ML15113A040
| ML15113A040 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 11/02/1981 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Duke Power Co |
| Shared Package | |
| ML15113A041 | List: |
| References | |
| DPR-38-A-102, DPR-47-A-102, DPR-55-A-099 NUDOCS 8111200484 | |
| Download: ML15113A040 (27) | |
Text
UNITED STATES LEAR REGULATORY COMMISS WASHINGTON, D. C. 20555 DUKE" POWER COMPANY DOCKET MO.
50-269 OCONEE NUCLEAR STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATInG LICENSE Amendment No. 102 License No. DPR-38
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The applications for amendment by Duke Power tompany (the licensee) dated September 8 and September 10, 1981, 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 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 iss uance of this amendment is in acccrdance with 10 CFR Part 51 of the Cormission's reaulations and all applicable requirements have been satis fied.
- 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 3.8 of Fac ility Operat ing 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.102 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
8111200464 811102 PDR ADOCX 05000269 P
- 3. This license amendment is effective as of the date of its issuance.
F R THE NUCLEAR REGULATORY COMMISSION On. Stolz, Chief er ting Reactors Branch #4, Di ision of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: November 2, 1981
o UNITED STATES LEAR REGULATORY COMMISSI U'
WASHINGTON, 0. C. 20555 DUKE POWER COM1PANY DOCKET NO. 50-270 OCONEE NUCLEAR STATION, UNIT NO. 2 AMEND1ENT TO FACILITY OPERATING LICENSE Amendment No.102 License No. DPR-47
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The applications for amendment by Duke Power tompany (the licensee) dated September 8 and September 10, 1981, 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 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 acccrdance with 10 CFR Part 51 of the Commiss.ion's reoulations. and all applicable requirements have been. satis fied.
- 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 3.B of Facility.Operating License No. DPR-47 is hereby amended to read as follows:
3.8 Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.102 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
-2
- 3. This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION ha.
Stolz, Chief ting Reactors Branch #4 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: November 2, 1981
UNITED STATES LEAR REGULATORY COMMISS WASHINGTON, D. C. 20555 DUKE-POWER COMPANY DOCKET NO.
50-287 OCONEE NUCLEAR STATION, UNIT NO. 3 A1ENDMENT TO FACILITY OPERATIfG LICENSE Amendment No. 99 License No..DPR-55
- 1. The Nuclear Regulatory Commission (the Commissi on) has found that:
A. The applications for amendment by Duke Power Company (the licensee) dated September 8 and September 10, 1981, 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 Comission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of, the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satis fied.
- 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 3.B of Facility Operating License No. DPR-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.. 99 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
F.s
- 3. This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Jn. Stolz, Chief 0 er ting Reactors Branch #4 Di sion. of Licensing,
Attachment:
Changes to the Technical Specifications Date of Issuance: November 2, 1981
ATTACHMENT TO LICENSE AMENDMENTS AMENDMENTNO.102 TO DPR-38 AMENDMENT NO.102 TO DPR-47 AMENDMENT NO. 99 TO DPR-55 DOCKETS NOS. 50-269, 50-270 AND 50-287 Replace the following pages of the Appendix "A" Technical Specifications with the attached pages.
The revised pages are identified by amendment numbers and contain vertical lines indicating the area of change.
Remove Pages Insert Pages 2.1-3d 2.1-3d 2.1-9.
2.1-9 2.1-12 2.1-12 2.3-2 2.3-2 2.3-3 2.3-3 2.3-5 2.3-5 2.3-6 2.3-6 2.3-7 2.3-7 2.3-10.
2.3-10 2.3-11 2.3-11 2.3-12 2.3-12 2.3-13 2.3-13 3.5-1 3.5-1 3.5-2 3.5-2 3.5-3 3.5-3 3.5-4 3.5-4 3.5-5 3.5-5 3.5-5a 3.5-5a 3.5-5b 4.5-3 4.5'-3
- 2.
The combination of radial and axial peak that causes central fuel melting at the hot spot.
The limit is 20.15 kw/ft for Unit 3.
Power peaking is not a directly observable.quantity, and, therefore, limits have been established on the bases of the reactor power imbalance produced by the power peaking.
The specified flow rates for Curves 1, 2 and 3 of Figure 2.1-2C. correspond to the expected minimum flow rates with. four pumps, three pumps and one pump in each loop, respectively.
The magnitude of the rod bow penalty applied to each fuel cycle is equal to or greater than the necessary burnup independent DNBR rod bow penalty for the ap plicable cycle minus a credit of 1% for the flow area reduction factor used in the hot channel analysis (4).
All plant operating limits are presently based on an original method of cal culating. rod bowing penalties that are more conservative than those that would be obtained with new approved procedures (4).
For Cycle 6 operation, this sub rogation results in a 10" DNBR margia, which is partially used to offset the reduction in DNBR due to fuel rod bowing..
The maximum thermal power for three-pump operation is 90.65 percent due to a power level trip produced by the flux-flow ratio 74.7 percent flow x 1.08 =
80.7 percent power plus: the maximum calibratiod and. instrument error (Reference 4).
The maximum thermal power for other coolant pump conditions are producedin a similar manner.
For each curve ofFigure 2.1-3C a pressure-temperature point above and to the left of the curve would result in a DNBR greater than 1.30 or a local quality at the point of minimum DNBR less than 22 percent forithat particular reactor coolant pump situation.
The curve, ofFigure 2.1-IC is the most restrictive of all possible reactor coolant pump-maximum thermal power combinations shown in Figure 2.1-3C.
References (1) Correlation of Critical Heat Flux ina Bundle Cooled by Pressurized Water, BAW-10000, March 1970.
(2)
Oconee 3, Cycle 3.
'Reload Report -
3AW-1453, August, 1977.
(3)
Amendment 1 Oconee 3, Cycle4 - Reload Report -
BAW-1486, June 12, 1978.
(4)
Oconee 3, Cycle 6
'Reload Report -
3AW-1634, ugust, 1980.
Amendments Nos. 102, 102 99 213d
THERMALPOWER LEVEL, 120
(-27, 112) 3 112' ACCEPTABLE 110' 4PUMP 202 OPERATION (37. 97 7) 27 902. 65(
3090. 65)
( 55,90. 2)
ACCEPTABLE 90 3&4 PUMP IOPERATION 80 37, 7 -35 2557
- 63. 285)
)703 (3.0.63. 26)
ACCEPTA8LE 80 2; 3&4 PUMP OPERATION 50 I
(37. 48.96) 55 41 46)
-40I UNACCEPTABLE 30 OPERATION:
20 UNACCEPTA8LE 2
I OPERATION 10 0 -50 20
-10 10 20 30 40 50 60 Reactor Power Imbaianca, CORE PROTECTION SAFETY LM ITS O3 CETUNIT 3N
~DK~UOCONEE. NUCLEAR STATION Figure 2.1-2C Amendments Nos. 102 102, &9 2.9
2400 ACCEPTABLE OPERATION/
2200 2000 800 1800 560 580 600 620 640 Reactor Cool ant Outl at Tmperature. F CURVE COOLANT FLOW. GPM
- POWER, PUMPS OPERATING TYPE OF M1T 374, 880 1005)*
112 4
ONBR 2
280, 035 (74 7%)
90, 65 3
ONBR 3
183,690 (49.05) 63 26 2
QUALI TY
=106.55 OF FIRST-CORE DESI GN FOW CORE PROTECTION SAFETY LIM TS UNIT 3 OCONEE NUCLEAR STATION Figure 2.1.30 Amendments Nos.10 2 102
& 99
During normal plant operation with all reactor coolant pumps operating, reactor trip is initiated when the reactor power level reaches 104.9% of rated power. Adding to this the possible variation in trip setpoints due to calibration and instrument errors, the maximum actual power at which a trip would be actuated could be 112%, which is more conservative-than the value used in the safety analysis.
(4)
Overpower:Trip Based on Flow and Imbalance The power level trip setpoint produced by the reactor coolant system flow is based on a powe-, to-f low ratio which has been established to accommodate the most severe thermal, transient considered in the. design, the. loss-of-coolant flow accident. from high power
- Analysis has. demonstrated-that the specified power-to-flow ratio is adequate. to prevent a-DNR of less than 1.3 should a low flow condition exist due-to any electrical malfunction.
The power level trip setpoint produced by the power-to-flow ratio provides both high power level and low flow protection in the event the-reactor power level increases or the reactor coolant flow rate decreases. The power level trip setpoint produced by the power-to-flow ratio provides. overpower DNB pro tection for-all modes of pump.operation. For every flow rate there is a maxi mum permissible power level,. and for every power level there is a minimum permissible low flow rate. Typical power level and Low flow rate combinations for the pump situations of Table 2.3-1A are as follows
- 1. Trip.would occur when four reactor coolant pumps are operating if power is 108% and reactor flow rate is 100%, or flow rate is 92.59% and power
..level is. 100%.
- 2. Trip would occur when three reactor coolant pumps are operating if power is 80.68% and reactor flow rate is 74.7% or flow rate is 69.4 and power level is 75%.
- 3. Trip would occur when one reactor coolant pump is operating in each loop (total of two pumps operating) if the power is 52.92 and reactor flow rate is 49.0% or flow rate is 45.37% and the power level is 49%.
The flux-to-flow ratios account for the maximum calibration and instrument errors and the maximum variation from the average value of the RC flow signal in such a manner that the reactor protective system receives a conservative indication of the RC flow.
For safet-r calculations. the maximum calibration and instrumentation arrors for the power level trip were-used.
The power-imbalance boundaries are established in order to prevent, reactor thermal limits from being. exceeded. These thermal limits are either power peaking kw/ft limits or DNBR limits.
The reactor power imbalance (power in the top half of core minus power in the bottom half of core) reduces the power level trip produced by the power-to-flow -ratio-such that the boundaries of Figure 2.3-2A-Unit 1 are produced. The power-to-flow ratio reduces the power 2.3-2B - Unit 2 2.3-2C - Unit 3 Amendments Nos.102, 102, 99 2.3-z
level trip and associated reactor power/reactor power-imbalance boundaries by 1.08% - Unit 1 for 1% flow reduction.
1.08%
- Unit 2 1.081' - Unit 3 Pump.lonitors 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 theipower 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 auclear over power trip setpoint. The trip setting limit shown in Figure 2.3-LA - Unit I 2.3 Unit 2 2.3-1C - Unit 3 for high reactor coolant system pressure (2300 AF) 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 lo pressure (11.14 To
-4706) trip (1800) psig a(1.1 Tout4 706) out d(1300) psig (1.1 Tt-4706) setpoints shown in Figure 2.3-lA have been established to maintain the DNB 2.3-13 2.3-IC ratio greater-than or equal to 1.3 for those design accidents that result in a pressure reductio.
(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 T 4746) out -
4746)
(11.14 Tout -
746) out Coolant Outlet Temnerature The high reactor coolant outlet temperature rito setting limit (618 0F) shown in Figure 2.3-LA has been established to orevent excessive core coolant Fu 2.3-13 2.3-1C temperatures in he operating range. Due r6 calibration and instrumentation errors, the safety analysis used a trip. setpoint of. 620 0F.
Reactor BUilding ?ressure The high reactor building -pressure trip setting limit (4 psig) provi-des positive assurance that a reactor trip will occur in the unlikely event of a.oss-of coolant accident, even in the absence of a low reactor coolant system pressure Ar p.
Amendments Nos.l102
,102 9
2.33 3
2400 23 00 PSI 200 ACCEPTABLE OPERATION 2100 UNACCEPTABLE OPERATION 2000 1900 P =1800 PSIG 1800 54F 0.-56 580:
Sao.
620
-640.
Reactor Ou tet Tmp erature, F PROTECTIVE SYSTEM MAXIMUM ALLOWA8LE SETPOINTS UNIT.1 OCONEE NUCLEAP STATION Figure 2.3-1 A Amendments Nos. 102 102 99 2.3-5
2400 2300 P3, T =1618 0F P. 23010 P SIG 4
2200 ACCEPTABLE OPERATION 2100 UNACCEPTABLE Z
OPERATION 2000 1900 P
800 PS 1T0 584F.
540 56 580 600 620 640 Reactor Outl e Temperature, F PROTECTIVE SYSTEM MAX IMUM ALLOWABLE SETPOINTS UNIT 2 onU OCONEE NUCLEAR STATION
- Figure 2.3 1 B Amendments Nos. 102
.102, 9
3-6
2400 2300 618 F P
230.0. PSIG ACCEPTABLE 2200 OPERATOON 2100 UNACCEPTABLE OPERATION 2000 1900 P
1800 PS 800 T
584F 540 560 580 600 620 640 Reactor Oul et TemperatureF PROTECTIVE SYSTEM MAXIMUM ALLOWABLE SETPOINTS UNIT 3 OCONEE NUCLEAR STATION Figure 2.3-IC Amendments Nos. 102 102 99 2.3-7
(HERMAL POWER LEVEL 8.0.10808 0
.)1 1
9032 100.
M
.864 CCEPTABLEN 4 PUMP OPERATOON 90T 24..87. 5.
39,80. 0
-8.80. 6)
- 13. 80o: 6)
UNACCUPTA8LE CCEPTABLE OPERATION UNACCEPTA8LE 3&4 PUMP OPERATION OPERATION 60 I(24
- 60. )
.39, 52. 6)
-8,
- 52. 3) 13,52. 9)
-50 CCEPTASLE
-40 2, 3&4 PUMP OP ERATJ N 24,32. 4
-30
-39.24.9)I 20
-40
-30
-20
-10 0
10 20 30 40 Power Imbalance, PROTECTIVE SYSTEM MAX IMUM ALLOWABLE SETPO INTS UNIT 3 OCONEE NUCLEAR STATION Fi gure 2.
3 2C Amendments Nos.
02 102 99 2.3-10
Table 2.3-lA Unit I Reactor Protective System Trip Setting Limits One Reactor oFour Reactor Three Reactor Coo-nt Pu Coolant Pumps Coolant Pumps Operati Operating Operating Each Loop (Operating Power (OeaigPwr (Ope ra ting Power Sudw CnRPS Segment Po00 raed RPSS~ct-0%Rtd
-75% Rated)
-49% Rated)
Bps
- 1.
Nuclear Power Max.
104.9 104.95.
(% Rated)
- 2.
Nuclear Power Max. Based 1.08 times flow 1.08 times flow 1.0* times flow Bypassed on Flow (2) and Imbalance, minus reduction minus reduction minus reduction
(% Rated)'
due to imbalance.
due. to imbalance.
due tioimane
- 3.
Nuclear Power Max Based NANA on Pump Monitors,
(%,Rated)
- 4.
High Reactor Coolant System 2300 2300 (4)
Pressure, psig, flax.
- 5.
Low Reactor Coolant System 1800 1800 1800 Bypassed Pressure, psig, Min.
- 6.
Variable Low Reactor Coolant (11.1 T
4706)()
(11.14 T 4706)(1)
(11.14 T 4706)(1)
Byjvussed System Pressure psig, Min.
out o)
- 7.
Reactor Coolant Temp. F., Max.
618 618 618 618
- 8.
High Reactor Building 4
4 4
4 Pressure, psig, Max.
(1) is in degrees Fahrenheit (0F).
out (2) Reactor Coolant System Flow, %.
(3) Administratively controlled reduction set only during reactor shutdown.
(4) Automatically set when other segments of the RPS are bypassed.
Table 2.3-in Unit 2 Ej Reactor Protective System Trip Setting Limits rt EO One Reactor Four Reactor Three Reactor Coolant Pump 0
Coolant Pumps Coolant Pumps Operating in Operating Operating Each Loop (Operating Power (Operating Power (Operating Power Shutdown RPS Segment
-100% Rated)
-75% Rated)
-49% Rated)
Bypass
- 1.
NuclearaPower Max.
104.9 104.9 104.9 5.0
(% Rated)
- 2.
Nuclear Power Hax. Based 1.08 times flow 1.08 times flow 1.08 times flow Bypassed on Flow (2) and Imbalance, minus reduction minus reduction minus reduction
(% Rated) due to imbalance due to imbalande due to imbalance
- 3.
Nuclear Power Hax. Based NA NA 55%
Bypassed on Pump Monitors. (f Rated)
High Reactor Coolant System 2300 2300 2300 1720 Pressure, psig, Hax.
- 5.
Low Reactor Coolant Systent 1800 1800 1800 Bypassed Pressure, psig, Min.
6 Variable Low Reactor Coolant (11.14 T
-,4706)
- 11. 14 Tu 4706)(1)
(11.14 T 4706)(1)
Bypassed System Pressure psig, Hin.
7.-
Reactor Coolant Temp. F., Max.
618 618 618 618
- 8.
High Reactor Building 4
4 4
4 Pressure, psig, Max.
(2)
T is in degrees Fahrenheit (oF).
(2)
Reactor Coolant System Flow, 7.
(3) Administratively controlled reduction set only during reactor shutdown.
(4) Automatically set when other segments of the RPS are bypassed.
Table 2.3-IC Unit 3 rt Reactor Protective System Trip Setting Limits One Reactor Four Reactor.
Three Reactor Coolant Pump Coolant Pumps Coolant Pumps Operating in Operating Operating Each Loop (Operating Power (Operating Power (Operating Power Shutdown R PS Segment
-100% Rated)
-75% Rated)
-49% Rated)
Bpass
- 1.
Nuclear Power ax.
104.9 104.9 104.9 5.0(3)
(% Rated)
- 2.
Nuclear Power Max. Based 1.08 times flow 1.08 times flow 1.08 times flow Bypassed on Flow (2) and Imbalance, minus reduction minus reduction minus reduction
(% Rated) due to imbalance due to imbalance due to imbalance Nuclear Power Hax. Based NA NA 55%
Bypassed0 on Pump Monitors, (% Rated) l4.
igh Reactor Coolant System 2300 2300 2300 1720(4)
Pressure, psig, Max.
- 5.
Low Reactor Coolant System 1800 1800 1800 Bypassed Pressure, psig, Min.
- 6.
Variable Low Reactor Coolant (11.14 To 4706)
(11.14 T 4706)
(11.14 T 4706)(".
Bypassed System Pressure,psig, Hin.
- 7.
Reactor Coolant Temp. F., Max.
618 618 618 618
- 8.
High Reactor Building 4
4 4
4 Pressure, psig, Max.
T T
is in degrees Fahrenheit (*F).
out (2) Reactor Coolant System Flow, %.
(3) Administratively controlled reduction set only during-reactor shutdown.
(4) Automatically set when other segments of the RPS are bypassed.
3.5 INSTRUMNTATION SYSTEMS 3.5.1 Operational Safety Instrumentation Applicability Applies; to unit instrumentation and control systems.
Objective To delineate the conditions of the unit instrumentation and safety circuits necessary to assure reactor safety.
Specifications 3.5.1.1 The reactor shall not be in a startup mode or in a critical state unless the requirements of Table 3.5.1-1, Column C are met.
3.5.1.2 In the event that the number of protective channels operable falls below the limit.given under Table 3.5.1-1, Column C; operation shall be limited as specified in Column D.
3.5.1.3 For on-line testing or in the event of a protective instrument or channel failure, a key-operated channel bypass switch associated with each reactor protective channel may be used to lock the channel trip relay in the untripped state. Status of the untripped state shall be indicated by a light. Only one channel bypass key shall be accessible for use in the control room. Only one channel shall be locked in this untripped state or contain a dummy bistable at any one time.
3.5.1.4 For on-line testing or maintenance during reactor power operation, a key-operated shutdown bypass switch associated with each reactor protective channel may be used in conjunction with a key-operated channel bypass switch as limited by 3.5.1.3.
Status of the shutdown bypass switch shall be indicated by a light.
3.5.1.5 During startup when the intermediate range instruments come on scale, the overlap between the intermediate range and the source range instrumentation shall not be less than one decade. If the overlap is less than one decade, the flux level shall not be greater than that readable on the source-range instruments until the one decade overlap is achieved.
3.5.1 6 In the event that one of the trip devices in either of the sources supplying power to the control rod drive mechanisms fails in the untripped state, the power supplied to the rod drive mechanisms through the-failed trip device shall be manually removed within 30 minutes. The condition will be corrected and the remaining trip devices shall be tested within eight. hours.
If the condition is not corrected and the remaining trip devices tested within the eight hour period, the reactor shall be placed in the hot shutdown condi tion within an additional four hours.
Amendments Nos. 102 - 102, & 99 3.5-1
Bases Every reasonable effort will be made to maintain all safety instrumentation in operation. A startup is not permitted unless three power range neutron instrument 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 pressure.
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. 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 C (Table 3.5.1-1). A tripped channel is considered to be operable. 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 isisolated from the reactor protective system is provided as a part of the teactor control 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 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 duringireactor power operation, reactor power and RCS pressure limits as specified in Table, 2.3-1A, B, or C are not applicable.
The source range and intermediate range auclear instrumentation overlap by
'one decade of neutron flux.
This decade overlap will be achieved.at 10-1o amps on the intermediate range instrument.
Power is normally supplied to the' 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 Amendments Nos 102 102 99
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.
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. 102 102 99 3.5-3
m TABLE 3.5.1-1 INSTRUMENTS OPERATING CONDITIONS 0
(A)
(B)
(C)
(D)
HINIMUM Operator Action If Conditions TOTAL NO.
CHANNELS CilANNELS Of Column C FUNCTIONAL UNIT OF CHANNELS TO THIP OPERABLE Cannot Be Met
.I Nuclear lastruwnentation 2
NA 1
Bring to hot shutdown within Intermediate Range 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (b)
Channels
- 2.
Nuclear Instrumentation 2
NA I
Bring to hot within Source Range Channels 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (b) (c)
- 3.
HPS Manual Pushbutton.
I Bring to hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
- 4.
RPS Power Range 4
2 3(a)
Bring to hot shutdown within Inst rument Channels 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
- 5.
2 3
Bring to hot shutdown within Temnperature Instrumen 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Channels
- 6.
RPS Pressure-Temperature 2
3 Bring to hot shutdown within Instruments Channels 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
- 7.
RPS Flux Imbalance 2
3 Bring.to hot shutdown within.
Flow Instrument Channels 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
- 8.
UPS Reactor Coolant Pressure
- a.
High Reactor Coolant 4
2 3
Bring to hot.shutdown within Pressure Instrument.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Channels
- b.
Low Reactor Coolant 42 3
Bring to hot shutdown within Pressure Channels 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
- 9.
RPS Power-Number of Pumps 4
2 3
Bring to hot shutdown within unastrunient Channels 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />'(h)
m*
TABLE 3.5.1-1 INSTRUMENTS OPERATING CONDITIONS (cont'd)
CT 0
(A)
(B)
(C)
(D)
HINITOUM Operator Action If Conditions
.TOTAL NO.
CHANNELS CHANNELS Of Colun C FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE Cannot Be Met
- 10.
RPS High Reactor Building 2
3 Bring to hot shutdown within Pressure Channels 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> II.
RPS Anticipatory Reactor Trip System (g)
Loss of Turbine 2
3 bring to hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
- b.
Loss of Hain Feedwater 4
2 3
Bring to hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
- 12.
ESF High Pressure In jection System and Reactor Building isolation (No-essential Systems)
- a.
2 3
Bring to hot shutdown within Pressure Instrument 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> e)
Channels Reactor Building 3
2 3
Bring to hot shutdown within 4 PSIG Instrument 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (e)
Channels
- c.
Manual Pushbutton 2
1 2
Bring to hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (e)
- 13.
ESF Low Pressure lijec tion System
- a.
2 3
Bring to hot shutdown within Pressure Instruiment 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (e)
Channels
TABLE 3.5. 1-1 INSTRUMENTs OPERATING CONDITIONS (cont'd) m (A)
()(C)
(D) 0 HINIHMi Operator Action If Conditions TOTAL NO.
CHANNELS ChANNELS Of Colum C
JFUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE CanInot Be He
- b.
Reactor Building 3
2 3
Bring to hot shutdown within 4 PS16 Instrumient 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (e)
Channels
- c.
Hanual hishbutton 2
1 2
Bring to hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (e)
- 14.
ESF Reactor Builiiag Isolation (Essential Systems)
& Reactor Building Cooling Systemu
- a.
Reactor Buil ding 3
2 3
Bring to hot shutdown within 4 PSIG Instru ment 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (e)
Channel
- b.
Manual Pushbuttou 2
1 2
Bring to hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (e)
- 15.
ESF Reactor Building Spray System
- a.
Reactor Building 3
2 3
Bring to hot shutdown within High Pressure 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (e)
Ins trumetu Channel
- b.
Manual Pushbutton 2
2 Bring to hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (e)
- 16.
Turbine Stop Valves 2
2 Bring to hot shutdown within Closure 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (f)
TABLE 3.5.1-1 INSTRUMENTS OPERATING CONDITIONS (cont' d)
NOTES:
(a) For channeltesting, calibration, or maintenance, one of the three minimum operable channels may be put into manual bypass leaving, a one out of two trip logic 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) (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 condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
(f) One channel may be inoperable for no more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> before going to the hot shutdown condition.
(g)
This requirement is applicable as follows:
Unit 1 -
following Sumer 1981 refueling outage Unit 2 - following Fall 1981 refueling outage Unit 3 -
April 1, 1981 (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.
Amendments Nos. 102, 102 99 3.5-5b
The High Pressure Injection System under normal operating conditions. has one pump operating. At least once per month, operation is rotated to another high pressure injection pump. This verifies that the high pressure injection pumps are operable.
The requirements of the Low Pressure Service Water System for cooling-water are more severe during normal operation than under accident codditions.
Rotation of the pump in operation on a monthly basis verifies that two pumps are operable-.
The low pressure injection pumps are tested singularly for operability by opening the borated. water storage tank outlet valves and the bypass valves in the borated water storage tank fill line.
This allows water to be pumped from the borated water storage tank through each of the injection lines and back to the tank.
Testing the manual operability of power-operated valves in the Low Pressure Injection System gives assurance that flow can be established in a timely manner even if the capability to operate a valve from the control room is lost.
With the reactor shut down, the valves-in each core flooding line are checked for operability by reducing the Reactor Coolant System Pressure until the indicated level in the core flood tanks verify that the check and, isolation valves have opened.
Power Operated Valves LP-17 and LP-18, are boundary valves between high pres sure and low pressure design piping.. As such, functional testing of these valves is performed during cold shutdown conditions when the Reactor Coolant System pressure is below the design pressure of the Low Pressure Injection System piping and the potential for over-pressurization of the low pressure system is eliminated. Check Valves CF-12, CF-14, LP-47, and LP-48 are located on the high pressure piping and therefore"cAn be leak tested with the Reactor Coolant System at hot shutdown conditions.
REFERENCE (1) FSAR, Section6' Amendments Nos. 02
,102
&.99 4.5-3