ML16138A756
| ML16138A756 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 04/08/1993 |
| From: | Matthews D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML16138A757 | List: |
| References | |
| NUDOCS 9304160158 | |
| Download: ML16138A756 (15) | |
Text
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. 199 License No. DPR-38
- 1. The Nuclear Regulatory Commission (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 December 8, 1992, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
- 2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 3.B of Facility Operating License No. DPR-38 is hereby amended to read as follows:
9304160158 930408 PDR ADOCK 05000269 p
-2 Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.199, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3. This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION David B. Matthews, Director Project Directorate 11-3 Division of Reactor Projects-I/Il Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance:
April 8, 1993
o0 UNITED STATES NUCLEAR REGULATORY COMMISSION o
WASHINGTON, D. C. 20555 DUKE POWER COMPANY DOCKET NO. 50-270 OCONEE NUCLEAR STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.199 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 December 8, 1992, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
- 2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 3.3 of Facility Operating License No. DPR-47 is hereby amended to read as follows:
-2 Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.199, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
- 3. This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION David B. Matthews, Director Project Directorate 11-3 Division of Reactor Projects-I/II Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance:
April 8, 1993
UNITED STATES 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.196 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 December 8, 1992, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
- 2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated.in the attachment to this license amendment, and Paragraph 3.B of Facility Operating License No. DPR-55 is hereby amended to read as follows:
-2 Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.196, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
- 3. This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION David B. Matthews, Director Project Directorate 11-3 Division of Reactor Projects-I/II Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance: April 8, 1993
ATTACHMENT TO LICENSE AMENDMENT NO.199 FACILITY OPERATING LICENSE NO. DPR-38 DOCKET NO. 50-269 AND TO LICENSE AMENDMENT NO.199 FACILITY OPERATING LICENSE NO. DPR-47 DOCKET NO. 50-270 AND TO LICENSE AMENDMENT NO.i96 FACILITY OPERATING LICENSE NO. DPR-55 DOCKET NO. 50-287 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the areas of change.
Remove Pages Insert Pages 1-4 1-4 3.5-2 3.5-2 3.5-5c 3.5-5c 4.1-2 4.1-2 4.1-3 4.1-3 4.1-4 4.1-4 4.1-8 4.1-8 4.1-8a 4.1-8a
00 1.5.5 Heat Balance Check A heat balance check is a comparison of the indicated neutron power and core thermal power.
1.5.6 Heat Balance Calibration An adjustment of the power range channel amplifiers output to agree with the core thermal power as determined by a heat balance on the secondary side of the steam generator considering all heat losses and additions.
1.5.7 Staggered Test Basis A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.
1.6 POWER DISTRIBUTION 1.6.1 Quadrant Power Tilt Quadrant power tilt is defined by the following equation and is expressed in percent.
100 x
Power in any core quadrant 1
Average power of all quadrants 1.6.2 Reactor Power Imbalance Reactor power imbalance is the power in the top half of the core minus the power in the bottom half of the core expressed as a percentage of rated power.
Imbalance is monitored continuously by the RPS using input from the power range channels.
Imbalance limits are defined in Specification 2.1 and imbalance setpoints are defined in Specification 2.3.
1.7 CONTAINMENT INTEGRITY Containment integrity exists when the following conditions are satisfied:
- a.
The equipment hatch is closed and sealed and both doors of the personnel hatch and emergency hatch are closed and sealed except as in b below.
- b.
At least one door of the personnel hatch and the emergency hatch is closed and sealed during refueling or during personnel passage through these hatches.
- c.
All non-automatic containment isolation valves and blind flanges are closed as required.
- d.
All automatic containment isolation valves are operable or locked closed.
- e.
The containment leakage determined at the last testing interval satisfies Specification 4.4.1.
Amendment No.199 (Unit 1)
Oconee Units 1, 2, & 3 1-4 Amnet o19(it)
Amendment No.199 (Unit 2)
Amendment No.196 (Unit 3)
Bases Every reasonable effort will be made to maintain all safety instrumentation in
-operation. A startup is riot permitted unless three power range neutron in strument channels and three channels each of the following are operable:
reactor coolant temperature, reactor coolant pressure, pressure-temperature, flux-imbalance flow, power-number of pumps, and high reactor building pres sure. The engineered safety features actuation system must have three analog channels and two digital channels functioning correctly prior to a startup.
Additional operability requirements are provided by Technical Specifications 3.1.12 and 3.4 for equipment which is not part of the RPS or ESFAS.
Operation at rated power is permitted as long as the systems have at least the minimum number of operable channels given in Column C (Table 3.5.1-1).
This is in agreement 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 reactor con trol system. Normal trip logic is two out of four.
The minimum number or operable channels required is three. While a bypassed channel is considered inoperable, a channel placed in the tripped condition is considered operable.
Thus, only one channel may be placed in bypass at any one time in order to maintain the minimum number of required channels. This results in a trip logic of two out of three.
It should be noted that an effective trip logic of one out of two can be achieved by placing one channel in bypass and one channel in the tripped condition.
The four reactor protective channels are 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 the control 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 pro vided 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 locked in 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 switch is operated for on-line testing or maintenance during reactor power operation, reactor power and RCS pressure limits as specified in Table 2.3-1 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 10-10 amps on the intermediate range instrument.
Power is normally supplied to the control rod drive mechanisms from two separate parallel 600 volt sources. Each voltage source and its associated breakers and SCR control relays comprise a trip system. Thus, the two trip systems and their associated trip devices form a 1-out-of-2 logic used twice which is referred to as a 1-out-of-2x2 logic.
Amendment No. 199 (Unit 1)
Oconee Units 1, 2, & 3 3.5-2 Amendment No.
199 (Unit 2)
Amendment No.
196 (Unit 3)
TABLE 3.5.1-1 INSTRUMENTS OPERATING CONDITIONS (cont'd)
NOTES:
(a)
For channel testing, calibration, or maintenance, the minimum of three operable channels may be maintained by placing one channel in bypass and one channel in the tripped condition, leaving an effective one out of two trip logic.
(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)
(Deleted)
(e)
If minimum conditions are not met within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after hot shutdown, the unit shall be in the cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
(f)
- 1.
Place the inoperable Reactor Trip Module output in the tripped condition within one hour or
- 2.
Remove the power supplied to the control rod trip devices associated with the inoperable Reactor Trip Module within one hour.
(g)
(Deleted)
(h)
The RCP monitors provide inputs to this logic. For operability to be met either all RCP monitor channels must be operable or 3 operable with the remaining channel in the tripped state.
(i)
- 1.
The power supplied to the control rod drive mechanisms through the failed CRD Trip Breaker shall be removed within one hour or
- 2.
With one of the CRD Trip Breaker diverse features (undervoltage or shunt trip device) inoperable, restore it to OPERABLE status in 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or place the breaker in trip in the next hour.
(j) 1.
With one SCR Control Relay inoperable in logic channel C or D, restore the inoperable SCR Control Relay to OPERABLE status in 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or remove power from the CRD mechanisms supplied by the inoperable channel's SCR Control Relay within the next hour.
- 2.
With two or more SCR Control Relays inoperable in logic channel C or D, remove power from the CRD mechanisms supplied by the inoperable channel's SCR Control Relay within one hour.
Amendment No. 199 (Unit 1)
Oconee Units.1, 2, & 3 3.5-5c Amendment No. 199 (Unit 2)
Amendment No. 196 (Unit 3)
..instrumentation err s induced by drift can be expecd to remain within acceptable tolerances if recalibration is performed at the intervals speci fied.
Substantial calibration shifts within a channel (essentially a channel failure) are revealed during routine checking and testing procedures. Thus, the minimum calibration frequencies set forth are considered acceptable.
Periodic use of the Incore Instrumentation System for power mapping is suffi cient to assure that axial and radial power peaks and the peak locations are controlled in accordance with the provisions of the Technical Specifications.
REFERENCE (1)
FSAR, Section 7.2.3.4.
(2)
BAW-10167A, "Justification for Increasing the Reactor Trip System On line Test Interval."
Amendment No. 199 (Unit 1)
Oconee Units 1, 2, & 3 4.1-2 Amendment No. 199 (Unit 2)
Amendment No. 196 (Unit 3)
Table 4.1-1 INSTRUMENT SURVEILLANCE REQUIREMENTS M
Channel Description Check Test Calibrate Remarks C
- 1. Protective Channel NA MO NA Coincidence Logic in the Reactor Trip Modules
- 2.
Control Rod Drive NA MO(1)
NA (1) This test shall independently Trip Breakers, SCR confirm the operability of the Control Relays E and F shunt trip device and the undervoltage device.
- 3. Power Range Amplifier ES(l)
NA (1)
(1) Heat balance check each shift.
Heat balance calibration whenever indicated core thermal power exceeds neutron power by more than 2 percent.
- 4. Power Range ES 45 Days MO(1)(2)
(1) Using incore instrumentation.
STB (2) Axial offset upper and lower chambers after each startup if not done previous week.
- 5.
Intermediate Range ES(1)
PS NA (1) When in service.
H3 36.
Source Range ES(1)
PS NA (1) When in service.
- 7.
Reactor Coolant ES 45 Days RF Temperature STB
- 8.
High Reactor Coolant ES 45 Days RF Pressure STB a0 o
- 9.
Low Reactor Coolant ES 45 Days RF Pressure STB E.E.E.
- 10. Flux-Reactor Coolant ES 45 Days RF C+ C_+
Flow Comparator STB
- 11. Reactor Coolant Pressure ES 45 Days RF Temperature Comparator STB
Table 4.1-1 (CONTINUED) 0 D Channel Description Check Test Calibrate Remarks (D
- 12. Pump-Flux Comparator ES 45 Days RF STB
- 13. High Reactor Building DA 45 Days RF Pressure STB
- 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 RF
- b. Reactor Building Pressure (4 psig)
ES MO RF
- 16. Low Pressure Injection NA MO NA Logic
- 17. Low Pressure Injection Analog Channels:
- a. Reactor Coolant M M Pressure ES MO RF
- b. Reactor Building Pressure (4 psig)
ES MO RF
- 18. Reactor Building Emergency NA MO NA Reactor Building isolation Cooling and Isolation includes essential systems System Logic (Essential Systems)
, 0
- 19. Reactor Building Emergency ES MO RF Cooling and Isolation System Analog Channel E..E Reactor Building C-i-Pressure (4 psig)
Table 4.1-1 (CONTINUED)
O 0
Channel Description Check Test Calibrate Remarks (D
Remarks (D
- 49. Emergency Feedwater MO NA RF Flow Indicators
- 50. PORV and Safety Valve MO NA RF Position Indicators
- 51. RPS Anticipatory NA 45 Days RF Reactor Trip System Loss STB of Turbine Emergency Trip System Pressure Switches
- 52. RPS Anticipatory Reactor Trip System Loss of Main Feedwater a) Control Oil Pressure NA 45 Days RF Switches STB 01 b) Discharge Pressure NA 45 Days RF Switches STB
- 53. Emergency Feedwater Initiation Circuits a) Control Oil Pressure NA MO RF Switches
(
(
(
b) Discharge Pressure NA MO RF Switches 220 (D
(54.
Containment High Range NA MO RF TMI Item II.F.1.3 C-+C- + C-+
Radiation Monitor 2f (RIA-57, 58)
OOZ0
- O
E O
5 E
E 0
-4
-4 E-4 2OJ) 0 I I z
Ev E
L 44 44 E
4w MM~~
E-W 0
E 4*)
14
- 3 0
> )
H 0 E Z l
w w
0 0
CO r -rl 4-4 E
0 O
0
)
MG)
C I
I I I
Amnden N.
99(Uit1 41
~~~.
0 n 4E n U i
1, 2
an 3
-7 4)0 00 0
0 5
Amnden No 19 (Uni
- 2) w w
A d
No.
19 (U
i
- 3)
.z) 0~~ ~ ~
~
w..
0y tvW 1
- 4)
E r
.,q
~~
~
~
~
0.C4 3
0
- 4.
1
- 11) f1 u
0
- 0.
44 0
V)~4
- r.
0,
5 V
Q s4 d41 ts4)
)
- 0) 4-)-14 4-j"1 4j -4
- 4)
-u
--4
-4a)r P
.- 4 9:C
- E-0C 0
tv w
a C 4),-
0
- 4)
- 0) 0 0
- 0.
00.1 0
0
- i U-E
)
1.-
L)M 0
0
-n M~ -C4' Q.Cn Ln V
0 in o
W Ocne nis1,2
.4and 3 4.-8 Amnden No. 19 (ni 1
1-iAedmn No.- 199I (U'-t 2)>
~.
Amnmn No 196 (Ui 3)'-
.