Text

Amends 165,165 & 162 to Licenses DPR-38,DPR-47 & DPR-55, Respectively,Correcting Typos
	ML16138A613
Person / Time
Site:	Oconee
Issue date:	12/11/1987
From:	Jabbour K; Office of Nuclear Reactor Regulation
To:	;
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ML16138A614	List: ML16138A613; ML16138A615;
References
	NUDOCS 8712220187
	Download: ML16138A613 (28)
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0 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. 165 License No. DPR-38

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

A. The application 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 August 15, 1985, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter 1; 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 amended by 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:

3.8 Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 16, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

8712220187 971211 PDR ADOCK 05000269 P

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

FOR THE NUCLEAR REGULATORY COMMISSION Kahtan N. Jabbour, Acting Director Project Directorate 11-3 Division of Reactor Projects -

I/II

Attachment:

Technical Specification Changes Date of Issuance: December 11, 1987 PI DRP-I/II P

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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. 165 License No. DPR-47

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

A. The application 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 August 15, 1985, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter 1; 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 amended by 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:

3.B Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 165, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

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

FOR THE NUCLEAR REGULATORY COMMISSION Kahtan N. Jabbour, Acting Director Project Directorate 11-3 Division of Reactor Projects -

I/II

Attachment:

Technical Specification Changes Date of Issuance:

December 11, 1987 PpI DRP-I/II 3/DRP-I/II OGC-Bethqsda PDII-3/DRP-I/II Mu ik~/rad H tis Acting Director 121d /87 11/1

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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. 162 License No. DPR-55

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

A. The application 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 August 15, 1985, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter 1; 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 amended by changes to the Technical Specifications as indicated in the attachments to this license amendment, and Paragraph 3.3 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. 162, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

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

FOR THE NUCLEAR REGULATORY COMMISSION Kahtan N. Jabbour, Acting Director Project Directorate 11-3 Division of Reactor Projects -

I/II

Attachment:

Technical Specification Changes Date of Issuance: December 11, 1987 PDI -

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ATTACHMENT TO LICENSE AMENDMENTS AMENDMENT NO. 165 TO DPR-38 AMENDMENT NO. 165 TO DPR-47 AMENDMENT NO. 162 TO DPR-55 DOCKET 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 number and contain vertical lines indicating the areas of change.

Remove Insert Page Page i

i iii iii 3.1-1 3.1-1 3.1-4 3.1-4 3.1-17 3.1-17 3.1-19 3.1-19 3.3-1 3.3-1 3.5-11 3.5-11 3.7-2 3.7-2 3.7-3 3.7-3 3.7-4 3.7-4 4.1-3 4.1-3 4.1-4 4.1-4 4.1-6 4.1-6 4.1-8 4.1-8 5.1-1 5.1-1 6.1-1 6.1-1 6.1-7 6.1-7 6.1-8 6.1-8 6.6-2 6.6-2 6.6-5 6.6-5

TABLE OF CONTENTS Section Page TECHNICAL SPECIFICATIONS 1

DEFINITIONS 1-1 1.1 RATED POWER 1-1 1.2 REACTOR OPERATING CONDITIONS 1-1 1.2.1 Cold Shutdown 1-1 1.2.2 Hot Shutdown 1-1 1.2.3 Reactor Critical 1-1 1.2.4 Hot Standby 1-1 1.2.5 Power Operation 1-1 1.2.6 Refueling Shutdown 1-1 1.2.7 Refueling Operation 1-2 1.2.8 Startup 1-2 1.3 OPERABLE 1-2 1.4 PROTECTIVE INSTRUMENTATION LOGIC 1-2 1.4.1 Instrument Channel 1-2 1.4.2 Reactor Protective System 1-2 1.4.3 Protective Channel 1-2 1.4.4 Reactor Protective System Logic 1-3 1.4.5 Engineered Safety Features System 1-3 1.4.6 Degree of Redundancy 1-3 1.5 INSTRUMENTATION SURVEILLANCE 1-3 1.5.1 Trip Test 1-3 1.5.2 Channel Test 1-3 1.5.3 Instrument Channel Check 1-3 Amendment No. 165 (Unit 1)

Amendment No. 165 (Unit 2)

Amendment No. 162 (Unit 3)

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 3.1.12 Reactor Coolant System Subcooling Margin Monitor 3.1.-24 3.2 HIGH PRESSURE INJECTION AND CHEMICAL ADDITION SYSTEMS 3.2.1 3.3 EMERGENCY 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 3.6 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 iii Amendment No. 165 (Unit 1)

Amendment No. 165 (Unit 2)

Amendment No. 162 (Unit 3)

3.1 REACTOR COOLANT SYSTEM Applicability Applies to the operating status of the reactor coolant system.

Objective To specify those limiting conditions for operation of the reactor coolant system components which must be met to ensure safe reactor operation.

Specification 3.1.1 Operational Components

a.

Reactor Coolant Pumps

1.

Whenever the reactor is critical, single pump operation shall be prohibited, single-loop operation shall be restricted to testing, and other pump combinations permissible for given power levels shall be as shown in Table 2.3-1.

2.

Except for test purposes and limited by Specification 2.3, power operation with one idle reactor coolant pump in each loop shall be restricted to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . If the reactor is not returned to an acceptable RC pump operating combination at the end of the 24-hour period, the reactor shall be in a hot shutdown condition within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

3.

The boron concentration in the reactor coolant system shall not be reduced unless at least one reactor coolant pump or one low pressure injection pump is circulating reactor coolant.

b.

Steam Generator

1.

One steam generator shall be operable whenever the reactor coolant average temperature is above 250 0F.

c.

Pressurizer Safety Valves,

1.

All pressurizer code safety valves shall be operable whenever the reactor is critical.

2.

At least one pressurizer code safety valve shall be operable whenever all reactor coolant system openings are closed, except for hydrostatic tests in accordance with the ASME Section III Boiler and Pressure Vessel Code.

3.1-1 Amendment No. 165 (Unit 1)

Amendment No. 165 (Unit 2)

Amendment No. 162 (Unit.3)

Bases - Units 1, 2 and 3 All components in the Reactor Coolant System are designed to withstand the effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are introduced by normal load transients, reactor trips, startup and shutdown operations, and inservice leak and hydrostatic tests. The various categories of load cycles used for design purposes are provided in Table 5.2-1 of the FSAR.

The major components of the reactor coolant pressure boundary have been analyzed in accordance with Appendix G to 10 CFR 50. Results of this analysis, including the actual pressure-temperature limitations of the reactor coolant pressure boundary, are given in BAW-1699 and BAW-1697.

The Figures specified in 3.1.2.1, 3.1.2.2 and 3.1.2.3 present the pressure-temperature limit curves for normal heatup, normal cooldown and hydrostatic tests respectively. The limit curves are applicable up to the indicated effective full power years of operation. These curves are adjusted by 25 psi and 10OF for possible errors in the pressure and temperature sensing instruments. The pressure limit is also adjusted for the pressure differential between the point of system pressure measurement and the limiting component for all operating reactor coolant pump combinations.

The cooldown limit curves are not applicable to conditions of off-normal operation (e.g., small LOCA and extended loss of feedwater) where cooling is achieved for extended periods of time by circulating water from the HPI through the core. If core cooling is restricted to meet the cooldown limits under other than normal operation, core integrity could be jeopardized.

The pressure-temperature limit lines shown on the figures specified in 3.1.2.1 for reactor criticality and one the figures referred to in 3.1.2.3 for hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50 for reactor criticality and for inservice hydrostatic testing.

The actual shift in RTNDT of the beltline region material will be established periodically during operation by removing and evaluating, in accordance with Appendix H to 10 CFR 50, reactor vessel material irradiation surveillance specimens which are installed near the inside wall of this or a similar reactor vessel in the core region, or in test reactors.

The limitations on steam generator pressure and temperature provides protection against nonductile failure of the secondary side of the steam generator. At metal temperatures lower than the RT of +600F, the protection against nonductile failure is achieved by limiting the secondary coolant pressure to 20 percent of the preoperational system hydrostatic test pressure.

3.1-4 Amendment No. 165 (Unit 1)

Amendment No. 165 (Unit 2)

Amendment No. 162 (Unit 3)

3.1.7 Moderator Temperature Coefficient of Reactivity Specification The moderator temperature coefficient shall not be positive at power levels above 95 percent of rated power.

Bases A non-positive moderator coefficient at power levels above 95% or rated power is specified such that the maximum clad temperatures will not exceed the Final Acceptance Criteria based on LOCA analyses. Below 95% of rated power the Final Acceptance Criteria will not be exceeded with a positive moderator temperature coefficient of +0.9 x 10-4 Ak/k/oF corrected to 95% rated power.

All other accident analyses as reported in the FSAR have been performed for a range of moderator temperature coefficients including +0.9 x 10 4 Ak/koF.

The moderator coefficient is expected to be zero or negative prior to completion of startup tests.

When the hot zero power value is corrected to obtain the hot full power value, the following corrections will be applied.

A.

Uncertainty in isothermal measurement The measured moderator temperature coefficient will contain uncertainty on the account of the following:

1.

+/-0.2 0F in the AT of the base and perturbed conditions.

2.

Uncertainty in the reactivity measurement of +/-0.1 x 10 4 Ak/k.

Proper corrections will be added for the above conditions to result in a conservative moderator coefficient.

B.

Doppler coefficient at hot zero power During the isothermal moderator coefficient measurement at hot zero power, the fuel temperature will increase by the same amount as the moderator. The measured temperature coefficient must be increased by 0.16 x 10 (Ak/k)/oF to obtain pure moderator temperature coefficient.

C.

Moderator temperature change The hot zero power measurement must be reduced by.09 x 104 (Ak/k)/oF. This corrects for the difference in water temperature at zero power (532 0F) and 15% power (5800 F) and for the increased fuel temperature effects at 15% power. Above this power, the average moderator temperature remains 580 0 F. However, the coefficient, must also be adjusted for the interaction of an average moderator temperature with increased fuel temperatures. This correction is

-.001 x 10 4 A i/AT power. It adjusts the 15%

3.1-17 Amendment No. 165 (Unit 1)

Amendment No. 165 (Unit 2)

Amendment No. 162 (Unit 3)

INTENTIONALLY BLANK 3.1-19 Amendment No, 165 (Unit 1)

Amendment No. 165 (Unit 2)

Amendment No. 162 (Unit 3)

3.3 EMERGENCY CORE COOLING, REACTOR BUILDING COOLING, REACTOR BUILDING SPRAY, AND LOW PRESSURE SERVICE WATER SYSTEMS Applicability Applies to the emergency core cooling, reactor building cooling, reactor building spray, and low pressure service water systems.

Objective To define the conditions necessary to assure immediate availability of the emergency core cooling, reactor building cooling, reactor building spray and low pressure service water systems.

Specification 3.3.1 High Pressure Injection (HPI) System

a. Prior to initiating maintenance on any component of the HPI system, the redundant component shall be tested to assure operability.

b. When the reactor coolant system (RCS), with fuel in the core, is in a condition with temperature above 350 0 F and reactor power less than 60% FP:

(1) Two independent trains, each comprised of an HPI pump and a flow path capable of taking suction from the borated water storage tank and discharging into the reactor coolant system automatically upon Engineered Safeguards Protective System (ESPS) actuation (HPI segment) shall be operable.

(2) Test or maintenance shall be allowed on any component of the HPI system provided one train of the HPI system is operable.

If the HPI system is not restored to meet the requirements of Specification 3.3.1.b(1) above within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the reactor shall be placed in a hot shutdown condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

If the requirements of Specification 3.3.1.b(1) are not met within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following hot shutdown, the reactor shall be placed in a condition with RCS temperature below 350 0F within an additional 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

c. For all Units, when reactor power is greater than 60% FP:

(1) In addition to the requirements of Specification 3.3.1.b(1) above, the remaining HPI pump and valves HP-409 and HP-410 shall be operable and valves HP-99 and HP-100 shall be open.

(2) Tests or maintenance shall be allowed on any component of the HPI system, provided two trains of HPI system are operable.

If the inoperable component is not restored to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , reactor power shall be reduced below 60% FP within an additional 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

3.3-1 Amendment No. 165 (Unit 1)

Amendment No. 165 (Unit 2)

Amendment No, 162 (Unit 3)

Bases Operation at power with an inoperable control rod is permitted within the limits provided. These limits assure that an acceptable power distribution is maintained and that the potential effects of rod misalignment on associated accident analyses are minimized. For a rod declared inoperable due to misalignment, the rod with the greatest misalignment shall be evaluated first. Additionally, the position of the rod declared inoperable due to misalignment shall not be included in computing the average position of the group for determining the operability of rods with lesser misalignments. When a control rod is declared inoperable, boration may be initiated to achieve the existence of 1% Ak/k hot shutdown margin.

The power-imbalance envelope defined in Figures 3.5.2-10 (Unit 1) 3.5.2-11 (Unit 2) 3.5.2-12 (Unit 3) is based on LOCA analyses which have defined the maximum linear heat rate (see Figure 3.5.2-16) such that the maximum clad temperature will not exceed the Final Acceptance Criteria. Corrective measures will be taken immediately should the indicated quadrant tilt, rod position, or imbalance be outside their specified boundary. Operation in a situation that would cause the Final Acceptance Criteria to be approached should a LOCA occur is highly improbable because all of the power distribution parameters (quadrant tilt, rod position, and imbalance) must be at their limits while simultaneously all other engineering and uncertainty factors are also at their limits.** Conservatism is introduced by application of:

a.

Nuclear uncertainty factors

b.

Thermal calibration

c.

Fuel densification power spike factors (Units 1 and 2 only)

d.

Hot rod manufacturing tolerance factors

e.

Fuel rod bowing power spike factors The 25% +/- 5% overlap between successive control rod groups is allowed since the worth-of a rod is lower at the upper and lower part of the stroke.

Control rods are arranged in groups or banks defined as follows:

Group Function 1

Safety 2

Safety 3

Safety 4

Safety 5

Regulating 6

Regulating 7

Xenon transient override 8

APSR (axial power shaping rods)

Actual operating limits depend on whether or not incore or excore detectors are used and their respective instrument calibration errors. The method used to define the operating limits is defined in plant operating procedures.

3.5-11 Amendment No. 165 (Unit 1)

Amendment No. 165 (Unit 2)

Amendment No. 162 (Unit 3)

1. Both 125 VDC instrumentation and control distribution centers (DCA and DCB);

2. All four 125 VDC instrumentation and control panelboards (DIA, DIB, DIC, and DID), including the associated isolating transfer diodes and diode monitors (ADA 1 & 2, ADB 1 & 2, ADC 1 & 2, ADD 1 & 2);

3. All four 120 VAC vital instrumentation power panelboards (KVIA, KVIB, KVIC, and KVID), including the associated static inverters;

4. The 240/120 VAC regulated power panelboard (KRA).

Additionally, the 125 VDC instrumentation and control batteries with an associated charger shall be operable as follows:

1. For operation of Unit 1 only, ICA or ICB, and 2CA or 2CB Unit 2 only, 2CA or 2CB, and 3CA or 3CB Unit 3 only, 3CA or 3CB, and ICA or 1CB

2. For operation of any two units, ICA or 1CB, 2CA or 2CB, and 3CA or 3CB.

3. For operation of all three units, five of the six batteries with their associated chargers.

(g) Both of the 125 VDC 230KV switching station batteries (SY-1, SY-2),

with associated chargers, distribution centers, and panel boards shall be operable.

(h) Both of the 125 VDC Keowee batteries (Bank 1 & 2) with as sociated chargers and distribution centers (IDA & 2DA) shall be operable.

(i) The level of Keowee Reservoir shall be at least 775 feet above sea level.

3.7.2 With the reactor heated above 2000F, provisions of 3.7.1 may be modified to allow the following conditions to exist:

(a) One of the two independent on-site emergency power paths, as defined in 3.7.1(b), may be inoperable for periods not exceeding 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months for test or maintenance, provided the alternate power path is verified operable within one hour of the loss and every eight hours thereafter.

(b) The circuits or channels of any single functional unit of the EPSL may be inoperable for test or maintenance for periods not exceeding 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , provided that:

1. The conditions of Table 3.7-1 for degraded operation are satisfied for that specific functional unit; and

2. The conditions of Table 3.7-1 for normal operation are satisfied for all other functional units.

3.7-2 Amendment No. 165 (Unit 1)

Amendment No. 165 (Unit 2)

Amendment No. 162 (Unit 3)

The circuits or channels of more than one functional unit of the EPSL may be inoperable only if:

1.

The inoperability results from a loss of power due to the in operability of a 125 VDC instrumentation and control panelboard (see 3.7.2(e) below); and

2.

The conditions of Table 3.7-1 for degraded operation are satisifed for the affected functional units.

If any event, if the reactor is subcritical, the inoperable circuit(s) or channel(s) shall be restored to operability and the conditions of Table 3.7-1 for normal operation shall be satisifed for all functional units before the reactor is returned to criticality.

(c) One 4160 volt main feeder bus may be inoperable for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

(d) One complete single string (i.e., 4160 volt switchgear (TC, TD, or TE),

600 volt load.center, (X8, X9, or X10), 600-208 volt XS1, XS2, or XS3),

and their loads) of each unit's 4160 volt Engineered Safety Features Power System may be inoperable for hours.

(e) One or more of the following DC distribution components may be in operable for periods not exceeding 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (except as noted in 3.7.2(g) below):

1.

One complete single string or single component (i.e., 125VDC battery, charger, distribution center, and panelboards) of the 125VDC 230KV Switching Station Power System.

2.

One complete single string or single component (i.e., 125VDC battery, charger, and distribution center) of the Keowee 125VDC Power System may be inoperable provided the remaining string of Keowee is operable and electrically connected to an operable Keowee hydro unit.

3.

One complete single string or single component (i.e., 125VDC battery, charger, distribution center, and associated isolating and transfer diodes) of any units 125VDC Instrumentation and Control Power System. Only one battery more than the number allowed to be inoperable per 3.7.1 (f) for the Station may be removed from service under this paragraph.

4.

One 125 VDC instrumentation and control panelboard and its associated loads, per unit, provided that no additional AC buses are made inoperable beyond the provisions of 3.7.2(a), (c), and (d), and provided that the conditions of Table 3.7-1 for normal operation are satisfied for all functional units of the EPSL before the 125 VDC instrumentation and control panelboard becomes inoperable. Additionally, the provisions of 3.7.2.(h) must be observed for the 120 VAC vital instrumentation power panelboard which is powered by the affected 125 VDC panelboard.

A one-time extension of inoperability for a period of 10 days per battery is granted to allow for installation of new Keowee batteries and battery racks.

Amendment No. 165 (Unit 1) 3.73 Amendment No. 165 (Unit 2)

Amendment No. 162 (Unit 3)

(f) For periods not to exceed 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months each unit's 125 VDC system may be separated from its backup unit via the isolating and transfer diodes.

(g) One battery each, from one or more of the following 125VDC systems may be simultaneously inoperable for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in order to perform an equalizer charge after the surveillance requirements of Specification 4.6.10.

1.

230 KV Switching Station 125VDC Power System

2.

Keowee Hydro Station 125VDC Power System

3.

Each unit's 125VDC Instrumentation and Control Power System, provided that the unit's remaining battery is operable. However, for operation of 1 or 2 units, no more batteries than those allowed to be inoperable per 3.7.1 (f) may be removed from service. For operation of 3 units, at least 4 or the 6 station IC& batteries shall be operable.

(h) One 120 VAC vital instrumentation power panelboard per unit and/or its associated static inverter may be inoperable for periods as specified below:

Maximum Allowed Period Panelboard of Inoperability KVIA 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months KVIB 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months KVIC 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months KVID 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months A single vital bus static inverter per unit may continue to be inoperable beyond the specified period, but no longer than 7 days total, provided that its associated 120 VAC vital instrumentation power panelboard is connected to the 240/120 VAC Regulated Power System and verified to be operable once every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

(i) 1.

A startup transformer may be inoperable for periods not exceeding 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months for test or maintenance, provided the underground feeder path, through transformer CT4; and to one 4160V standby bus is verified operable within one hour of loss and every eight hours thereafter. The remaining operable startup transformers can be shared between units within the same 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of the above startup transformer being determined inoperable. Prior to exceeding 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , they shall be aligned and connected such that each one is providing a path for power to one and only one unit.

2.

In the event that a startup transformer becomes inoperable for unplanned reasons, then one unit shall be in cold shutdown within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months with its loads powered from the standby buses. The remaining operable startup transformers can be shared between units within the same 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of the above startup transformer being determined inoperable. Prior to exceeding 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , they shall be aligned and connected such that each one is providing a path for power to one and only one unit.

Amendment No. 165 (Unit 1) 3.7-4 Amendment No. 165 (Unit 2)

Amendment No. 162 (Unit 3)

Table 4.1-1 INSTRUMENT SURVEILLANCE REQUIREMENTS Channel Description Check Test Calibrate Remarks

1. Protective Channel NA MO NA Coincidence Logic

2. Control Rod Drive NA MO NA Trip Breaker

3.

Power Range Amplifier ES(l)

NA (1)

(1) Heat balance check each shift. Heat balance calibration whenever indi cated core thermal power exceeds neutron power by more than 2 percent.

4. Power Range ES MO MO(1)(2)

(1) Using incore instrumentation.

(2) Axial offset upper and lower chambers after each startup if not done pre vious week.

5. Intermediate Range ES(1)

PS NA (1) When in service.

6. Source Range ES(1)

PS NA (1) When in service.

7.

Reactor Coolant ES MO RF M M 0 Temperature

8.

High Reactor Coolant ES MO RF Pressure

9. Low Reactor Coolant ES MO RF Pressure 0N 01 a%

10. Flux-Reactor Coolant ES MO RF Flow Comparator rtrt rt

11.

Reactor Coolant Pressure ES MO RF Temperature Comparator

Table 4.1-1 (CONTINUED)

Channel Description Check Test Calibrate Remarks

12.

Pump-Flux Comparator ES MO RF

13.

High Reactor Building DA MO RF Pressure

14.

High Pressure Injection &

NA MO NA Includes Reactor Building Reactor Building Isolation Isolation of non-essential Logic (Non-essential systems) systems

15.

High Pressure Injection Analog Channels:

a. Reactor Coolant Pressure ES MO RE

b. Reactor Building Pressure (4 psig)

ES MO RE

16.

Low Pressure Injection NA MO NA Logic

17.

Low Pressure Injection Analog Channels:

a. Reactor Coolant Pressure ES MO RF

b. Reactor Building Pressure (4 psig)

ES MO RF

trt r

18.

Reactor Building Emergency NA MO NA Reactor Building isolation Cooling and Isolation includes essential systems System Logic (Essential Systems) t-t-n~

19.

Reactor Building Emergency ES MO RF Cooling and Isolation System Analog Channel Reactor Building Pressure (4 psig)

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Amendment No. 165 (Unit 2) mendmnent N9. 162 (Unit 3)

Table 4.1-1 (CONTINUED)

Channel Description Chedk Test Calibrate Remarks

49. Emergency Feedwater MO NA RF Flow Indicators

50. PORV and Safety Valve MO NA RF 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 I

00

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5' DESIGN FEATURES 5.1 SITE 5.1.1 The Oconee Nuclear Station is approximately eight miles northeast of Seneca, South Carolina. Figure 2.1-4 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 exclusion area as defined above.

REFERENCE (1) FSAR, Chapter 2 (2) Technical Specification 3.10 5.1-1 Amendment No. 165 (Unit 1)

Amendment No. 165 (Unit 2)

Amendment No. 162 (Unit 3)

6.0 ADMINISTRATIVE CONTROLS 6.1 ORGANIZATION, REVIEW, AND AUDIT 6.1.1 Organization 6.1.1.1 The Station Manager shall be responsible for overall facility operation and shall delegate in writing the succession to this responsibility during his absence.

6.1.1.2 In all matters pertaining to actual operation and maintenance and to these Technical Specifications, the Station Manager shall report to and be directly responsible to the Vice President, Nuclear Production Department, through the General Manager, Nuclear Stations. The organization is shown in Figure 6.1-2.

6.1.1.3 The station organization for Operations, Technical Services.

Maintenance, Station Services, and Integrated Scheduling shall be functionally as shown in Figure 6.1-1. Minimum operating shift requirements are specified in Table 6.1-1.

6.1.1.4 Incorporated in the staff of the station shall be personnel meeting the minimum requirements encompassing the training and experience described in Section 4 of ANSI/ANS-3.1-1978, "Selection and Training of Nuclear Power Plant Personnel" except for the Station Health Physicist, the Superintendent of Operations and the Operating Engineer.

The Station Health Physicist shall have a bachelor's degree in a science or engineering subject or the equivalent in experience, including some formal training in radiation protection, and shall have at least five years of professional experience in applied radiation protection of which three years shall be in applied radiation protection work in one of Duke Power Company's nuclear stations.

A qualified individual who does not meet the above requirements, but who has demonstrated the required radiation protection management capabilities and professional experience in applied radiation protection work at one of Duke Power Company's multi-unit nuclear stations, may be appointed to the position of Station Health Physicist by the Station Manager, based on the recommendations of the System Health Physicist and as approved by the General Manager, Nuclear Stations.

The Superintendent of Operations shall have a minimum of eight years of responsible nuclear or fossil station experience, of which a minimum of three years shall be nuclear station experience. A maximum of two years of the remaining five years of experience may be fulfilled by academic training, or related technical training, on a one-for-one time basis. The Superintendent of Operations shall hold or have held a Senior Reactor Operator license.

6.1-1 Amendment No. 165 (Unit 1)

Amendment No. 165 (Unit 2)

AMend'ent N6, 162 (Unit 3)

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6. 1-7 Amendment No. 165 (Unit 1)

Amendment No. 165 (Unit 2)

Amendn~ent No. 162 (Unit 3)

Executive Vice President Engineering, Construction and Production Vice President Director Nuclear Production DepartmentNuclear Safety Review Board General Manager Nuclear Station*

Manager Oconee Nuclear Station

Responsible for Fire Protection Program OCONEE NUCLEAR STATION MANAGEMENT ORGANIZATION CRART Figure 6.1-2 6.1-8 Amendment No.

165 (Unit 1)

Amendment No. 165 (Unit 2)

Amendment No. 162 (Unit 3)

'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 during the reporting period.

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 members of the public due to their activities inside the unrestricted area boundary during the reporting 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 calender 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.

total container volume (cubic meters),

b.

total curie quantity (determined by measurement or estimate),

c.

principal radionuclides (determined by measurement or estimate),

d.

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

e.

number of shipments, 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.

6.6-2 Apendment No.

165 (Unit 1)

Amendment No. 165 (Unit 2)

Amendment No. 162 (Unit 3)

6.6.3 Special Reports 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 specification:

a.

Auxiliary Electrical Systems, Specification 3.7

b.

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

c.

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

d.

Fire Protection and Detection Systems, Specification 3.17

e.

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

f.

Reactor Building Surveillance, Containment Leakage Tests, Specification 4.4.1

g.

Structural Integrity Surveillance, Tendon Surveillance, Specification 4.4.2.2

h.

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

i.

Dose Calculations (40 CFR 190), Specification 4.21 6.6-5 Amendment No. 165 (Unit 1)

Amendment No. 162 (Vnit 2)

Amendment No. 162 (Unit 3)

ML16138A613

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