional tect of this instrument menns the ir.jection of a cimulated signal into the inotnment (not primary sensor) tn verify the proper instrment channel response alarm and/or initiating action.

((s); 'Ihis instnament vill be calibrated every three months by means of a built in current source, and each j

renseling outage with a known radioactive source.

(7) Surveillance also to be performed on containnent isolation function of this instrumentation at the l

specified intervals.

Amendment No. 2 3 63

~ - _ _

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS C.

Scram Insertion Times C.

Scram Insertion Times 1.

The average scram insertion time, During each Operating cycle, each based on the de-energization of the operable control rod shall be sub-l scram pilot valve solenoids at time jected to scram ti:ne tests from the zerc, of all operable control rods fully withdrawn position.

If testing in the reactor power operation con-is not accomplished during reactor dition shall' be no greater than:

power operation, the measured scram 1

insertion times shall be extrapolated

% Inserted From Avg. Scram Insertion to the reactor power operation condi-

. Fully Withdrawn Times (sec) tion utilizing previously determined 5

0.375 correlations.

20 0.900 50 2.00 90.

3.50 2.

The average of the scram insertion times for the three fastest control rods of all groups of four. control rods in a two by two array shall be no greater than:

Percent of Rod Length Inserted Seconds 5

0.398 20 0.954 50 2.120 90 3.80 1

3.3/4.3 53 1

-Amendment No. 3 i.

Bases Continued 3.3 and 4.3:

The scram times for all control rods will be determined during each operating cycle. The weekly control rod exercise test. serves as a periodic check against deterioration of the control rod system and also

. verifies the ability of the control rod drive to scram since if a rod can be moved with drive pressure, it will scram because of higher pressure applied during scram.

Allowing for monthly exercising of one rod in any two by. two array is consistent with the bases for local and overall core reactivity insertion ratesL assumed in the transient analyses discussed above. The frequency of exercising the control rods under the conditions of two or more control' reds out of sersi provides even further assurance of the reliability of the remaining control rods.

. The occurrence of scram times within the limits, but significantly longer than the average, should be

. viewed as an indication of a systematic problem with control rod drives especially if the number of drives ' exhibiting such scram times exceeds six, the allowable number of inoperable rods.

D.

Control Rod' Accumulators The basis for this specification was not described in the FSAR and, therefore, is presented in its entirety.

Requiring no more than one inoperable accumulator in any nine-rod square array is based on a series of KY PDQ-4 quarter core. calculations of a cold, clean core.

The worst case in a nine-rod withdrawal sequence resulted in a k,gg<1.0 -- other repeating rod sequences with more rods withdrawn resulted in k gg>1.0.

At reactor pressures in excess of 800 psig, even those control rods with inoperable ac,cumulators will be able to meet required scram insertion times due to the action of reactor

-pressure. In addition, they may be normally inserted using the control-rod-drive hydraulic system.

Procedural control will assure' that control rods with inoperable accumulators will be spaced in one-in-nine

= array rather than grouped together.

~

E.

Reactivity Anomalies During each-fuel cycle excess operating reactivity varies as fuel depletes and as any burnable poison i

in supplementary control is burned. The magnitude of this excess reactivity is indicated by the integrated worth of control rods inserted into the core, referrei to as the control rod inventory in the core. As fuel burnup progresses, anomalous behavior in the excess reactivity may be detected by comparison of actual rod inventory at any. base equilibrium core state to predicted rod inventory at that state.

Rod inventory predictions can be normalized to actual initial steady state rod patterns to minimize calcula-4 tional uncertainties. Experience with other operating BWR's indicates that the control rod inventory should be predictable to the equivalent of one per cent in reactivity.

3.3/4.3 Bases 91 4

Amendment No. 3 4

4

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMLdTS S.

Reactor Vessel Temperature and Pressure B.

Reactor Vessel Temperature and Pressure

'1.

.During in-service hydrostatic or leak test-1.

During in-service hydrostatic or leak

~ing, the' reactor vessel shell temperatures testing when the vessel pressure is

~

specified in 4.6.B.1-shall be at or.above above 312 psig, the following' temper-the higher of the temperatures shown on the atures shall be recorded at least every two curves of Figure 3.6.2.where the dashed 15 minutes :

curve, "RPV Beltline Region," is increased by the expected shift in RT r a Figure a.

Reactor vessel shell adjacent NDT 3.6.1.

to shell flange.

b.

Reactor vessel bottom head.

2. ' During heatup by non-nuclear means (except with the reactor vessel 2.

Test specimens representing the vented), cooldown - following nuclear reactor vessel, base weld, and weld shutdown, or low level physics tests heat af' acted zone metal shall be the reactor vessel shell and fluid installed in the reactor vessel temperatures specified in 4.6.A shall adjacent to the vessel wall at the be at or above.the higher of the core midplane level. The material

temperatures of Figure-3.6.3 where the sample program shall conform to dashed curve, "RPV Beltline Region,"

ASTM E 185-66.

Samples shall be is increased by the expected shift in withdrawn at one fourth and three RT fr m Figure 3.6.1.

fourths service life. Analysis of NDT the first sample shall include a 3.

During' all operation with a critical quantitiative determination of the reactor, other than for low level copper and phosphorous content, physics tests or at times when the reactor vessel is vented, the reactor 3.

Neutron flux wires shcil be installed vessel shell and fluid temperatures in the reactor vessel adjacent to the specified in 4.6.A shall be at or reactor vessel wall at the core mid-above the higher of the - temperatures plane level. The wires shall be removed of Figure 3.6.4 where the dashed curve, and tested during the first refueling "RPV Beltline Region," is increased outage to experimentally verify the by the expected shift in RT fr a calculated value of neutron fluence at NDT Figure 3.6.1.

one fourth of the beltline shell thickness that is used to determine the NDTT shift from Figure 3.6.1.

3.6/4.6 122 Amendment No. -3 t

3.0 LWITING CONDITIONS FOR OPERATION 4.0 SURVELLLANCE REQUIREMENTS i

5.

Snubbers may be added to safety related The required inspection interval shall not be systems without prior. License Amendment lengthened more than one step at a time, to Table 3.6.1 provided that a revision to Table 3.6.1 is included with the next Snubbers may be categorized in two groups, license amendment request.

" accessible" or " inaccessible" based on their accessibility for inspection during reactor operation. These two groups may be inspected independently according to the above schedule.

2.

All hydraulic snubbers whose seal materials are other than ethylene propylene or other material that has been demonstrated to be compatible with the operating environment shall be visually in-spected for operability every 31 days.

3.

Once each Operating Cycle, a representative sample l

of 10 hydraulic anubbers or approximately 10% of the hydraulic snubbers, whichever is less, shall be functionally tested for operability including verification of proper piston movement, lock up, and bleed. For each unit and subsequent unit found inoperable, an additional 10% or ten hydraulic snubbers shall be so tested until no more failures are found or all units have been tested.

Snubbers designated in Table 3.6.1 as being especially dif ficult to remove or located in High Radiation Areas during shutdown are exempt i

from this requirement.

4.

Snubbers may be reclassified as being in or out of High Radiation Areas during shutdown in Table 3.6.1 based on the most recent radiation survey provided that a revision to Table 3.6.1 is included with the next license enemdment request.

3.6/4.6 130 Anendment No. 3

^

Bases Continued 3.6 and 4.6:

D.

Coolant Leakage The former 15 gpm limit for leaks from unidentified sources was established assuming such a leakage was coming from the primary system. Tests have been conducted which demonstrate that a relationship exists between the size of a crack and the probability that the crack will propagate.

From the crack the probability size a leakage rate'can be determinej5. For a crack size which gives a leakage of 5 gpm, of rapid propagation is less than 10 Thus, an unidentified leak of 5 gpm when assumed to be from the primary system had less than one chance in 100,000 of propagating, which provides adequate margin.

A leakage of 5 gpm is detectable and measurable..The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period allowed for determination of leakage is also based on the low probability of the crack propagating.

The capacity of the drywell sump pumps is 100 gpm and the capacity of the drywell equipment drain tank pumps is also 100 gpm.

Removal of 25 gpm from either of these sumps can be accomplished with consider-able margin.

E.

Safety / Relief Valves Testing of all required safety / relief vavles each refueling outage ensures that any valvo deterioration is detected. A tolerance value of 1% for safety / relief valve setpoints is specified in Section III of the ASME Boiler and Pressure Vessel Code. Analyses have been performed with all valves assumed set 1%

higher (1108 peig +.1%) than the nominal setpoint; the 1375 psig code limit is not exceeded in any case.

The safety / relief valves are used to limit reacter vessel overpressure and fuel thermal duty.

The. required safety / relief valve steam flow capacity is determined by analyzing the transient accompany-ing the main steam flow stoppage resulting from a postulated MSIV Closure from a power of 1670 Mwt.

The analysis assumes a multiple-failure wherein direct scram (valve position) is n'eglected. Scram is assumed to be from indic-xt means (high flux).

In this event, the safety / relief valve capacity is assumed to be 83.2% of the full power steam generation rate.

1 3.6/4.6 BASES 150 Amendment No. 3

H. ~ Shock Suppressora (Snubbers) (contd.')

Examination 'of defective snubbers at reactor facilities and material tests performed at several laboratories has shown that millable gum polyurethane deteriorates rapidly under the temperature and moisture conditions present 'in many snubber: locations. Although molded polyurethane exhibits greater resistance to these conditions, it. also may be unsuitable for application in the higher temperature environments. Data are not currently available to precisely define _ an upper temperature limit for the molded polyrethane. Lab tests f and in plant experience indicate that seal materials are available, primarily ethylene propylene compounds, which should give satisfactory performance under the most severe conditions expected in reactor installations.

To further increase the assurance of' snubber reliability, functional tests should be performed once each operating cycle. These tests will incude stroking-of the snubbers to verify proper piston movement,

' lockup and bleed. Ten percent or ten snubbers, whichever is less, represents an adequate sample for such tests. Observed. f ailures on these samples should require. testing of additional units.

Snubbers in s

High Radiation Areas or those especially dif ficult to remove need not be selected for functional tests

'provided operability was previously verified. Snubbers are considered especially dif ficult to remove if they (1) have a rated capacity greater than 50,000 lb, (2) are located greater than 5 feet above the adjacent platform, or (3) located greater than 3 feet below the adjacent pla t fo rm.

Y 3.6/4.6~

BASES 155 Amendment No. 3.

J

. 3'.0. - LIMITINC " CONDITIONS FOR ' OPERATION 4.0.. SURVEILLANCE REQUIREMENTS C.

Secondary Containment C.

Secondary Containment 1.

Except as.specified in 3.7.C.2 and 1.

Secondary containment surveillance shall

. 3.7.C.3, Secondary Containment Integrity be performed as indicated below:

shall be maintained during all modes of o

' plant operation.,

a.

Secondary containment capability to i

maintain at least a 1/4 inch of water

2.. Secondary Containment Integrity is not vacuum under calm wind (2 < u < mph) required when all of the following con-conditiens with a filter train flow ditions are satisfied:

rate of <4,000 scfm, shall be dem-onstrated at each refueling outage

-a.

the reactor is suberitical and prior to re fueling. Verification

. Specification 3.3.A is set.

that each automatic damper actuates to its -isolation position shall l b...The reactgr water ' temperature is be performed at each refueling outage below 212 and.the reactor coolant and af ter maintenance, repair or replace-system is vented.

ment work is performed on the damper er its associated actuator, control circuit.

c.

No activity is being performed which or power circuit.

.can reduce the shutdown margin below that specified in Specification 3.3. A s

d.

The fuel cask or irradiated fuel is not being moved within the reactor building.

3.

_ With an inoperable secondary contain-ment ' isolation damper, restore the inop..rable damper-to operable status or isolate the affected duct by use of a closed damper or

-blind flange within eight hours.

4.

If Specifications 3.7.C.1 through

- 3.7.C.3 cannot be met, initiate a normal. orderly shutdown and have the reacto; in-the Cold Shutdova condition l

-withie 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Alterations of the 3.7/4.7 169 Amendment No. 3

3.0 LIMITING CONDITIONS FOR OPRATION 4.0 SURVEILLANCE REQUIREMENTS D.

Primary Containment Isolation Valves reactor core, operations with a

-potential for reducing the shutdown 1.

The primary containment isolation valve margin below that specified in surveillance shall be performed as follows:

specification 3.3.A, and handling of irradiated fuel.or the fuel cask a.

At least once per operating cycle the in the Secondary Containment are to l

be immediately suspended if secondary operable isolation valves that are power operated and automatically containment integrity is not main-initiated shall be tested for simulated

tained.

automatic initiation and closure times.

b.

At least once per operating cycle the D.

Primary Containment Isolation Valves primary system instrument Itne flow 1.

During reactor power operating conditions, e eck 7alves shall be tested for proper all isolation valves listed in Table 3.7.1 ope ra t ton, and all primary system instrument line c.

At least once per quarter flow check valves shall be operable except as specified in 3.7.D.2.

(1) All normally open power-operated isolation valves (except for the main steam line power-operated isolation valves) shall be fully closed and reopened.

3.7/4.7 170 Amendment No. 3.

p t

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVElLLANCE REQUIREMENTS

.c.

For the diesel generators to be c.

At least once each Operating Cycle considered operable, there shall be-during shutdown simulate a loss of of fsite a minimum of 26,250 gallons'of diesel power in conjunction with an ECCS actuation fuel.(7 days supply for 1 diesel gen-test signal, and:

erator at full load) 'in the diesel oil storage tank.

1.

Verify de-energization of the emergency

/-

busses and load shedding from the emer-gency busses.

2.

Verifying diesel starts from ambient conditions on the auto-start signal and is ready to accept emergency loads wit in ten seconds, energizes the emergency busses with permanently connected loads, energizes the auto-connected emergency loads in proper time sequence, and operates for greater than five minutes while its generator is loaded with the emergency loads, d.

During the monthly generator test, the diesel fuel oil transfer pump and diesci oil service pump shall be operated.

e.

Once a month the quantity of diesel fuel available shall be logged.

f.

Once a month a sample of diesel fuel shall be taken and checked for quality.

3.9/4.9 202 I

Anendment No. 3

3.-0 ' LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 4.

Station Battery System 4

Station Battery System If one of the. two 125 V battery systems or a.

Every week the specific gravity the 250 V battery system is made or found and voltage of the pilot cell to be inoperable for any reason, an orderly and temperature of the adjacent shutdown of the reactor will be initiated cells and overall battery voltage and the reactor water temperature shall be shall be measured, reduced to less than 212 F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless such battery systems are sooner made operable.

b.

Every three months the measure-ments chall be made of voltage of each cell to nearest 0.01 volt, specific gravity of each cell, and temperature of every fifth cell, c.

Every refueling outage, the station batteries shall be sub-l jected to a rated load discharge test.

Determine specific gravity i

o and voltage of each cell af ter the discharge.

5.

24V Battery Systems 5.

24V Battery Systems From and after the date that one of the two a.

Every week the specific gravity and 24V battery systems is made or found to be voltage of the pilot cell and tempera-l inoperable for any reason, refer to Specifi-ture of adjacent cells and overall caton 3.2.for appropriate action.

battery voltage shall be measured.

b.

Every three months the measurements shall be made of voltage of each cell to nearest 0.01 volt, specific gravity of each cell, and temperature of every fifth cell.

3.9/4.9-Amendment No. 3

?03 e

P Rl". l DENT SENIOR VICE PRt.SIDENT POWER SUPPLY VICE PRESIDENT-VICE PRESIDENT-PIANT ENGINEERING AND CONSTRUCTION

^

CENERAL MA*:AGER HEADQUARTERS NUCLEAR PLANTS NUCLEAR SERVICES MANAGER OF PLANT MANAGERS NUCLEAR SUPPORT NUCLEAR PLANTS l

SERVICES I

i I

l SAC ADMINISTRAr10N g

I l

I GENERAL SUPERINTENDENT SAFETY AUDIT I

COMMITTEE (SAC)

OPERATIONAL QA 1

I I

I 8

i_ _ AUDIT &

_i REVIEW FIGURE 6.1.1

'NSP CORPORATE ORGANIZATIONAL HEl.AT10NSillP TO ON-SITE OPERATING ORGANIZATION 6.1 Amendment No. 3-h-

4>

9*

%#:;>49

/////

.s%

$l/// \\$q)Y

// Y $Ykkf

+e+y,,,/

%'44

.e v <e _

TEST TARGET (MT-3) l.0 if a Ha y "2j I!!E I.l

["2 EM

!11 1.25 1.4 g

\\

d%

/

4 x%4

• $F86,,

'4/<y h

s &,,,/////

4

%@'q)de

///

AMs

\\VNp,,fv 4/,

//47g cc.e vo g,

e

'f+,,9, (4,,Njh V'^

IMAGE EVALUATION TEST TARGET (MT-3) l.0 gmgu y a gg I.l

[* lilM 1.8 1.25 1.4 1.6

=m

/

6" 4

\\

1. %p,,,

f4

• SNV b//

'4A..b 5,,

PINIT OPERATIONS hAlW.lER*

COMMIITCE I

I I

r --- -

PIANT SUPERINTLTDENT-SITERINTENDENT SUPERIllTEMDENT PIRIT SUPERINTENDElG-OPERATION &

SECURITY &

QUALITY-ENGINEERING &

MAINTDIANCE*

SERVICES-ENGINEERING

• RADIATION PROTECTION *(LS0)

I 1

IRAINING SUPPORT FOR QA I

SUPERVISOR AND QC FUNCTION J

SUPERINTENDENT SUPERINTENDENT SUPERINTENDENT SUPERINTENDENT OF SUPERINTENDENT OF RADIATION TECHNICAL ~

OPERATICES MAINTENANCE

• OPERATIONS *(I.S0)

PROTFCTION

• ENGINFFRTNG

• ENCTMFFETMC*

i SHIFT SUPERVISOR (ISO) i MECllANICAL &

LEAD PIANT EQUIP-TEGINICAL ENGINEERS FOR TECHNICAL SUPPORT EECTRICAL MENT & REACTOR SUFFORT &

1:UCLEAR ENGINEER.

FOR OPERATION, MAINTENANCE OPERATOR (ID)

RADIATION ING, INSTRUMENTA.

MAINTENANG, SUR-

. CROUP PROTECTION TION, CONTROLS, VEILLANCE, & TEST-PIANT EQUIPMDg SPECIALISTS COMPUTER; INSTRU-ING MENT & CONTROLS

& REACTOR OPERATOR (14)

ASSISTANT PIANT Code:

• Key Supervisor MONTICELLO NUCEAR GENERATING PIANT I

Licensed Operator Functional Organization for EQUIP OPERATOR LSO Licensed Senior Operator On-Site Operating Group PLANT ATIDiDA.T FIRE BRIGADE FIGURE 6.1.2 g.tBERS 235 (AS REQUIRED)

Amendment No. 3

6.2 -Review and Audit Organizational units for the review and audit of facility operations shall be constituted and have the responsibilities and authorities outlined below:

A.

Safety Audit Committe1 (SAC)

P.ee Safety Audit Comin' tee provides the independent review of plant operations from a nuclear

. safety standpoint.

Audits of plant operation are enducted under the coRizance of the SAC.

1.

Membership a.

The SAC shall consist of at least five (5) persons.

b.

The SAC Chairman shall be an HSP representative, not having line responsibility for operation of the the plant, appointed by the Vice Prealdent - Powe r Product ion.

Other members shall be appointed by the Vice President - Power Production or by such other person as he may designate. The Chainaan shall appoint a Vice Chairman froin the SAC 4

membership to act in his absence.

c.

No more than two members of the SAC shall be from groups holding fine responsibility for operation of the plant.

d.

A SAC member may appoint an alternate to serve in his absence, with concurrence of the Chairman. No more than one alternate shall serve on the SAC at any one time.

The alternate member shall have voting rights.

2.

Quali fications a.

The SAC members should collectively have the capability required to review activities in the following areas: nuclear power plant operations, nuclear engineering, chemistry and. radiochemis try, metallurgy, instrumentat ion and control, radiological saf ety,

mechanical and electrical engineering, quality assurance pract ices, and other appropriate fields associated with the unique characteristics of the nuclear power plant.

6.2 237 Amendment No. 3

f.

. Investigation of all events which are required by regulation or technical

.spectfications"to be reported to NRC in writing within 24 hourn.

.g.

Revisions to the Facility Emergency Plan, the Facility. Security Plan, and the Fire Protection Program, i

h.

Operations Committeo minutes.to determine if matters considered by that Committee involve unreviewed or unresolved nafety questions.

I.

Other nuclear safety'esatters referred to the SAC by the Operations Committee, plant management or company management.

J.

All recognized indications of an unanticipated deficiency in some aspect of l

design or operation of' safety-related structuren, systems, or components.

k.

Reports of special inspections and audita conducted in accordance with specification 6.3.

1

~6.

Audit

'Ihe operation of the nuclear power plant shall be audited formally under the cognizance of the. SAC to assure safe facility operation.

a.

Audits of selected aspects of plant operation, as delineated in Paragraph 4.4

'of ANSI N18.7-1972, shall be performed with a frequency commensurate with their

. nuclear safety significance and in a manner to assure that. an audit of all nuclear safety-related activitlen in completed within a period of two yearn.

.'Ihe audits shall be performed in accordance with appropriate written IONtructions and procedures.

b.-

Periodic review of the audit program should be performed by the SAC at least twice a year to assure its adequacy.

Written reports of the audits shall'he reviewed by the Vice Prealdent - Power

.c.

Production, by the SAC at a scheduled meeting, and by members of management having responsibility in the areas audited.

6.2 Amendment No. 3 W

4

,w

e O

7.

Authority The SAC shall be advisory to the Vice President - Power Production.

I 8.

Records Minutes shall be prepared and retained for all scheduled meetings of the Safety Audit Committee. The minutes shall be distributed within one month of the meeting to the Vice

- Presidert - Power Production, the General Manager Nuclear Plants, each member of the SAC and others designated by the Chairman or Vice Chairman. There shall be a formal approval of the minutes.

9.

Procedures A written charter for the SAC shall be prepared that contains:

Subjects within the purview of the group.

a.

b.

Responsibility and authority of the group.

c.

Mechanisms for convening meetings.

d.

Provisions of use of specialists or subgroups.

~

Authority to obtain access to the nuclear power plant operating record files e.

and operating personnel when assigned audit functions, Requirements for distribution of reports and minutes prepared 'by the group to f.

others in the MSP Organization.

a 6/2 24u Amendment No. 3 4

1 A.

Operations Committee (OC) l.

Membership The Operations Committee shall consist of at ' least six (6) members drawn from the key super-visors of the on-site. supervisory staf f.

The Plant Manager shall serve as Chairman of the DC and shall appoint a.Vice Chairman from the DC membersip to act in his absence.

2. ' Meeting Frequency

'The Operations Committee will meet on call by the Chairman or as requested by individual members and at least monthly.

i 3.

Quorum 1

A quorum shall include a majority of the permanent members, including the Chairman or Vice Chairman 4.

Responsibilities - The following subjects shall be reviewed by the Operations Committee:

l a.

Proposed tests and experiments and their results, b.

Modifications to plant systems or equipment as described in Final Safety Analysis Report and having nuclear safety significance or which involve an unrevie'ved safety question as defined in 10 CFR 50.';9.

c.

Proposals which would effect permanent changes to normal and emergency operating i

procedures and any other proposed changes or procedures that are determined by

{

the Plant Manager to af fect nuclear safety.

d.

Proposed changes to the Technical Specifications or operating license, i

l

'e.

All reported or suspected violations of Technical Specifications, operating license requirements, administrative procedures, or operatinh Procedures.

Results of investi-l gations, including evaluation and recommendations to prevent recurrence, will be I

reported, in writing, to the General Manager Nuclear Plants and to the Chairman of the Safety Audit Committee.

6.2 241 Amer: bent No. ~3 t

f.

All events which are required by regulations or Technical Specifications to be reported to NRC in writing within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, g.

Drills on emergency procedures (including plant evacuation) and adequacy of communi-

, cation with off-site support groups, h.

All procedures required by these Technical Specifications, including implementing procedures of the Emergency Plan and the Security Plan, shall be reviewed with a frequency commensurate with their safety significance but at an interval of not more than two years.

i. Perform special reviews and investigations, as requested by the Safety Audit Com-mittee.

5.

Authority The OC shall be advisory to the Plant Manager.

In the event of disagreement between the recommendations of the OC and the Plant Manager, the course determined by the Plant Manager to be the more conservative will be followed. A written summary of the disagreement will be sent to the General Manager Nucl'ar Plants and the Chairman of the SAC for review.

6.

Records Minutes shall be recorded for all meetings of the OC and shall identify all documentary material reviewed. The minutes shall be distributed to each member of the OC, the Chairman and each member of the Safety Audit Committee, the General Manager Nuclear Plants and others designated by OC Chairman or Vice Chairman.

i 7.

Procedures ll written charter for the OC shall be prepared that contains:

Responsibility and authority of the group a.

b.

Content and method of submission of presentations to the Operations. Committee 6.2 242 Amendment No. 3 4

4

d c.

Mechanism for scheduling meetings d.

Meeting agenda e.

Use of subcommittee

- f.

Review and approval, by members, of DC actions g.

Distributien of minutes 6.3 Special Inspections and Audits A.

An independent fire protection and loss prevention inspection and audit shall be performed annually utilizing either qualified of f-site Northern States Power

- Company personnel or an outside fire protection consultant.

B.

An inspection and audit by an outside qualified fire protection consult 1nt shall be pe r fo rmed a t intervals no greater than three years.

6.4 Aa* ion to be Taken if a Safety Limit is Exceeded If a Safety Limit is exceeded, the reactor shall be shut down immediately. An immediate report shall be made to the Commission and to the Ceneral Manager Nuclear Plants, or his designated alternate in his absence. A complete analysis nf the circumstances leading up to and resulting from the situation, together with recommendations by the Operations Committee, shall also be prepared. This report shall be submitted to the Commission, to the General Manager Nuclear Plants and the Chairman of the Safety Audit Committee within 14 days of the occurrence.

Reactor operation shall not be resumed until authorized by the U.S. Nuclear Regulatory Commission.

i 6.2 - 6.4 243 Amendment No. 3