Text

Amends 174,174 & 171 to Licenses DPR-38,DPR-47 & DPR-55, Respectively.Amends Consist of Changes to Tech Specs in Response to Applications Dtd 841008,880106 & 0315
	ML16138A665
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Site:	Oconee
Issue date:	06/06/1989
From:	Matthews D; Office of Nuclear Reactor Regulation
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ML16138A666	List: ML16138A665; ML16293A747;
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pS REGU UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 DUKE POWER COMPANY DOCKET NO. 50-269 OCONEE NUCLEAR STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 174 License No. DPR-38

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

A. The applications for amendment to the Oconee Nuclear Station, Unit 1 (the facility) Facility Operating License No. DPR-38 filed by the Duke Power Company (the licensee) dated October 8, 1984, January 6, 1988 and March 15, 1988, as supplemented or revised August 27, 1985, January 30, June 27, August 13, and September 19, 1986; January 18, May 13, September 16 and December 29, 1988; and May 17, 1989, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D. The issuance of this license 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 satisfied.

2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachments to this license amendment, and Paragraph 3.B. of Facility Operating License No. DPR-38 is hereby amended to read as follows:

P CK 0

PDC

-2 3.B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 174, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

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

FOR THE NUCLEAR REGULATORY COMMISSION vid B. Matthews, Director Project Directorate 11-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Technical Specification Changes Date of Issuance: June 6, 1989

-2 3.B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 174, are hereby incorporated in the license. The licensee shall operate the facility in-accordance with the Technical Specifications.

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

FOR THE NUCLEAR REGULATORY COMMISSION Original Signed By:

David B. Matthews David B. Matthews, Director Project Directorate 11-3 Division of Reactor Projects -

I/Il Office of Nuclear Reactor Regulation

Attachment:

Technical Specification Changes Date of Issuance:

June 6, 1989 OFFICIAL RECORD COPY Lz; I I-3 P 4 II-3 SB RXB NRJSICB r5fRo LWiens:bd Hodges SNewberry 05 /88,p 0 5//88 5 t /88 05/cW88 05/Z1/89 OGC PD I DMatthews O05/1b/89

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 DUKE POWER COMPANY DOCKET NO. 50-270 OCONEE NUCLEAR STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 174 License No. DPR-47

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

A. The.applications for amendment to the Oconee Nuclear Station, Unit 2 (the facility) Facility Operating License No. DPR-47 filed by the Duke Power Company (the licensee) dated October 8, 1984, January 6, 1988 and March 15, 1988, as supplemented or revised August 27, 1985, January 30, June 27, August 13, and September 19, 1986; January 18, May 13, September 16 and December 29, 1988; and May 17, 1989, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D. The issuance of this license 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 satisfied.

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

-2 3.B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 174, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

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

FO THE NUCLEAR REGULATORY COMMISSION avid B. Matthews, Director Project Directorate 11-3 Division of Reactor Projects -

I/II Office of Nuclear Reactor Regulation

Attachment:

Technical Specification Changes Date of Issuance:

June 6, 1989

-2 3.B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 174, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

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

FOR THE NUCLEAR REGULATORY COMMISSION Original Signed By.

David B. Matthews David B. Matthews, Director Project Directorate 11-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Technical Specification Changes Date of Issuance:

June 6, 1989 OFFICIAL RECORD COPY LA; I1-3 P

II-3 NR B

4 N4SICB MJlUb~

LWiens:bd eWHodges S ewbe 05

/88 05/

/88 O5 88 05/- 88 05/7j/89 OG PDIA 05 i/*8 j1Matthews

0 UNITED STATES 0

NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 DUKE POWER COMPANY DOCKET NO. 50-287 OCONEE NUCLEAR STATION, UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 171 License No. DPR-55

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

A. The applications for amendment to the Oconee Nuclear Station, Unit 3 (the facility) Facility Operating License No. DPR-55 filed by the Duke Power Company (the licensee) dated October 8, 1984, January 6, 1988 and March 15, 1988, as supplemented or revised August 27, 1985, January 30, June 27, August 13, and September 19, 1986; January 18, May 13, September 16 and December 29, 1988; and May 17, 1989, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D. The issuance of this license 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 satisfied.

2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachments to this license amendment, and Paragraph 3.B. of Facility Operating License No. DPR-55 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.171, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

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

FO THE NUCLEAR REGULATORY COMMISSION David B. Matthews, Director roject Directorate 11-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Technical Specification Changes Date of Issuance:

June 6, 1989

-2 3.B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 171, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

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

FOR THE NUCLEAR REGULATORY COMMISSION Original Signed By; David B. Matthews David B. Matthews, Director Project Directorate 11-3 Division of Reactor Projects -

I/II Office of Nuclear Reactor Regulation

Attachment:

Technical Specification Changes Date of Issuance:

June 6, 1989 OFFICIAL RECORD COPY LA II-3 P. DII-3 NRR.

LB RR: RXB NR(ICB MR0 LWiens:bd

.k~ra-ken WHodges SNewberry 05/ /88 05S/

88 05/ /88 05A//88 05/W89 08 yPDII-3 DMatthews 05"05/-7989

ATTACHMENT TO LICENSE AMENDMENT NO. 174 FACILITY OPERATING LICENSE NO. DPR-38 DOCKET NO. 50-269 AND TO LICENSE AMENDMENT NO. 174 FACILITY OPERATING LICENSE NO. DPR-47 DOCKET NO. 50-270 AND TO LICENSE AMENDMENT NO. 171 FACILITY OPERATING LICENSE NO. DPR-55 DOCKET NO.

50-287 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the areas of change.

Remove Page Insert Page iii iii iv iv V

v vi vi 3.1-24 3.5-44 3.5-44 3.5-45 3.5-45 3.5-46 3.5-46 3.15-1 3.15-1 3.15-2 3.15-2 4.1-8 4.1-8 4.1-8a 4.1-8a 4.1-9 4.1-9 4.12-1 4.12-1

Section Page 3.1.1 Operational Component 3.1-1 3.1.2 Pressurization, Heatup 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 Temperature Coefficient of Reactivity 3.1-17 3.1.8 (Intentionally Blank) 3.1-19 3.1.9 Low Power Physics Testing Restrictions 3.1-20 3.1.10 Control Rod Operation 3.1-21 3.1.11 Shutdown Margin 3.1-23 3.2 HIGH PRESSURE INJECTION AND CHEMICAL ADDITION SYSTEMS 3.2-1 3.3 EMERGENCY CORE COOLING, REACTOR BUILDING COOLING, 3.3-1 REACTOR 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 3.5-31 Actuation Setpoints 3.5.4 Incore Instrumentation 3.5-33 3.5.5 Radioactive Effluent Monitoring Instrumentation 3.5-37 3.5.6 Accident Monitoring Instrumentation 3.5-44 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 OCONEE -

UNITS 1, 2, & 3 Amendment No. 174 (Unit 1)

Amendment No. 174 (Unit 2)

Amendment No. 171 (Unit 3)

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 CONTROL ROOM PRESSURIZATION AND FILTERING SYSTEM AND 3.15-1 PENETRATION ROOM VENTILATION SYSTEMS 3.16 HYDROGEN PURGE SYSTEM 3.16-1 3.17 FIRE PROTECTION AND DETECTION SYSTEMS 3.17-1 4

SURVEILLANCE REQUIREMENTS 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 4.4-1 4.4.2 Structural Integrity 4.4-14 4.4.3 Hydrogen Purge System 4.4-17 4.4.4 Reactor Building Purge System 4.4-20 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 System 4.5-10 4.5.4 Low Pressure Injection System Leakage 4.5-12 4.6 EMERGENCY 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 4.8-1 OCONEE -

Units 1, 2, & 3 iv Amendment No. 174 (Unit 1)

Amendment No. 174 (Unit 2)

Amendment No. 171 (Unit 3)

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 PRESSURIZATION AND 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 5

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 OCONEE -

UNITS 1, 2, & 3 v

Amendment No. 174 (Unit 1)

Amendment No. 174 (Unit 2)

Amendment No. 171 (Unit 3)

LIST OF TABLES Table No.

Page 2.3-1 Reactor Protective System Trip Setting Limits Units 1,2 and 3 2.3-7 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 3.5-39 Conditions 3.5.5-2 Gaseous Process and Effluent Monitoring Instrumentation 3.5-41 Operating Conditions 3.5.6-1 Accident Monitoring Instrumentation 3.5-45 3.7-1 Operability Requirements for the Emergency Power 3.7-14 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 4.4-1 List of Penetrations with 10CFR50 Appendix J Test Requirements 4.4-6 4.11-1 Radiological Environmental Monitoring Program 4.11-3 4.11-2 Maximum Values for the Lower Limits of Detection (LLD) 4.11-5 4.11-3 Reporting Levels for Radioactivity Concentrations in Environmental Samples 4.11-8 4.17-1 Steam Generator Tube Inspection 4.17-6 6.1-1 Minimum Operating Shift Requirements with Fuel in Three Reactor Vessels 6.1-6 OCONEE -

UNITS 1, 2, & 3 vi Amendment No. 174 (Unit 1)

Amendment No. 174 (Unit 2)

Amendment No. 171,(Unit 3)

3.5.6 Accident Monitoring Instrumentation Applicability Applies to accident monitoring instrumentation.

Objective To ensure that sufficient information is available on selected plant parameters to monitor and assess such parameters following an accident.

Specifications 3.5.6.1 The accident monitoring instrumentation shown in Table 3.5.6-1 shall be operable per applicability indicated in the Table. The provisions of Technical Specification 3.0 do not apply.

3.5.6.2 In the event that the number of accident monitoring instrumentation channels falls below the limit given in Table 3.5.6-1 Column A; operation shall be limited as specified in Column B.

Bases The operability of the accident monitoring instrumentation for accident conditions as appropriate ensures that sufficient information is available on selected plant parameters to monitor and assess these variables following an accident.

RCS subcooled margin is directly indicated in the control room. Core subcooled margin is indicated on both ICC plasma displays, the 0AC video, and a digital control board meter. Loop A subcooled margin is indicated on one ICC plasma display, the 0AC video, and a digital control board meter. Loop B subcooled margin is indicated on the other ICC plasma display, the OAC video, and a digital control board meter. The 0AC video and the digital control board meters are redundant displays of the same signal.

The operability requirements of the Reactor Coolant System subcooling margin monitors ensures that sufficient information is available to the operators to provide prompt recognition of saturated conditions in the primary coolant system and advanced warning of the approach to inadequate core cooling.

Guidance for these requirements was provided by the NRC letter of July 2, 1980, and derived from the implementation of the TMI-2 lessons learned program.

Temperature indications from all 24 qualified core exit thermcouples can be displayed on the 0AC. 12 qualified core exit thermcouples per train will input to each train of process electronics and can be displayed on the respective ICC plasma display.

OCONEE -

Units 1, 2, & 3 3.5-44 Amendment No. 174 (Unit 1)

Amendment No. 174 (Unit 2)

Amendment No. 171 (Unit 3)

Table 3.5.6-1 ACCIDENT MONITORING INSTRUMENTATION C) i-(A)

(B)

(C)

Required Operable Instrument Channels Act ion Appli cabl ity

1.

Containment Pressure Monitor (PT-230, -231) 2 of 2 1

Above hot shutdown re

2.

Containment Water Level Monitor Wide Range (LT-90, -91) 2 of 2 2

Above hot shutdown

3.

Containment High-Range Radiation Monitor (RIA-57, -58) 2 of 2 2

Above hot shutdown

4.

Containment Hydrogen Monitor (MT-80,

-81) 2 of 2

.2 Above hot shutdown

5.

Wide Range Hot Leg Level (RC-LT0123, RC-LT0124) 2 of 2 3

Above hot shutdown

6.

Reactor Vessel Head Level (RC-LT0125, RC-LT0126) 2 of 2 3

Above hot shutdown

7.

Qualified Core Exit Thermocouple Trains 2 of 2 (a) 2 Above hot shutdown U'

8.

Subcooling Monitors 2 (b) 4 When RCS temperature a~

3ot1 (D (D (D 0

>300o CL 0- 0 MJ (D.

(D

3

=

C+~

-+

Table 3.5.6-1 (CONTINUED)

ACCIDENT MONITORING INSTRUMENTATION ACTIONS Action 1:

If one channel is inoperable, the channel shall be restored to operable status within 7 days, or the unit shall be in hot shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

If two channels are inoperable, at least one channel shall be restored to operable status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , or the unit shall be in hot shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Action 2:

If one channel is inoperable, the channel shall be restored to operable status within 30 days, or the unit shall be in hot shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

If two channels are inoperable, at least one channel shall be restored to operable status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , or the unit shall be in hot shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Action 3:

If one channel is inoperable, the channel shall be restored to operable status within 7 days, or a report shall be submitted to the Commission within the next 30 days outlining the cause of the inoperability and the plans and schedule for restoring the channel to operable status.

If two channels are inoperable, at least one channel shall be restored to operable status within 7 days, or the unit shall be in hot shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Action 4:

If one of the required channels is inoperable, at least one channel shall be restored to operable status within 30 days or the unit shall be in hot shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and below 300OF within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

If two of the required channels are inoperable, at least one channel shall be restored to operable status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or the unit shall be in hot shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and below 300oF within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

NOTES (a) 5 of 12 qualified core exit thermocouples must be operable per train for a train to be considered operable.

(b) Operable subcooling margin monitors must consist of:

1)

One direct indication for 1 of 2 RCS hot legs and one direct indication for the core; or

2)

One direct indication for each RCS hot leg.

OCONEE -

UNITS 1, 2, & 3 3.5-46 Amendment No. 174 (Unit 1)

Amendment No. 174 (Unit 2)

Amendment No. 171 (Unit 3)

3.15 Control Room Pressurization and Filtering System and Penetration Room Ventilation System Applicability Applies to the Unit 1 and 2, and Unit 3 control room pressurization and filtering systems and the penetration room ventilation system.

Objective To define the conditions necessary to assure operability of the control room pressurization and filtering system and the immediate availability of the penetration room ventilation systems.

Specification 3.15.1 Penetration Room Ventilation Systems

a.

Two trains of the penetration room ventilation systems shall be operable at all times when containment integrity is required or the reactor shall be shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months with the following excep tion:

(1) If one of two trains of a penetration room ventilation system is made or found to be inoperable for any reason, reactor operation is permissible only during the succeeding seven days provided that all active components of the other train of the penetration room ventilation system shall be demonstrated to be operable within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and daily thereafter.

3.15.2 Control Room Pressurization and Filtering Systems

a.

With the reactor above hot shutdown conditions both outside air booster fans shall be operable.

(1) If one outside air booster fan is inoperable, restore the inoperable fan to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or the unit shall be in hot shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

(2) If both outside air booster fans are inoperable, restore at least one inoperable fan to operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or the unit shall be in hot shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

OCONEE -

UNITS 1, 2, & 3 3.15-1 Amendment No. 174 (Unit 1)

Amendment No. 174 (Unit 2)

Amendment No. 171 (Unit 3)

b.

With the reactor above hot shutdown conditions and both outside air booster fans operable, the control room pressurization and filtering systems shall be capable of maintaining a positive pressure within the control room.

(1) If the above requirements of Specification 3.15.2.b are not met within 30 days, the unit shall be in hot shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c.

With the reactor above hot shutdown conditions, both filter trains shall be operable.

(1) If one filter train is inoperable, restore the inoperable filter train to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or the unit shall be in hot shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

(2) If both filter trains are inoperable, restore one inoperable filter train to operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or the unit shall be in hot shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

d.

The provisions of Specification 3.0 do not apply.

Bases A single train of reactor building penetration room ventilation equipment retains full capacity to control and minimize the release of radioactive materials from the reactor building to the environment in post-accident conditions.

The control room pressurization and filtering system is comprised of two separate outside air booster fans with prefilter/HEPA/carbon filter trains, two redundant control room air handling unit fans, and associated ductwork.

The system is designed to protect the control room operators from the effects of accidental release of radioactive effluents or toxic gases in the Turbine or Auxiliary Building.

Protection is provided by pressurizing the control room with filtered outside air to prevent inleakage of radioactive effluents or toxic gases from the Turbine or Auxiliary Building only. Specification 3.15.2.b applies to all instances where the reactor is above hot shutdown and the system is judged incapable of maintaining the control room at a positive pressure or, if during refueling frequency testing per Specification 4.12.1.b the system is demonstrated to be incapable of maintaining the control room at a positive pressure,.

OCONEE -

UNITS 1, 2, & 3 3.15-2 Amendment No. 174 (Unit 1)

Amendment No. 174 (Unit 2)

Amendment No. 171 (Unit 3)

Table 4.1-1 (CONTINUED)

C n Channel Description Check Test Calibrate Remarks a rn m

49.

Emergency Feedwater MO NA RF I

Flow Indicators

50.

PORV and Safety Valve MO NA RF v4 Position Indicators

51.

RPS Anticipatory NA MO RF Reactor Trip System Loss of Turbine Emergency Trip System Pressure.Switches

52. RPS Anticipatory Reactor Trip System Loss of Main Feedwater a) Control Oil Pressure NA MO RF Switches b) Discharge Pressure NA MO RF Switches c,

53. Emergency Feedwater Initiation Circuits a) Control Oil Pressure NA MO RF Switches b) Discharge Pressure NA MO RF Switches

=

54. Containment High Range NA MO RF TI'MI Iein II.F. 1.3 Radiation Monitor o 0 0(RIA-57,

58)

C+ C+ C+

C-Table 4.1-1 (CONTINUED) o Channel Description Check Test Calibrjate Remarks

55.

Containment Pressure MO NA AN TMI Item II.F.1.4 Monitor (PT-230, 231)

56.

Containment Water Level MO NA RIF TMI Item II.F.1.5 Monitor-Wide Range (LT-90, -91)

57.

Containment Hydrogen NA MO AN TMI Item II.F.1.6 Monitor (MT-80,-81)

58. Wide Range Hot Leg Level NA RF RF

59. Reactor Vessel Head Level NA RF RF

60.

Core Exit Thermocouples MO NA RF 00

61.

Subcooling Monitors MO RF RF ES -

Each Shift QU -

Quarterly DA -

Daily AN -

Annually WE -

Weekly PS -

Prior to startup, if not performed previous week MO -

Monthly NA -

Not Applicable RF - Refueling Outage C0 C C F

oo 4i

-J CCC C+.

rJ.

CJ

Table 4.1-2 MINIMUM EQUIPMENT TEST FREQUENCY Item Test Frequency I.

Control Rod Movement Movement of Each Rod Monthly

2.

Pressurizer Safety Valves Setpoint Each Refueling

3.

Main Steam Safety Valves Setpoint Each Refueling 4.

4.

Refueling System Interlocks Functional Prior to Refueling

5.

Main Steam Stop Valves (1)

Movement of Each Stop Monthly Valve

6.

Reactor Coolant System (2)

Evaluate Daily Leakage

7.

Condenser Cooling Water Functional Each Refueling System Gravity Flow Test

8.

High Pressure Service Functional Monthly Water Pumps and Power Supplies

9.

Spent Fuel Cooling System Functional Prior to Refueling

10.

High Pressure and Low (3)

Vent Pump Casings Monthly and Prior Pressure Injection System to Testing

11.

Emergency Feedwater Functional Each Refueling Pump Automatic Start and Automatic Valve Actuation Feature Applicable only when the reactor is critical.

(2) Applicable only when the reactor coolant is above 200oF and at a steady state temperature and pressure.

(3)

Operating pumps excluded.

(4) Number of safety valves to be tested each refueling shall be in accordance with ASME CodesSection XI, Article IWV-3511, such that each valve is tested at least once every 5 years.

OCONEE -

UNITS 1, 2, & 3 4.1-9 Amendment No. 174 (Unit 1)

Amendment No. 174 (Unit 2)

Amendment No. 171 (Unit 3)

4.12 CONTROL ROOM PRESSURIZATION AND FILTERING SYSTEM Applicability Applies to control room pressurization and filtering system components Objective To verify that these systems and components will be able to perform their design functions.

Specification 4.12.1 Operating Tests

a.

Control room outside air booster fan system tests shall be performed quarterly. These tests shall consist of an external visual inspection, a flow measurement for each unit and pressure drop measurements across each filter bank. Pressure drop across pre-filter shall not exceed 1 inch H20 and pressure drop across HEPA shall not exceed 2 inches H20. Fan motors shall be operated continuously for at least one hour, and all louvers shall be proven operable.

b.

On a refueling frequency, verify the system maintains the control room at a positive pressure with both outside air booster fans on during system operation.

4.12.2 Filter Tests On a refueling frequency, for the Unit 1 and 2 and the Unit 3 control room an in-place leakage test using DOP on HEPA units and Freon-112 (or equivalent) on carbon units shall be performed at design flow on each filter train. Removal of 99.5 percent DOP by each entire HEPA filter unit and removal of 99.0 percent Freon-112 (or equivalent) by each entire carbon adsorber unit shall constitute acceptance performance. These tests must also be performed after any maintenance which may affect the structural.integrity of either the filtration system units or of the housing.

Bases.

The purpose of the control room pressurization filtering system is to protect the control room operators from the effects of accidental release of radioac tive effluents or toxic gases in the Turbine Building or Auxiliary Building only. The system is designed with two 50 percent capacity filter trains each of which consists of a prefilter, high efficiency particulate filters, carbon filters, booster fans, air handling unit fans, and associated ductwork to pressurize the control room with outside air.

Since these systems are not normally operated, a periodic test is required to insure their operability when needed. Quarterly testing of this system will show that the system is available.

Refueling frequency testing of the installed carbon adsorber stage and abso lute filters will verify the leak integrity of the cleanup system. Refueling frequency testing will also verify the ability of the system to maintain the control room at a positive pressure to minimize infiltration of hazardous effluents.

OCONEE -

UNITS 1, 2, & 3 4.12-1 Amendment No. 174 Unit 1 Amendment No. 174 Unit 2 Amendment No. 171 (Unit 3)

ML16138A665

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