Text

Amends 125,125 & 126 to Licenses DPR-38,DPR-47 & DPR-55, Respectively,Revising Radiological Effluent Tech Specs for Compliance W/App I of 10CFR50
	ML15113A077
Person / Time
Site:	Oconee
Issue date:	01/16/1984
From:	Stolz J; Office of Nuclear Reactor Regulation
To:	; Duke Power Co
Shared Package
ML15113A078	List: ML15113A077; ML15113A079;
References
	DPR-38-A-125, DPR-47-A-125, DPR-55-A-122, DPR-55-A-126 NUDOCS 8402030045
	Download: ML15113A077 (74)
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o

-UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 DUKE POWER COMPANY DOCKET NO.

50-269 OCONEE NUCLEAR STATION, UNIT NO.

1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 125 License No. DPR-38

1. The Nuclear Regulatory Commission (the Commission) has found that:

The application for amendment by Duke Power Company (the licensee)

A. dated February 9, 1983, as.supplemented February 28, 1983, and April 28, 1983, 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 fdclity will operate in conformity with the application, the pro visions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satis fied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in.the attachment to this license amendment and paragraph 3.B of Facility Operating License No.

DPR-3 8 is hereby amended to read as follows:

3.B Technical Specifications The Technical Specifications. contained in Appendices A and B, as revised through Amendment No. 125 are hereby incorporated in the license. The licensee shall operate'the facility in accordance with the Technical Specifications.

8402030045 840116 PDR ADOCK 05000269 P.

PDR

-2

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION fJo n-F. Stol z, Chief 0 erating Reactors nch fivision of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: January 16, 1984

0 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 DUKE POWER COMPANY DOCKET NO.

50-270 OCONEE NUCLEAR STATION, UNIT NO.2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment.No.125 License No. DPR-47

1. The Nuclear Regulatory Commission (the Commission) has found that:

The application for amendment by Duke Power Company (the licensee)

A. dated February 9, 1983, as supplemented February 28, 1983, and April 28, 1983, 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 TO CFR Chapter I; B. The facility will operate in conformity with the application,. the pro visions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satis -____

fied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to-this license amendment and paragraph 3.B of Facility Operating License No. DPR-47 is hereby amended to read as follows:

3.8 Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 125 are hereby incorporated in the license. The licensee shall operate-the facility in accordance with the Technical Specifications.

-2

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION o

F. Stolz, Chief 0

ratina Reactors anch #4 ivision of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: January 16, 1984

t RUNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 DUKE POWER COMPANY DOCKET NO.

50- 287 OCONEE NUCLEAR STATION, UNIT NO.

3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 122 License No. DPR-55

1.

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

The application for amendment by Duke Power Company (the licensee)

A. dated February 9, 1983, as supplemented February 28, 1983., and April 28, 1983, 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 TO CFR Chapter I; B. The facility will operate in conformity with the application, the pro visions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satis-_

fied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 3.8 of Facility Operating License No.

DPR-55 is hereby amended to read as follows:

3.8 Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment. No. 122 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specificatibns.

-2

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION Jo F. Stol z, Chief 0 eratina Reactors B ch #4 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: January 16, 1984

ATTACHMENTS TO LICENSE AMENDMENTS AMENDMENT NO.125 TO DPR-38 AMENDMENT NO.125 TO DPR-47 AMENDMENT NO.122 TO DPR-55 DOCKETS NOS. 50-269, 50-270 AND 50-287 Replace the following pages of the Appendix "A" Technical Specifications with the attached pages.

The revised pages are identified, by amendment numbers and contain vertical lines indicating the area of change.

Remove Pages Insert Paces iii iii iv IV V

V v-a vi vi 1-5.

1-5 1-6 3.5-34 3.5-35 3.5-36 3.5-37 3.5-38 3'.5-39 3.5-40 3.9-1 3.9-1 3.9-2 3.9-2 3.9-3 3.9-3 3.9-4 3.9-5 3.10-1 3.10-1 3.10-2 3.10-2 3.10-3 3.10-3 3.10-4 3.10-4 3.11-1 3.11-1 3.11-2 4.1-1 4.1-1 4.1-2 4.1-2 4.1-5 4.1-5

4. T-TO0

4. T'-1 0 4.1-10 4.1-11 4.1-12 4.1-12 4.1-13 4.1-13 4.1-14 4.1-15 4.1-16 4.1-17 4.1-18

-2 Remove Pages Insert Pages 4.11-1 4.11-1 4.11-2 4.11-2 4.11-3 4.11-3 4.11-4 4.11-4 4.11-5 4.11-5 4.11-6 4.11-7 4.11-8 4.21-1 4.21-2 5.1-T 5.1-1 6.1-Ta 6.1-la 6.1-2 6.1-2 6.T-3 6.1-3 6.1-3a 6.1-4 6.1-4 6.1-5 6.1-5 6.1-5a 6.4-1 6.4-1 6.4-2 6.5-2 6.5-2 6.6-1 6.6-1 6.6-2 6.6-2 6.6-3 6.6-3 6.6-6 6.6-6 6.6-6a.

6.6-7 6.6-7 6.6-8 Delete 6.6-9 Delete 6.3-1

Section Page 1.5.4 Instrument Channel Calibration 1-3 1.5.5 Heat Balance Check 1-4 1.5.6 Heat Balance Calibration 1-4 1.6 POWER DISTRIBUTION 1-4 1.6.1 Quandrant Power Tilt 1-4 1.6.2 Reactor Power Imbalance 1-4 1.7 CONTAINMENT INTEGRITY 1-4 1.8 RADIOLOGICAL.EFFLUENT CONTROL 1-5 1.8.1 Source Check 1-5 1.8.2 Offsite Dose Calculation Manual (ODCM) 1-3 1.8.3 Process Control Program (PCP) 1-5 1.8.4.

Solidification 1-5 1.8.5 GAseous Radwaste Treatment System 1-5 1.8.6 Ventilation Exhaust Treatment System 1-5 1.8.7 Purge-Purging 1-5 1.8.8 Venting 1-6 1.8.9 Member(s) of the Public 1-6 1.8.10 Unrestricted Area 1

2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1-1 2.1 SAFETY LL4ITS, REACTOR CORE 2.1-1 2.2 SAFETY LIMITS -

REACTOR COOLANT SYSTE4 PRESSURE 2.2-1 2.3 LIMITING SAFETY SYSTEM SETTINGS, PROTECTIVE INSTRUMENTATION 2.3-1 3

LIMITING CONDITIONS FOR OPERATION 3.0-1 3.0 LIMITING CONDITION FOR OPERATION 3.0-1 3.1 REACTOR COOLANT SYSTEM 3.1-1

'1 Amendments Nos'. 1Z5, 125, & 122

Section Page 3.1.1 Ooerational Comvonents 3.1-1 3.1.2 Pressurization, HeatuD, and Cooldown Limitations 3.1-3 3.1.3 Minimum Conditions for Criticality 3.1-8 3.1.4 Reactor Coolant System Activity 3.1-10 3.1.5 Chemistry 3.1-12 3.1.6 Leakage 3.1-14 3.1.7 Moderator Temtrature Coefficient of Reactivity 3.1-17 3.1.8 Single Loon Restrictions 3.1-19 3.1.9 Low Power Physics Testing Restrictions 3.1-20 3.1.10 Control Rod Ooeration 3.1-21 3.1.11 Shutdown Margin 3.1-23 3.1.12 Reactor Coolant System Subcooling Margin Monitor 3.1-24 3.2 HIGH PRESSURE INJECTION AND CHEMICAL ADDITION SYSTEMS 3.2-1 3.3 EMERGENCY GRE COOLING, REACTOR BUILDING COOLING, REACTOR 3.3-1 BUILDING SPRAY AND LOW PRESSURE SERVICE WATER SYSTEMS 3.4 SECONDARY SYSTEM DECAY HEAT REMOVAL 3.4-1 3.5 INSTRUMENTATION SYSTEMS 3.5-1 3.5.1 Operational Safety Instrumentation 3.5-1 3.5.2 Control Rod Group and.Power Distribution Limits 3.5-6 3.5.3.

Engineered.Safety Features Protective System 3.5-28 Actuation Setnoints 3.5.4 Incore Instrumentation 3.5-30 3.5.5 Radioactive Effluent Monitoring Instrumentation 3.5-34 3.6 REACTOR BUILDING 3.6-1 3.7 AUXILIARY ELECTRICAL.SYSTEMS 3.7-1 3.8 FUEL LOADING AND REFUELING 3.8-1 3.9 RADIOACTIVE LIQUID EFFLUENTS 3.9-1 ii T Amendments Nos.

125, 125, & 122

Section Page 3.10 RADIOACTIVE GASEOUS EFFLUENTS 3.10-1 3.11 SOLID RADIOACTIVE WASTE 3.11-1 3.12 REACTOR BUILDING POLAR CRANE AND AUXILIARY HOIST 3..12-1 3.13 SECONDARY SYSTEM ACTIVITY 3.13-1 3.14 SNUBBERS 3.14-1 3.15 PENETRATION ROOM VENTILATION SYSTEMS 3.15-1 3.16 HYDROGEN PURGE SYSTEM 3.16-1 3.17.

FIRE PROTECTION AND DETECTION SYSTEMS 3.17-1 4

SURVEILLANCE REQUIRLENTS 4.0-1 4.0 SURVEILLANCE STANDARDS 4.0-1 4.1 OPERATIONAL SAFETY REVIEW 4.1-1 4.2 STRUCTURAL INTEGRITY OF ASME CODE CLASS 1, 2 AND 3 4.2-1 COMPONENTS 4.3.

TESTING FOLLOWING OPENING OF SYSTEM 4.3-1 4.4 REACTOR BUILDING 4.4-1 4.4.1 Containment Leakage Tests 4-1 4.4.2 Structural Integrity 4.4.3 Hydrogen Purge System 4.4-17 4.5 EMERGENCY CORE COOLING SYSTEMS AND REACTOR BUILDING 4.5-1 COOLING SYSTEMS PERIODIC TESTING 4.5.1 Emergency Core Cooling Systems 4.5-1 4.5.2 Reactor Building Cooling Systems 4.5-6 4.5.3 Penetration Room Ventilation Svstm 4.5-10 4.5.4 Low Pressure Injection System Leakage 4.5-12 4.6.

MERGENCY POWER PERIODIC TESTING 4.6-1 4.7 REACTOR CONTROL ROD SYSTEM TESTS 4.7-1 4.7.1.

Control Rod Trip Insertion Time 4.7-1 4.7.2 Control Rod Program Verification 4.7-2 4.8 MAIN STEAM STOP VALVES 4.8-L Amendments Nos.

125,125

,& 122 IV

Section Page 4.9 EMERGENCY FEEDWATER PUMP AND VALVE PERIODIC TESTING 4.9-1 4.10 REACTIVITY ANOMALIES 4.10-1 4.11 RADIOLOGICAL ENVIRONMENTAL MONITORING 4.11-1 4.12 CONTROL ROOM FILTERING SYSTEM 4.12-1 (INTENTIONALLY BLANK) 4.13-1 4.14 REACTOR BUILDING PURGE FILTERS.AND SPENT FUEL.POOL 4.14-1 VENTILATION SYSTEM 4.15 IODINE RADIATION MONITORING FILTERS 4.15-1 4.16 RADIOACTIVE MATERIALS SOURCES 4.16-1 4.17 STEAM GENERATOR TUBING SURVEILLANCE 4.17-1 4.18

SNUBBERS, 4.18-1 4.19 FIRE PROTECTION AND DETECTION SYSTEM 4.19-1 4.20 DELETED PER AMENDMENTS 109, 109, AND 106 4.21 DOSE CALCULATIONS 4.21-1 DESIGN FEATURES 5.1-1 5.1 SITE 5.1-1 5.2 CONTAINMENT 5.2-1 5.3 REACTOR 5.3-1 5.4 NEW AND SPENT FUEL STORAGE FACILITIES 5.4-1 6

ADMINISTRATIVE CONTROLS 6.1-1 6.1 ORGANIZATION, REVIEW, AND AUDIT 6.1-1 6.1.1.

Organization 6.1-1 6.1.2 Technical Review and Control 6.1-2 6.1.3 Nuclear Safety Review Board.

6.1-3a 6.2 ACTION TO BE TAKEN IN THE EVENT OF A REPORTABLE OCCURRENCE 6.2-1 6.3 ACTION TO BE TAKEN IN THE EVENT A SAFETY LIMIT IS EXCEEDED 6.3-1 6.4 STATION OPERATING PROCEDURES 6.4-1 V

Amendments Nos.

125k 125,& 122

Section Page 6.5 STATION OPERATING RECORDS 6.5-1 6.6 STATION REPORTING REQUIREMENTS 6.6-1 6.6.1 Routine Renorts 6.6-1 6.6.2 Non-Routine Revorts 6.6-5 6.6.3 Soecial Renorts 6.6-7 6.7 ENVIRONMENTAL QUALIFICATION 6.7-1 6.8 OFFSITE DOSE CALCULATION MANUAL (ODCM) 6.8-1 v-a Amendments Nos.125

, 125,& 122

LIST OF TABLES Table No.

Page 2.3-LA Reactor Protective System Trip Setting Limits - Unit 1 2.3-11 2.3-13 Reactor Protective System Trip Setting Limits -

Unit 2 2.3-12 2.3-IC Reactor Protective System Trip Setting Limits -

Unit 3 2.3-13 3.5-1-1 Instruments Operating Conditions 3.5-4 3.5-1 Quadrant Power Tilt Limits 3.5-14 3.5.5-1 Liquid Effluent Monitoring Instrumentation Operating 3.5-36 Conditions 3.5.5-2 Gaseous Process and Effluent Monitoring Instrumentation 3.5-38 Operating Conditions 3.7-1 Operability Requirements for the Emergency Power 3.7-13 Switching Logic Circuits 3.17-1 Fire Protection & Detection Systems 3.17-5 4.1-1 Instrument Surveillance Requirements 4.1-3 4.1-2 Minimum Equipment Test Frequency 4.1-9 4.1-3 Minimum Sampling Frequency and Analysis Program 4.1-10 4.1-4 Radioactive Effluent Monitoring Instrumentation 4.1-16 Surveillance Requirements 4.2-1 Oconee Nuclear Station Capsule Assembly Withdrawal 4.2-3 Schedule at Crystal River Unic No. 3 4.4-1 List of Penetrations with 10 CFR 50.Appendix J Test 4.4-6 Requirements 4.11-1 Radiological Environmental Monitoring Program 4.11-3 4.11-2 Maximum Values for the. Lower Limits of Detection (LLD)

.11-3 4.11-3 Reporting Levels for Radioactivity-Concentrations in 4.11-8 Environmental Samples 4.17-1.

Steam Generator Tube Inspection 4.17-6 6.1-1 Minimum.Operating Shift Requirements with Fuel in Three 6.1-6 Reactor Vessels Amendments Nos.

125 125

& 122vi

L.8 RADIOLOGICAL EFFLUENT CONTROL 1.8.1 Source Check A Source Check is the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.

1.8.2 Offsite Dose Calculation -Manual (ODCMI)

The Offsite Dose Calculation Manual is a manual containing the methodology and parameters to be used in the calculation of offsite doses due to radio active gaseous and liquid effluents and in the calculation of gaseous and liquid effluent monitoring instrumentation alarm/trip setpoints and in the conduct of environmental radiological monitoring.

1.8.3 Process Control Program (PCP)

The Process Control Program is a procedurethat shall contain the sampling, analysis, and formulation determination by which solidification of radioactive liquid waste is assured.

1.8.4 Solidification Solidification shall be the immobilization of wet radioactive wastes such as evaporator bottoms, spent resins, sludges, and reverse osmosis concentrates as a result of a process of. thoroughly mixing the waste type with a solidifi cation agent(s) to form a. free standing monolith with chemical and physical characteristics specified in the Process Control Program (PCP).

1.8.5 Gaseous Radwaste Treatment System A Gaseous Radwaste Treatment System is.-any system designed. and installed to reduce radioactive gaseous effluents by collecting.primary coplant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to-the environ ment.

1.8.6 Ventilation Exhaust Treatment System A Ventilation Exhaust Treatment System is any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsor bers and/or KEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment.

Engineered Safety Features (ESF) atmospheric cleanup systems are not consider ed to be Ventilation Exhaust Treatment System components.

1.8.7 Purge-Purging Purge or Purging is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

1-5 New Page Amendments-Nos. 125,12r,&

122

1.8.8 Venting Venting is the controlled process of discharging air or gas from a confinement to maintain temperature. Dressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not provided or required during Venting. Vent, used in system names, does not imply a venting process.

1.8.9 Member(s) Of The Public Members(s) Of The Public shall include all persons who are not occupationally associated with the plant. This category does not include employees of the utility, its contractors or its vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries.

This category does include persons who use portions of the site for recrea tional, occupational or other purposes not associated with the plant.

1.8.10 Unrestricted. Area An Unrestricted Area shall be any area at or beyond the site boundary to which access is not controlled by the licensee for purposes of protection of in dividuals from exposure to radiation and radioactive materials or any area within the site boundary used for residential quarters or industrial,. com merical institutional.and/or recreational purposes.

1-6 New Page Amendments Nos.

125,.125., & 122

3.5.5 Radioactive Effluent Monitoring Instrumentation Applicability Applies to radioactive liquid effluent, gaseous effluent, and gaseous process monitoring instrumentation.

Specifications 3.5.5.1 Liquid Effluents

a.

The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.5.5-1 shall be operable with their alarm/trip setpoints set to ensure that the limits of Speci fication 3.9.1 are not exceeded.

b.

If a radioactive liquid effluent monitoring instrumentation channel alarm/trip setpoint is less conservative than required, without delay suspend the release of radioactive liquid ef fluents monitored by the affected channel, or declare the channel inoperable, or change the setpointso it is acceptably conservative.

c.

In the event that the number of operable radioactive liquid effluent monitoringinstrumentation channels falls below the limit given under Table 3.5.5-1, Column A, action shall be as shown in Column B. Exert best efforts to return-the instru ments to operable status within 30 days: and,.if unsuccessful, explain in the next Semiannual Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner-.

3.5.5.2 Gaseous Process and Effluents.

a.

The radioactive gaseous process and effluent monitoring instru mentation channels shown in Table 3.5.5-2 shall be operable~

with their alarm/trip setpoints set to ensure that the limits of Specification 3.10.1 are not exceeded.

b.

If a radioactive gaseous effluent monitoring instrumentation channel alarm/trip setpoint is less conservative than required, without delay suspend the release of radioactive gaseous ef fluents monitored by the affected channel or declare the chan ael inoperable, or change the setpoint so it is acceptably conservative.

c.

In the event that the number of radioactive gaseous process or effluent monitoring instrumentation channels falls below the limit given under Table 3.5.5-2, Column A,, action shall be takenas shown in Column B. Exert best efforts to return the instruments to operable status within 30 days and, if unsuc cessful, explain in the next Semiannual Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner.

3.5-34 New Page Amendments Nos.

125 125, & 122

3.5.5.3 Setpoints The setpoints ghall be determined in accordance with the methodology described in the ODCMi and shall be recorded. Setpoint correction may be permitted with out declaring the channel inoperable.

3.5.5.4 The provisions of Technical Specification 3.0 do not apply.

Bases The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases. The alarm/trip setpoints for these instruments shall be calculated in accordance with NRC approved methods in the ODC1 to assure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20.

The operability and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63,.and 64 of Appendix A to 10 CFR Part 50.

The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases.

The alarm/trip setpoints for these instruments shall be calculated in accordance with NRC approved methods in the ODCM to assure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20.

This instrumentation also includes provisions for monitoring (and controlling) the concentration of potentially -explosive gas mixtures in the waste gas holdup system. The operability and use of this.

instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

3.5-35 New Page Amendments Nos. 125, 125

& 122

Table 3.5.5-1 LIQUID EFFLUENT-MONITORING INSTRLENTATION OPERATING CONDITIONS A

B INSTRUMENT MINIUM OPERATOR ACTION IF OPERABLE MINIMUM NUMBER OF CHANNELS APPLICABILITY OPERABLE CHANNELS IS NOT MET

1.

Monitors Providing Automatic Termina tion of Release Liquid Radwaste Efflu ent Line Monitors 1 RIA-33 1(a),

Turbine Building Sump 1 RIA-54 (Units 1 & 2) 1 (b) 3 RIA-54 (Unit 3) 1 (b)

2.

Monitors not Providing Automatic Termination of Release Low Pressure Service Water 1 RIA-35 1

(d) 2 RIA-35 1

(d) 3 RIA-35 1

(d)

3.

Flow Rate Measuring Devices Liquid Radwaste Effluent Line 1

(c)

Keowee Hydroelectric Station Tailrace Dis charge NA NA NA

4.

Continuous Composite Sampler

3.Chemical Treat ment Pond Composite Sampler and Sampler Flow Monitor (Turbine Building Sumps Effluent) 1 (d)

At all times.

- Flow determined from number of hydro units operating; if hydro is not operating, leakage flow, which is measured periodically, is used.

3.5-36 New Page Amendments Nos.125,125

&122

Table 3.5.5-1 NOTES (a) Effluent releases may continue provided that prior to initiating a release:

1.

Two independent samples are analyzed in accordance with Specification 3.9 and;

2.

Two independent data entry checks for release rate calculations and valve lineups of the effluent pathway are conducted.

Otherwise, suspend release of radioactive effluents by this pathway.

(b) Effluent releases may continue provided that prior to each discrete release of the sump, grab samples are collected and analyzed for gross radioactivity (beta and/or gamma) at a lower limit of de tection of at least 10 pCi/ml.

(c) Effluent releases may continue provided flow rate is estimated at least once per four hours during actual releases.

(d) Effluent releases may continue provided that grab samples are collected and analyzed for gross radioactivity (beta and/or gamma) at a lower limit of detection of at least 10' pCi/ml every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

3.5-37 New Page Amendments Nos.125,

125, & 122

Table 3.5.5-2 GASEOUS PRQCESS AND EFFLUENT MONITORING INSTRUELNTATION OPERATING CONDITIONS A

B INSTRUMENT MINIMUM OPERATOR ACTION IF OPERABLE MINIM NUMBER OF CHANNELS APPLICABILITY OPERABLE CHANNELS IS (PER NOT MET RELEASE PATH)

1.

Waste Gas Holdup Tanks

a.

Noble Gas Activity Monitor - Providing Alarm and Automatic Termination Of Release (RIA-37,

38) 1 (a)

b.

Effluent Flow Rate Monitor (Waste Gas 1

(b)

Discharge Flow)

2.

Unit Vent Monitoring System

a.

Noble Gas Activity Monitor Providing Alarm and Automatic Termination of Con tainment Purge Re lease (RIA -45) 1 (a)

b.

Iodine Sampler 1

(4)

c.

Particulate Sampler 1

(d)

d.

Effluent Flow Rate Monitor (Unit Vent 1

(b)

Flow)

e.

Sampler Flow Rate Monitor 1

(e)

f.

Effluent Flow Rate Monitor (Containment Purge) 1 (b)

3.

Interim Radwaste Building Ventilation Monitoring System

a.

Noble Gas Activity Monitor (RIA - 53) 1 (c)

b.

Iodine Sampler#

1 (d)

c.

Particulate Sampler#.1 (d) 3.5-38 New Page Amendments Nos.

125,125

, & 122

Table.3.3.5-2 (Cont'd)

GASEOUS PROCESS AND EFFLUENT MONITORING INSTRUMENTATION OPERATING-CONDITIONS A

B INSTRUMENT MINIMUM OPERATOR ACTION IF OPERABLE MINIMUM NUMBER OF CHANNELS APPLICABILITY OPERABLE CHANNELS IS (PER NOT MET RELEASE PATH)

d.

Effluent Flow Rate Monitor (Interim 1(b)

Radwaste Exhaust)#

e.

Sampler Flow Rate Monitor#

1(e)

4.

Hot Machine Shop Ventilation Monitoring System

a.

Iodine Sampler#

1 (d)

b.

Particulate Sampler# 1 (d)

c.

Effluent Flow Rate Monitor (Hot Machine 1

(b)

Shop Exhaust)#

d-Sampler Flow Rate Monitor#

1 (e)

'At all times.

During waste gas holdup tank releases and/or containment purge operation.

Effective upon installation of equipment.

3.5-39 New Page Amendments Nos. 125,125

& 122

Table 3.5.5-2 NOTES (a)

Effluent releases from waste gas tanks or containment purges may con tinue provided that prior to initiating a release:

1.

Two independent samples are analyzed and;

2.

Two independent data entry checks for release rate calculations and valve lineups of the effluent pathway are conducted and; Effluent release from ventilation system or condenser air ejectors may continue provided that grab samples are taken once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and these samples are analayzed for gross activity (beta and/or gamma) within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or continuously monitor through the unit vent. Otherwise, suspend release of radioactive effluents via this pathway.

(b) Effluent releases may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

(c) Effluent releases may continue provided grab samples are taken once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and these samples are analayzedfor gross activity (beta and/or gamma) within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

(d) Effluent releases may continue provided samples are continuously col lected with auxiliary sampling equipment for periods not to exceed 7 days and analyzed.within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of the end of sample collection.

(e) Alarms indicating low flow may be substituted for flow measuring devices.

3.5-40 New Page Amendments Nos. T25 125

&.122

3.9 RADIOACTIVE LIQUID EFFLUENTS Applicability Applies at all times to the controlled release of all liquid waste discharged from the site which may contain radioactive materials, except as noted.

Appendix I dose limits for radioactive liquid effluent releases (T.S.

3.9.2) are applicable only during normal operating conditions which include expected operational occurrences, and are not applicable during unusual operating con ditions that result in activation of the Oconee Emergency Plan.

Objective To establish conditions for the controlled release of radioactive liquid effluents.

To implement the requirements of 10 CFR 20, 10 CFR 50.36a, Appendix A to 10 CFR 50, Appendix I to 10 CFR 50, 40 CFR 141 and 40 CFR 190.

Specification 3.9.1 Concentration

a.

The concentration of radioactive material released at anytime from the site-boundary for liquid effluents to Unrestricted Areas (denoted in Figure 2.1-4(a) of the Oconee Nuclear Station Final Safety Analysis Report) shall be limited. to the concen tration specified in 10 CFR Part 20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases.

For dissolved or entrained noble gases-the con centration shall be limited to 2 x 10 pCi/ml total activity.

b.

If the concentration of radioactive material released in liquid effluents to Unrestricted Areas exceeds the above.Specified limits, without delay restore the concentration to within the above limits.

3.9.2 Dose

a.

The dose or dose commitment to a.Member Of The Public from radioactive materials in liquid effluents. to Unrestricted Areas shall.be limited to:

1) during any calendar quarter:

< 4.5 mrem to the total body (15 mrem to any organ and;

2) during any calendar year:

< 9 mrem to the total body

< 30 mrem to any organ.

3.9-1 Amendments Nos. 125,125

, & 122

b.

If the calculated dose from.the release of radioactive materials in liquid effluents exceeds any of the above limits, except during unusual operating conditions that result in activation of the Oconee Emergency Plan, and in lieu of any other report required by Section 6.6.2, a report shall be submitted within 30 days from the end of the quarter during which the release occurred, to the regional NRC Office which ihcludes the following:

1.

Cause(s) for exceeding the limit(s)

2.

A description of the program of corrective action initi ated to:

reduce the releases of radioactive materials in liquid effluents, and.to keep these levels of radio active materials in liquid effluents in compliance with the above limits, or as low as reasonably achievable.

3.

Results of radiological analyses of the drinking water source and the radiological impact on finished drinking water supplies with regard to the requirements of 40 CFR 141.

3.9.3 Liquid Waste Treatment

a.

The appropriate subsystems of the liquid radwaste treatment system shall be used to reduce the radioactive materials in liquid waste prior to their discharge., if the projected dose due to liquid effluent releases to unrestricted areas,.when averaged over 31 days would exceed 0.18 mrem to the total body or 0.6 mrem to any organ.

b.

If radioactive liquid waste is discharged without treatment and in excess of the above limit, a report shall be submitted with in 30 days to the-regional NRC Office which includes the following:

1.

Cause of equipment or subsystem inoperability.

2.

Corrective action to restore equipment and prevent re currence.

3.9.4 Chemical Treatment Ponds (CTP 1 and 2)

a.

The quantity of radioactive material in the Chemical Treatment Ponds (CTP) shall be limited so that, for all radionuclides identified, excluding noble gases and tritium, the sum of the ratios of activity (in curies) to the limits in 10 CFR 20, Appendix B,. Table II, Column 2 shall not exceed 1.7 x 10'.

A 1.7 x 105 j

Cj Amendments Nos.

125 125

& 122 3.9-2

where Aj =.pond inventory limit for single radionuclide 'j' (curies)

Cj = 10 CFR 20, Appendix B, Table II, Column 2, concentration for single radionuclide 'j' (curies)

b.

After a primary to secondary leak is detected, the initial batch of used Powdex resin shall not be transferred to the CTP.

No batch of used powdex resin shall be transferred to the CTP unless the sum of the ratios of the activity of the radionuclides identified in the preceeding batch from any powdex cell in the same unit is less than 0.1% of the limit identified in 3.9.4.a.

I 1 1.0 x 10 j, Aj where Qj radionuclide activity in the batch Aj pond inventory limit for radionuclide 'j'

c.

The radionuclide inventory per batch of used powdex resin transferred., averaged over the transfers of the previous 13 weeks, shall. not exceed 0.01% of the pond radionuclide inven tory limit. If this average exceeds 0.01% of the pond radionu clide inventory limit, then a report will be submitted within 30 days to the Regional NRC Office.describing the reason or reasons for exceeding the objective and plans for future operktion.

Decay of radionuclides may be taken into account in determining inventory levels.

Qj

+ Qj

+

Qj

+ Qjn. <

.01% x Aj 1 2 (n-1) aI.

where Qj = activity or radionuclide 'j' in the batch a = number of batches transferred to the chemical treatment ponds during the previous 13-week period.

3.9.5 Liquid Holdup Tanks

a.

The quantity of radioactive material contained :in each out side temporary tank shall be limited to less than or equal to 10 curies, excluding. tritium and dissolved or entrained noble gases. Tanks included in this specification are those outdoor tanks that are not surrounded by liners, dikes, or walls-capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system.

b.

The quantity of radioactive material contained in each of the outside-temporary tanks shall be determined to be-within the above limit by analyzing a representative sample of the tanks contents at least once per 7 days when radioactive materials are being added to the tank.

Amendments Nos. 125, 125,. & 122 3.9-3

c.

If the quaditity of radioactive materiailin any outside tem-porary tank exceeds the above limit, suspend all additions to radioactive material to the tank without delay.

3.9.6 The provisions of Technical Specification 3.0 do not apply.

Bases The concentration specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the site to unrestricted areas will be less than the concentration levels specified in 10 CTR Part 20, Appendix B, Table II. The concentration limit for noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protec tion (ICRP) Publication 2.

The dose specification is provided to assure that the release of radioactive material in.liquid effluents will be kept "as low as is -reasonably achievable."

Also, for fresh water sites with drinking water supplies which can be potentially affected by plant operations, there is reasonable assurance that the operation of the facility will not result in radionuclide concentrations in the finished drinking water that are in excess of the requirements of 40 CFR 141.

The dose calculations in the ODCM implement the requirements in Section III.A of Appendix I.

that conformance. with the giviip, nf Appeandix T istLa shown by calculational procedures based on models and data such that the actual exposure of an indivi dual through appropriate pathways is unlikely to be substantially underestimated.

Section IV of Appendix I of 10 CFR 50 states that-the licensee is permitted the flexibility of operation during unusual operating conditions, to assure the public is provided with a dependable source of power when compatible with considerations of health and safety of the public.

Section I of Appendix I of 10 CFR 50 states that this appendix provides specific numerical guides for-- design objectives and limiting conditions for operation, to assist holders of licenses for light-water-cooled nuclear power reactors in meeting the re quirements to keep releases of radioactive material to unrestricted areas as low as practical, and reasonably achievable, during normal reactor operations, including expected operational occurrences. Using the flexibility granted during unusual operating conditions, and the stated applicability of the de sign objectives for the Oconee Nuclear Station, Appendix I dose limits for radioactive liquid effluent releases (T.S. 3.9.2), are concluded to be not applicable during unusual operating conditions that result in the activation of the Oconee Emergency Plan.

For units with shared radwaste-treatment systems, the liquid effluents from the shared system are proportioned among the units sharing that system.

The requirements that the appropriate portions of this.system be used when speci fied provides assurance that the releases of radioactive materials in liquid ef fluents will be kept "as low as is reasonably achievable."

This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix-A to 10 CFR Part 50 and design objective Section II.D of Appendix A to 10 CFR.Part 50.

3.9-4 Amendments Nos. 125

,125

& 122

The inventory limits of - hechemical treatment ponds are based on limiting-the consequences of an uncontrolled release of the pond inventory. The short term rate limit (2 mrem/hr) of 10CFR20.105 is applied to 10CFR20.106 in the following expression:

Aj x

106 pCi x

gal 2 mrem/hr x 8760 hr 1.3 x 100 gal curie 3785 ml 500 mrem/yr yr Cj Aj 1.7 x-10s Cj where Aj = pond inventory limit for radionuclide j' (curies)

Cj = 10CFR20 Appendix B, Table II, Column 2 concentration for radionuclide 'j' 1.3 x 106 gal = estimated volume of smaller chemical treatment pond The batch limits provide assurance that activity input to the CTP will be min imized.

3.9-5 New Page Amendments Nos.

125, 1251

& 122

3.10 RADIOACTIVE GASEOUS EFFLUENTS Applicability Applies at all times to the controlled release of all gaseous waste discharged from the station which may contain radioactive materials.

Objective To establish conditions for the controlled release of radioactive gaseous effluents.

To implement the requirements of 10CFR20, 10CFR50.36a, Appendix A to 10CFR50, Appendix I to 10CFR50, and 40CFRI90.

Specifications 3.10.1 Dose Rate

a.

The instantaneous dose rate at the site (exclusion area) boundary for gaseous effluents (Figure 2.1-4(a) of the Oconee Nuclear Station Final Safety Analysis Report) due to radioactive materials released in gaseous effluents from the site shall be limited to the following values:.

1.

The dose rate limit for noble gases shall be:

< 500 mrem/yr to the total body

< 3000 mrem/yr to the skin and;

2.

The dose rate limit for all radioiodines and for all radioactive materials in particulate form and radionuclides other than noble gases with half-lives greater than 8 days shall be < 1500 mrem/yr to any organ.

b.

If the dose rate exceeds the above limits, without delay de crease the release rate to,within the above limits.

3.10.2.

Dose

a.

The air dose due to noble gases released in gaseous effluent from the site shall be limited to the following:

1.

During any calendar quarter:

< 15 mrad for gamma radiation

< 30 mrad for beta radiation

2.

During any calendar year:

< 30 mrad for gamma radiation

< 60 mrad for beta radiation 3.10-1

b.

The dose to a Member Of The Public from radioiodines, tritium and radioactive materials in particulate form with half-lives greater than 8 days in.gaseous effluents released from the site, shall be limited to the following:

1.

During any calendar quarter:

< 22.5 mrem to any organ

2.

During any calendar year:

< 45 mrem to any organ.

C.

If the calculated dose from these gaseous effluents exceeds any of above limits in lieu of any other report required by Specifi cation.6.6.2, a report shall be submitted within 30 days from the end of the quarter during which the release occured to the regional INRC Office which includes the following:

1.

Cause(s) for exceeding the limit(s);

2.

A description of the program of corrective action-initi ated to:

reduce the releases of radioactive materials in gaseous effluents, and to keep these levels of radio active materials in gaseous effluents in compliance with the above limits or as low as reasonably achievable.

3.10.3 Gaseous Radwaste Treatment

a.

The Gaseous Radwaste Treatment System shall be used to reduce the noble gases. in gaseous wastes prior to their discharge,. if the projected gaseous effluent air dose due to gaseous effluent releases from the site, when averaged over 31 days exceeds 0.6 mrad for gamma radiation and 1.2 mrad for beta radiation..

b.

The Ventilation.Treatment Exhaust System shall be used to re duce radioactive materials other than noble gases in gaseous waste prior to their discharge when the projected doses due to effluent releases to unrestricted areas when averaged over 31 days would exceed 0.9 mrem to any organ. This does not apply to the Auxiliary Building Exhaust System since it is not "treated" prior to release.

c.

If radioactive gaseous waste is discharged without treatment for more than 31 days and in excess of the above limits in lieu of any other report required by specification 6.6.2, a report shall be submitted within 30 days to the regional NRC Office which includes the following:

1.

Cause of equipment or subsystems inoperability

2.

Corrective action to restore equipment and prevent recurrence Amendments Nos. 125, 125

& 122 3.10-2

3.10.4 Waste Gas Holdup Tanks

a.

The quantity of radioactivity contained in each waste gas hold up tank shall be limited to < 3.8E+05 curies noble gases (considered as Xe-133).

b.

Daily, when radioactive materials are being added to a waste gas holdup tank, the quantity of radioactive material contained in the tank being filled shall be determined.

c.

If the quantity of radioactive material in any waste gas hold up tank exceeds the above limit, without delay suspend all additions of radioactive material to the tank and within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , reduce the tank contents to within the above limit.

3.10.5 Used Oil Incineiation Used oil, contaminated by radioactivity,.may be incinerated in the Station auxiliary boiler provided releases do not exceed one-tenth of one percent (0.1%) of the limits in Technical Specification 3.10.2.b.2.

3.10.6 The provisions of Technical Specifications 3.0 do not apply.

Bases Specification 3.10.1 is provided to assure that the dose rate at anytime at the exclusion area boundary from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR Part 20 for unrestricted areas.

The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table II. These limits provide reasonable assurance that radioactivity material discharged in gaseous effluents will not result in the exposure of an individual in an unrestricted area, either within or outside the exclusion area boundary, to annual average concentrations exceeding the limits specified in Appendix B, Table II of 10 CFR Part 20 (10 CFR Part 20.106(b)).

For individuals who may at times be within the exclusion area boundary, the occupancy of the individual will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the exclusion area boundary. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to an individual at or beyond the exclusion area boundary to < 500 mrem/year to the total body or to < 3000 mrem/year to the skin. These release rate limits also restrict, at all times, the corresponding thyroid rate above back ground to an. infant via the milk animal-milk-infant pathway to < 1500 mrem/year for the nearest milk animal to the plant.

Specification 3.10.3 is provided to implement the requirements of Appendix I, 10 CFR Part 50. The specification provides the required operating flexibility and at the same time implement the-guides set forth in Appendix I to assure that the. releases of radioactive material in gaseous effluents will be kept "as low as is reasonably achievable."

Surveillance requirements are imple mented to meet the requirements of Appendix I. Calculational procedures based on models and data show that the actual-exposure of an individual through the appropriate pathways is unlikely to be substantially underestimated.

3.10-3 Amendments Nos.

125, 125

, & 122

The ODCM calculational methods for calculating the doses due to the actual release rates of the subject materials will be consistent with the methodology provided in Regulatory Guide 1.109, "Calculating of Annual Doses to Man from Routine Releases of Reactor Effluents for the Pur-pose of..Evaluating Compliance with 10 CFR Part 50, Appendix I, "Revision I, October 1977 and Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors."

Equations in the ODCM are provided for determining the actual doses based upon the historical average atmospheric conditions. The release rate specifications for radioiodines, radioactive material in particulate form and radionuclides other than noble gases are dependent on the existing radionuclide pathways to man, in the unrestricted area.

The pathways which are examined in the develop ment of these calculations are:

1) individual inhalation of airborne radio nuclides, 2) deposition of radionuclides onto green leafy vegetation with.

subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, and 4) deposition on the ground with subsequent exposure of man.

The requirement that the appropriate portions of these systems be used when specified provides reasonable assurance.that the release of radioactive ma terials in gaseous effluents will be kept "as low as is reasonably achievable."

This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and design objective Section IID of Appendix I to 10 CFR Part 50.

Restricting the quantity of radioactivity contained in each waste gas holdup tank provides assurance that in the event of an unedntrolled release of the tanks contents, the resulting total body exposure to an individual at the nearest exclusion area boundary will not exceed 0.5 rem.

3.10-4 AAmo 4mnt Nor_

1 19C T7 P)Z T Ioo

9 0

3.11 SOLID RADIOACTIVE WASTE Applicability Applies to the processing and packaging of radioactive solid waste prior to shipment from the site.

Specification 3.11.1 Solid Radioactive Waste

a.

The Solid Radwaste System shall be used in accordance with a Process Control Program, for the solidification of wet radio active wastes.

Prior to the shipment of containers of radio active wastes from the site, radioactive wastes shall be pro cessed and packaged to ensure meeting the requirements of 10 CFR Part 20, 10 CFR Part 71, and Federal and State regulations governing the disposal of radioactive wastes.

b.

If the requirements of 10CFR Part 20 and/or 10CFR Part 71 are not satisfied, suspend shipments of defectively packaged solid radioactive wastes from the site.

3.11.2 Process Control Program The Process Control Program shall be used to verify the Solidification of at least one representative-test specimen from at least every tenth batch of each type of wet radioactive waste to be solidified.

1.

Solidification

a.

If any test specimen fails to verify Solidification, the Solidification of.the batch under test shall be suspended until such time as additional test specimens can be obtained, alternative Solidification parameters. can be determined in accordance with the Process Control Program, and a subsequent test verifies Solidification. Soldification of the batch may then be resumed using the alternative Solidification-para meters determined by the Process Control Program.

2.

Process Control Program

b.

If the initial test specimen from a batch of waste fails to verify Solidification, the Process Control Program shall pro vide for the collection and. testing of representative test specimens.from each consecutive batch of the same type of wet waste until at least 3 consecutive initial test specimens demonstrate Solidification. The Process Control Program shall be modified as required to assure Solidification of subsequent batches of waste.

3.11.3 The provisions of Technical Specification 3.0 do not apply.

Bases The solid radwaste system will be used whenever solid radwastes require pro cessing and packaging prior-to being shipped offsite. This specification 3.11-1 Amendments Nos..

125, 125, & 122

implements the requirements of 10 CFR Part 50.36a and General Design Criterion 60 of Appendix A to 10 CFR Part 50.

The process parameters included in estab lishing the Process Control Program may include, but are not limited to waste type, waste pH, waste/liquid/solidification agent/catalyst ratios, waste oil content, waste principal chemical constituents, mixing and curing times.

3.11-2 New Page Amonnt 125.15

& 1?

4.1 OPERATIONAL SAFETY REVIEW Applicability Applies to items directly related to safety limits and limiting conditions for operation.

Objective To specify-the frequency and type of surveillance to be applied to unit equip ment and conditions.

Specification 4.1.1 The frequency and type of surveillance required for Reactor Protec tive System and Engineered Safety Feature Protective System instru mentation shall be as stated in Table 4.1-1.

4.1.2 The frequency and type of surveillance required for selected equip ment shall be as stated in Table 4.1-2.

4.1.3 Required sampling should be performed as detailed in Table 4.1-3.

4.1.4 The frequency and type of surveillance required for radioactive effluent monitoring instrumentation shall be as stated in Table 4.1-4.

4.1.5 Using the Incore Instrumentation System, a power map shall be made to verify expected power distribution at periodic intervals not to exceed ten effective full power days.

Bases Failures such as blown instrument fuses, defective indicators, and faulted amplifiers are, in many cases, revealed by alarm or annunciator action. Com parison of output and/or state of independent chainels measuring the same variable supplements this type of built-in surveillance. Based on experience in operation of both conventional and nuclear systems, when the unit is in operation, the minimum checking frequency stated is deemed adequate for're actor system instrumentation.

Calibration is performed to assure the presentation and acquisition of accurate information. The nuclear flux (power range) channels amplifiers are calibrated.

(during steady-state operating conditions) when indicated neutron power exceeds core thermal power by more than two percent. During non-steady-state operation, the nuclear flux channels amplifiers are calibrated daily to compensate for instrumentation drift and changing rod patterns and core physics parameters.

Calibration checks are also performed following significant changes in core conditions (power level and control rod positions) in order to assure that the core thermal power indication during non-steady-state operations does not exceed the indicated neutron power by more than the tolerance (4% FP) assumed in ;he safety analysis for significant duration (e.g.,.4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ).

Channels subject only to "drift" errors induced within the instrumentation itself can tolerate longer intervals between calibrations. Process. system 4.1-1 Amendmefnts Nos.

125 125

& 122

instrumentation errors induced by drift can be expected to remain within acceptable tolerances if recalibration is performed at the intervals speci fied.

Substantial calibration shifts within a channel (essentially a channel failure) are revealed during routine checking and testing procedures.

Thus, the mini mum calibration frequencies set forth are considered acceptable.

Periodic use of the Incore Instrumentation System for power mapping is suffi cient to assure that axial and radial power peaks and the peak locations are controlled in accordance with the provisions of the Technical Specifications.

REFERENCE (1) FSAR, Section 7.2.3.4.

4.1-2 Amendments Nos.125

, 125, & 122

TABLE 4.1-1 (Continued)

Channel Description Check Test Calibrate Remarks

20. Reactor Building Spray NA MO NA System Logic

21.

Reactor Building Spray NA MO RF System Analog Channel Reactor Building High Pressure

22.

Pressurizer Temperature ES NA RF

23.

Control Rod Absolute ES NA RF(2)

(1) Check with Relative Position Indicator.

Position (2) Calibrate rod misalignment channel.

24.

Control Rod Relative ES(R)

NA F(2)

(1) Check with Absolute Position Indicator.

Posit i on (2) Calibrate rod misalignment channel.

25.

Core Flood Tanks kA

a.

Pressure ES NA RF 1b.

Level ES NA RF

26. Pressurizer Level

-ES NA RF

27.

Letdown Storage Tank DA NA RF Level

28.

Delete

29.

High and Low Pressure NA NA RF Injection Systems Flow Channels

TA 111h 4. 1-3

=5 Iiit iiIII mumIij Sam Hug L-reiencVy And~ Aalysi s_ £ram CL It Cu Check Fre( icucy (A

I Reactor Coolant

a.

Gammma Isotopic Analysis

a.

3 Liimms/wveek*

b.

lHadjochemical Analysis for Sr 89, 90 1).

Moll tiIIy*

Ln

d.

Gross Beta Activity (1)

d.

3 timems/wet-k*

-e.

ChenmistLry (CI, IT arid 02)

C.

5 tjmmes/wceIk*

f.

Boron Concentration r.2tm (II

g.

Ers.Alphia Activity

g.

fiomh Iy*

It.

E lDeteriinatjon (2)

It.

semi-annna Ii

2.

Borated Water Storage Itoron Concntration Weekly", atid after each Tank Water Sample ma ketip C)

3.

Co rc F] ood ing Ta nk Boron Concentration Hoith I y* amid after each makeup

4.

Spent Fuel Pool Witter Boron Conicentration Hotimi-

aild a f Ler sammp I e eachi makeuip

5.

(YFSG or Final Feedwater

a.

Gross Beta Activity

a.

Wee k I y 1).

Gammia Isotopic Analysis (3)

6.

Coimceiitratd Boric Acid Boron ColiceiLration Twice weekly*

Not appl i cale if reactor i s iii a-coldm shutdown comid ition for a period exceedinmg the samipI.i iig I rcqmatIcy.

~"~jm Ii :aB eonl y whmenii l

is iii Lime reactour.

IkA 1I i cable omly whmen hiel -is in wil storage iii thme spent fute I poolI

(D (D

ion(5)

V.I11I 1

Pet:.

.(I in cm

-11cic PrUtilG lim vlit r

I C op.iaI r

h S m l rC 4

rd-b 1

j i

m (4))

t. init v'cmat Saiilsl oil,

.s.

titilate Spe~i tt ( 4)

Coi I. ill s oaitor, weekly

a.

<10 -'

pt/

1f 1-M3

)

(1311s141t,: Wilst(v Gasi el y (4)riii t

< 10'- gCi/t A 1-131)

I.Oeii k~~ eanttr Hhiu i i iig Il. Va rLi cIa. v s (B)l (I)s

,x Aux 6 ia u

/I io Vvlt ai at muss, Spuerit Fut.

I jI Cv-144 and tlo-99 (I

Weekly Composite 5x IeIVeratitat iota.. Air 1a Ejectors I (2)

O~lur Prilseilual Gaimmd (2)

Weekly Composite 8

()

I'(IiI L11 Ia s( 7)

(3)

Gross Alpha~ Activity (Ii.

tloaIs~ly, using compoSite (3)

< 10 Ili' c

samples of one week (4)

Hadioclsemical Asaiysj!"

(4)

Ouaamterly Composite (4)

< 10 jit i /(-a Sr 89, 90 1c.

Gases by Prmincipal Gammsa~ 1 C.

Weekly Gratai Sample C.

PC0 pi (-ae

4. Tr itIun aim.

Weekly G~rabj Sample d1.

<Io~ Pci/ci

lisa,. ~ ~ F h aaalai IS: I4.maia~,1aa -a~ I Is I. *'li

a Ct Ifaml Ie'ist! ul ejacliIsitih.

hO ll/

iastasall T I I 11 11.

Giab, sinil-jae prioir to rv-

11.

Il fil RI. eal-aa I o 11111 11rg

.j.

lit I 1a.a I 6.aamaiia 1:aai ters

a.

Gab~

Siampsle eiauc purge

6 11 hasa

m.

4 1 a lit: 1 '

a 111'.11 1 a1 11 1.

a IaI basaI.

Grall Saimpale eachI purge It.

lo l0 it I.

'1.

I-owtee-Ihy-lit 11;11 ai a.~a e.- k~sa I ow kate5 Aisit a I I y

10.

D~eeI r14 II.

lackwasa Ha' vIs v a rig Tanaks Pr a ass ilil a:iaaaas aia tLi ers ina-Gaab SampleI prior Lo ti sajig d s~i viuolsle gases release of each hatch

12. 93 Chia I Iratasent Posai.

VI Pa li Ia Ila I Gsaaluaaa Emi t ters.

a.

Montlly from

a.

Ce-144.siad bl~a xl 0(' po: /I IlI Couilosie (10)s OL ht-a- Gasua Niic I I sJe-.

Sx10-7 pCi/smi D~issolIveal Gase..

1-131 <l1O6li: a/all ha.

lHaaiucaewj cal Anialysis 1..

Qiiarterly from

b.

'SXlO 5 pci/sal for So-s Sr-89.,Sr-90, Fe-5aS curmposite Sample(9)<0"icii'fo e5 T I I uam

c.

Mionthly from (to.~

<10-5 PCi/Mi composite sample

i ah.

Gross Allahba ActilViy

d.

Mtounthly from (10 I

10~ Pci /all compajosite sample(O

TABLE 4.1-3 NOTES (1)

When radioactivity level is-greater than 10 percent of the limits of Specification 3.1.4, thesampling frequency shall be increased to a minimum of once each day.

(2) E determination will be started when gross gamma activity analysis in dicates greater than 10 pCi/ml and will be redetermined for each 10 pCi/ml increase in gross gamma activity analysis thereafter.

A radio chemical analysis for this purpose shall consist of a quantitative measurement of 95 percent of, the radionuclides in the reactor coolant with half lives greater than 30 minutes. This is expected to consist of gamma isotopic analysis of the primary coolant, including dissolved gaseous activities, radiochemical analysis for Sr-89 and Sr-90, and tritium analysis.

(3) When gross beta activity increases by a factor of two above background, iodine concentrations will bedetermined by gamma isotopic analysis and performed thereafter when the gross beta activity increases by 10 percent.

(4) Samples shall be changed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing (on after removal from sampler).

(5) The LLD is.defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal.

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

4.66 s LLD =

b E

V 2.22 x 100 Y

exp (-Adt)

Where:

LLD is the "a priori" lower limit of detection as defined above (as microcurie per unit mass.or volume),

sbis the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute),

E is the counting efficiency (as counts per disintegration),

V is the sample size (in units of mass or volume),

4.1-13 Amendments Nos. 125,125, & 122

TABLE 4.1-3 NOTES (Continued) 2.22 x 106 is the number of disintegrations per minute per microcurie, Y is the fractional radiochemical yield (when applicable),

X is the radioactive decay constant for the particular radio nuclide, and At is the elapsed time between midpoint of sample collection and time of counting (for plant effluents, not environmental samples).

Typical values of E, V, Y,_and At should be used in the calculation.

It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability'of a measure ment system and not as an a posteriori (after the fact) limit for a particular measurement.

(6) The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides:

Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported.

(7) The principal gamia emitters for which the LLD specification applies exclusively are the, following radionuclides:

Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144 for particulate emissions. This list does not mean that only these nuclides.are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported.

(8) The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with Specification 3.10.1, 3.10.2.a and 3.10.2.b.

(9) A composite sample is one in.which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen which is repre sentative of the liquids released.

(10) To be representative of the quantities and concentrations of radioactive materials in-liquid effluents, samples shall be collected continuously in proportion to the rate of flow of the effluent stream. Prior to analyses, all sampies taken for the composite shall be thoroughly mixed in order for the composite sample to be representative of the effluent.release.

4.1-14 New Page Amendments Nos. 125

, 125,

& 122

TABLE 4.1-3 NOTES (Continued)

(11) A batch release is the discharge of liquid wastes of-a discrete volume.

Prior to sampling for analyses, each batch shall be isolated, and then thoroughly mixed, to assure representative sampling.

4.1-15 New Page Amendments Nos.

125, 125, & 122

TlABLE.4.1-4 l(AI OAC'll VE E,FlIJI-.N'I' HONITORI NG INSTRIUMENTAT[ON SUIRVEIL.LANCE HEqIIREMEN'fS CIIANNLKL CHIANNELI

RESPONSE

SOURCE CHANNE I, FUNCTIONAL I SRM'TCIIECK(4)

CHECK CALI BRATION TEST'

1.

hi (liid RHdwas Le Lffluient Line

a.

Lfh te i ine tionilor (I RIA-33)

DA AN QuO.I It. E IIItet Fl ow Rate Honitor

NA AN NA LnC.

Hliujil~i Flow Device

NA AN NA S2.

'futrbizie Gui Id lg Stimp

y. Stmp H~iLr(IIA-54)

DA MO AN(3)

QU(2)

b.

Hjiuuiial Flow Device

NA AN NA 7 3.

Low Pressire Service Water

a.

Efflumeii Line Monitor (RIA-35)

D)A MOAN(3) qu(l)'

1).

Hiin i uin Flow Device NA AN NA 4.

113 Chem ical Trva [went Pond Couilos i Le SanyaI e r DA NA AN NA 5.

Waste Gas Ho Idtip SystLem

.i Noble (;is Activity Monitor P rov i d i ag Al aria and Au [omat i c Termiinaiona of Release (NIA-37,

-38)

DA AN(3)

IUl I. Effhwteta Flow Hate Moni [or (Wastte Gas Discharge Flow)

NA AN NA 6;.

Uhti t. Venl Motti torisig it.

Nole Gas Activity Ionti [or (RIA-lp5)

DA Ho AN(3)

QIJ(2) 1).

I oi ae Sampl er D iA NA NA NA C, Pa rt icuate SamplIe r D)A NA NA NA

-u ~

11 FIlluia~lt lIOW Rilte Monkitor (MLtt Vent F I w)

D A NA AN NA ID Fl intmim low Dt-v i.cc lDA NA AN NA V.

TABLE 4.1-4(CONTINUED)

HAD I Acrm.i yEFFLUENT MION I TOR INC,.INSTRtUMENTAT7I ON SURVEI LLANCE REQIIREMENIS CHANNEL

.CIIANNEl.

RESPONSE

SOURCE CHANNEL IFUNCTIONAL.

Ns'mHLIENT

-ChECK (4)

CHECK CALl BRATION TS

7.

Interim Iii Utwaste Bui lding VeitLi Jlton Moit orinig it Noble Gas Activity Monijior (HIA-53)

DA NO AN (3)

()

1V modinae Samapler#i DA NA NA NA C.

Plarticaitate Samipler#/

DA NA NA NA' d.

E Cf luicat Flow Rate Honi br

( Interin Hadwaste ExhausL)I D

IA NA AN NA C.

HfliI~imuIII Flow Device#I DA NA AN NA 7 8.

111 Mahinei i Shop

a.

Iodim! Sampler/#

DA NA NA NA

1)

Pa n ctiiiaLe Sampled/

DA NA NA NA C.

E-l fli ient Flow Rate floni br (Hlol Ma chi hae Shop Exliaus1)1/

DA NA AN NA

d.

HfliIIIIIini Flow DeCvice/I DA NA AN NA

Dcluriung each release vi a thiis pathway.

lEf fect i e upo)01 iustLa I Ia Li os of eq i pinch L.

FreItincyNolat jolt DA Da~ i h y HO)

Mon 1hl1y IT Comapleted prior to each release QU

-Qiiarteriiy AN.-

Ammaia I I y NA

-Not Applicable

TABLE 4.1-4 (Continued)

TABLE NOTATION (1) The Channel Functional Test shall also demonstrate that automatic isola tion of this pathway and control room alarm annunciation occurs if any of the following conditions exists:

1.

Instrument indicates measured levels above the alarm/trip setpoint.

2.

Circuit failure (downscale only).

(2) The Channel Functional Test shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists.

1.

Instrument indicates measured levels above the alarm setpoint.

2.

Circuit failure (downscale only).

(3) The initial Channel Calibration shall be performed using one or more of the. reference standards certified by the National Bureau of Standards or using standards that have been obtained from suppliers that par ticipate in measurement assurance activities with NBS.

The standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent Channel Calibration sources, that have been related to the initial calibration shall be used.

(Op erating plants may substitute previously established.calibration pro cedures for this requirement).

(4) The Channel Response Check shall consist of verifying indications during periods of. release. Channel Response Check shall be made at least once per 24-hours on days on which continuous, periodic, or batch releases are made.

Amendments Nos. 125

, 125

, & 122 4.1-18 New Page

4.11 RADIOLOGICAL ENVIRONMENTAL MONITORING Applicability Applies to the surveillance of the station environ for radiation and radioactive materials attributable to station operation and effluent releases.

Soecification 4.11.1 Radiological Environmental Monitoring Program

a.

The radiological environmental monitoring samples shall be collected in accordance with Table 4.11-1 and shall be analyzed pursuant to the requirements of Tables 4.11-1, 4.11-2 and 4.11-3.

b.

If the radiological environmental monitoring program is not conducted as required, a description of the reason for not con ducting the program as required and plans to prevent a recurrence shall be included in the Annual Radiological Environmental Operating Report. Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, seasonal unavailability, or to malfunction of automatic sampling equipment. If the latter, every-effort shall be made to complete corrective action. prior to the end of the next sampling period.

c.

If samples become permanently unavailable from any of the re quired sample locations, the locations from which samples were unavailable may then be deleted from the program provided re placement samples were obtained and added to the environmental monitoring program, if available. These new locations will be identified in the next semi-annual report..

4.11.2 Land Use Census

a.

A land use census shall be conducted and shall identify the location of the nearest milk animal and the nearest residence in each of the 16 meterological sectors within a distance of five miles.

Broad leaf vegetation sampling shall be performed at the site boundary in the direction sector with the highest D/Q in lieu of the.garden census.

b.

If a land use census identifies a location which yields a cal-'

culated dose or dose commitment (via the same exposure pathway) greater than a location from which samples are currently being obtained pursuant to Specification 4.11.1, then the new location shall be added to the radiological environmental monitoring program within 30 days. The sampling.location having the lowest calculated.dose or dose commitment (via the same exposure pathway) may be deleted from this monitoring program after October 31 of the year in which this land use census was conducted. These new locations will be identified in the next semi-annual report.

4.11-1 Amendments Nos..125,125, & 122

C.

The land use census shall be conducted during the growing season at least once per 12 months using that information that will provide the best results, such as by a door-to door survey, aerial survey, or by consulting local agricul ture authorities. The results of the land use census shall be included in the Annual Radiological Environmental Operat ing Report.

4.11.3 Interlaboratory Comparison Program

a.

Analyses shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program which has been approved by the NRC.

b.

If these analyses are not performed as required, report cor rective actions in the Annual Radiological Environmental Ope rating Report.

c.

A summary of the results obtained as part of the above required Interlaboratory Comparison Program and in accordance with the methodology and parameters in the ODCM shall be included in the Annual Radiological Environmental Operating Report.

4.11.4 The provisions of Technical Specification 3.0 do not apply.

Bases.

The environmental monitoring program required by this specification provides measurements of radiation and of radioactive materials in those exposure path ways and for those radionuclides which lead to the highest potential radiation exposures of individuals resulting from the station operation.

This monitoring program thereby supplements the radiological effluent monitoring program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and modeling of program will be effective for at least the first three years of commercial operation.

Following this period, program changes may be initiated-based on operational experience.

The detection capabilities required by Table 4.11-1 are considered optimum for routine environmental measurements in industrial laboratories. The specified lower limits of detection correspond to less than the 10CFR50, Appendix I, design objective dose-equivalent to 45 mrem/year for atmospheric releases to the most sensitive organ and individual.

The land use census specification is provided to assure that changes in the use of unrestricted areas are identified and that modifications to the monitor ing program are made if required by the-results of this census.

The requirements for participation in an Interlaboratory Comparison Program is

-provided to assure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are per formed as part of a quality assurance program for environmental monitoring.n order to demonstrate that the results are reasonably valid.

Amendments Nos.

125

, 125, & 122 4.11-2

TABLE 4. 11-1 EAI~O1.GI AL ENVIRONMIENTAL. MONITORI NC PROGHA CD C+Exposuire Pathlway Nuibe r of Sampl.ing anrd Type atli(

IFrcqu.eairy and/or SanpIe Sample Loain"Coll recjueucy of AnaIsisc (A

.1.

A I BORNEi ra a.

ad io od ie c Li1oca Iions Continuous operatIion of Radi10 mline caniistecr.

andi ParLiculaLes sampler wiLh) sample coi1-Gammua isotop)i c aiia ly r

leclion as required by sis for ]-E)1 or eachi dusl loading bhut at sample.

-Parni ctiale least once per 7days.

s amplIe r.

Gamma isotopic' Q0 analysis on each sample.

2.

1)1IRECTfION RADIAT ION

40) Local ions Countiotis integra lion

- Gaiiuia dlose oii eachn wilth collection at dosimeter.

-least once per 92 (lays.

3.

WATEMORNE

a.

Suirface 2 Locations Coiiposi Le'- sample col-Giamma isotopic analysis lected over a period of each comipsite sample of < 31 days, by locatioii.

rr itLi in analyses of: composite camlp Ic at le-ast ojice per 92 days.,

/

b.

D)rii k iig 31 Loca istis Compos ite-sample Cot1-CGross 1wLa aud gamma I ected over a period isotop ic aal ys is ot, of < 31 flays.

-each comphosite cSample.

Tr L uima maayi of o7mposi~ s campl acit 1least oice jw r 92 clays.

'.UIUJpos itv C

amp Ies shial b e col ecLed hjy coll 1ectinig an al ipiloL at intierval s ]iot exceeding 2 hiouirs.

Coa i

i ati(

id iIc-i Ie circ ofa yeis si u tu onlyif difrn fromu cotlecLion FrU lny

TABLE 4.11-1 (Continued)

RA)IOLOGICAL ENVIRONMENTAL HONITOR I NG PROGRAN Exposure PaLhway Number of Samp I ing and Type and Frespsecy and/or Sampl)e Sample Locat i.ons**

Collection Frequency of Analysis**

LA

c.

Sediment from 2 Locations.

At least once per 184 Gamma isotopic analysis Shoreline days.

of each sample.

4.

INGESTION

a.

Milk 3 Locations At least once per 15 Ganuna isotopic and 1-131 days when animals are analysis of each sample.

on pasture; at least once per 31 days at other Limes.

b.

Fist 2 Locations At least once per 184 Gammata isotopic analysis (lays.

on edible portion.

One sample of each of Lte following species:

1.

Bass

2.

Catfish

c.

Broad-leaf 2 Locations At least once per 31 Ganma isotopic analysis.

Vegetation days.

Composite samples shall be collected by collecting an aliquot at intervals not exceeding 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

l ocaLionis are identified is the ODCM

- Frequeny of ainalysis staLted only iif different from collecLion frequency.

TiABILE4.11-2 IIAXIMUH VALUES FOR THlE LOWER LItIITS OF DEJECTION (LLD)ac (n1 ParLictilaie.

Broadlea f Wa ter or Gas Fish Hilk Vegetation cieh Aalis (pCi/1)~/u)

(pC i/kr, we L (PC j/I)

___(jCi/kgwet.)

pikjr gross beLa 4

32000S

-I 54 130

(.3 l

130 59 30 260 Fe 58, 60 15 130 65, 30 260 9Zr 30 95 151)I 1.3 1s1 7 x 10 2 160

134, 1-37 CS 15,18 5,6 x 10-2 130,150 15,18 60,80 15,80 140 Ila 60 60 140.,

L5 IS

TABLE 4.11-2 (Continued)

TABLE NOTATION a

The LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample with 95% probability of detection and with 5% probability of falsely concluding that a blank observation represents a "real" signal.

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

LLD=

4.66 sb E

V 2.22 Y

exp(-Adt) where.

LLD is the lower limit of detection as defined above (as pCi per unit mass or volume) sb is the standard deviation of, the background counting rate or. of the counting rate of a blank sample as appro priate (as counts per minute)

E is the counting efficiency (as counts per disintegration)

V is the sample size (in units of mass or volume) 2.22 is the number of disintegrations per minute per picocurie Y is the fractional radiochemical yield (when applicable) x is the radioactive decay'constant for the particular radionuclide.

At is the elapsed time between sample collection (or end of the sample collection period) and time of counting Typical values of E, V, Y, and At should be used in the cal culation.

The LLD is defined as an a priori (before the fact) limit represent ing the capability of a measurement system and not as a posteriori (after the fact) limit for a particular measurement.

Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions.

Occasionally background fluctuations, unavoidably small sample sizes, the presence of inter ferring auclides, or other uncontrollable circumstances, may render these LLDs unachievable.

In such cases, the contributing factors will be identified and described in the. Annual Radiological Environ mental Operating Report.

4.11-6 New Page Amendments Nos. 125,..125, & 122

TABLE 4.11-2 (Continued)

TABLE NOTATION b

LLD for gamma isotopic analysis for 1-131 in drinking water samples.

Low level 1-131 analysis on drinking water will not be routinely performed because the calculated dose from 1-131 in drinking water at all locations is less than 1 mrem per year. Low level I-131 analyses will be performed if abnormal releases occur which could reasonably result in > 1 pCi/liter of 1-131 in drinking-water. For low level analyses of 1-131 an LLD of. 1 pCi/liter will be achieved.

c Other peaks which are measurable and identifiable, together with the radionuclides in Table 4.11-2, shall be identified and reported.

Amendments Nos. 125,

125,

& 122 4.11-7 New Page

m TABLE 4.11-3

'D*R PI1 N

.EES FOR RADI OACTI(VI TY CONCENTRATIONS IN ENV IRONMENTAL SAMPLES 0

Reporting LeVels Water Airborne Particulate Fish Milk Broad Leaf Vegetation (pLi rGa inO (pCi/K&Iset)

(Pci/I)

(pCj/K

-Lt) 11-3 2 x 104 Mna-.54.

I x 10-1 3 x 104 Fe-59 4 k 102 1, X 104 4

Co-60 3 x 102

.1 x10 Zji-65 3 x 102 2 x10 Zr-Nb-95 4 x 102~

1-11

.21.0

.3 1x 102 C%-1:34 30 1(

I x Io3 60 1 x 103:

Cs 7 50

.20 2 x 103 70(

2 x 103 1a -La-14 0 2 x 102

.3 x 102 09

'A4or drinkinig water samnples.

Tis i s 40 CFI? PartL 141 value.

l Aow I t-vel1 J-131 analyses are perf ormed.

4.21 DOSE CALCULATIONS Applicability Applies to the projected and cummulative dose contributions from all radio active liquid and gaseous effluents.

Specification 4.21.1 Dose From All Sources The annual (calendar year) dose or dose commitment to any Member Of The Public due to releases of radioactivity and to radiation from uranium fuel cycle sources shall be limited to less than or equal to 25 mrems to the total body or any organ, except the thyroid, which shall be.limited to less than or equal to 75 mrems.

a.

With the calculated doses.from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Specification 3.9.2.a, 3.10.2.a, or 3.10.2.b, cal culations should be made including direct radiation contribu tions from the reactor units and from outside storage tanks to determine whether the above limits of Specification-4.21.1 have been exceeded.-

If such is the case in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days, pursuant to Specific'ation 6.6.3, a-Special Report that defines the corrective action to be taken.to reduce subsequent releases to prevent redurrence of exceeding the above limits and includes the schedule for achieving conformance with the above limits. This Special Report, as defined in 10 CFR Part 20.405c, shall include an analysis that estimates the radiation exposure (dose) to a Member Of The Public from uranium fuel cycle sources, (including all effluent pathways and direct radiation, for the calendar year that includes the release(s) covered by this report. It shall also describe the levels of radiation and concentration of radioactive material involved, and the cause of the exposure levels or concentrations.

If the estimated dose(s) exceeds the above limits, and if the release condition resulting in violation of 40 CFR Part 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR Part 190.

Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete.

b.

The provisions of Technical Specification 3.0 do not apply.

Amendments Nos.

125, 125, & 122 4.21-1 New Page

4.21.2 Dose Due to Liquid Effluents

a.

Monthly, cummulative dose contributions from liquid effluents shall be determined in accordance with the Offsite Dose Calcu lation Manual.

4.21.3 Dose Due to Gaseous Effluents

a.

Monthly, cummulative dose contributions from gaseous effluents shall be determined in accordance with the Offsite Dose Calcu lation Manual.

Amendments Nos.

125,125

, & 122 4.21-2 New Page

5 DESIGN FEATURES 5.1 SITE 5.1.1 The Oconee Nuclear Station is approximately eight miles northeast of Seneca, South Carolina. Figure 2-3 of the Oconee FSAR shows the plan of the site. The minimum distance from the reactor center line to the boundary of the exclusion area and to the outer boundary of the low population zone as defined in 10 CFR 100.3.,

shall be one mile and six miles respectively.

5.1.2

.For the purposes of satisfying 10 CFR Part 20, the "Restricted Area," for gaseous release purposes only, is the same as the ex clusion area as defined above.

REFERENCE (1) FSAR, Chapter 2 (2) Technical Specification 3.10.

Amendments Nos.

125,125-.. & 122 5.1-1

years of the remaining -five years of experience may be fulfilled by aca'demic training,- o? related technical training on a one-for-one time basis.

The Operating Engineer shall hold a Senior Reactor Operator license.

6.1.1.5 Retraining and replacement of station personnel shall be in accordance with Section 5.5 of the ANSI/ANS-3.1-1978, "Selection and Training of Nuclear Power Plant Personnel."

6.1.1..6 A training, program for the. fire brigade shall meet or exceed the requirements of Section 27 of the NFPA Code-1975, except that training sessions may be held quarterly.

6.1.1.7 The two functions of the Shift Technical Advisor, namely accident assessment and operating experience assessment, are fulfilled in the following manner:

a.

An experienced SRO, who has been instructed in additional academic subjects, will be assigned on-shift to provide the accident assessment capability.

b.

The operating experience assessment function will be provided by the Station -Safety-Review-Group.

Amendments Nos. 125, 125

& 122 6.1-la New Page

6.1.2 Technical Review and Control 6.1.2.1 Activities

a.

Procedures required by Technical Specification 6.4 and other procedures which affect station nuclear safety, and changes (other than editorial or typographical changes) thereto, shall be prepared by a qualified individual/organization. Each such procedure, or procedure change, shall be reviewed by an individual/group other than the individual/group which prepared the procedure, or procedure change, but.who may be from the same organization as the individual/group which prepared the procedure, or procedure change. Such procedures and procedure changes may be-approved for temporary use by two members of the station.staff, at least one of whom holds a Senior Reactor Operator's License on the unit(s) affected.

Procedures and procedure changes shall be approved prior to use or within seven days of receiving temporary approval for use by the station Manager; or by the Operating Superintendent, the Technical Services Superintendent or the Maintenance Superintendent, as previously designated by the Station Manager.

b.

Proposed changes to the Technical Specifications shall be prepared by a qualified individual/organization. The preparation of each proposed Tech nical Specifications change shall be reviewed-by an individual/group other than the individual/group which prepared the proposed change, but who may be:from.the same organization.as.the individual/group which prepared the' proposed change. Proposed changes to the Technical Specifications shall be approved by the Station Manager.

c.

Proposed modifications to station nuclear safety-related structures, systems and components shall be designed by a. qualified individual/

organization. Each such modification shall be reviewed by an individual/

group other than the individual/group which designed the modification, but who may be from the same organization as the individual/group which designed the modificationw. Proposed modifications to station nuclear safety-related structures, systems and components shall be-approved prior to implementation by the Station.Manager; or by the Operating Superintendent, the Technical Services Superintendent, or the Maintenance Superintendent, as previously designated by the Station Manager.

d.

Individuals responsible for reviews performed in accordance with 6.1.2.1.a, 6.1.2.1.b, and 6.1.2.1.c shall be members of the station supervisory staff, previously designated by the Station Manager to perform such reviews.

Each such review shall include a determination of whether or not additional, cross-disciplinary, review is necessary. If deemed necessary, such review shall be performed by the appropriate designated station review personnel.

e-.

Proposed tests and experiments which affect station nuclear safety and are not addressed in.the FSAR or Technical Specifications shall be reviewed by the Station Manager; or by the Operating Superintendent, the Technical Services Superintendent or the Maintenance Superintendent, as previously designated by the Station Manager.

Amendments Nos. 125

, 125

, & 122 6.1-2

f.

Incidents reportable pursuant to Technical Specification 6.6.2.1 and vio lations of Technical Specilitations shall be investigated and a report prepared which evaluates the occurrence and which provides recommendations to prevent recurrence.

Such reports shall be approved by the station Manager and transmitted to.the Vice President, Nuclear Production Department, or his designee; and to the Director of th6 Nuclear Safety Review Board.

g.

The Station Manager shall assure the performance of special reviews and investigations, and the preparation and submittal of reports thereon, as requested by the Vice President, Nuclear Production Department.

h.

The station security program', and implementing procedures, shall be re viewed at least annually. Changes determined to be necessary as a result of such review shall be approved by the Station. Manager and transmitted to the Vice President, Nuclear Production Department,.or his designee; and to the Director of the Nuclear Safety Review Board.

i.

The station emergency plan, and implementing procedures, shall.be reviewed at least annually. Changes determined to be necessary as a result of such review shall be approved by the Station Manager and transmitted to the Vice President, Nuclear Production Department, or his designee; and the Director of the Nuclear Safety Review Board.

j.

The Station Manager shall assure that an independent fire protection and loss prevention inspection and audit shall be performed annually utilizing qualified off-site personnel and that an inspection and audit by a qualified fire consultant shall be performed at intervals no greater than three years..

k. Unplanned onsite releases of radioactive material to the environs shall be investigated and a report prepared which evaluates the occurrence and which provides recommendations to prevent recurrence. Such reports shall be approved by the Station Manager and transmitted to the Vice President, Nuclear Production Department, or his designee; and to the Director of the Nuclear Safety Review Board.

1. Proposed changes to the Offsite Dose Calculation Manual (ODCM) shall be prepared by a qualified individual/organization.

Each proposed change shall be reviewed by an individual/group other than the individual group which prepared the proposed change, but who may be from the same organiza tion as the individual/group which prepared. the proposed change. Pro posed. changes to the ODCM shall be approved by the Station Manager prior to implementation.

6.1.2.2 Records Records of the above activities shall be maintained.

Amendments Nos. 125,125, & 122 6.1-3

6.1.3 Nuclear Safety Review Bo-ard 6.1.3.1 Function The NSRB shall function to provide independent review and audit of designated activities in the areas of:

a.

Nuclear power plant operations

b.

Nuclear Engineering

c.

Chemistry and radiochemistry

d.

Metallurgy

e.

Instrumentation and control

f.

Radiological safety

g.

Mechanical and.electrical engineering

h.

Administrative control and quality assurance practices 6.1.3.2 Organization

a.

The Director, members and alternate members of the NSRB shall be formally appointed by the Vice President, Nuclear Production Department, and shall have an academic degree in an engineering or physical science field; and in addition, shall have a minimum of five years technical experience, of

--- which: a minimum. of three years shall be in one or more areas givenn 6.1.3.1.

b.

The NSRB shall. be composed of at least five members, including the Director, Members of the NSRB may be from the Nuclear Production Depart ment, from other departments within the Company or from external to the Company. A maximum of one member of the NSRB may be from the Oconee Nuclear Station staff.

c.

Consultants may be utilized by the NSRB to provide expert advice to the NSRB, as determined necessary by-the Director of the NSRB.

d.

Staff assistance may be provided to the NSRB in order to promote the proper, timely and expeditious performance of its functions.

e.

The NSRB shall meet at least once per six months. The period between such meetings shall not exceed eight months.

f.

A quorum of the NSRB shall consist of the Director, or his designated alternate, and at least two other NSRB members or alternate members.

No more than a:minority of the quorum shall have line responsibility for operation of Oconee Nuclear Station.

New Page Amendments Nos. 125,125,& 122 6.1-3a-

6.1.3.3 Subjects Requiring Review The following subjects shall be reported to and reviewed by the NSRB:

a.

The safety evaluations for (1) changes to procedures, equipment or systems, and (2) tests or experiments completed under the provisions of 10 CFR 50.59(a)(1) to verify that such actions did not constitute an unreviewed safety question.

b.

Proposed changes to procedures, equipment or systems which involve an un reviewed safety question as defined in 10 CFR 50.59.

c.

Proposed tests or experiments which involve an unreviewed safety question as defined in 10 CFR 50.59.

d.

Proposed changes in Technical Specifications or the Facility Operating.

Licenses.

e.

Violations of applicable statutes, codes, regulations, orders, Technical Specifications, license requirements, or of internal procedures or.

instructions having nuclear safety significance.

f.

Significant operating abnormalities or deviations from normal and ect~gg.4_

performance of station equipment that affect nuclear safety.

g.

Incidents that ire the-subject of non-routine:reports submitted to the Commission.

h.

Quality Assurance Department audits relating to station operations and actions taken in response to these audits.

Amendments Nos.

125, 125, & 122

.1-4

6.1.3.4 Audits Audits of station activities shall be performed under the cognizance of the NSRB. These audits shall encompass:

a.

The conformance of station operation to provisions contained within the Technical Specifications and applicable facility operating license conditions at least once per year.

b.

The performance, training and qualifications of the station staff at least once per year.

c.

The results of actions taken to correct deficiencies occurring in equip ment, structures, systems or methods of operation that affect nuclear safety at least once per six months.

d.

The performance of activities required by the quality assurance program to meet the criteria of Appendix B to 10 CFR 50 at least once per two years.

e.

The station emergency plan and implementing procedures at least once per 12 months.

f.

The station security plan and implementing procedurzs at least once per 12 months.

g.

Any other area of' station operation considered appropriate by the NSRB or the.Vice President, Nuclear Production Department.

h.

The station fire protection program and implementing procedures at least once per 24 months.

i.

The Offsite Dose Calculation Manual and implementing procedures at least once per 24 months.

j.

The Radiological Environmental Monitoring Program and the cesults thereof at least once per 12 months.

k.

The Process Control Program and implementing procedures for solidifica tion of radioactive wastes at least once per 24 months.

1.

The performance of activities required by the Quality Assurance Program to meet the criteria of Regulatory Guide 1.21 Revision 1, June 1974 and Regulatory Guide 4.1 Revision 1, April 1975 at least once per 12 months.

Amendments Nos.

125,125, & 122 6.1-5

6.1.3.5 Responsibilities and Authorities

a.

The NSRB shall report to and advise the Vice President, Nuclear Produc tion Department on those areas of responsibility specified in Specifica tions 6.1.3.3 and 6.1.3.4.

b.

Minutes shall be prepared and forwarded to the Vice President, Nuclear Production Department, and to the Executive Vice President, Power Opera tions, within 14 days following.each formal meeting of the NSRB.

c.

Records of activities performed in accordance with Specifications 6.1.3.3 and 6.1.3.4 shall be maintained.

d.

Audit reports. encompassed by Section 6.1.3.4 shall be forwarded to the Vice President, Nuclear Production Department, and to the Executive Vice Presi dent, Power Operations and to the management position responsible for the areas audited within 30 days of completion of each audit.

Amendments Nos. 125,125, &122 6.1-5a New Page

6.4 STATION OPERATING PROCEDURES Soecification 6.4.1 The station shall be operated and maintained in accordance with approved pro cedures.

Written procedures with appropriate check-off lists and instructions shall be provided for the following conditions:

a.

Normal startup, operation, and shutdown of the complete facility and of all systems and components involving nuclear safety of the facility.

b.

Refueling operations.

C.

Actions taken to correct specific and foreseen potential malfunc tions of systems or components involving nuclear safety and radia tion levels, including responses to alarms, suspected primary system leaks and abnormal reactivity changes.

d.

Emergency procedures involving potential or actual release of radioactivity.

e.

Preventive or corrective maintenance which could affect nuclear safety or radiation exposure to personnel.

f.

Station survey following an earthquake.

g.

Personnel radiation protection procedures.

h.

Operation of radioactive waste management systems.

i.

Control of pH in recirculated coolant after loss-of-coolant accident. Procedure shall state that pH will be.measured and the addition of appropriate caustic to coolant will commence within 30 minutes after switchover to recirculation mode of core cooling to adjust the pH to a range of 7.0 to 8.0 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

j.

Nuclear safety-related periodic test procedures.

k.

Long-term emergency core cooling systems.

Procedures shall include provision for remote or local operation of system components necessary to establish high and low pressure in jection within 15 minutes after a line break.

1.

Fire Protection Program implementation.

m,.

Offsite Dose Calculation Manual implementation.

a.

Process Control Program implementation.

Amendments Nos. 125, 125

, & 122 6.4-1

6.4.2 A respiratory protective program approved by the Commission shall be in force.

Amendments Nos. 125, 125, & 122

6. 4-2 New Page

h.

Bj-product material inventory records.

i.

Minutes of Nuclear Safety Review Board Meetings.

j.

Training records.

k.

Test results, in units of microcuries, for leak tests performed pur suant to Specification 4.16.

1.

Radioactive liquid effluent, gaseous effluent, and gaseous process monitoring instrumentation alarm/trip setpoints.

Amendments Nos. 125

, 125, & 122 6.5-2

6.6 STATION REPORTING REQUIRMENTS 6.6.1 Routine Revorts In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Regional Administrator Region II unless otherwise noted.

6.6.1.1 Startup Report A summary report of unit startup and power escalation testing shall be sub mitted following (1) receipt of an operating license, (2) amendment to the facility license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly alered the nuclear, thermal, or hydraulic performance of the unit. Startup reports shall be submitted (1) within 90 days following completion of the startup test pro gram, (2) 90 days following resumption or commencement of commercial power operation, or (3) nine months following initial criticality, whichever occurs first.

If a startup report does not cover'all three events, i.e., initial criticality, completion of the startup test program and resumption or commence ment of commercial.power operation supplementary reports shall be submitted at least every three months until all three events are completed.

6.6.1.2 Monthly Operating Report Routine reports of operating statistics and shutdown experience shall be sub mitted on a monthly basis to the Director, Office of Management Information and.Program Control, U.S. Nuclear Regulatory Commission, Washington, D.C.,

20555, with a copy to the appropriate Regional Office, to be submitted by the fifteenth of each month following the calendar month covered by the report.

6.6.1.3 Personnel Exposure.and Monitoring Report Prior to March 1 of each year, a tabulation shall be submitted to the NRC of the number of station, utility and other personnel (including contractors) receiving exposures.greater than 100 mrem/yr and their associated man-rem exposure according to work and job functions, e.g., reactor operations and surveillance, inservice.

inspection, routine maintenance, special maintenance (describe maintenance),

waste processing, and refueling. The dose assignment to various duty functions may be estimates based on-pocket dosimeter, TLD, or film badge measurements.

Small exposures totalling less than 20% of the individual total dose need not be accounted for.

In the aggregate, at least 80% of the total body dose received from external sources shall be assigned to specific major work functions.

6.6.1.4 Radioactive Effluent Release Report Routine Radioactive Effluent Release Reports covering the operating of the unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year.'

Amendments Nos. 125, 125, & 122

The Radioactive Effluent Release Reports shall include a summary of the quan tities of radioactive liquid and-.gaseous effluents and solid waste released from the station.

The Radioactive Effluent Release Reports shall include a summary of the meteorological conditions concurrent with the release of gaseous effluents during each quarter.

The Radioactive Effluent Release Reports shall include an assessment of the radiation doses from radioactive effluents to individuals due to their activities. inside the unrestricted area boundary during the report period.

All assumptions used. in making these assessments (e.g., specific activity, exposure-time and location) shall be included in these reports.

The Radioactive Effluent Release Reports shall include the following infor mation for all unplanned releases to unrestricted areas of-radioactive ma terials in gaseous. and liquid effluents:

a.

A description of the event and equipment involved.

b.

Cause(s) for-the unplanned release.

c.

Actions taken to prevent recurrence.

d.

Consequences of the unplanned release.

The Radioactive Effluent Release Reports shall include an assessment of radia tion doses from the radioactive liquid and gaseous effluents released from the station during each calendar quarter. In addition, the unrestricted area boundary maximum noble gas gamma air and beta air doses shall be evaluated.

The annual average meteorological conditions shall be used for determining the gaseous pathway doses. Approximate and. conservative approximate methods are acceptable. The assessment of radiation doses shall be performed in ac cordance with the Offsite Dose Calculation Manual.

The Radioactive Effluent Release Reports shall include the following infor-mation for each type of solid waste shipped offsite during the report period:

a.

container volume,

b.

total curie quantity (determined by measurement or estimate),

c.

principalradionuclides (determined by measurement or estimate),

d.

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

e.

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

and

f.

solidification agent (e.g., cement, or other approved agents (media)).

The Radioactive Effluent Release Reports shall include a list and description of unplanned releases from the site. to Unrestricted Areas of radioactive.

materials-in gaseous and liquid effluents made during the reporting period.

Amendments Nos. 125 125, & 122-6.6-2

The Radioactive Effluent Release Reports shall include any changes made during the reporting period to the Offsite Dose Calculation Manual (ODCM),

as well as a.listing of new locations for dose calculations and/or environmental monitor ing identified by the land use census pursuant to Specification 4.11.2.

The Radioactive Effluent Release Recort to be submitted 60 days after January 1 of each year shall also include an assessment of radiation doses to the likely most exposed Member Of The Public from reactor releases and other nearby uranium fuel cycle sources (including doses from primary effluent pathways and direct radiation) for the previous calendar year to show conformance with 40 CFR 190, Environmental Radiation Protection Standards for Nuclear Power Operation. Methods for calculating the dose. contribution from liquid and gaseous. effluents are given.ia the.ODCM.

6.6.1.5 Radiological Environmental Monitoring Routine radiological environmental operating reports covering. the operation of the unit during the previous calendar,year shall be submitted prior to May 1 of each year.

The Annual Radiological Environmental Operating Report shall include summaries, interpretations, and statistical evaluation of the results of the-radiological environmental surveillance activities for the report. period, including a com parison with preoperational studies, operational controls (as appropriate),

and previous environmental surveillance reports and an assessment of the ob served impacts of the. plant operation on the environment.

The reports shall.

also include the results of the land use censuses required by Specification 4.11.

If.harmful effects are detected by the monitoring, the report shall.

provide an analysis of the problem and a planned course-of action to alleviate the problem.

The Annual.Radiological Environment Operating Report shall include a summary of the results obtained as part of the required Interlaboratory Comparison Program and in accordance with the ODCM. Alternatively, participants in the EPA cross-check program shall provide the EPA program code designation for the unit.

The Annual Radiological Environmental Operating Report shall include sum marized and tabulated results of the radiological environmental.samples re quired-by Specification 4.11 taken during.the report period. In the event that some results are not available for inclusion with the report, the re port. shall be submitted noting and explaining the reasons for the missing re sults. The missing data shall be submitted as soon as. practical in a supple mentary report.

The initial report shall also include the following:

a summary description of the radiological environmental monitoring program including sampling methods for each sample type, size and physical characteristics-of each sample type, sample preparation methods, analytical methods, and measuring equipment used; a map of all sampling, locations keyed to a table giving distances and direc tions from one reactor;. and.the result of land use censuses required by Specification 4.11. Subsequent reports. shall describe all. substantial changes in these aspects.

6.6-3 Amendments Nos.125..,2

& 122'

(9)

Performance of structures, systems, or components that requires remedial action or corrective measures to prevent operation in a manner less con servative than assumed in the accident analyses in the safety analysis report or technical specifications bases; 'or discovery during unit life of conditions not specifically considered in the safety analysis report or Technical Specifications that require remedial action or corrective measures to prevent the existence or development of an unsafe condition.

b.

Thirty-Day Written Reports The types of events listed below shall be the subject of written reports to the Regional Administrator, Region II, within 30 days of discovery of the event.

(Copy to the Director, Office of Management Information and Program Control.)

(1) Reactor protection system or engineered safety feature instrument settings which are found to be less conservative than those established by the technical specifications but. which do not prevent the fulfillment of the functional requirements of affected systems.

(2)

Conditions leading to operation in a degraded mode permitted by a limiting condition for operation or shutdown required by a limiting condition for operation.

(3) Observed inadequacies in the implementation of administrative or procedural controls during operation of. a unit which could. cause reduction of degree of redundancy provided. in the Reactor Protective System or Engineered Safety Feature Systems.

(4) Occurrence of radioactive material contained in liquid or gaseous holdup tanks in excess of that permitted by the limiting condition for operation established in the technical specifications.

(3) An unplanned offsite release of 1) more than 1 curie of radioactive material in liquid effluents,. 2) more than 150 curies of noble gas in gaseous effluents or 3) more than 0.05. curies of radioiodine in gaseous effluents.

The report of an unplanned offsite release of radioactive material shall include the following information:

1.

A description of the event and equipment involved.

2.

Cause(s) for the unplanned release.

3.

Action taken to prevent recurrence.

4.

Consequences of the unplanned release-.

(6) Measured levels-of radioactivity in an environmental sampling medium determined to exceed the reporting level values of Table 4.11-3 when averaged over any calendar quarter sampling period.

-hen more than one of the radionuclides in Table 4.11-3 are detected in the sampling medium, this report shall be submitted if:

concentration (1) concentration (2)

++

>1.0 limit level (1) limit level (2)

Amendments Nos.

125 125

& 122 6.6-6

When radionuclides other than those in Table 4.11-3 are detected and are the result of plant effluents, this report shall be sub mitted.if the potential annual dose to an individual is equal to or greater than the calendar year objectives of Specifications 3.9 and 3.10. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report.

6.6.2.2 Environmental Monitoring

a.

If individual milk samples show 1-131. concentrations of 10: picocuries per liter or greater, a plan shall be submitted within one week advising the NRC of the proposed action to ensure the plant related annual doses will be within the:design objective of 45 mrem/yr to the thyroid of any individual.

b.

If milk samples collected over a calendar quarter show average concen trations of 4.8 picocuries per liter or greater, a plan shall be sub mitted within 30 days advising the NRC of the proposed action to ensure the plant related annual doses will be within the design objective of 45 mrem/yr to the thyroid of any individual.

Amendments Nos. 125 125

& 122 6.6-6a New Page

6.6.3 Special Reoorts Special reports shall be submitted to the Regional Administrator Region II, within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specifications:

a.

Single Loop Restrictions, Specification 3.1.8

b.

Auxiliary Electrical Systems, Specification 3.7 c.,

Radioactive Liquid Effluents, Dose, Specification 3.9.2 Liquid Waste Treatment, Specification 3.9.3 Chemical Treatment Ponds, Specification 3.9.4

d.

Radioactive Gaseous Effluents, Dose, Specification 3.10.2 Gaseous Radwaste Treatment, Specification 3.10.3

e.

Fire Protection and Detection Systems, Specification 3.17

f.

Reactor Coolant System Surveillance, Inservice Inspection, Specification 4.2.1 Reactor Vessel.Specimen, Specification 4.2.4

g.

Reactor Building Surveillance, Containment Leakage Tests, Specification 4.4.1

h.

Structural Integrity Surveillance, Tendon Surveillance, Specification 4.4.2.2

i.

Radiological Environmental. Monitoring Program, Specification 4.11.1 Land Use Census, Specification 4.11.2

j.

Dose Calculations (40 CFR 190), Specification 4.21 Amendments Nos.

125,125

, & 122 6.6-7

6.S OFFSITE DOSE CALCULATION ANWUAL (ODCM) 6.8.1 The ODC1 shall describe the methodology and parameters to be used in the cal culation of offsite doses due to radioactive gaseous and liquid instrumenta tion alarm/trip setpoints consistent.with the applicable LCO's contained in these Technical Specifications.

The ODCM shall be submitted to the Commission. at the time of proposed Radio logical Effluent Technical Specifications and shall be-subject to review and approval by the Commissionprior to implementation.

6.8.2 Any changes to the ODCM shall be made by the following method:

1.

Shall be submitted to the Commission by inclusion in the semi annual Effluent Release Report for the period in which the change(s) was made and shall contain:

a.

sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information. Information submitted should consist of a package of those pages of the ODCM to be changed with each page numbered and provided with an approval and date box, together with appropriate analyses or evaluations justifying the change(s);

b.

a determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and

c.

documentation of the fact that the change has been reviewed in accordance with Technical Specification 6.1.2.1.(l) and found acceptable by the Station Manager.

2.

Shall become effective upon review and acceptance by the Station tanager after confirmation of receipt unless otherwise acted upon by the Commission through written notification to the licensee.

Amendments Nos. 125

,125, & 122 6.8-1 New Page

ML15113A077

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