ML20072L379
| ML20072L379 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 08/24/1994 |
| From: | Kuo P Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20072L381 | List: |
| References | |
| NUDOCS 9408310235 | |
| Download: ML20072L379 (20) | |
Text
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UNITED STATES
[ [(
' )j NUCLEAR REGULATORY COMMISSION
~ !
WASHINGTON, D.C. 20655-0001
%...../
BALTIMORE GAS AND ELECTRIC COMPANY DOCKET NO. 50-317 CALVERT CLIFFS NUCLEAR POWER PLANT UNIT NO. 1 AMENDMENT TO FACILITY OPERATING l! CENSE Amendment No.193 License No. DPR-53 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Baltimore Gas and Electric Company (the licensee) dated May 27, 1994, complies wtth 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-53 is hereby amended to read as follows:
9408310235 940824 PDR ADOCK 05000317 P
b
. 2.
Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment' No.193, 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 the date of its issuance and shall be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION e
v-P.. T. Kuo, Acting Director projectDirectorateI-1
,0ivision of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: August 24, 1994 i
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UNITED STATES
[. l, j
NUCLEAR REGULATORY COMMISSION
'e WASHINGTON, D.C. 20S56-0001
/
~ BALTIMORE GAS AND ELECTRIC COMPANY DOCKET NO. 50-318 CALVERT CLIFFS NUCLEAR POWER PLANT. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.170 License No. DPR-69 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Baltimore Gas and Electric Company (the licensee) dated May 27, 1994, 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 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 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-69 is hereby amended to read as follows:
' 2.
Technical Soecifications The Technical ' Specifications contained in Appendices A and B, as revised through Amendment No.170,'are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
i 3.
This license amendment is effective as of the date of its issuance and shall be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION 0- 6== - l' & ~ -
P. T. Kuo, Acting Director Project Directorate I-l Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation j
l Attachment-Changes to the Technical Specifications Date of Issuance: August 24, 1994
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4 ATTACHMENT TO LICENSE AMENDMENTS AMENDMENT NO. 193 FACILITY OPERATING LICENSE N0. DPR-53 AMENDMENT NO. 170 FACILITY OPERATING LICENSE NO. DPR-69 DOCKET NOS. 50-317 AND 50-318 Revise Appendix A as follows:
Remove Paaes Insert Paaes 3/4 3-6 3/4 3-6 3/4 3-7 3/4 3-7 3/4 3-19 3/4 3-19 3/4 3-21 3/4 3-21 3/4 3-22 3/4 3-22 B 3/4 3-1 B 3/4 3-1 i
I
.c n,-
g TABLE 4.3-1 y
G g
REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREMENTS n
z C
CHANNEL MODES IN WHICH Zi 7
CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE E
FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED M
8 Z
1.
Manual Reactor Trip NA NA S/Um NA g
2.
Power Level - High j
a.
Nuclear Power S
Dm,MW N
,Q Q
1, 2 l
b.
AT Power S
DN,R Q
1 l
3.
Reactor Coolant Flow - Low S
R Q
1, 2 l
g 4.
Pressurizer Pressure - High S
R Q
1, 2 l
5.
Containment Pressure - High S
R Q
1, 2 l
w 6.
Steam Generator Pressure - Low S
R Q
1, 2 l
7.
Steam Generator Water Level -
S R
Q 1, 2 l
Low 8.
Axial Flux Offset S
R Q
1 l
9.
a.
Thermal Margin / Low Pressure S R
Q 1, 2 l
N b.
Steam Generator Pressure S
R Q
1, 2 l
R Difference - High k
- 10. Loss of Load NA NA S/UN NA E
w l
52 TABLE 4.3-1 (Continued)
EE f'
REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE REQUIRENENTS EE e,
E:
CHANNEL MODES IN WHICH
$3
- q CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE EE v'
FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED
- Ri e
ac
- 11. Wide Range Logarithmic Neutron Flux S
RfD S/UIU 1, '2 52 and, 3, 4, 5 c:
{}
Monitor-
{j
- 12. Reactor Protection System Logic NA NA Q and S/U(U 1, 2 l
Matrices
- 13. Reactor Protection System Logic NA NA Q and S/U(U 1, 2 l
Matrix Relays
- 14. Reactor Trip Breakers NA NA M
1, 2 and
- Gb if a
n W
.if 43
i Q
TABLE 4.3-2 R
G g
ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS w
G n
C CNANNEL MODES IN WHICN 4
4 CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE E
4 FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED R
"i C
- D 1.
SAFETY INJECTION (SIAS) d E
~
a.
Manual (Trip buttons)
NA NA R
NA b.
Containment Pressure - High S
R Q
1,2,3 c.
Pressurizer Pressure - Low S
R M(1)g)(3) 1, 2. I d.
Automatic Actuation Logic NA NA 1, 2, 3 2.
R i
a.
Manual (Trip buttons)
NA NA R
NA Y
b.
Containment Pressure - High S
R M g(6) 1,2,3 l
il G
c.
Automatic Actuation Logic NA NA 1, 2, 3 3.
CONTAINMENT ISOLATION (CIS)'
a.
Manual CIS (Trip buttons)
NA NA R
NA b.
Containment Pressure - High S
R M(]1 U) 1, 2, 3 l
c.
Automatic Actuation Logic NA NA 1, 2, 3 N
a et E
t w
b
t g
TABLE 4.3-2 (Continued)
G P
g ENGINEER,D SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS w
b CHANNEL MODES IN WHICH 7
CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE E
FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED E
5 C
g 4.
MAIN STEAM LINE ISOLATION (SGIS)
~
a.
Manual SGIS (MSIV Hand Switches NA NA R
NA and Feed Head Isolation Hand Switches) b.
Steam Generator Pressure - Low S
R 1,2,3 l
M gN U
c.
Automatic Actuation Logic NA NA 1, 2, 3 g
5.
CONTAINMENT SUMP RECIRCULATION (RAS)
Y a.
Manual RAS (Trip Buttons)
NA NA R
NA b.
Refueling Water Tank - Low NA R
1,2,3 l
Mg) c.
Automatic Actuation Logic NA NA 1, 2, 3 6.
CONTAINMENT PURGE VALVES ISOLATION a.
Manual (Purge Valve Control NA NA R
NA Switches) b.
Containment Radiation - High Area S
R Q
6**
l
[
Monitor E
il i
a E
i
=
w m
I L
g TABLE 4.3-2 (Continued)
R G
g ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS w
b CHANNEL MODES IN WHICH q
CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE E
FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED i5i "i
C 7.
LOSS OF POWER y
=
~
a.
4.16 ky Emergency Bus Undervoltage NA R
Q 1,2,3 l
(Loss of Voltage) b.
4.16 kv Emergency Bus Undervoltage NA R'
Q 1, 2, 3 l
(Degraded Voltage) 8.
CVCS ISOLATION' R
West Penetration Room / Letdown NA R
Q 1,2,3,4 l
Y Heat Exchanger Room Pressure - High N
9.
Manual (Trip Buttons)
NA NA R
NA b.
Steam Generator Level - Low S
R Q
1,2,3 c.
Steam Generator AP - High S
R Mg) 1, 2, 3 d.
Automatic Actuation Logic NA NA 1, 2, 3 E
a an m
3/4.3 INSTRUMENTATION TABLE 4.3-2 (Continued)
TABLE NOTATION Containment isolation of non-essential penetrations is also initiated by SIAS (functional units 1.a and 1.c).
Must be OPERABLE only in MODE 6 when the valves are required OPERABLE and they are open.
(1)
The logic circuits shall be tested manually at least once per 31 days.
(2)
SIAS logic circuits A-10 and B-10 shall be tested monthly with the l
exception of the Safety Injection Tank isolation valves. The SIAS i
logic circuits for these valves are exempted from testing during operation; however, these logic circuits shall be tested at least once per 18 months during shutdown.
(3)
SIAS logic circuits A-5, and B-5 are exempted from testing during operation; however, these logic circuits shall be tested at least j
once per 18 months during shutdown.
(4)
CIS logic circuits A-5 and B-5 are exempted from testing during operation; however, these logic circuits shall be tested at least once per 18 months during shutdown.
(5)
SGIS logic circuits A-1 and B-1 are exempted from testing during operation; however, these logic circuits shall be tested at least once per 18 months during shutdown.
(6)
CSAS logic circuits A-3 and B-3 are exempted from testing during operation; however, these logic circuits shall be tested at least once per 18 months during shutdown.
CALVERT CLIFFS - UNIT 1 3/4 3-22 Amendment No. 193
3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 PROTECTIVE AND ENGINEERED SAFETY FEATURES (ESF)
INSTRUMENTATION The OPERdBILITY of the protect %e and ESF instrumentation systems and bypasses ensure that 1) the associated ESF action and/or reactor trip will be initiated when the parameter monitored by each channel or combination thereof exceeds its setpoint, 2) the specified coincidence logic is maintained, 3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and 4) sufficient system functional capability is available for protective and ESF purposes from diverse parameters.
The OPERABILITY of these systems is required to provide the overall reliability, redundance and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions.
The integrated operation of each of these systems is consistent with the assumptions used in the accident analyses.
The surveillance requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards.
The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability. The quarterly frequency for the channel functional tests for these systems is based on the analysis presented in the NRC-approved Topical Report CEN-327, "RPS/ESFAS Extended Test Interval Evaluation," as supplemented.
The measurement of response time at the specified frequencies provides assurance that the protective and ESF action function associated with each channel is completed within the time limit assumed in the accident analyses.
No credit was taken in the analyses for those channels with response times indicated as not applicable.
Response time may be demonstrated by any series of sequential, overlapping or total channel test measurements provided that such tests demonstrate the total channel response time as defined.
Sensor response time verification may be demonstrated by either 1) in place, onsite or offsite test measurements or 2) utilizing replacement sensors with certified response times. The response time limits are contained in UFSAR Chapter 7, and updated in accordance with 10 CFR 50.71(e).
3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 Radiation Monitorino Instrumentation The OPERABILITY of the radiation monitoring channels ensures that 1) the radiation levels are continually measured in the areas served by the individual channels and 2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded.
CALVERT CLIFFS - UNIT 1 B 3/4 3-1 Amendment No. 193
Q TABLE 4.3-1 R
G g
REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREMENTS w
E n
C CHANNEL MODES IN WHICH q
CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE E
FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED i
=
1.
Manual Reactor Trip NA NA S/U )
NA y
U g
Z 2.
Power Level - High j
N D ),M
), QUI f2 I3 a.
Nuclear Power S
Q 1, 2 l
D ),R Q
1 l
U b.
AT Power S
3.
Reactor Coolant Flow - Low S
R Q
1, 2 l
g 4.
Pressurizer Pressure - High S
R Q
1, 2 l
5.
Containment Pressure - High S
R Q
1, 2 l
w 6.
Steam Generator Pressure - Low S
R Q
1, 2 l
7.
Steam Generator Water Level -
S R
Q 1, 2 l
Low 8.
Axial Flux Offset S
R Q
1 l
9.
a.
Thermal Margin / Low Pressure S R
Q 1, 2 l
Ng b.
Steam Generator Pressure S
R Q
1, 2 l
Difference - High
- 10. Loss of Load NA NA S/U )
NA U
'.E O
i t
4 g
TABLE 4.3-1 (Continued)
R G
g REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREMENTS w
--e n
E C
CHANNEL MODES IN WHICN 4
Q CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE E
FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED '
E i
x
- 11. Wide Range Logarithmic Neutron Flux S
R(5)
S/U )
fl 1, 2 y
and, 3, 4, 5 y
c*
Monitor
- 12. Reactor Protection System Logic NA NA Q and S/U )
1, 2 l
II
- to Matrices
- 13. Reactor Protection System Logic NA NA Q and S/U(1) 1, 2 l
Matrix Relays
- 14. Reactor Trip Breakers NA NA M
1, 2 and
- M o.
~
an
~~
O 1
1 r
~
,,s
.y..
g TABLE 4.3-2 R
G g
ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS w
n E!
C CHANNEL MODES IN WHICH q
CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE E
FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED i5i "I
C 3m 1.
SAFETY INJECTION (SIAS) y x
a.
Manual (Trip buttons)
NA NA R
NA b.
Containment Pressure - High S
R Q
1,2,3 c.
Pressurizer Pressure - Low S
R Wg)W 1, 2, 7 d.
Automatic Actuation Logic NA NA M
1, 2, 3 2.
R a.
Manual (Trip buttons)
NA NA R
NA Y'
b.
Containment Pressure - High S
R 1, 2, 3 l
M gN D
G c.
Automatic Actuation Logic NA NA 1,2,3 3.
CONTAINMENT ISOLATION (CIS)#
a.
Manual CIS (Trip buttons)
NA NA R
NA 1,2,3 l
b.
Containment Pressure - High S
R M gN U
c.
Automatic Actuation Logic NA NA 1,2,3 N
a r+
E O
Q TABLE 4.3-2 (Continued)
R G
P ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION. SURVEILLANCE REQUIREMENTS w
GE n
E; CHANNEL HODES IN WHICH Q
CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE E
FUNCTIONAL UqI_I CHECK CALIBRATION TEST REQUIRED M
5 i
e 4.
MAIN STEAM LINE ISOLATION (SGIS) a.
Manual SGIS (MSIV Hand Switches NA NA R
NA and Feed Head Isolation Hand Switches) b.
Steam Generator Pressure - Low S
R 1,2,3 l
M gN U
c.
Automatic Actuation Logic NA NA 1,2,3 g
5.
CONTAINMENT SUMP RECIRCULATION (RAS)
\\"
a.
Manual RAS (Trip Buttons)
NA NA R
NA b.
Refueling Water Tank - Low NA R
1,2,3 l
Mg) c.
Automatic Actuation Logic NA NA 1, 2, 3 6.
CONTAINMENT PURGE VALVES ISOLATION a.
Manual (Purge Valve Control NA NA R
NA Switches)
Q 6"
l b.
Containment Radiation - High S
R
[
Area Monitor h
an o
~
I g
TABLE 4.3-2 (Continued)
R G
ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS E
n C
CHANNEL MODES IN WHICH 7
CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE E
FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED j5i 5
c D
5 7.
LOSS OF POWER 5
z a.
4.16 kv Emergency Bus Undervoltage NA R
Q 1,2,3 l
(Loss of Voltage) b.
4.16 kv Emergency Bus Undervaltage NA R
Q 1,'2, 3 l
(Degraded Voltage) 8.
CVCS ISOLATION R
West Penetration Room / Letdown NA R
Q 1, 2, 3, 4 l
l Y
Heat Exchanger Room Pressure - High i
9.
Manual (Trip Buttons)
NA NA R
NA b.
Steam Generator Level - Low S
R Q
1, 2, 3 c.
Steam Generator AP - High S
R Mg) 1,2,3 d.
Automatic Actuation Logic NA NA 1, 2, 3
@it et i!F O
Is l
\\
3/4.3 INSTRUMENTATION TABLE 4.3-2 (Continued)
TABLE NOTATION Containment isolatioil of non-essential penetrations is also initiated t
by SIAS (functional units 1.a and 1.c).
Must be OPERABLE only in MODE 6 when the valves are required OPERABLE and they are open.
(1)
The logic circuits shall be tested manually at least once per 31 days.
(2)
SIAS logic circuits A-10 and B-10 shall be tested monthly with the l.
j exception of the Safety Injection Tank isolation valves. The SIAS logic circuits for these valves are exempted from testing during operation; however, these logic circuits shall be tested at least once per 18 months during shutdown.
(3)
SIAS logic circuits A-5 and B-5 are exempted from testing during operation; however, these logic circuits shall be tested at least once per 18 months during shutdown.
(4)
CIS logic circuits A-5 and B-5 cre exempted from testing during operation; however, these logic circuits shall be tested at least once per 18 months during shutdown.
(5)
SGIS logic circuits A-1 and B-1 are exempted from testing during operation; however, these logic circuits shall be tested at least once per 18 months during shutdown.
(6)
CSAS logic circuits A-3 and B-3 are exempted from testing during operation; however, these logic circuits shall be tested at least once per 18 months during shutdown.
CALVERT CLIFFS - UNIT 2 3/4 3-22 Amendment No. 170
3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 PROTECTIVE AND ENGINEERED SAFETY FEATURES (ESF)
INSTRUMENTATION The OPERABILITY of the protective and ESF instrumentation systems and bypasses ensure that 1) the associated ESF action and/or reactor trip will be initiated when the parameter monitored by each channel or combination thereof exceeds its setpoint, 2) the specified coincidence logic is maintained, 3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and 4) sufficient system functional capability is available for protective and ESF purposes from diverse parameters.
The OPERABILITY of these systems is required to provide the overall reliability, redundance and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions.
The integrated operation of each of these systems is consistent with the assumptions used in the accident analyses.
The surveillance requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards. The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability. The quarterly frequency for the channel functional tests for these systems is based on the analysis presented in the NRC-approved Topical Report CEN-327, "RPS/ESFAS Extended Test Interval Evaluation," as supplemented.
The measurement of response time at the specifled frequencies provides assurance that the protective and ESF action function associated with each channel is completed within the time limit assumed in the accident I
analyses. No credit was taken in the analyses for those channels with response times indicated as not applicable.
Response time may be demonstrated by any series of sequential, overlapping or total channel test measurements provided that such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either 1) in place, onsite or offsite test measurements or 2) utilizing replacement sensors with certified response times.
The response time limits are contained in UFSAR Chapter 7, and updated in accordance with 10 CFR 50.71(e).
3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 Radiation Monitorina Instrumentation The OPERABILITY of the radiation monitoring channels ensures that 1) the radiation levels are continually measured in the areas served by the individual channels and 2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded.
CALVERT CLIFFS - UNIT 2 B 3/4 3-1 Amendment No. 170