ML20127K319

Amend 9 to License DPR-22 (Change 18) Correcting Typos

ML20127K319
Person / Time
Site:	Monticello
Issue date:	04/10/1975
From:	Ziemann D Office of Nuclear Reactor Regulation
To:
Shared Package
ML20127K314	List: ML20127K310 ML20127K319 ML20127K328 ML20127K336
References
NUDOCS 9211200372
Download: ML20127K319 (40)

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ATTAQMDrr 7t) LICENSE A>"mnispyy % o

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OWKiB NO. :18 TO 1MB TBCHNICAL SPECIPICATIONS PROVISIONAL OPPJtATING LICEME NO. DPR-22 DOCKET NO. 50-263 l

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Replace the existing pages of the Technical Specifications listed below with the attached ravised pages bearing,the same numbers, except as otherwise noted. Changed areas on these pages ar's shown by marginal linest i

Pgs, i - vii Pg. 108 Pg. viii - Delete Pg. 116 Pg. 38 Pt. 117 i

Pg. 39 Pg. 118

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Pg. 40 Pg. 118A - Add Pg. 43 Pg. 132 Pg. 45 Pg. 133 Pg. 59 Pg. 147A Pg. 62 Pg. 148 Pg. 63 Pg. 157 Pg. 78 Pg. 170 Pg. 85A Pg. 172 Pg. 89 Pg. 181 Pg. 94 Pg. 183 Pg. 104 l

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Foran AEC-315 (Rev. 9 53) AECM 0240 W u. s. movanmusut enswTine arrics: nov4.sae-see l

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Technical Sp.cifications We Technical Specifications contained in Appendix A, as revised, are hereby incorporated.in the license.

4% The licensee shall,eperate'the" fact'lity=in accordance sw-with the Technica1' Specifications,ias revised by issued changes thereto through Change No. 18.

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

POR THE NUCLEAR REGULATORY COMMISSION knol signed by I)ennis L. Demans

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Dennis L. Ziemann, Chief Operating neactors Branch #2 Division of Reactor Licensing i

Attachment:

i Change No. 18 to the Technical Specifications Date of Issuance:

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WORTHERN STATES POWER 00MPANY DOCKET No. 50-263

)ONTICELLO NUCLEAR GENERATING FLAFF AlfENDifENT TO PROVISTORAL OPERAT NG'LYCENSE M 4/ w we

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u.,..uuay Amendment No. 9 l

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License No. DPR-22 l

The Nuclear Regulatory Commission (the Cometasion) having found that:

1.

The application for amendment by the Northern States Power A.

company (the licensee) dated November. 15, 1974, complies with 4

the standards and requirements of the Atomic Energy'Act of 1954, as amended (the Act), and the Consaission's rules and regulations set forth in 10 CTR Chapter I; The facility v111' operate in conformity with the application, B.

the provisions of the Act, and the rules and regulations of f

the Commission; l

There is reasonable assurance (1) that the activities authoris:ed C.

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 Connaission's regulations; and The issuance of this n '==nt will not be inimical to the D.

common defense end security or to the health and safety of the public.

Accordingly,-the license is==maded by a change _to the Technical 2.

4 Specifications as indicatedin the attachment to'this' license amenduent and Paragraph 3.B. of Provisional Operating License No. DPR-22 is hereby amended to read as follows:

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' Northern States Power ' cpany APR 10 D cc w/cnclosures:

Tne Environmental Conservation Arthur henquist, Esquire Vice Presid2nt

-Law Lit >rary Northcrn Slates Power Company Minneapolis Public Librar/

300 ::icollet Mall 414 :,'1 collet Mall Minneapuus, :!innesota 55401 Minneapolis, Minnesota 55401 Mr. D.

S. Douelas, Auditor Cerald Charnalf Wri,ht County licard of Con.missioners Shaw, Pittman, Potts, Trowbridge P

Duffalo, Minnesota 55313 and Madden 910 - 17th Street, M. W.

Washingter., D.

C.

20036 Warren R. I.auson,'!.

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Secretary and Enocutive officer l'nive rsiti' Campur lioward J. Vo;;el, Esquirc

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Minneapolis, Minnesota 55440 Le nt f:eunsel 2730 :;ean Parimav

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" nnesot a 3541G "r. Carv Mili n federal Activitic

Iranch Pavironmental Protection A"ency Steve Cadler, P.

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1 N. Wacker Drive, noom su 2120 Carter /-venuo St. Paul, "innesota 5510S Ch ic, ;;o, a11nois 60006 Mr. Daniel L.

Ficker Assistant City Attorney 63S City lla11 St. Paul,'linnesota 55102 Ken D ugan, Director-City of Ft. Paul Pollution I

Control Services 100 East 10th Street St. Paul, innesota 55103 Sandra S. Gardebring Special Assistant Attorney General Counsel tor Mirnesota Pellution Control Agency 1935 W. County Road D2 i

Roseville, Minnesota 35113 Anthony Z.

Roisman, Esquire Berlin, Roisman and Kessler 1712 N Street, N. W.

i Washington, D.

C.

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i TABIE OF CONTENTS j

Page 1

1.0 DEFINITIONS 6-2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 6

2.1 and 2.3 Fuel Cladding Integrity 13 1

2.1 Bases 18 2.3 Bases 23 2.2 and 2.4 Reactor Coolant System 24 2.2 Bases 26 2.4 Basen 28 3.0 LIMITING CONDITIONS FOR OPERATION AND 4.0 SURVEILIANCE REQUIREMENT 3 18 28 3.1 and 4.1 Reactor Protection System 37 3.1 Bases 43 4.1 Bases 47 3.2 and 4.2 Protective Instrumentation 47 A.

Primary Containment Isolation Functions 48 B.

Emergency Core Cooling Subsystems Actuation

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Control Rod Block Actuation 48 D.

Air Ejector Off-Cas Systen i

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Reactor Building Ventilation Isolation and 49 I

Standby Gas Treatment System Initiation 3.2 Bases 64 4.2 Eases 71-3 3 and 4.3 Control Rad System 75 A.

Reactivity Limitations 75 B.

Control Rad Withdrtwal 76 C.

Scram Insertion Times 79 D.

Contml Rod Accumulators 80 E.

Reactivity Anomalies 81 3.3 and 4.3 ibses 82 3.4 and 4.4 Standby Liquid Control System 68 18 A.

Nortaal Ope mtion 88 B.

Operation with Inoperable Components 89 C.

Volume-Concentration Requirements 90 3.4 arat 4.4 Bases 94 3 5 and 4.5 Core and Containment. Cooling Systems 96 A.

Core Spmy System 96 B.

LPCI Subsystem 98 C.

EliR Service Water Systa 101

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'HPCI System 103 I

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Automatic Pressure Relief System 104

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RCIC System 106 I

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Minimu;n Core and Containment Cooling System Availability

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Recirculation Systema 108A I

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Average Planar IRGR 108A

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Incal IllGR 1088 l

i 3.5 Bases 109 j

t 4.5 Bases 114 1

3.6 and 4.6 Primary System Boundary 115 l

A.

Thermal Limitations.

115 18 t

B.

Pressurization Temperature 116.

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Coolar.t Chemistry 116

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Coolant Leakage 118 k

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Safety / Relief Valves 119 F.

Structural Integrity 120 G.

Jet Pumps 120-i 3.6 and 4.6 Bases 130 5

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E 3.7 and 4.7 Containment Systems 139 A.

Primary Containment 139 B.

Standby Cas Treatment System 148 C.

Secondary Containment 150 D.

Primary Containment Isolation Valves 151 3.7 Bases 156 4.7 Bases 161 3.8 and 4.8 Radioactive Effluents 168 A.

Airborne Effluents 168 B.

Mechanical Condenser Vacuum Pump 170B C.

Lioald Effluents 171 IO D.

Radtaactive Liquid Storage 173 E.

Augmented Off-Gas System 173 F.

Environmental Monitoring Program 173A 3.8 and 4.8 Bases 177 3.9 and 4.9 Auxiliary Electrical Systems 180 A.

Operational Requirements for StartuP 180 B.

Operational Requirements for Continued Operation 181 i

1.

Offsite Power (Line) 181 2.

Offsite Power (Transfomers) 182 iv

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Standby Diesel Generators 182 4.

Station Battery Systems 183 3.9 Bases 185 4.9 Bases 186 3.10 and 4.10 Refueling 187 A.

Refueling Interlocks 187 B.

Core Monitoring 188 C.

Fuel Storage Pool Water Level 188 D.

Movement of Fuel 188 E.

Extended Core and Control Rod Drive Maintenance 188A 3.10 and 4,10 Bases 189 5.0 DESIGN FEATURES 190 18 6.0 ADMINISTRATIVE CONTROLS 192 6.1 Organization 192 6.2 Review and Audit 195 6.3 Actions to be taken in the Event of an Abnormal Occurrence 201 in Plant Operation 6.4 Action to be taken if a Safety Limit is Exceeded 201 6.5 Plant Operating Procedures 202 6.6 Plant Operating Records 2(9 6.7 Plant Reporting Requirements 211 v

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f I-LIST OF FIGURES

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2.1.1 Fuel Cladding Integrity Safety Limit 10 2 3.1 APRM Flow Hererenced Scram and IOd Block Trip Settings 11 l

232 Relationship Between Peak Heat Flux ard Power for Peaki'ng Factor of 3.08 la 4.1.1

'M' Factor - Graphical Aid in the Selection of an l

Adequate Interval Between Tests 46 4.2.1 System Unavailability 74 3.4.1 Sodium Pentadorate Solution Volume - Concentration Requirements 92 3.4.2 Sodium Pentaborate Solution Temperature Requicements 93 4.6.1 Minimum Reactor Pressurization Temperature

- 122 4.6.2 Chloride Stress Corrosion Test Results @ SOOP 123 18 6.1.1 NSP Corporate Organizational Relationship to On-site Operating Organization 193 6.1.2 Functional Organization for On-site Operating Group 194 i

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3.1.1 Reactor Protection System (Scram) Instrument Requirements 30 4.1.1 Scram Inetrument Functional Tests - Minimum Functional Test Frequencies 34 for Safety Instrumentation and Control Circuit; 4.1.2 Scram Instrument Calibration - Minimum Calibration Frequencies for 36 Reactor Protection Instrument Channels 3.2.1 Instrumentation that Initiates Primary Containment Isolation Functions 50 3.2.2 Instrumentation that Initiates Emergency Core Cooling 53' Systems L

3.2.3 Instrumentation that Initiates Rod Block 57 3.2.4 Instrumentation that Initiates Reactor Building Ventilation 60 Isolation and Standby Cas Treatment System Initiation 18 f

3.2.5 Trip Functions and Deviations 69 4.2.1 Minimum Test and Calibration Frequency for Core Cooling, Rod Block 61 and Isolation Instrumentation 4.6.1 In-Service Inspection Requirements for Monticello 124 3.7.1 Primary Containment Isolation 153 4.8.1 Sample Collection and Analysis Manticello Nuclear Plant - Environmental 174 Manitoring Program 4

6.5.1 Protection Factors for Respirators 206 vii wi.

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31 Three AIM 4 instrument channels are provided for each protection trip system. APRM's #1 and 83 operate contacts in one subchannel, and AFRI!'s 8~2 anti 83 operate ccntacts in the other subchannel.

APRM's 84,85, and 86 are arranged similarly in the other protection trip system. Each protection trip system has one mom APRM than is necessary to meet the minimum number required. This allows the bypassing of one APRM per protection trip syste= for maintenance, testing, or calibration.

Ad-t ditional IR!i channels have also been provided to allow for bypassing of one such channel in each trip system.

I The bases for the scram settings for the IR1, AIRM, high reactor pressure, reactor lov vater level, turbine control valve fast closure, and turbine stop valve closure are discussed in Specifications 2 3 and 2.4.

Instrincentation (pressure switches) in the drywell are provided to detect a loss of coolant accident and initiate the emergency core cooling equipment. This instrumentation is a backup to the water

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level instrumentation which is discussed in Specification 3 2.

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The control. rod drive scram system is designed so that all of the water which is discharged from the reactor by the scram can be accommodated in the discharge piping. A part of this piping is an Instru-ment volume which accommo<iates in excess of 32 gallons of water and is the low point in the piping.

No credit was taken for this volume in the design of the discharge piping as concerns the amount of water which must be accommodated during a scram.

During normal operation the discharge volume is empty; however, should it fill with water, the water discharged to the piping from the reactor could nct be acco m tated which wou]d resd t in slow scram times or partial or no control rod insertion.

18 l To preclude this occurence, level switches have been provided in the instrument volume which alarm and scram the reactor when the volume of water in the discharge volume receiver tank reaches 32 gallons.

As indicated utiove, there is sufficient volume in the piping to acconmodate the scram without impair-ment of the scram times or amount of insertion of the control rods. This function shuts the reactor down while sufficient vclume remains to accorrmodate the discharged vuter and precludes the situation

-in which a scram would be required but not be able to perform its function adequately.

Loss of condenser vacuum occurs when the condenser can no longer handle the heat input.

Loss of

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3.1 condenser vacuum initiates a clo:.ure of t he turbine stop valves and turbine bypass valves which eliminates the heat input to the conitenser.

Closure of the turbine stop and bypass valves causes a pressure transient, neutron flux rise, and an increase io surface heat flux.

To prevent the clad safety limit f rom being exceeded if this occurs, a re.ictor scram occurs on turbine stop j

valve closute.

The turbine stop valve closure scram f unc t ion alone is adequate to prevent.the clad safety limit from being exceeded in the event of a turbine trip t ransient without bypass.

i Reference FSAR Section 14.5.1.2.2 and supplerental information submitted February 13, 1973.

The condenser low vacuum scram is a back-up to t:e stop valve closure scram and causes a scram before the stop valves are closed and thus the rcsalting t ransient is less severe.

Scram occurs 1

at 2 3" IIg vacuum, stop valve closure occurs at 20" Ilg vacuum, and bypass closure at 7" lig vacuum.

High radiation levels in the main steamline tunnel above that due to the normal nitrogen and j

oxygen' radioactivity is an indication of leaking fuel.

A scram is initiated whenever such radiation level exceeds ten times normal full power background. The purpose of this scram is to i

reduce the source of such radiation to the ext.ent necessary to prevent excessive. release of 18l radioactive materials.

The main steamline isolation valve closure scram is set to scram when the isolation valves are 610% closed from full open.

This scram anticipates the pressure and flux transient, which would transient is insignificant.

occur when the valves close. By scramming at this setting the resultant ~

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Reference Section 14.5.1.3.1 FSAR and supplemental infc rmation submitted February 13, A reactor mode switch is provided which actuates or bypasses the various scram functions appropriate to the particular plant operating status.

Reference Section 7.7.1 FSAR.

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i The manual scram function is active in all modes, thus providing for a manual means of rapidly inserting control rods during all modes of reactor operation.

,The IRM systea provides protection against excessive power levels and short reactor periods in the 3.1 BASES 39

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31 start-up and intermed5 ate power ranges.

Ref. Section 7.h.4 FSAR. A source range monitor (SRM) system is also provided to supply additional neutrcn level infor1ation during start-up but has no scram functions.

Ref. Section 7.h.3 FSAR. Thus, the I!M is required in the " Refuel" and "Startup" modes.

In the power range the APRM system provides required protection.

Ref. Section 7.4.5 2 FSAR. Thus, the IhM system is not required in the "Run " mode. The AFRM's cover only the power range, the IRM's provide adequate coverage in the start-up and intermediate range, and there-fore, the APRM's are not required for the " Refuel" or "Startup" modes.

1 The high reactor pressure, high drywell pressure, and reactor low water level scrams are required for all modes of plant operatiou unleas the reactor is suberitical and depressurized.

They are, therefore, required to be operational for all modes of reactor operation except in the " Refuel" mode with the reactor suberitical and reactor temperature less than 212 F as allowed by Note 3.

18 The scram discharge volume high level trip function is required for all modes with the exception that it may be bypassed in the " Refuel Mode" under the provisions of Table 3.1.1, allowable by-pass condition (a).

In order to reset the safety system after a scram condition, it is necessary to drain the scram discharge volume to clear this scram input condition. This condition usually follows any scram, no matter what the initial cause might have been. Since all of the control rods are completely inserted following a scram it is permissible to bypass this condition because a control rod block prevents withdrawal as long as the switch is in the bypass condition for this function.

To permit plant operation to generate adequate steam and pressure to establish turbine seals and condenser vacuum at relatively low reactor power, the main condenser vacuum trip is bypassed until e

600 psig. This bypass also applies to the main steam isolation valves for the same reason.

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Section 7.7.1.2 FSAR.

An automatic bypass of the turbine control valve fast closure scram'and turbine stop valve closure sc ram is effective whenever the turbine first stage pressure is less than 307. of its rated value.

This insures that reactor thermal power is less than 457. of its rated value.

Closure of.these valves from such a low initial power Icvel does not constitute a threat to the integrity of any barrier to the release of radioactive material.

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4.1 A. The minimum functional testing frequency used in this specification is based cn a reliability analysis using the concepts developed in reference (1). This concept was specifically adopted to the one out of two taken twice logic of the reactor protection system fcr Monticello. The analysis shows that the sensors are primarily responsible for the reliability of the reactor protection system. This analysis makes use of unsure failure rate experience at conventional and nuclear power plants in a relintility model for the system. An " unsafe failure" is defined as one which negates channel operability, and which, due to its nature, is revealed only when the channel is functionally tested or attempts to respond to a real signal. Failures such as blown funes, ruptured bourdon tutes, faulted amplifiers, faulted cables, etc., which result in

" upscale" or "downscale" readings on tre reactor instrumentation are " safe" and will be easily-recognized by the operators during operation because they are revealed by an alam or a scram.

The 13 scrum sensor channels listed in Table 4.1.1 are divided into three groups (A.,

E., and C.)

and are defined on Table 4.1.1.

The sensors that make up group (A) are specifically selected from among the whole family of I

industrial on-off sensors that have earned an excellent reputation for reliable operation.

Actual history on this class of sensors operating in nuclear power plants shows four failures e

18l in 472 sensor years, or a failure rate of 0.97 X 10-6/hr.

During ' design, a goal of 0.99999 probability of success (a t the 507. confidence level) was adopted to assure that a balanced and adequate design is achieved.

The probability of success is primarily a function of the sensor failure rate and the test in te rva l. A three-month test interval was planned for group.

(A) sensors.

This is in keeping with good operating practice, and satisfies the design goal for the logic configuration utilized in the Reactor Protection System.

To satisfy the long-term objective of maintaining an adequate level of safety throughout the plant lifetime, a minimum goal of 0 9999 at the 95% confidence level is proposed. With the (l..out of 2)

X (2) logic, this requires that each sensor have an availability of 0 993 at the 95% confidence 4.1 BASES 43 y,

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- hour monitoring interval. for' the analog devices, as assumed above, and a weekly test intervall for the bi-stable trip circuits, the design reliability goal of 0 99999 is attained with ample j

margin. The test frequency of once per week has developed principally-on' the basis. of past practice.-

and good judgment,'and nothing has developed to indicate that the frequency should change.

I Group' (C) devices are active only during a given portion of the ' operational' cycle.. For example, j

the.IRM is active. during startup and inactive during full-power operation. Thus, the' only ' test l i

l that is meaningful is the one: performed just prior to shutdown'or startup; i.e.,

the tests that i

are perfomed just prior to use of the instrument.

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Calibration frequency of the instrument channel is divided into two groups as.. defined on Table l

4.1.2.

Experience with passive type' instruments indi. cates that a yearly' calibration'is adequate.. For l

those devices which employ amplific rs,etc., drift specifications call for drift to be less' than-

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0 5 1/ month; i.e., in the period of a month 'a drif t of 0 5% would occur and thus provide for adequate margin. For the APRM system drift of electronic apparatus is not the only-consideration j

in determining a calibration frequency. Change 'in power' distribution and loss of chamber sensitivity dictate a calibration every three days. Calibration en this frequency assures plant operation at. '

l or below thermal limits.

4 B.

The peak heat flux shall be checked once per day to determine if the APRM ocram requires adjust -

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ment. This will nomally be done by checking the LPRM readings. Only a small' number of control..

l rods are moved daily and thus the peaking factors are not expected to change 'significantly,.thus j.

a daily check of the peak heat. flux is adequate.

(1) Reliability of Engineered Safety Features as a Function of Testing Frequency. I. M Jacobs,-

Nuclear Safety, Vol. 9, No. 4, l July-Aug.1968, pgs. 310-312.

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Notes:

i Upon ' discovery that minimum re' uirements for the number of operable or operating trip -systems.

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or. instrument channels are not satisfied actions shall be initiated to:

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

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(b) Place the plant under the specified required conditions using nomal' operating procedures.-

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(7) here must be a total of at least 4 operable or operating APIM channels.

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1. Inequired conditions when minimum conditions for operation are not satisfied.

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l A.

Reactor in Shutdown mode.

B.

No rod withdrawals pe mitted while in Refuel or Startup mode.

C.

Reactor in k n mode.

t D.

No rod withdrawals pemitted while in.the Run mode.

E.

Pcwer on IIM range or below and reactor in Startup, Refuel, or Shutdown mode.

• • Allowable Dypass Conditions

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SIM Detector-not-fully-inserted rod block may be bypassed when the: SIM channel ' count rate is a.

l 100 cps.or when -all IIM range switches are above Ibsition 2.

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IIM Downscale rod block may be bypassed when the IIM range switch is-in' the lowest range. position.

RBM Upscale and Downscale rod blocks may be bypassed below 30% rated power.

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18 d.

SRM Upacale block may be bypassed when associated IRM range switches are above Position 6..

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Rod Block and Isolation Instrumentation Ins t rument Channel Test (3)

Calibration (3)

Senser Check (3) 3.

Steam Line low Pressure Note 1 Once/3 months None 4.

Ste zm Line liigh Radiation Once/ week (5)

Note 6 Once/ shift HPCI ISO!ATION 1.

Steam Line High Flow Note 1 Once/3 months Nome 2.

Steam Line High Temperature Note 1 Once/3 months None RCIC iSCIATION 1.

Steam Line liigh Flow Note 1 Once /3 months None l 33 2.

Steam Line fligh Temperature Note 1 Once/3 months None HEACTOR BLIIIDING VENTIIATION 1.

Radiation Monitors (Plenum)

Note 1 Once/3 months Once/shif:

2.

Radiation Monitors (Refueling Floor)

Note 1 Once/3 months (4)

OFF-CAS ISOLATION 1.

Radiation Monitors (Air Ejectors)

Notes (1,5)

Nota 6 Once/ shift NOTES:

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

62 32/4.2 m

g y

q w

m' 9

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1-e-1M 4

i Table 4.2.1 - Continued NOTES:

[

(2) Calibrate prior to normal shutdown and start-up and thereaf ter check once per shif t and test once per week until no longer required. Calibration of this instrument prior to normal shutdown means ad.lustment of channel trips so that they correspond, within acceptable range and accuracy, to a 18 simulated signal injected into the instrument (not primary sensor).

In addition, IRM gain adjustment will be perfomed, as necessary, in the APR l/ Iici overlap region.

(3) Functional tests, calibrations and sensor checks are not required when the systems are not required to be i

operable or are tripped. If tests are missed, they shall be performed prior to returning the systems to an operable status.

(4) L'henever fuel handling is in process, a sensor check shall be performed once per shift.

(5) A Functional test of this instrument means the injection of a siculated signal into the instrument (not primary sensor) tn verify the proper instrument channel response alarm and/or initiating action.

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

63 3.2/4.2 we

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v' ' - - -'-

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y

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVETIJ.ANCE REQUIREMENTS (iv)

The rod block function of the rod worth minimizer shall be verified by attempting to with-draw an out-of-sequence control rod beyond the block point.

(b) Whenever the reactor is in the (b)

If the rod worth minimizer is. inoperable startup or run mode below 10%

while the reactor is in the startup or rated thermal power, no control run mode below 10% rated thermal power rods shall be moved unless the and the second independent operator rod worth minimizer is operable or engineer is being used, he shall or a second independent operator verify that all rod positions are or engineer verifies that the correct prior to cocunencing withdrawal or insertion of each rod group, operator at the reactor console jg is following the control rod I f' program. The second operator may be used as a substitute for an inoperable rod worth minimizer during a startup only if the rod worth minimizer fails af ter with-d rawal o f a t least twelve control rods.

~

4.

Control rods shall not be withdrawn 4.

Prior' to control rod withdrawal for for startup or refueling unless at startup or during refueling verify least two source range channels have that at least two source range an observed count rate equal to or channels have an observed count rate greater than three counts per second, of at least three counts per second.

5.

Whenever the Engineer, Nuclear, deter-5.

Whenever the Engineer, Nuclear, deter-mines that a limiting control rod mines that a limiting control rod pattern pattern exists, withdrawal of desig-exists, an instrument functional test nated control rods shall be permitted of the RWM shall be performed prior to only when the RWM system is operable, withdrawal'of the designated rod (s) and daily thereafter.

3,3/4.~3 78 6

Bases Continued 3.3 and 4.3:

This is adequate and conservative when compared with the typical time delay of about 210 milli-seconds estimated f rom scram test results. Approximately the first 90 milliseconds of the time 33 l Interval results f rom the sensor and circuit delays; at this point the pilot scram solenoid is deenergized. Approximately 120 milliseconds later control rod motion is estimated to begin.

IIowever, to be conservative, control rod motion is not assumed to start until 200 milliseconds later. This value was included in the transient analyses and is included in the allowable scram insertion times of Specifications 3.3.C.1 and 3.3.C.2.

3.4/4.3 BASES 85A

j i

r 4

i 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIRD4DITS b

s systems and pump demineralized -

water into-the reactor vessel.

This' test checks explosion of I

the charge associated with the tested system, proper operation of the valves and pump capacity.

Both systems shall be tested-and inspected, including each explosion valve in ~ the course of two operating cycles, i

b.

Explode one of two primer assemblies manufactured in the same batch to verify proper' function.

Then install, as a 18 replacement, the second primer assembly in the' explosion valve of the ' system tested for operation.

c.

Test that the. setting of the system pressure relief valves is between'1350 and.1450 psig.

.B.

Operation with Inoperable Components B.

Surveillance with' Inoperable Components j

From and af ter the date that a, redun-When a component becomes. inoperable, its j

dant component is made or found to be redundant component shall be demonstrated to be operable immediately and daily thereafter.

4 3.4/4.4 '

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Bases 3.4-and 4.4:

A.

The design objective of the standby liquid control system is to provide the capability of bringing the reactor from full power to a cold, xenon-free shutdown assuming that none of the withdrawn con-trol rods can be inserted.

To meet this objective, the liquid control system is designed to inject a quantity of boron which produces a concentration of 900 ppm of boron in the reactor core in less than 125 minutes.

900 ppm boron concentration in the reactor core is required to bring the reactor from full j

power to a 3*/. A k subcritical condition considering the hot to cold reactivity swing, xenonfpoisoning and an additional 257. boron concentration margin for possible imperfect mixing of-the chemical solution 18 in the reactor water and. dilution from the water in the cooldown circuit. A minimum, net quantity of 1400 gallons of solution having a 21.47. sodium pentaborate concentration is required to meet this shut-down requirement.

4 The time requirement (125 minutes) for insertion of.the baron solution was selected to override the rate of reactivity insertion due to cooldown of the reactor following the xenon poison peak.

The maximum net storage volume of the boron solution is 2895 gallons.

(250 gallons are contained below the pump unct ion and, t he re fore, have not been uned in the net qinntities above. )

Boron concentration, solution temperature, and volume (including check of tank Aeater and pipe l

heat tracing system) are checked on a frequency to assure a high reliability of operation of the -

system should it ever be required.

Experience with pump operability demonstrates that testing at a three-month interval is adequate to detect if failures have occurred.

The only, pract.ical time to test the standby liquid control system is during a refueling outage and by initiation from local stations. Components of the system are checked periodically as described above and make a functional test of the entire system on a frequency of less than once each refuel-ing outage unnecessary. A test of explosive charges from one manufacturing batch is made to assure that the replacement charges for the tested system are satisfactory. A continual check of the firing circuit continuity is provideti by pilot lights in the control room.

The relief valves in the standby liquid control system protect the system piping and positive dis-placement pumps which are nominally designed for 1500 psi from overpressure.- The pressure relief valves discharge back to the standby liquid control solution tank.

3.4/4.4 BASES 94 t

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30 LIMITING CONDITIOriS FOR OPEl<ATIOII 4.0 SURVEILIAIICE HEQUIREMENTS 3

To be considered operable, the IIPCI i

system shall meet the following conditions:

a.

The IIPCI shall be capable of delivering 3,000 gpm into the reactor vessel for the reactor pressure range of 1120 psig 18 to 150 psig.

b.

The condensate storage tanks shall contain at least 75,000 gallons of condensate water.

c.

The controls for automatic transfer of the !!PCI pump suction from the condensate storage tank to the suppression chamber shall be operable.

4.

If the requirements of 3.5.D.1-2 cannot be met, an orderly reactor chutdown shall be initiated immediately 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.

E.

Automatic Pressure Relief System E.

Surveillance of the Automatic Pressure Relief System shall be performed as follows:

104 3.3/4.5

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4

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l 30 LIMITING CONDITIO!E FOR OPERATION 4.0 SURVEILIANCE. REQUIREMENTS l

P

.j 3

When irradiated fuel is in the reactor t

vessel and reactor coolant temperature is less than 212 F, all low pressure core

^

and containment cooling subsystems may be inoperable provided no work is being done i

which has the potential for draining the p

i reactor vessel except as allowed by specification 3 5.G.4 below.

i 4.

When irradiated fuel is in the reactor I

vessel and the vessel head is removed, the suppression chamber may be drained

+

i completely and no more than one control rod drive housing or instrument thimble 18l opened at any one time provided that the spent fuel pool gates are open and i-the fuel pool water level is maintained

.at a level of greater than or equal to 33 feet.

H.

Deleted F

t i

E

[

108 i

1

.35/4~.5 j

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILI.ANCE REQUIREMENTS r

B.

Pressurization Temperature B.

Pressurization Temperature 1.

The reactor vessel shall be vented and 1.

Reactor vessel shell temperature and power operation shall not be conducted reactor coolant pressure shall be unless the reactor vessel temperature is permanently recorded whenever the shell equal to or greater than that shown in temperature is below 220*F and the reactor vessel is not vented.

Figure 4.6.1.

2.

The reactor vessel head bolting studs shall 2.

When the reactor vessel head bolting studs not be under tension unless the tempera-are tightened or loosened, the reactor ture of the vessel head flange and the vessel head flange and head temperature head are > 70*F.

shall be permanently recorded.

3.

A neutron flux dosimeter and material samples shall be Installed in the reactor vessel adjacent to the vessel wall at the core midplane level. The material sample program shall conform to ASTM E 185-66. The neutron flux dosimeter shall be removed during the first refueling outage and tested to verify or adjust the calculated values of neutron fluence used to determine the vessel NDTT (Nil Ductility Transition Temperature) from Fig. 4.6.1.

A material surveillance specimen sample shall be removed when the estimated change in NDT temperature is 50*F, but priorg to tge time when the estimated fluence is 10 n/cm, or at the 15 years operational interval whichever first occurs.

C.

Coolant Chemistry C.

Coolant Chemistry 1.

(a) A sample of reactor coolant shall be 1.

The steady stat e radiciodine concentra-18 tion in the reactor coolant shall not taken at least every 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> and exceed 5 microcuries of I-131 dose 18 equiv-l ent per gram of water.

116 3.6/4.6 y

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i 3.0 LIMITINC CONDITIONS FOR OPERATION 4.0 SURVEILI ANCE REQUIREMENTS f

i analyzed for radioactive iodines of l

I-131 through I-135 during power

,{

operation.

In addition, where steam i

jet air ejector monitors indicate an increase in radioactive gaseous effluents 18 of 25 percent or 5000 pCi/sec, whichever i

i I

is greater,.during steady state reactor operation a reactor coolant sample shall i

l be taken and analyzed for radioactive 1

fodines.

1 I

(b)

Isotopic analysis of reactor coolant samples shall be made at least once per month.

(c) Whenever the steady state radioiodine r

concentration of prior operation is greater than I percent but less than ac percent of Specification 3.6.C.1, i

l

- uample of reactor coolant shall be l

taken within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of any. reactor l

L startup and analyzed for radioactive-i' iodines of 1-131 through I-135.

18 (d) Whenever the steady state radioiodine concentration of prior operation is l

greater than 10 percent of Section 3.6.C.1, a sample of reactor coolant shall be taken prior to any reactor l

startup and analyzed for radioactive fodines of I-131 through I-135 as

]

well as the coolant sample and analyses required by Specification 4.6.C.I.(c) s above.

i 4

117

)

3.6/4.6

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i I

f 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIRDiENTS i

2.

(a) The reactor coolant wat.er shall not 2.

During startup and at steaming rates exceed the following limits with below 100,000 pounds per hour, a' sample l

steaming rates less than 1b0,000 of reactor coolant shall be taken every pounds per hour except as specified four hours and analyzed for-conductivity in 3.6.C.2.b.

and chloride content.

18 Conductivity 5 paho/cm Chloride ion 0.1 ppm

{

(b) For reactor startups the maximum value for conductivity shall not exceed 10 pmho/cm and the maximum value for 4

chloride ion concentration shall not exceed 0.1 ppm for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after placing the reactor in the power operating condition.

3.

Except as specified in 3.6.C.2.b above, the 3.

(a) With steaming rates greater thar.

reactor coolant water shall not exceed the or equal to 100,000 lbs, per hcur, a following limits with steaming rates greater reactor coolant sample shall be taken than or equal to 100,000 lbs. per hour.

at least every 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> and when thJ 3

4 continuous conductivity monitors in-j Conductivity 5 paho/cm dicate abnormal conductivity (other 18 Chloride ion 0.5 ppm chan short-term spikes) and analyzed for conductivity and chloride ion 4.

If Specifications 3.6.C 1 through 3.6.C.3 content.

are not met, an orderly shutdown shall be 18 initiated and the reactor shall be in the (b) When the continuous conductivity cold shutdawn condition within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

monitor is inoperable, during pcwer operation, a reactor coolant sample j

18 should be taken at least once per shift and analyzed. for conductivity I

and chloride ion content.

}

l 3.6/4.6 118 1

)

lj 1

4 i

i 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS D.

Coolant Leakage D.

Coolant Leakage j

ll Any time irradiated fuel is in the reactor vessel, Reactor coolant system leakage intu the and reactor coolant temperature is above 212 F, drywell shall be checked and recorded reactor coolant leakage into the primary contain-at least once per day.

ment from unidentified sottrees shall not exceed 5 gpm.

In addition, the total reactor coolant system leakage into the primary containment shall not exceed 25 gpa.

If these conditions cannot be met, initiate an orderly shutdown and have the j

reactor placed in the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

il i

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118A 3.6/4.6 e

Bases Continued 3.6 and 4.6 C.

Coolant Chemistry A steady state radiciodine concentration limit of 5 pCi of I-131 dose equivalent per gram of water in the reactor coolant system can be reached if the gross radioactivity in the gaseous effluents are near their limits or there is a failure or prolonged shutdown of the cleanup demineralizer.

In the event of a steam line rupture outside the drywell, the NRC staf f calculations show the resultant radiological dose at the nearest site boundary (465 m) to be less than 30 Rem to the thyroid. This dose was calculated on the basis of the radiciodine concentration limit of 5 pCi of I-131 dose equivalent per gram of water, atmospheric diffusion from an equivalent elevated release of 30 meters under fumigation conditions for Pasquill type F 1 meter /see wind speed and a steam line isolation valve closure time of five seconds with a steam / water mass release of 36,000 pounds.

The reactor coolant sample will be used to assure that the limit of Specification 3.6.C.1 is not exceeded.

The radioindine concent ration would not be expected to change rapidly during steady state operation over a

[

period of 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />.

In addition, the trend of the radioactive gaseous effluents, which is continuously monitored, is a good indicator of the trend of the radiofodine concentration in the reactor coolant. When a significant increase in radioactive gaseous effluents is indicated, as specified, an additional reactor coolant sample shall be tak(n and analyzed for radioactive iodine.

Whenever an isotopic analysis is performed, a reasonable effort will be made to determine a significant percentage of those contributors representing the total radioactivity in the reactor coolant sample.

Usually at least 80 percent of the total gauma radioactivity can be identified by the isotopic analysis.

It has been observed that radiolodine concentration can change rapidly in the reactor coolant during transient reactor operraions such as reactor shutdown, reactor power changes, and reactor startup if failed fuel is j

present. !w viscified, additional reactor coolant sample: shall be taken and analyzed for reactor operations in which st m:

tate radioiodine concentrations in the reactor coolant indicate various levels of iodine releases frue. e fuel. Since the radiolodine concentration in the reactor coolant is not continuously l

measured, reactor coolant sampling would be.. effective as a means to rapidly detect gross fuel element failures. However, some capability to detect gross fuel element failures is inherent in the radiation monitors in the off-gas system and on the main steam line.

i Materials in the primary system are primarily 304 stainless steel and zircaloy. The reactor water chemistry i

limits are established to prevent damage to these materials. The limit placed on chloride ccncentration is to prevent stress corrosion cracking of the stainless steel.

132 3.6/4.6

..4

~.,.

x

Bases Continued 3.6 and 4.6:

When conductivity is in its proper normal range (approximately 10 kmho/cm during reactor startup and Spaho/cm during power operation), pH and chloride and other impurities affecting conductivity must also be within their normal range. When and if conductivity becomes abnormal, then chloride measurements are made to determine whether or not they are also out of their normal operating values. This would not necessarily be the case.

Conductivity could be high due to the presence of a neutral salt, e.g., Na2SO, which would 4

not have an effect on pH or chloride.

In such a case, high conductivity alone is not a cause for shutdown.

j In some types of water-cooled reactors, conductivities are in fact high due to purposeful addition of additives.

In the case of EWRs however, no additives are used and where neutral pH is maintained, I

conductivity provides a very good measure of the quality of the reactor water.

Significant changes therein provide the operator with a warning mechanism so he can investigate and remedy the condition causing the change before limiting conditions, with respect to variables affecting the boundaries of the reactor coolant, are exceeded. Methods available to the operator for correcting the off-standard condition include operation of the reactor cleanup system reducing the' input of impurities and placing the reactor in the cold shutdown condition. The major benefit of cold shutdown is to reduce the temperature dependent corrosion rates and provide time for the cleanup system to reestablish the purity of the rea7 tor coolant.

During startup periods, which are in the category of less than 100,000 pounds per hour, conductivity may exceed 5pmho/cm because of the initial evolution of gases and the initial addition of dicsolved metals.

During this period of time when the conducrivity exceeds Sumho (other than shure tum spikes), samples will be taken to assure the chloride concentration is less than 0.1 ppm.

The conductivity of the reactor coolant.is continuously monitored. The samples of the coolant which are taken every 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> will serve as a reference for calibration of these monitors and is considered adequate to assure accurate readings of the monitors.

If conductivity is within its normal range, chlorides and other impurities will also be within their normal ranges. The reactor coolant samples will also be used to determine the chlorides. Therefore, the sampling frequency is considered adequate to detect long-term changes in the chloride ion content.

Isotopic analyses required by Specification 4.6.C.1(b) may be performed by a gamma scan and gross beta and alpha determination.

3.6/4.6 133

4.0 SURVEILLANCE REQUIREMENTS 3.0 LIMITING CONDITIONS FOR OPERATION b.

When the position of any drywell-d.

One position alarm circuit can be inoperable providing that the redundant position alarm suppression chamber vacuum breaker valve circuit is operable. Both position alarm is indicated to be riot-fully closed at a time when such closure is required, the circuits may be inoperable for a period not to exceed seven days provided that all vccuum drywell to suppression chamber differential breakers are operable, pressure decay shall be demonstrated to be less than that:shown on Figure 3.7.1 immediately and following any evidence of subsequent operation of the inoperable valve.until the. inoperable valve is restored to a normal condition.

When both position alarm circuits are made c.

or found to be inoperable, the control panel indicator light status shall be recorded daily to detect changes in the vacuum breaker position.

5.

Oxygen Concentration 5.

Oxygen concentration Whenever inerting is required, the primary a.

After coczpletion of the startup test containment oxygen concentration shall be program and demonstration of plant measured and recorded on a weekly basis, electrical output, the primary ec,uts-ment atmosphere shall be reduced to less than 5% oxygen with nitrogen gas whenever the reactor is in the run 18 mode, except as specified in 3.7.A.S.b.

b.

Within the 24-hour period subsequent to placing the reactor in the run mode following shutdown, the containment atmosphere oxygen concentration shall be reduced to less than 5% by weight, and maintained in this condition. Deinerting may consnence 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to leaving the run mode for a

~

reactor shutdown.

147A

~

~

i 3.0 1.IMITING CONDITIONS FOR OPERATION 4.0 SURVEIID.HCE REQUIPDTEPS 6.

If the specifications of 3.7. A cannot be r.et, the reactor shall be placed in a cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

1 B.

Standby Cas Treatment Syster.

B.

Standby Cas Treatment System 1.

Except as specified in 3.7.B.3 below, both circuits of the standby gas treat-E**

"Y**""""

cent system shall be operable at all shall be performed as indicated below:

tic.es when secondary containment integrity is required.

a.

At Ic3St nce Per operating cycle it shall be demonstrated that:

4 (1) Pressure drop across the combined high-efficiency and charcoal filters is less than 7.0 inches of water, and i

(2) Inlet heater output is at least 15 kv.

b.

18 Within 30 days of the beginning of each refueling outade, whenever a filter is changed whenever :=rk is performed that could affect filter systems efficiency, and at intervals not to exceed six months between refueling outages, it shall be demonstrated that:

(1) The removal efficiency of the installed i

particulate filters for particles having a mean diameter of 0.7 microns shall be 4

~

3.7/k.7 148 1

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a 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLCCE REQUIREMENTS i

]

C.

Secondary Containment C.

Secondary Containment 1

l 1.

Secondary containment integrity shall be 1.

Secondary containment surveillance shall j

maintained during all nodes of plant be performed as indicated below:

l j

operation except when all of the following conditions are met.

i j

a.

The reactor is subcritical and Specifi-a.

Secondary containment capability to cation 3.3.A is met, maintain at least a 1/4 inch of water vacuum under calm wind (2 < u < 5 r ph) i conditions with a filter train flow

+

rate of 64,000 scfm, shall be dem-4 onstrated at each refueling outage prior to refueling. This surveillance

'I testing should be reported in the semiannw*l operating reports.

l

{

b.

The reactor water temperature is below

[

j 212 and the reactor coolant system is vented.

i f

i i

i c.

No activity is being performed which can reduce the shutdown margin below that specified in Specification 3.3.A.

4 I

150 3.7/4.7 1.

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3.0 LIMITING COI;DITIONS FOR OPERATICII k.O SURVEILLANCE REQUIREMHITS d.

The fuel cask or irradiated fuel is 4

not being moved within the reactor t

building.

Primary Contairment Isolation Valves D.

Primary Containment Isolation Valves D.

1.

Whenever the reactor is in the run mode, 1.

The primary containment isolation valves all isolation valves listed in T ble 3.7.1 surveillance shall be performed as follows:

and all primary system instrument line q

flow check valves shall be operable except a.

At least once per operatind cycle the as specified in 3.7.D.2.

operable isolation valves that are power operated and automatically initiated shall be tested for simulated

.[

automatic initiation and closure times.

j b.

At least once per operating cycle the primary system instrument line flow l

check valves shall be tested for proper f

operation.

i i

c.

At least once per quarter (1) All normally open power-operated i

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

s 151 37/4.7 I

O l

j r

i 3.0 LII".ITIliG CONDITIOUO FOR CPERATIQU 4.0 SURVEILLANCE RDQUIRDG2TfS 1

+

c.

At least once per quarter-Continued (2) With the reactor power less than 75% of rated, trip main steam isolation valves (one at a time) and verify closure time.

d.

At least once per week the main steam-line power-operated isolation valves shall be exercised by partial closure and subsequent reopening.

2.

In the esent any isolation valve specified 2.

Whenever an isolation valve listed in in Table 3.7.1 becomes inope rab le, reactor Table 3.7.1 is inoperable, the position of 18 operation in the run mode may continue at least one fully closed valve.in each line provided at least one valve in each line having an inoperable valve shall be recorded having an inoperable valve is closed.

daily.

18l 3.

If Specification 3.7.D 1 and 3.7.D.2 cannot I

be met, initiate normal orderly shutdown i

and have reactor in the cold shutdown l

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

t 4

i h7/4.7 152 l

e i

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 6.

The release rates of radioactive 3.

Gaseous release of tritium shall be

{

particulates with half-lives greater calculated on a quarterly basis from than 8 days shall not exceed 8 percent tritium concentration of the condensate.

]

of the limit in Specification 3.8.A.5 Vaporous tritium shall be calculated j

averaged over any calendar quarter.

from a representative sample. The i

sum of these two values shall be re-7.

If the maximum release rate limits of Spec-ported as the total tritium release.

ifications 3.8.A.1, 3.8.A.3, or 3.8.A.5 are 1

not met following a routine surveillance 4.

Radiciodine and radioactive particulates witt.

check, an orderly shutdown shall be half lives greater than 8 days released from initiated and the reactor shall be in the the off-gas stack and reactor building vent cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, shall be continuously sampled. Station record.

of release of all radioiodine 131 and partic-8.

If the limits of Specification 3.8.A.2, ulates with half lives greater than 8 days j

3.8.A.4, or 3.8.A.6 are exceeded, shall be maintained on tie basis of all stack j

appropriate corrective action such as an and vent cartridges counted.

The charcoal orderly reduction of power shall be cartridges shall be counted weekly when the r

initiated to bring the releases within measured release rate of radioiodine 131 l

these limits.

activity is less than the limit of Specifi-cation 3.8. A.4; otherwise stack cartridges 9.

If the release rates exceed four percent -

shall be counted daily if the stack I-131 of the limits in Specification 3.8.A.1 contribution exceeds 50% of the Ihnit of averaged over any calendar quarter or. two Specification 3.8. A.4 ' and vent cartridges percent of the limits in Specifications 18 shall be counted daily if the vent 1-131 gg l 3.8.A.3 or J.8.A.5 averaged over any contribution exceeds 50% of Specification r

calendar quarter, the following actions 3.8.A.4.

The particulate filters shall be shall be taken:

counted weekly when the measured release j

rate of particulate radioactivity with half-lives greater than 8 days is less than the 3

limit of Specification 3.8 A.6; otherwise stack filters shall be counted daily if the stack particulate contribution exceeds 50%

of the limit of Specification 3.8. A.6 and vent filters counted daily if the vent particulate contribution exceeds.30% of the ihmit of Specification 3.8. A.6 170

'O 3.0 LIMITING CONDITIONS FOR OPHIATION 4.0 SURVEILIJd!CE REQUIBD4EITIS c.

Isotopic analyses including detemination of tritium of representative batches of liquid effluent shall be performed and recorded at least once per quar-ter.

Each batch of effluent released shall be counted for gross alpha and beta activity and the results recorded. At least once per month a gar:=a scan of representative batches of effluent shall be perfarned and recorded to detemine the c:c.::n energy I>eaks of these batches. If energy peaks other than those determined by the previous isotopic analyses are found, a new set of isotopic d

analyses shall be performed and record ed.

d.

Grab samples shall be taken from the discharge canal monthly and analyzed for tritium and significant isotopes.

e.

Deleted.

yg 172 3.8/4.8

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1 i

I i

3.0 IlMITINO CONDITION FCR OPERATION k.O SURVEIIlld3CE REQUIRD4H;TS

?

1 l

2.

Both diesel generators are operable and capable of feedirg their designated 4160 j

volt buses.

I 3

A second source of off-site power ( reserve transformer lAR) is fully operational ar.1 energized to carry power to the plant kl60V ac buses.

i 4

(a) kl60V Buses #15 and 816 are enerd zed.

i (b) L80V Load Centers E103 and flok are l

i enerd zed.

5 All station 24/L8,125, and 29 volt batteries are charged and in service, and associated battery chargers are operable.

B.

k' hen the mode switch is in Run, the avail-ability of electric power shall be as spec; ified in 3.9. A, except as specified in 3.9.B.1, 3.9.B.2, 3.9.B.3 and 3.9.B.4 or the reac'.or f

18 shall be placed in the cold shutdown condition r

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

1.

Transmission Lines From and af ter the date that incoming power is available from only one line, reactor operation is permissible only i

during the succeeding seven days unless an additional line is sooner placed in 181 3 9A.9 F

y m-, -,

r

..---, -, -+.-

4

0 ----- -..

!.. O SURVEILI/J;CE REQUIRD4 BITS 30 I.IMITING CONDITIONS FOR OPULATIGN During each refueling outage, c.

For the diesel generators to be the conditiot.s under which the c.

considered operable, there shall be diesel generators are required a minimum of 26,250 gallens of diesel will be simulated and tests cc.n-fuel (7 days supply for 1 diesel gen-ducted to demonstrate that they erator at full load) in the diesel will start and be ready to accept the oil storage tank.

emergency load within ten seconds.

d.

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

Once a month the quantit ~ of e.

diesel fuel available'st.

be logged.

f.

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

t 4.

Station Battery Systems 4.

Station Battery System Every week the specific gmvity a.

If one of the two 125 V battery systems or the and voltage of the pilot cell 18 250 V battery system is made or found to be and temperature of ad,Jacent cells and overall battery voltage shall inoperable for any reason, an orderly shut-down of the reactor will be initiated and be measured.

the reactor water temperature shall be l

0 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.

t 183 3.9/4.9

.