Text

Amends 92,92 & 89 to Licenses DPR-38,DPR-47 & DPR-55, Respectively,Revising Tech Specs Re High Pressure Trip Setpoint & Pressurizer Power Operated Relief Valve Setpoint & Adding License Conditions & Tech Specs Re TMI-2 Lessons
	ML15112A986
Person / Time
Site:	Oconee
Issue date:	01/28/1981
From:	Reid R; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML15112A987	List: ML15112A986; ML15112A988; ML15112A989;
References
	NUDOCS 8102200786
	Download: ML15112A986 (41)
	v • d • e

PiA Rfti,4 R

UNITED STATES o

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., 92 License No. DPR-33

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

A. The applications for amendment by Duke Power Company (the licensee) dated May 21, 1979; October 2, 1930, as supplemented October 30, 1980; and Octo ber 20., 1980, 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 applications, the pro visions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorizedby 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 reoulations and all applicable requirements have been satis fied.

2. Accordingly, Facility Operating License No. DPR-38 is hereby amended by revising paragraph 3.B. and adding paragraphs 3.H., 3.I., and 3.J. as follows and by changing the Technical Specifications as indicated in the attachment to this license amendment:

3.B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 92 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

8102200

-2 3.H.

Systems Integrity The licensee shall implement a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as low as practical levels.

This program shall include the-following:

1. Provisions establishing preventive maintenance and periodic visual inspection requirements, and

2. Integrated leak test requirements for each systed at a fre quency not to exceed refueling cycle intervals.

3.1.

Iodine Monitorina The licensee shall implement a program which will ensure the capa bility to accurately determine the airborne iodine concentration in vital areas under accident conditions. This program shall include the following:

1. Training of personnel,

2. Procedures for monitoring, and

3. Provisions.for maintenance of sampling and analysis equip ment.

3.J.

Backup Method for Determining Subcooling Margin The licensee shall implement a program which will ensure the capability to accurately monitor the Reactor Coolant System'subcooling margin. This program shall include the followiig:

1. Training of personnel, and

2. Procedures for monitoring.

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

FOR THE NUCLEAR REGULATORY COMMISSION Robert W. Reid, Chief Operating Reactors Branch #4 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: January 28, 1981

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.92 License.No. DPR-47

1.

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

A. The applications for amendment by Duke Power Company (the licensee) dated May 21, 1979; October 2, 1980, as supplemented October 30, 1980; and Octo ber 20. 1980, comply with the standards and reouirements 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 applications, the Dro 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 reaulations and all applicable requirements have been satis fied.

2. Accordingly, Facility Operating License No. DPR-47 is hereby amended by revising paragraph 3.B. and adding paragraphs 3.H., 3.1., and 3.J. as follows and by changing the Technical Specifications as indicated in the attachment to this license amendment:

3.B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 92 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

-2 3.H. Systems Integrity The licensee shall implement.a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as low as practical levels.

This program shall include the following:

1. Provisions establishing preventive maintenance and periodic visual inspection requirements, and

2. Integrated leak test requirements for each systen at a fre quency not to exceed refueling cycle intervals.

3.1.

Iodine Monitorina The licensee shall implement a program which will ensure the capa bility to accurately determine the airborne iodine concentration in vital areas under accident conditions.

This program shall include the following:

1. Training of personnel,

2. Procedures for monitoring, and

3. Provisions for maintenance of sampling and analysis equip ment.

3.J.

Backup Method for Determining Subcooling Margin The licensee shall implement a program which will ensure the capability to accurately monitor the Reactor Coolant System subcooling margin. This program shall include the following:

1. Training of personnel, and

2. Procedures for monitoring.

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

FOR THE NUCLEAR REGULATORY COMMISSION Robert W. Reid, Chief Operating Reactors Branch #4 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance:

January 28, 1981

UNITED 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. 89 License No. DPR-55

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

A. The applications for amendment by Duke Power Company (the licensee) dated May 21, 1979; October 2, 1980, as supplemented October 30, 1980; and Octo ber 20, 1980, 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 in10 CFR Chapter I; B. The facility will operate in conformity with the applications,.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 wil.l 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, Facility Operating License No. DPR-55 is hereby amended by revising paragraph 3.B. and adding paragraphs 3.H., 3.1.., and 3.J. as follows and by changing the Technical Specifications as indicated in the attachment to this license amendment:

3.B.

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 89 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

-2

3.

H.

Systems Integrity The licensee shall implement a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as low as practical levels.

This program shall include the following:.

1. Provisions establishing preventive maintenance and periodic visual inspection requirements, and

2. Integrated leak test requirements for each systan at a fre quency not to exceed refueling cycle intervals.

3.1.

Iodine Monitorina The licensee shall implement a program which will ensure the capa bility to accurately determine the airborne. iodine concentration in vital areas under accident conditions. This program shall include the following:

1. Training of personnel

2. Procedures for monitoring, and

3. Provisions for maintenance of sampling and analysis equip ment.

3.J.

Backup Method for Determining Subcooling Margin The licensee shall implement a program which will ensure the capability to accurately monitor the Reactor Coolant System subcooling margin.

This program shall include the following:

1. Training of personnel, and

2. Procedures for monitoring.

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

FOR THE NUCLEAR REGULATORY COMMISSION Robert W. Reid, Chief Operating Reactors Branch #4 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: January 28, 1981

ATTACHMENT TO LICENSE AMENDMENTS AMENDMENT NO. 92 TO DPR-38 AMENDMENT NO. 92 TO DPR-47 AMENDMENT NO.

89 TO DPR-b5 DOCKETS NOS.

50-269, 50-270 AND 50-287 Revise Appendix A as follows:

Remove Pages Ie Pages iii ii i

\\'

iv vi vi 2.2-1 2.2-1 2.3-3 2.3-3 2.3-5 2.3-5 2.3-6 2.3-6 2.3-7 2.3-7 2.3-11 2.3-11 2.3-12 2.3-12 2.3-13 2.3-13 3.1-23 3.1-24 3.4-1, 3.4-1 3.4-2 3.4-2 3.5-2 3.5-2.

3.5-3 3z5-3 3.5-4 3.5-4

3. 5-5 3.5-5 3.5-5a 3.5-28 3.5-28 4.1-1 4.1-1 4.1-4 4.1-4 4.1-7 4.1-7 4.1-8 4.1-8 4.1-9 4.1-9 4.9-1 4.9-1 6.1-1 6.1-1 6.1-2 6.1-2 6.1-6 6.1-6 6.1-6a 6.1-7 6.1-7 6.6-6 6.6-6

Section Paze 1.5.4 Instrument Ch.Annel Calibration 1-3 1.5.5 Beat Balance Check1 1.5.6 Beat Balance Calibraion 1-4 1.6

?OWR DISTRIBUTION' 1-4 1.7 CONTAIVIU=

fLvnmG IT 1-4

.2 SAY LL41TS MM L~i= M SAFETI. SrD( SMT1flGS

2. 1-1 2.1 SAFET

LLMITS, REACTOR CORE2.

2.2 SAFETY LflMIT S, REACTO R COOLANT sYsTY PRESSURE 2.2-1 2.3 LflITING SAIMT SYSTM SETTM2S, PR.OTECTIVE 2.3-1 INSTRVU(MTAT7ON 3

L2=1 I.

COD=-IONS -9OR O'vJU.TIN 310-1 3.0 LIMITING CONDIION FOR OPERAT~iON 3.0-1 3.1 REACTOR COOLANT SYST7 31.1-1 3.1.1 Oyerational Cmonents 3.1-1 3.1.2

?ressurizatiofl.

Heatuo, and Codown Liitat+/-onS

3. 1-3 3.1.3 Mini= Conditions for Criticality

3. 1-8 3.1.4 Reactor Colnt system Ativ4lt?

3.1-10 3.1.5 Cbemistry 3.1-12 3.1.6 Leakage 3.1-14 3.1.7 Moderator Tertue Coefficient of Reativty 3.1-17 3.1.8 Single oo Restrictions, 3.1-19 3.1.9 bow Poer

?vsics T stinx Restiction; 3.1-20 3.1.10 Co~ntrol Iod Overtionl312 3.1.11 (ITE1TIONALLY BLANVK)

.12

~

faactor Coolan System S~o~-l agn~'rtr312 3.2r S*G PSSURE LVECTION -UM CMIXICAL ADDITION 35.STVIS 3.2-1

3. 3

=RGZCY CORE COOLING.,

REACTOR BUILDING COOLING, REACTOR 3.3-1 BUILDING SRAY AM LOW. PRESSU SER.VICZ WATEE. S7ST=

Amendments Nios.

92, 92, &39

Section Page 3.4 UUNDARY 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.

Control Rod Group and Power Distribution Limits 3.5-6 3.5.3 neere afetyFeatures Protective System 3.5-28 Actuation Setpoints 3.5.4 Incore Instrumentation 3.5-30 3.6 REACTOR BUILDING 3.6-1 3.7 AUXILIARY ELECTRICAL SYSTEMS 3.7-1 3.3 FUEL LOADING AND REFUELING 3.8-1 3.9 RELEASE OF LIQUID RADIOACTIVE WASTE 3.9-1 3.10 RELEASE OF GASEOUS RADIOACTIVE WASTE 3.10-1 3.11 MAXIMUM POWER RESTRICTION 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 SHOCK SUPPRESSORS (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 REQUIREMENTS 4-1 4.0 SURVEILLANCE STANDARDS 4-1 4.1 OPERATIONAL SAFETY REVIEW 4.1-1 4.2 STRUCTURAL INTEGRITY OF ASME CODE CLASS 1, 2 AND 3 COMPONENTS 4.2-1 4.3 TESTING FOLLOWING OPENING OF SYSTEM 4.3-1 4.4 REACTOR.BUILDING 4.4-1 Amendments Nos. 92, 92& 89

Section Page 4.4.1 Containment Leakage Tests 4.4-1 4.4.2 Structural integrity 44-6 4.4.3 Hydrogen Purge System 4.4-10 4.5 EMERGENCY CORE COOLING SYSTEMS AND REACTOR 4.5-1 BUILDIN1G COOLING..SYSTEMS PERIODIC TESTING 4.5.1 Emergency Core Cooling Systems 4.5-1 4.5.2 Reacrtor Building Cooling Systems 4.5-6 Penetration Room Ventilation Sste.

m 4.5-10

.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 Trio Insertion Time 4.7-1 4.7.2 Control Rod Program Verification 4.7-2 4.8 MAIN STEAM STOP VALVES 4.8-1 4.9 EMERGENCY FEEDWATER PUMP AND VALVE PERIODIC 4.9-1 TESTING 4.10 REACTIVITY ANOMALIES 4.10-1 4.11 ENVIRONMENTAL SURVEILLANCE 4.11-1 4.12 CONTROL ROOM FILTERING SYSTEM 4.12-1 (INTENTIONALLY BLANK) 4.13-1 4.14 REACTOR BUILDING PURGE FILTERS AND THE SPENT 4.14-1 FUEL POOL VENTILATION SYSTEM 4.15 IODINE RADIATION MONITORING FILTERS 4.15-1 4.16 RADICACTIVE MATERIALS SOURCES 4.16-1 4.17 STEAM GENERATOR TUBING SURVEILLANCE 4.17-1 4.18 HYORAULiC SHOCK SUPPRESSORS (SNUBBERS) 4.18-1 4.19 FIRE PROTECTION AND DETECTION SYSTEM 4.19-1 4.20 REACTOR VESSEL INTERNALS VENT VALVES

.20-1 Amendments Nos.

92, 92& 89 iv

LIST OF TABLES Table No.

Page 2.3-1A Reactor Protective System Trip Setting Limits -

Unit 1 2.3-11 2.3-1B Reactor Protective System Trip Setting Limits -

Unit 2 2.3-12 2.3-1C Reactor Protective System Trip Setting Limits - Unit 3 2.3-13 3.5-4 3.5.1-1 Instruments Operatinq Conditions 3.5-1 Quadrant Power Tilt Limits 3.5-14 3.17-1 Fire Protection & Detection Systems 3.17-3 4.1-1 Instrument Surveillance Requirements 4.1-3 4.1-2 Minimum Equipment Test Frequency 4.1-9 4.1-3 Minimu1 Sampling Frequency 4.1-10 4.2-1 Oconee Nuclear Station Capsule Assembly Withdrawal Schedule 4.2-3 at Crystal River Unit No. 3 4.11-1 Oconee Environmental Radioactivity Monitoring Program 4.11-3 4.11-2 Offsite Radiological Monitoring Program 4.11-4 4.11-3 Analytical Sensitivities 4.11-5 4.18-1 Safety Related Shock Suppressors (Snubbers) 4.18-3 6.1-1 Minimum Operating Shift Requirements with Fuel in Three 6.1-6 Reactor Vessels 6.6-1 Report of Radioactive Effluents

6.

6-8 Amendments Nos.

9 92-89 vi

2.2 SAFETY LIMITS -

REACTOR COOLANT SYSTEM PRESSURE Applicability Applies to the limit on reactor coolant system pressure.

Objective To maintain the integrity of the reactor coolant system and to prevent the release of significant amounts of fission product activity.

Specification 2.2.1 The reactor coolant system pressure shall not exceed 2750 psig when there are fuel assemblies in the reactor vessel.

2.2.2 The setpoint of the pressurizer code safety valves shall be in accordance with ASIME, Boiler and Pressurizer Vessel Code,Section III, Article 9, Summer 1967.

Bases The reactor coolant system serves as a barrier to prevent radionuclides in the reactor coolant from reaching the atmosphere.

In the event of a fuel cladding failure, the reactor coolant system is a barrier against the release of fission products. Establishing a system pressure limithelps to assure the integrity of the reactor coolant system. The maximum transient pressure allowable i the reactor coolant system pr ure vessel under the ASME code,Section III, is 110% of design pressure.

The maximum transient pressure allowable in the reactor coolant system piping, valves, and fittings under USAS Section B31.7 is 110% of design pressure. Thus, the safety lim of 2750 psig (110% of the 2500 psig design pressure) has been established..

The settings, the reactor (jgh pressure trip (2300 psig) and thepressurizer safety valves (2500 psig) have been established to assure never reaching the reactor coolant system pressure safety limit. The initial hydrostatic test was conducted at 3125 psig (125% of design pressure) to verify the integrity of the reactor coolant system. Additional assurance that the Reactor Coolant pressure does not exceed the safety limit is provided by setting the pressurizer electromatic relief valve at 2450 psig.

REFERENCES (1) FSAR, Section 4 (2) FSAR, Section 4.3.10.1 (3) FSAR, Section 4.2.4 Amendments Nos. 92 92, & 89 2.2-1

level trip and associated reactor power/reactor power-imbalance boundaries by 1.08% - Unit 1 for 1% flow reduction.

1.08%

Unit 2 1.08%

- Unit 3 P'umD Monitors The pump monitors prevent the minimum core DNBR from decreasing below 1.3 by tripping the reactor due to the loss of reactor coolant pump(s).

The circuitry monitoring pump operational status provides redundant trip protection for DNB by tripping the reactor on a signal diverse from that of the power-to-flow ratio.. The pump monitors also restrict the power level for the number of pumps in operation.

Reactor Coolant System Pressure During a startup accident from low power or a slow rod withdrawal from high power, the system high pressure setpoint is reached before the nuclear over power trip setpoint.

The trip setting limit shown in Figure 2.3-A -

Unit 1 2.3 Unit 2 2.3-1C - Unit 3 for high reactor coolant system pressure C.300 psig) has been established to maintain the system pressure below the safety limit (2750 psig) for any design transient. (1)

The low pressure (1800) psig and variable low pressure (11.14 T

-4706) trip (1800) psig (11.14 TUt-4706 (1800) Psis406 out (1800 psig(11.14 Tout-4 7 06 )

sIepoints shown in Figure 2.3-LA have been established to maintain the DNB 2.3-13 2.3-1C ratio greater than or equal to 1.3 for those design accidents that result in a pressure reduction.

(2,3)

Due to the calibration and instrumentation errors, the safety analysis used a variable low reactor coolant system pressure trip value of (11.14 Tout - 4746)

(11.14 T

- 4746)

(11.14 Tout -

4746) out Coolant Outlet Temperature The high reactor coolant outlet temperature trip setting limit.(619o shown in Figure 2.3-LA has been established to prevent excessive core coolant 2.3-13 2.3-1C temperatures in the operating range.

Due to calibration and instrumentation errors, the safety analysis used a trip setpoint of 62007.

Reactor Building Pressure The high reactor building pressure trip setting limit (4 psig) provides positive assurance that a reactor trip will occur in the unlikely event of.a loss-of coolant accident, even La the absence of a low reactor coolant system pressure trip.

-Amendments N~os. 92, 92 &89 2.3-3

4 2400 2300 P= 2300 PSIG T

619. F 2200 ACCEPTABLE OPERATION 2100 2000 UNACCEPTABLE OPERATION 1900 P = 1800 psig 1800 T

584'F 540 560 580 600 620 640 Reactor Out It.TIc!rature OF PROTECTIVE SYSTEM MAXIMUM ALLOWABLE SETPOINTS UNIT 1 mee OCONEE NUCLEAR STATION Figure 2.3-lA Amendments Nos.

92

, 92 & 89

-5

2400 2300 P30 P 2300 PSIG T 619 F 2200 ACCEPTABLE OPERATION 2100 UNACCEPTABLE 2000 OPERATION 1900 P. = 800 psi 1800 540 560 580 600 620 640 Reactor Outlet Temperature, *F PROTECTIVE SYSTEM MAXIMUM ALLOWABLE SETPOINTS UNIT 2 OCONEE NUCLEAR STATION Amendments Nos.

92 92 & 89 2.3-6 Figure 2.3-1B

2400 2300 P z 2300 PSIG T= 6 F

2200 ACCEPTABLE OPERATION 2100 UNACCEPTABLE OPERATION 2000 1900 P

1800 psg 1800 Ta 584F 540 560 580 600 620 640 Rsactor Outlet Teagerature. F PROTECTIVE SYSTEM MAXIMUM ALLOWABLE SETPOINTS 2.3-7 UNIT 3 Amendments Nos.

92, 92 & 89 utnwo OCONEE NUCLEAR STATION Figure 2.3-1C

Table 2.3-IA Unit I rt Reactor Protectivp Syae. Trip Setting Limits 0

Wl One Reactor Four Reactor Three Reactor Coolant Pump Coolant Pumps Coolant Puamps Operating In Operating Operating Each Loop (Operating Power (Operating Power (Operating Power Shutdown a.0 II' seents 19% Rated)

-175% Rated

-49%. Rated) _yjgs

1. Nuclear Power hax.

105.5 105.5 105.5 5.0 (f Rated) 00

2. Nuclear Power tHax. Based 1.08 times flow 1.08 times flow 1.08 times flow Bypassed on Fino (2) and Imbalance.

minus reduction minus reduction minus reduction

(% Rated) due to Imbalance due to imbalance due to Imbalance

3. Nuclear Power Hax. Based NA NA 55%

Bypassed on Pump IHonitors,

(% Rated)

4.

uigh Reactor Coolant 2300 2300 2300 l12o System Pressre,- puig, Hax.

5. Lou Reactor Coolant 1800 1800 1800 Bypassed System Pressure, paig. Hin.

6. Variable Low Reactor (1l.14Tou -4106)(1)

(II.14Tou 406)

(l0.14T

-4106)

Bypassed Coolant system Pressure psig, hin.

Reactor Coolant Temp.

F.. Hax.

619 619 619 619 t.' High Reactor Building 4

4 4

Pressure, paig, Hax.

(1) T is in degrees Fahrenheit ().

(2) Reactor Coolant System Flow, %.

(3) Administratively controlled reduction set only during reactor shutdown.

(4) Automatically met when other segments of the RPS are bypassed.

Table 2.3-18 Unit 2 Reactor protective System Trip Setting Limits M

One Reactor

.:j Four Reactor Three Reactor Coolant Paump A

Coulant Pumps Coolant Pumps Operating in z

Operating Operating achl Loop 0

(Operating Power (Operating Power (Operating Power Shutdo n (A

RPS Segment

-100% Rated)

-75% Rated)

-49% Rated)

Byga 0

Nuclear Power Hax.

105.5 105.5 105.5 5.0

(% Rated)

2.

Nuclear Power fax. Based 1.08 times flow 1.08 times flow 1.08 times flow Bypassed on Flow (2) and Imbalance, minus reduction minue reduction minus reduction

(% Rated) due to imbalance due to imbalance due to imbalance

3.

Nuclear Power. Hax.

Based NA NA 55%

Bypassed on Pump Honitors,

(% Rated)

4.

uagh Reactor Coolant System 2300 2300 2300 1720(4)

Pressure, psig, fax.

kCj

5.

Low Reactor Coolant System 1800 1800 1800 Bypassed Pressure, psig, hiln.

6.

Variable Low Reactor Coolant (1114 Tout -

4706)(1)

(1114 Tout 4706)(1)

(11.14 Tout 4706)(1)

Bypassed System Pressure psig Hn.

H

7.

Heaetor Coolant Temp. F., Hax.

619 619 619 619

8.

High Reactor Building 4

4 4

4 Pressure, psig, fax.

()Tu t iin degrees Fahrenheit ("F).

(2)

Reactor Coolant System Flow, %.

(3)

Administratively controlled reduction set only during reactor shutdown.

(4)

Autoatically set when other segments of the RPS are bypassed.

.I!

I.,

AX Iio Is.~A

)li

!jm 1im ltE:

-iod 'Rind fll'olo I ot(Oa 1

m~j~~i if'!IP (1j11v

%tg7) qt 10Wditi, 1in USVN VN P""i

" JPOJ Jv-ll)II4 1

rm leim I

wl a1e pilf 01i' ;?li

" U101~ g I m li.'~ti til ii

~

l mm iti1 ?.lq).Im *IlIidlvIm I'(5~m

. O IIlldp l sPIlilt')

311PC q i mj hipp 1

0:

101 (7)) A9ij.f Jij)0 110 00 moll otllj

!o-11~l lltt~

80-rat.1.

Ri.1

-xrJ 3010.1,1 Jiil)P0N i

0 a~r" U) 1T l~m~l

INTENTIONALLY BLANK Amendments Nos. 92, 92 & 89

3. -23

3. L.12 Reactor Coolant System Subcooling Margin Monitor Specification 3.1.12.1

a. The Reactor Coolant System subcooling monitors shall be operable when the average RCS coolant temperature is above 300 0F.

b. If one monitor is inoperable, the monitor shall be restored to operable status within seven 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 .

c. If both of the subcooling monitors are inoperable because of an outage of the operational computer, and the computer is out of service for less than four hours, and the backup method for determining subcooling margin is available, then a capability to. determine subcooling margin is. available and a report pursuant to Specificationi 6.6.2 is not required.

d. If both of the subcooling monitors are inoperable, then restore at least one monitor to operable status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in atleast.hot shutdown within h e

next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Bases The operability requirements of the Reactor Cooiant. System subcooling margin monitors ensures that sufficient informationxis 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.

Guid ance 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..

Amendments No. 92 92 & 89 3.1-24

3.4 SECONDARY SYSTEM DECAY HEAT RE2OVAL Applicability Applies to the secondary system requirements for, removal of reactor decay heat.

Objective To specify minimum conditions necessary to assure the capability to remove decay heat from the reactor core.

Specification 3.4.1 Emergency Feedwater System The reactor shall not be heated above 250 0F unless the following conditions are met:

a. Three emergency feedwater pumps (one steam-driven pump capable of being powered from an operable steam supply system and two-motor-driven pumps), and associated initiation circuitry, shall be operable.

b. Two 100% emergency feedwater flow paths shall be operable. Each flow path shall have at least one flow indicator operable.

C. If one emergency feedwater pump or emergency feedwater flow path is inoperable then, restore it to operable status within 60 hours6.944444e-4 days 0.0167 hours 9.920635e-5 weeks 2.283e-5 months . Otherwise, the unit shall be in a hot shutdown condition withinn a additional 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and below 250 0F in another 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

3.4.2 The 16 steam system safety valves shall be operable.

3.4.3 A minimum of 72,000 gallons of water per operating unit shall be available in the upper surge tank, condensate storage tank, and hotwell.

3.4.4 The emergency condenser circulating water system shall be operable.

3.4.5 The controls of the emergency. feedwater system shall be independent of the Integrated Control System.

Bases The Main Feedwater System and the Turbine Bypass System are normally used for decay heat removal and cooldown above 2500F.

Feedwater makeup is supplied by operation of a hotwell pump condensate booster pump and a main feedwater pump.

The Emergency Feedwater (EFW) System assures sufficient feedwater supply to the steam generators of each unit, in the event of loss of the main Feedwater System, to remove energy stored in the core and primary coolant.

The EFW System is designed to provide a sufficient secondary side steam generator heat sink to enable cooldown from reactor power operation down to cold shutdown conditions.

3.4-1 Amendments Nos.

92,92 & 8B9

A 100% emergency feedwater flowpath shall be considered to be either:

1) the s team-driven turbine pump, associated valves and piping capable of feeding either steam generator or 2) both motor-driven pumps, associated valves and piping each capable of feeding the associated steam generator.

One flow indicator or steam generator level indicator per steam generator is sufficient to provide indication of emergency feedwater flow to the steam generators and to confirm emergency feedwater system operation.

In the event that at least one indicator per steam generator is not available, then the flowpath to this steam generator is considered to be inoperable.

The.=i System is designed to start automatically in the event of loss of both main feedwater pumps or low main feedwater header pressure. The EFW System will supply sufficient feedwater for approximately five-hour cooldown at a flowrate of at least 720 gpm to enable the Reactor Coolant System to reach conditions at which the Decay Heat Removal System may be operated.

Two motor-driven emergency feedwater pumps are installed in each unitin addition to the steam-driven emergency feedwater pump. The motor-driven pumps are powered from diverse emergency power supplies.

All automatic initiation logic and control functions are independent from the Integrated Control System (ICS).

Normally, decay heat is removed by steam relief through the turbine bypass system to the condenser. Condenser cooling water flow is provided by a siphon effect from Lake Keowee through the condenser.for final heat rejection to the Keowee Hydro Plant tailrace. Decay heat can also be removed from the steam generators by steam relief through the main steam relief valves.

The minimum amount of water in the upper surge tank, condensate storage tank and hotwell is the amount needed for 11 hours1.273148e-4 days 0.00306 hours 1.818783e-5 weeks 4.1855e-6 months of operation per unit. This is based on the conservative estimate of normal makeup being 0.5% of throttle flow. Throttle flow at full load, 11,200,000 lbs/hr., was used to calculate the operation time. For decay heat removal the operation time with the volume or water specified would be considerably increased due to the reduced throttle flow.

The total relief capacity of the 16 steam system safety valves is 13,105,000 lbs/hr.

REFERENCE FSAR, Section 10 Amendments Nos.

92, 92 & 89

Bases Every reasonable effort will be made to maintain all safety instrumentation in operation. A startup is not permitted unless three power range.neutron instru ment channels and two channels each of the following are operable:

four reactor coolant temperature instrument channels, four reactor coolant flow instrument channels, four reactor coolant pressure instrument channels, four pressure temperature instrument.channels, four flux-imbalance flow instrument channels, four power-number of pumps instrument channels, and high reactor building pressure instrument channels. The engineered safety features actuation 'system must have two analog channels functioning correctly prior to a startup. Addi tional operability requirements are provided by Technical Specifications 3.1.12 and 3.4 for equipment which are not part of the RPS or ESFAS.

Operation at rated power is permitted as long as the systems have at least the redundancy requirements of Column B (Table 3.5.1-1).

This is in agree-.

ment with redundancy and single failure criteria of IEEE-279 as described in FSAR Section 7.

There are four reactor protective channels. A fifth channel that is isolated from the reactor protective system is provided as a part of the reactorcontrol system. Normal trip logic is two out of four. Required trip logic for the power range instrumentation channels is two out of three. Minimum.trip.logic on other channels is one out of two.

The four reactor protective channels were provided with key operated bypass switches to allow on-line testing or maintenance on only one channel at a time during power.operation. Each channel is provided alarm and lights to indicate when that channel is bypassed. There will be one reactor protective system bypass switch key permitted in thecontrol room. That key will be under the administrative control of the Shift Supervisor. Spare keys will be maintained in a locked storage accessible only to the station Manager.

Each reactor protective channel key operated shutdown bypass switch is provided with alarm and lights to indicate when the shutdown bypass switch is being used.

There are four shutdown bypass keys in the control room under the administrative control of the Shift Supervisor. The use of a key operated shutdown bypass switch for on-line testing or maintenance during reactor power-operation has no significance when used in conjunction with a key operated channel bypass switch since the channel trip relay is lockedin the untripped state.

The use of a key operated shutdown bypass switch alone during power operation will cause the channel to trip.

When the shutdown bypass switchnis operated for on-line testing or maintenance during reactor power operation, reactor power and RCS pressure limits as specified in Table 2.3-IA, B, or C are not applicable.

The source range and intermediate range nuclear instrumentation overlap by one decade of neutron flux.

This decade overlap will be achieved at.10-amps on the intermediate range instrument.

Power is normally supplied tothe control rod drive mechanisms from"two separate parallel 600 volt sources.

Redundant trip devices are employed in each of these sources.

If any one.of these trip devices fails in the untripped state on-line repairs to the failed device, when practical, will be made and the remaining trip devices will be tested.

Four hours" is ample time to test the remaining trip devices and in many cases make on-line repairs.

Amendments Nos., 92, 92 & 89

Containment isolation valves on non-essential systems are isolated by diverse signals from high containment pressure and low reactor coolant system pres sure devices.

The systems considered to be non-essential include:

1. Letdown line

2. RC Pump seal return line

3.

Quench Tank sample line

4. Quench Tank gaseous vent

5. Reactor Building purge lines

6. Reactor Building sump drain line

7. Reactor Building atmosphere sample line

8. Pressurizer sample line

9. OTSG sample line

10.

OTSG drain line Containment isolation valves on essential systems are isolated by high con tainment pressure only. The systems considered to be essential include:

1. Component cooling to RC pumps

2. Low pressure service water cooling to RC pump motor REFERENCE FSAR, Section 7.1 Amendments Nos.

92, 92 & 89 3.5-3

TABLE 3.5.1-1 INSTRUMENTS OPERATING CONDITIONS (A)

(B)

(C)

Minimum Minimum Operator Action If Conditions Operable Degree Of Of Column A and B Functional Unit Channels Redundancy Cannot Be Met

1. Nuclear Instrumentation 1

0 Bring to hot shutdown within Intermediate Range 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (b)

Channels

2.

Nuclear Instrumentation 1

0 Bring to hot shutdown within Source Range Channels 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (b)(c)

3. RPS Manual Pushbutton 1

0 Bring to hot shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

4. RPS Power Range 3(a) l(a)

Bring to hot shutdown within Instrument Channels 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

5. RPS Reactor Coolant 2(d) 1 Bring to hot. shutdown within Temperature Instrument 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Channels

6.

RPS Pressure-Temperature 2(d)

Bring to hot shutdown within Instruments Channels 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s:

7. RPS Flux Imbalance 2

1 Bring to hot shutdown within Flow Instrument Channels 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

8.

RPS Reactor Coolant Pressure

a. High Reactor Coolant 2

1 Bring to hot shutdown within Pressure Instrument 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Channels

b. Low Reactor Coolant 2

1 Bring to hot shutdown within Pressure Channels 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

9. RPS Power-Number of Pumps 2

1 Bring to hot shutdown within Instrument Channels 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

10. RPS High Reactor Building 2

1 Bring to hot shutdown within Pressure Channels 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Amendments Nos.

92, 92 & 89 3.5-4

TABLE 3.5.1-1 INSTRLMENTS OPERATING CONDITIONS (cont'd)

(A)

(B)

(C)

Minimum Minimum Operator Action If Conditions Operable Degree of Of Column A and B Functional Unit Channels Redundanc Cannot Be Met

11.

ESF High Pressure Injection System &

Reactor Building Isolation (Non-essential Systems)

a. Reactor Coolant 2

1 Bring to hot shutdown within Pressure Instru-12 hours (e) ment Channels.

b. Reactor Building 2

1 Bring to hot shutdown within 4 PSIG Instrument 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (e)

'Channels

c. Manual Pushbutton 2

1 Bring to hot shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (e)

12.

ESF Low Pressure In

ection System

a. Reactor Coolant 2

1 Bring to hot shutdown within Pressure Instrument 1.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months (e)

Channels Reactor Building 2

1 Bring to hot shutdown within SIG Instrument.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (e)

Channels

c. Manual Pushbutton 2

1 Bring to hot shutdown wirhin 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (e)

13.

ESP Reactor Building Isolation (Essential Systems) & Reactor Building Cooling System

a. Reactor Building 2

2Brng to hot shutdown Weih)n PSIG Instrumet 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (e)

Channel 1Briag to ho shutdown w -h im

b. Manual Pushbutton 2

12 thoo t so 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (e) 4..

ES-Reactor Building.

Spray System a

Reactor Building Bring to hot shu do wnwi-'n High oressure astru:ment :hannel Amendments Nos.

92, 92 & 89

TABLE 3.5.1-1 INSTRUMENTS OPERATING CONDITIONS (Cont'd)

(A)

Minimum (B)

(C)

Operable Minimum Operator Action If Conditions Analog Degree Of Of Column A and B Functional Unit Channels Redundancy Cannot Be Met

b. Manual Pushbutton 2

1 Bring to hot shutdown withn 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (e)

15. Turbine Stop Valves 2

1 Bring to hot shutdow n wi:b n Closure 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (f)

(a) For channel testing, calibration, or maintenance, the minimum number of operable channels may be two and a degree of redundancy of one for a maximum of four hours.

(b) When 2 of 4 power range instrument channels are greater than 10% rated power, hot shutdown is not required.

(c) When 1 of 2 intermediate range instrument channels is greater than 10-10 amps, hot shutdown is not required.

(d) Single loop operation at power (after testing and approval by the NRC/.DOL) is not permitted unless the operating channels are the two receiving Reactor Coolant Temperature from operating loop.

(e) If minimum conditions are not met within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after-hot shutdown, the unit shall be in the cold shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

(f) One operable channel with zero minimum degree of redundancy is allowed for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months before going to the hot shutdown condition.

Amendments Nos.

92, 92 &89 3.5-5a

3.5.3 Engineered Safety Features Potective System Actuation Setpoints Applicability This specification applies to the engineered safety features protective system actuation setpoints.

Objective To provide for automatic initiation of the engineered safety features protective system in the event of a breach of RCS integrity.

Specification The engineered safety features protective actuation setpoints and permissible bypasses shall be as follows:

Functional Unit Action Setpoint High Reactor Building Reactor Building Spray 30 psig Pressure High-Pressure Injection 4 psig Reactor Building Isolation (Non-essential Systems)

Low-Pressure Injection

_4 psig Start Reactor Building Cooling & Reactor Building Isolation (Essential Systems) 4 psig Penetration Room Ventilation 4 psig Lower Reactor Coolant High Pressure Injection(1) 1500 psig System Pressure

& Reactor Building Isolation (Non-essential Systems)

(2)

Low Pressure Injection(2 500 psig (1) May bebypassed below 1750 psig and is automatically reinstated aboved 1750 psig.

(2) May be bypassed below 900 psig and is automatically reinstated above 900 psig.

Bases High Reactor Building Pressure The basis for the 30 psig, and 4Ipsig setpoints for the high pressure signal is.toestablish a setting which would be rached immediatey inthe event oz a DBA, cover the entire spectrum of break sizes andyet be far enough above normal operation maximum internal pressure to prevent spurious initiation.

Low Reactor Coolant Sstem Pressure The basis for the 1500 psig low reactor coolant pressure setpoint for high pressure injection initiation and 500 psig for low pressure injection is to Amendments Nos.

92, 92 & 89 3.5-28

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 Protective System and Engineered Safety Feature Protective System instrumentation shall be as stated in Table 4.1-1.

4.1.2 Equipment and sampling test shall be performed as detailed in Tables 4.1-2 and 4.1-3.

4.1.3 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.

Comparison of output and/or state of independent channels measuringthe 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 reactor 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 opera tion, 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 do.es not ex ceed the indicated neutron power by more than the tolerance. (4% FP) assumed in the 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 it self can tolerate longer intervals between calibrations. Process system instru mentation errors induced.by drift can be expected to remain within acceptable tolerances if recalibration 'is performed at the intervals specified.

4.

1-1 Amendments Nos.

92, 92 & 89

0.

0

-4

-4>,

S4 4

Amnmet Nos 92 2-9

.6 40 a

-A L41 Amenment No.

9292 89

.1-

(D.

-Iai~

hiiidooCac t~

aIiate Remariks

.IOV asid Saflet y Va I ye MOl NA

'RF P115 1 o.

I Ilsd j Cai ts C\\)

ES Each Shift QU

-Quarterl1y D)A -

Daily AN -

AnnualIly WE

-Weekly PS -Prior to startup, if not performed previous week MO M onthly NA

-Not Applicable RF Refueling Outage 4z:

0)

0 Table 4.1-2 MINIUM EQUIPMENT TEST FREQUENCY Item Test Frequency

1. Control Rod Movement (Movement of Each Rod Monthly

2. Pressurizer Safety Valves Sepin ach Refueling (4)

3. Main Steam Safety Valves Setpoint

-Each Refueling

4. Refueling System Interlocks Functional Prior to Refueling Main Steam Stop Valves Movement of Each Stop Monthly Valve (2)

6. Reactor Coolant System Evaluate Daily Leakage 7

Condenser Cooling Water Functional Each Refueling System Gravity Flow Test

3. High Pressure Service Functional Monthly Water Pumps and Power Supplies

9.

Spent Fuel Cooling System Functional Prior to Re tueling

10.

High Pressure and Low (3)

Vent Pump Casings Monthly and Prio Pressure Injection System to Testjrng 1I.

Emergency Feedwater Functional Each Refueling

?ump Automatic S:art and Automatic Valve Actuation Feature

12. RCS Subcooling Functional Each Refuelin onitor Applicable only when the reactor is critical.

(2)

Applicable only when :he reactor coolant is abcve 000 F and at a stead state temperature and pressure.

(3) Operating pumps excluded.

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

Amendments Nos.

92, 92, & 89 4.1-9

4.9 EMERGENCY FEEDWATER PU2P AND VALVE PERIODIC TESTING ADp1icabilicy Applies to the periodic testing of the turbine-driven and motor-driven emergency feedwater pumps and associated valves.

Obj ective To verify that the emergency feedwater pumps and associated valves are operable.

Specification 4.9.1 Pump Test Monthly, the turbine-driven and motor-driven feedwater pumps shall be operated on recirculation to the upper surge tank for a minimum or one hour.

4.9.2 Valve Test Quarterly, automatic valves in the :emergency feedwater flow oath will be determined to be operable in accordance with the applicable edition of the ASME Boiler and Pressure Vessel Code,Section XI.

4.9.3 Acceotance Criteria These tests shall be considered satisfactory if control board indication and visual observation of the equipment demonstrates that all components have operated properly.

Bases The monthly testing frequency is sufficient to verify that the emergency feed water pumps are operable. Verification of.correct operation is made both from the control room instrumentation and direct visual observation of the pumps.

The parameters which are observed are detailed in the applicable edition of the ASYE Boiler and Pressure Vessel Code,Section XI.

REFERENCES (1) FSAR, Section 10.2.2 (2) FSAR, Section 14.1.2.8.3 Amendments Nos. 92, 92 & 89 4.9-1

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 opera tion and shall'delegate in writing the succession to this responsi bility 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., Steam Production, through the Manager, Nuclear Production. The organization is shown in Figure 6.1-2.

.6.1.1.3 The station organization for Operations, Technical Services and Maintenance 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 Site Health Physicist.

The Site Health Physicist shall have a bachelor's degree in a science or engineering subject or theequivalent 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 protec tion work at one of Duke Power Company's multi-unitnuclear stations, may be appointed to the position of Site Health Physicist by the sta tion Manager, based on the recommendations of the System Health Physi cist 'and as approved by the Manager, Nuclear Production.

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 -he following manner:

Amendments Nos.

92, 92 & 89

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

b. Several engineers, familiar with plant operations and.represeating diverse technical backgrounds will be assigned to provide the operating experience assessment.

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 orocedure 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 MIanaoer, or by the.Operating Superintendent,, the Technical Services Suverintendent 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 aical Specifications change shall be reviewed by an iadividual/group other than the individual/group which.prepared the proposed change, bu. who may be from the same organization as the individual/group which prepared the proposed change.. Proposed changes to the Technical Specificatioas shall be approved by the station lanager.

c.

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

Each such modification shall be~ reviewed by an individu~al/

group other than the individual/group which designed the modification, but ou may be from the same.organization asthe individual/group which designed the modification. Proposed modifications to station nuclear safety-related structures, systems and components shall be. approved prior to iorolementation by. the station lfanager; or by the Operating S up erinat-n d

enat, the Technical Services Superintendent, or the.)ifaintenance Superintendeat.

as previously designated oy the station Manager.

1.

-dividuals responsible for reviews performed in accordence with 6.1.2.1 6.17

.b and 6.1.2..c shall be members of the station suvervlisory sta Z orev'ious~vy designated by the station >anager to perform such reviews.

7-ach such revi~ew shall include a determination of wdether or not additionaL, cross-d-isc-ipL4-iar7, review is necessary.

If deemed necessar-7.

suca revieCw salbe -,eriformed ':-. the approprate desig7nated station rev~iew persoDnei,.

e.

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

lmendments Nos.

92, 2 & 39 6-2

TlAB1LE 6.1I-1I MINI MUMI OPERAT ING SIFTl REQII I REMENTrS

()(With Fuel in the Three Rea ctor Vessels) 0 koOne Two All All r\\

Uni t Units Units Uni ts Op e r at i ng Ope rati Itg*11 Ope ra tL ng Shutdown.

Se It r L, peratsor 4

4 3

NRea or OpertoreI

4.

5pr ot 3

Alote CA 0 (1 shltdowi Onl1y one S E eqO rL e if tod uts are operated f rotone Cotol Room.

ADDIIONAL QUI viENTS 1.- One licensed operator per unit shall be in the Control Room at all times when there is fuel in the reactor vessel.

2. Two licensed operators shall be in the Control Room during startup and scheduled shutdown of a reactor.

3. At least one licensed operator shall be in the reactor building when fuel handling operations in the reactor building are in progress.

4. An operator holding a Senior Reactor Operator license and assigned no other operational duties shall be in direct charge of refueling-operations.

5. At least one person per shift shall have sufficient training to perform routine health physics requirements.

6. If the computer for a reactor is inoperable for more than eight hours, an operator, in addition to those required above, shall supplement the shift crew.

A fire brigade of 5 members shall be maintained on site at all times.

This excludes 3 members of the minimum operating. shift requirements that are required to be present in the. control rooms.

Amendments Nos 92, 92 & 89 6.1-6a

I-D Manager

=4 Oconee Niic tear Station (D

S l1.*n e

r I

_Meh ni a Adiusti ive rehla evcsOeatlgN nea Sc IVI.CCS SuEcr I prnenen SupMintendet i e nd III-orcii SlO.

mix_

ntumn n

~' LaLion SaSC f Le.

Health El pr u Iectrical 11.hsltSupervisorsin iue.IM

~~~~~~~~~H Eng e Ir~C t..Eit e i neeriiI I

Shift shShfft Tfechn~ical

&rt(..D~KPW~ OCONEE NUCLEAR STATION TATION ORGANIZATION CHART IFiGUIRE 6.1-1 NLIc tea I Sie ty

(9) Performance of structures, systems, or components that requires remedial action or corrective measures to orevent 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 Director, Office of Inspection and Enforcement, 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 proce dural controls during operation of a unit which could cause reduction of degree of red.undancy.prcvided in the Reactor Protective System or Engineered Safety Feature Systems.

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 15 mrem/yr to the thyroid of any individual.

b. If milk samples collected over a calendar quarter show average concentrations of 4.8 picocuries per liter or greater, a plan shall be submitted within 30 days advising the NRC of the proposed action to ensure the plant related annual doses will be within the design objective of 15 mrem/yr to the thyroid of-any indivi dual.

c. If, during any annual report period, a measured level of radioactivity in any environmental medium other than those associated with gaseous radioiodine releases or liquid effluent releases exceeds ten times the control station value, a written notification will be submitted within one week advising the NRC of this condition.

This notification should include an evaluation of any release conditions, environ mental factors, or other aspects necessary to explain the anomalous result.

d. If, during any annual report period, a measured level of radioactivity in any environmental medium associated with liquid effluent releases exceeds 50 times the control station value for sampling.points at or upstream of location 000.7 or ten times the control station value for sampling points downstream of location 000.7, a writtennotification will be submitted within one week advising the NRC of this condition. This notification should include an evaluation of any release condi tions, environmental factors, or other aspects necessary to.explain'the anomalous results.

Amendments Nos.

92, 92 & 89 6.6-6

ML15112A986

Contents

Text

Attachment:

Attachment:

Attachment:

Navigation menu

ML15112A986

Text

Attachment:

Attachment:

Attachment:

Navigation menu

Search