ML20084R670
| ML20084R670 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 05/30/1995 |
| From: | Hebdon F NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20084R674 | List: |
| References | |
| NUDOCS 9506090368 | |
| Download: ML20084R670 (12) | |
Text
'6 f* K'?
g-4 UNITED STATES g
j NUCLEAR REGULATORY COMMISSION t
WASHINGTON, D.C. 20066 4001
,o TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-327 SEQUOYAH NUCLEAR PLANT. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE l
Amendment No.199 License No. DPR-77 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated April 6, 1995, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Comission'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 Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon 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 Comission's regulations and all applicable requirements have been satisfied.
l i
9506090368 950530 PDR ADOCK 05000327 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-77 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, to be implemented within 45 days.
FOR THE NUCLEAR REGULATORY COMISSION
~
Frederick J. Hebdon, Director Project Directorate 11-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
thy 30,1995
ATTACHMENT TO LICENSE AMENDMENT N0,109 FACILITY OPERATING LICENSE NO. DPR-77 DOCKET NO. 50-327 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages.
The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.
REMOVE INSERT i
3/4 3-11 3/4 3-11 i
3/4 3-13 3/4 3-13 B3/4 3-1 83/4 3-1 l
i y
v e---
r-
TABLE 4.3-1 REACTOR TRIP SYSTEN INSTRUNENTATION SURVEILLANCE REQUIRENENTS ui CHANNEL NODES IN WHICH S
CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE IS g
FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED
[
1.
Nanual Reactor Trip N.A.
N.A.
S/U(1)and R(9) 1, 2, and
- 2.
Power Range, Neutron Flux 5
D(2),(3)
Q 1, 2
-l and Q(6) 3.
Power Range, Neutron Flux, N.A.
R(6)
Q 1, 2 High Positive Rate 4.
Power Range, Neutron Flux, N.A.
R(6)
Q 1, 2 High Negative Rate 5.
Intermediate Range, S
R(6)
S/U(1) 1, 2, and
- Neutron Flux 6.
Source Range, Neutron Flux S(7)
R(6)
N and S/U(1) 2, 3, 4, 5, and
- 7.
Overtemperature Delta T S
R Q
1, 2 8.
Overpower Delta T S
R Q
1, 2 9.
Pressurizer Pressure--Low S
R Q
1, 2 10.
Pressurizer Pressure--High S
R Q
1, 2 11.
Pressurizer Water Level--High S
R Q
1, 2 k
12.
Loss of Flow - Single Loop S
R Q
l l
[
- 13. Loss of Flow - Two Loops S
R N.A.
1 l
l
- 14. Steam Generator Water Level--
l E
Low-Low l
A.
Steam Generator Water Level-- S R
Q 1, 2 T
Low-Low (Adverse)
B.
Steam Generator Water Level-- S R
Q 1, 2 Low-Low (EAN)
C.
RCS Loop AT S
R Q
1, 2
~
D.
Containment Pressure (EAN)
S R
Q 1, 2 i
1 Table 4.3-1 (Continued)
NOTATION With the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal.
If not performed in previous 31 days.
(1)
(2) -
Heat balance only, above 15% of RATED THERMAL POWER. Adjust channel if absolute difference greater than 2 percent.
Compare incore to excore AXIAL FLUX DIFFERENCE above 15% of RATED (3)
THERMAL POWER. Recalibrate if the absolute difference greater than or equal to 3 percent. The frequency of this surveillance is every 31 EFPD. This surveillance is not required to be performed until 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> after thermal power is it 15% RTP.
Deleted.
(4)
Each train or logic channel shall be tested at least every 62 days on (5) a STAGGERED TEST BASIS. The test shall independently verify the OPERABILITY of the undervoltage and automatic shunt trip circuits.
Neutron detectors may be excluded from CHANNEL CALIBRATION.
(6)
(7) -
Below P-6 (Block of Source Range Reactor Trip) setpoint.
Deleted.
(8)
(9) -
The CHANNEL FUNCTIONAL TEST shall independently verify the operability of the undervoltage and shunt trip circuits for the manual reactor trip function.
(10) -
Local manual shunt trip prior to placing breaker in service. Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS.
l (11) -
Automatic and manual undervoltage trip.
i SEQUOYAH - UNIT 1 3/4 3-13 Amendment No. 54, 114, 141, 199 i
l e
3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 PROTECTIVE AND ENGINEERED SAFETY FEATURES (ESF)
INSTRUMENTATION The OPERA 8ILITY of the protective and ESF instrumentation systems and interlocks ensure.that 1) the associated ESF action and/or reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches 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, redundancy 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 Engineered Safety Features System interlocks perform the functions indicated below on increasing the required parameter, consistent with the setpoints listed in Table 3.3-4:
P-11 DeMats the manual block of safety injection actuation on low pressurizer pressure.-
4 P-14 Trip of all feedwater pumps, turbine trip, closure of feedwater isolation valves and inhibits feedwater control valve modulation.
On decreasing the required parameter the opposite function is performed at reset setpoints.
The surveillance requirements specified for these systems ensure that the l
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 surveillance for the comparison of the incore to the excore Axial Flux Difference is required only when reactor power is 2 15 percent. The 96 hour0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> delay in the first performance of the surveillance after reaching 15 percent reactor thermal power (RTP), following a refueling outage, is to achieve a higher power level and approach Xenon stability. The surveillance is typically performed when RTP is 2 30 percent to ensure the results of the evaluation are more accurate and the adjustments more reliable. The frequency of 31 EFPD is to allow slow changes in neutron flux to be better detected during the fuel cycle.
l i
SEQUOYAH - UNIT 1 B 3/4 3-1 Amendment No. 141, 199
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4 UNITED STATES 3
j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2065H001 I'
TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-328 SE0VOYAH NUCLEAR PLANT. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.190 License No. DPR-79 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated April 6, 1995, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Comission'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 i
safety of the public, and (ii) that such activities will be conducted in compliance with the Comission'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 Comission's regulations and all applicable requirements have been satisfied.
4 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-79 is hereby amended to read as follows:
(2) Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised 'through Amendment No.190, 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, to be implemented within 45 days.
FOR THE NUCLEAR REGULATORY COMMISSION (A
Y Frederick J. Hebdon, Director Project Directorate II-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: thy 30,1995 i
ATTACHMENT TO LICENSE AMENDMENT NO.100 FACILITY OPERATING LICENSE NO. DPR-79 QQ[KET NO. 50-328 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages.
The revised pages are identified by the captioned amendment number ard contain marginal lines indicating the area of change.
REMOVE INSERT 3/4 3-11 3/4 3-11 3/4 3-13 3/4 3-13 B3/4 3-1 B3/4 3-1 l
TABLE 4.3-1 m
m REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS h
CHANNEL MODES FOR WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE IS g
FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED
[
1.
Manual Reactor Trip N.A.
N.A.
S/U(1)and R(9) 1, 2, and
- 2.
Power Range, Neutron Flux S
D(2),(3)
Q 1, 2 l
and Q(6) 3.
Power Range, Neutron Flux, N.A.
R(6)
Q 1, 2 High Positive Rate 4.
Power Range, Neutron Flux, N.A.
R(6)
Q 1, 2 High Negative Rate g
5.
Intermediate Range, Neutron Flux S
R(6)
S/U(1) 1, 2, and *
[
6.
Source Range, Neutron Flux S(7)
R(6)
M and S/U(1) 2, 3, 4, 2.
5, and
- 7.
Overtemperature AT S
R Q
1, 2 8.
Overpower AT S
R Q
1, 2 9.
Pressurizer Pressure--Low S
R Q
1, 2 g
k 10.
Pressurizer Pressure--High S
R
.Q 1, 2 11.
Pressurizer Water Level--High S
R Q
1, 2 5
12.
Loss of Flow - Single Loop S
R Q
l g;
13.
Loss of Flow - Two Loops S
R N.A.
1 E
- 14. Steam Generator Water Level--
M Low-Low A. Steam Generator Water Level--
S R
Q 1, 2 5
Low-Low (Adverse)
B. Steam Generator Water Level--
S R
Q 1, 2 3
Low-Low (EAM)
C. RCS Loop AT S
R Q
1, 2 D. Containment Pressure (EAM)
S R
Q 1, 2
- l
Table 4.3-1 (Continued)
NOTATION With the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal.
If not performed in previous 31 days.
(1)
Heat balance only, above 15% of RATED THERMAL POWER. Adjust channel
.(2) if absol,ute difference greater than 2 percent.
Compare incore to excore AXIAL FLUX DIFFERENCE above 15% of RATED (3)
THERMAL POWER. Recalibrate if the absolute difference greater than or equal to 3 percent. The frequency of this surveillance is every 31 EFPD. This surveillance is not required to be performed until 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> after thermal power is 2 15% RTP.
I (4) -
Deleted.
Each train or logic channel shall be tested at least every 62 days (5) on a STAGGERED TEST BASIS. The test shall independently verify the OPERABILITY of the undervoltage and automatic shunt trip circuits.
Neutron detectors may be excluded from CHANNEL CALIBRATION.
(6)
Below P-6 (Block of Source Range Reactor Trip) setpoint.
(7)
(8) -
Deleted.
The CHANNEL FUNCTIONAL TEST shall independently verify the (9) operability of the undervoltage and shunt trip circuits for the manual reactor trip function.
j (10) -
Local manual shunt trip prior to placing breaker in service.
Each train shall be tested at least every 62 days on a i
STAGGERED TEST BASIS, (11) -
Automatic and manual undervoltage trip.
SEQUOYAH - UNIT 2 3/4 3-13 Amendment No. 46, 104, 132, 190
s E
3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION The OPERABILITY of the Reactor Trip and Engineered Safety Features Actuation Systems instrumentation and interlocks ensure that 1) the associated action and/or reactor trip will be initiated when the parameter monitored by each channel or combination thereof~ reaches 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 from diverse parameters.
The OPERABILITY of these systems is required to provide the overall reliability, redundancy 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 Engineered Safety Feature Actuation System interlocks perform the functions indicated below on increasing the required parameter, consistent with the setpoints listed in Table 3.3-4:
P-11 Defeats the manual block of safety injection actuation on low pressurizer pressure.
P-14 Trip of all feedwater pumps, turbine trip, closure of feedwater isolation valves and inhibits feedwater control valve modulation.
On decreasing the required parameter the opposite function is performed at reset setpoints.
The surveillance for the comparison of the incore to the excore Axial Flux Difference is required only when reactor power is 2 15 percent. The 96 hour0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> delay in the first performance of the surveillance after reaching 15 percent reactor thermal power (RTP), following a refueling outage, is to achieve a higher power level and approach Xenon stability. The surveillance is typically performed when RTP is 2 30 percent to ensure the results of the evaluation are more accurate and the adjustments more reliable. The frequency of 31 EFPD is to allow slow changes in neutron flux to be better detected during the fuel cycle.
SEQUOYAH - UNIT 2 B 3/4 3-1 Amendment No. 132, 190
.