ML17320A880
| ML17320A880 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 11/22/1983 |
| From: | Varga S Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17320A881 | List: |
| References | |
| NUDOCS 8312200046 | |
| Download: ML17320A880 (15) | |
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UNITED STATES e
NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 INDIANA AND MICHIGAN ELECTRIC COMPANY DOCKET NO. 50-315 DONALD C.
COOK NUCLEAR PLAI'IT UNIT NO.
1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 76 License No.
DPR-58 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Indiana and Michigan Electric Company (the licensee) dated January 22,
- 1982, as supplemented by letter dated July 3, 1983, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; 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 P
E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
83i2200046 83ii22 PDR ADOCK 050003i5 P
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No.
DPR-58 is hereby amended to read as follows:
(2)
Technical S ecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.
76
, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
4.
The change in Technical Specifications is to become effective within 30 days of issuance of this amendment.
In the period between issuance of the amendment and the effective date of the new Technical Specifications, the licensee shall adhere to the Technical Spdcifications for the systems, components, or operation existing at the time.
The period of time between changeover of systems, components, or operation shall be minimized or compensated for by suitable temporary alternatives.
This license amendment is effective as of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
,S' n
ar Operating Reactor ranch No.
1 Division of Licensing
Attachment:
Changes to Technical Speci ficati ons Date of Issuance:
November 22, lg83
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 INDIANA AND MICHIGAN ELECTRIC COMPANY DOCKET NO. 50-316 DONALD C.
COOK NUCLEAR PLANT UNIT NO.
2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.
57 License No.
DPR-74 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Indiana and Michigan Electric Company (the licensee) dated January 22,
- 1982, as supplemented by letter dated July 3, 1983, complies with the standards and
. requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; 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 amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No.
DPR-74 is hereby amended to read as follows:
(2)
Technical S ecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.
57
, are hereby incorporated in the license.
The licensee.shall operate the facility in accordance with the Technical Specifications.
3.
The change in Technical Specifications is to become effective within 30 days of issuance of this amendment.
In the period between issuance of the amendment and the effective date of the new Technical Specifications, the licensee shall adhere to the Technical Specifications for the systems, components, or operation existing at the time.
The period of time between changeover of systems, components, or operation shall be minimized or compensated for by suitable temporary alternatives.
4.
This license amendment is effective as of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
Attachment:
Changes to Technical Specifications Date of Issuance:
November 22, 1983 a
Operating Reactors 8
a h No. I Division of Licensing
ATTACHMENT TO LICENSE AMENDMENTS AMENDMENT NO. 76 TO FACILITY OPERATING LICENSE NO. DPR-58 AMENDMENT NO.
57 TO FACILITY OPERATING LICENSE NO.
DPR-74 DOCKET NOS. 50-315 AND 50-316 Revise Appendix A as follows:
Remove Pa es - Unit 1 3/4 3-11 3/4 3-12 3/4 3-13 3/4 3-14 3/4 3-26a Insert Pa es - Unit 1
3/4 3-11*
3/4 3-12 3/4 3-13*
3/4 3-14 3/4 3-26a Remove Pa es - Unit 2 3/4 3-11 3/4 3-12 3/4 3-13 3/4 3-14 3/4 3-25a Insert Pa es - Unit 2
~
3/4 3-11 3/4 3-12*
3/4 3-13 3/4 3-14*
3/4 3-25a
- Included for convenience only
CD n
n CD CD PC I
FUNCTIONAL.UNIT TABLE 3.3-2 Continued REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES
RESPONSE
TIME 12.
Loss of Flow - Single Loop (Above P-8) 13.
Loss of Flow - Two Loops (Above P-7 and below P-8) 14.
Steam Generator Water Level--Low-Low 15.
Steam/Feedwater Flow Mismatch and Low Steam Generator Water Level 16.
Undervoltage-Reactor Coolant Pumps 17.
Underfrequency-Reactor Coolant Pumps 18.
Turbine Trip A.
Low Fluid Oil Pressure B.
Turbine Stop Valve 19.
Safety Injection Input from ESF 20.
Reactor Coolant Pump Breaker Position Trip.
< 0.6 seconds
< 0.6 seconds
< 1.$ seconds NOT APPLICABLE 1
2 seconds
< 0.6 seconds NOT APPLICABLE NOT APPLICABLE NOT APPLICABLE NOT APPLICABLE
TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS Cl C)
C)
?C
'I 0
FUNCTIONAL UNIT 1,
Power Range, Neutron Flux 3.
Power
- Range, Neutron Flux, High Positive Rate 4.
Power Range, Neutron Flux, High Negative Rate 5.
Intermediate
- Range,
. 'Neutron Flux 6.
Source
- Range, Neutron Flux 7.
Overtemperature
)I)T 8.
Overpower hT 9.
Pressurizer Pressure--Low 10.
Pressurizer Pressure--High 11.
Pressurizer Mater Level--High 12.
Loss of Flow - Single Loop CHANNEL CHECK N.A.
N.A.
N,A.
CHANNEL CALIBRATION N.A.
D(2), M(3) and q(e)
R(6)
R(6)
R(e)
R(e)
CHANNEL FUNCTIONAL TEST S/U(l)
M S/U(1)
M. and S/U(l)
MODES IN MHICH SURVEILLANCE
~
RE UIRED N.A.
1, 2
1, 2
1 j 2 1,2and*
2(7), 3(7),
d and 5
)
1 ~
2 1
2 1
2
CD TABLE 4. 3-1 Continued REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS n
n CD CD 7CI FUNCTIONAL UNIT 13.
Loss of Flow - Two Loops 14.
Steam Generator Water Level-Low-Low 15.
Steam/Feedwater Flow Mismatch and Low Steam Generator Water Level 16.
Undervoltage - Reactor Coolant Pumps 17.
Underfrequency - Reactor Coolant Pumps 18.
Turbine Trip A.
Low Fluid Oil Pressure B.
Turbine Stop Valve Closure 19.
Safety Injection Input from ESF CHANNEL CHECK N.A.
N.A.
N.A.
N.A.
N.A.
CHANNEL CALIBRATION R
N.A.
N.A.
N.A.
CHANNEL FUNCTIONAL TEST N.A.
S/U(1)
S/U(l)
M(4)
MODES IN WHICH SURUEILLANCE tttlltt
~
1 1,
2 1,
2 1,
2 1,
2 1,
2 20.
21.
22., Automatic Trip Logic
/
Reactor Coolant Pump Breaker Position Trip Reactor Trip Breaker N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
M(5) and S/U(l)
M(5)
N.A.
)
2*
~ 1 2*
TABLE 4.3-1 Continued NOTATION (1)-
(2)
(3)
(4)-
(5)-
(6)
(7)
With the reactor trip system breakers Hosed and the control rod drive system capable of rod withdrawal.
If not performed in previous.? days.
Heat balance'nly, above 15K of RATED THERMAL POWER.
P Compare incore'o excore axial imbalance above 15K of RATED THERMAL POWER.
Recalibrate if absolute diffe) ence
> 3 percent.
Manual ESF functional input check:ev'ery 18 months.
Each train tested every other month.
Neutron detectors may be excluded from CHANNEL CALIBRATION.
Below P-6 (BLOCK OF SOURCE RANGE REACTOR TRIP} setpoint.
0.
C.
COOK-UNIT 1
I 3/4 3-14 Amendment No. 76 and 57 0
TABLE 3.3-4 Continued ENGINEERED SAFETY-FEATURE ACTUATION SYSTEH INSTRUHENTATION TRIP SETPOINTS FUtlCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES 6.
tlOTOR DRIVEN AUXILIARYFEEDHATER PUMPS a.
Steam Generator Hater 17K of narrow range Level -- Low-Low instrument span each steam generator
> 16K of narrow range instrument span each steam generator b.
4 kv Ous Loss of Voltage c.
Safety Injection d.
Loss of t1ain Feedwater Pumps 7.
TURBINE DRIVEN AUXILIARY FEEDMATER PUHPS a.
Steam Generator Hater Level -- Lolv'-Low I
3196 volts with a 2-second delay.
Not Applicable tlot Applicable
> 17K of narrow range instrument span each steam generator 3196, +10, -36 volts with a 2+.2 second delay Not Applicable Not Applicable
> 16K or narrow range instrument span each steam generator Reacto Coolant Pump Bus Und~rvoltpge LPSS OF POWER a>
4 kv Ous Loss of Voltage b
4 kv Bus Degraded Voltage
> 2750 Voltseach bus 3196 volts with a 2-second delay 3596 volts with a 2.0 min. time delay
> 2725 Voltseach bus 3196, +10, -36 volts with a 2+.2 second delay 3596, +36, -18 volts with a 2.0 minute.+ 6 second time delay
TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS FUNCTIONAl UNIT l.
Power
- Range, Heutron Flux 3.
Power Range, Neutron Flux, High Positive Rate 4,
Power Range, Neutron Flux, High Negative Rate CHANNEL CHECK H.A.
H.A.
H.A.
CHANNEL CHANNEL FUNCTIONAL CALIBRATION TEST N.A.
S/U(1.)
D(2), M(3)
M and j(6)
R(8)
M R(6)
MODES IN MHICH SURVEILLANCE RE U(RED
~
, H.A.
I l,i2 1,
2 1,
2 5.
Intermediate
- Range, Neutron Flux 6.
Source
- Range, Neutron Flux 7.
Overtemperature hT 8.
Overpower dT 9,
Pressurizer Pressure Low 10.
Pressurizer Pressure--High ll.
Pressurizer Mater Level--High a
12.
Loss of Flow - Single Loop
/
a7
~ a S
R(6)
R(6)
S/U(1) 1, 2 and *
(4 and S/U(l) 2(7); 3(7),
4 an+
M 1,
2 1,
2 1,
2 1,
2 ls 2
TABLE 4.3-1 Continued REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS FUNCTIONAL UNIT 13.
Loss of Flow - Two Loops 14.
Steam Generator Water Level--
Low-Low CHANNEL CHECK CHANNEL CALIBRATION R
CHANNEL FUNCTIONAL.
TEST N.A.
MODES IN WHICH SURVEILLANCE RE UIRED 1,
2 15.
Steam/Feedwater Flow Mismatch and Low Steam Generator Water L'evel 1,
2 16.
Undervoltage - Reactor Coolant Pumps 17.
Underfrequency - Reactor Coolant Pumps 18.
Turbine Trip A.
Low Fluid Oil Pressure B.
Turbine Stop Valve Closure 19.
Safety Injection Input from ESF 20.
Reactor Coolant Pump Breaker Position Trip 21.
Reactor Trip Breaker 22.
Automatic Trip Logic N.A.
N.A.
N.A..
N.A.
N.A.
N.A.
N.A.
N.A.
R N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
S/U(1)
S/U(1)
M(4)
R R(5) and S/U(l)
M(5) 1, 2
1, 2
1, 2
N.A.
1, 2*
1, 2*
TABLE 4.3-1 Continued NOTATION With the reactor tr'ip system breakers closed and the control rod drive system capable of rod withdrawal.
(1)
If not performed in previous 7 days.
(2)
Heat balance only, above 15% of RATED THERMAL POWER.
Adjust channel 'if absolute difference
> 2 percent.
(3)
Compare incore to excore axial offset 'above 15'X of RATED THERMAL POWER.
Recalibrate if absolute difference
> 3 percent.
(4)
Manual ESF functional input check every 18 months.
(5)
Each train tested every other month.
(6)
Neutron detectors may be excluded from CHANNEL CALIBRATION.
(7)
Below P-6 (BLOCK OF SOURCE RANGE REACTOR TRIP) setpoint.
D.
C.
COOK - UNIT.2 3/4 3<<13 Amendment No. 76 and 57
INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2.1 The Engineered -Safety Feature Actuation System (ESFAS) instrumenta-tion channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of,. Table
- 3. 3-4 and with RESPONSE TIMES as shown in Table 3.3-5.
APPLICABILITY:
As shown in Table
- 3. 3-3.
ACTION:
a.
With an ESFAS instrumentation channel trip setpoint less conserva-tive than the value shown in the Allowable Values column of Table 3.3-4, declare the channel inoperable and apply the applicable ACTION requirement of Table 3.3-3 until the channel is restored to OPERABLE status with the trip setpoint adjusted consistent with the Trip Setpoint value.
b.
With an ESFAS instrumentation channel inoperable, take the ACTION shown in Table 3.3-3.
SURVEILLANCE RE UIREMENTS 4.3.2.1.1 Each ESFAS instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations for the MODES..and at the frequencies shown in Table 4.3-2.
4.3.2.1.2 The logic for the interlocks shall be demonstrated OPERABLE during the automatic actuation logic test.
The total interlock function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by interlock operation.
4.3.2.1.3 The ENGINEERED SAFETY FEATURES
RESPONSE
TIME of each ESFAS function shall be demonstrated to be within the limit at least once per 18 months.
Each test shall include at least one logic train such that both logic trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once per N times 18 months where N is the total number of redundant channels in a specific ESFAS function as shown in the "Total No. of Channels" Column of Table 3 ~ 3 3
0.
C.
COOK - UNIT 2 3/4 3-14
TABLE 3.3-4 Continued ENGINEEREO SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS Cl C)
FUNCTIONAL UNIT 6.
MOTOR DRIVEN AUXILIARYFEEOMATER PUMPS a.
Steam Generator Mater Level -- Low-l>w TRIP SETPOINT
> 21X of narrow range instrument span each steam generator ALLOWABLE VALUES R 20K of narrow range instrument span each steam generator b.
C.
4 kv Bus Loss<of Voltage Safety Injection 3196 volts with a 2 second delay Not Applicable 3196, +18, -36 volts with a 2 'a 0.2 second delay Not Appl>cable 2
CL 0)
Ft d.
Loss of Hain Feedwater Pumps TURBINE DRIVEN AUXILIARY FEEDMATER PUHPS-a.
Steam Generator Mater Level -- Low-Low b.
Reactor Coolant Pump Bus Undervoltage
- 8. 'OSS OF POWER a.
4 kv Bus Loss of Voltage b.
4 kv Bus Degraded Voltage Not Applicable R 21X of narrow range instrument span each steam generator R 2750. Volts--each bus 3196 volts with a 2 second delay 3596 volts with a 2.0 minute time delay Not Applicable R 20X of narrow range instrument span each steam generator R 2725 Voltseach bus 3196, +1G, -36 volts with a 2.t:0.2 second delay
~ +36,
-,3 8 vol ts ~rith a 2.0 minute i 6 second time delay