ML20072F251
| ML20072F251 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 03/16/1983 |
| From: | Adensam E Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20072F254 | List: |
| References | |
| NUDOCS 8303230431 | |
| Download: ML20072F251 (19) | |
Text
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TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-327, SE000YAH NUCLEAR PLANT, UNIT 1 AMENDHENT TO FACILITY OPERATING LICENSE Amendment No. 27 License No. DPR-77 1.
The Nuclear Regulatory Commission (the Connission) has found that:
A.
The application for amendment to the Sequoyah Nuclear Plant, Unit 1 (the facility) Facility Operating License No. DPR-77 filed by the Tennessee Valley Authority (licensee), dated September 17, 1982, and supplemented December 29, 1982, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's regulations as set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the license, as amended, the provisions of the Act, and the rules and regulations of the Com-mission; C.
There is reasonable assurance (1) that the activities authorized by this aendment 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 inuance 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 ir, 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 Appendix A Technical Specifications as indicated in the attachments to this license acend-aent 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 Appendix A, as revised through Amendment No. 27, are hereby incorporated into the license.
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A 2-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 Elinor G. Adensam, Chief Licensing Branch No. 4 Division of Licensing
Attachment:
Appendix A Technical Specification Changes Date of Issuance: March 16, 1983
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The licensee shall operate the facility in accordance with the Technical Specifications.
3.
Within 90 days after the effective date of this -*ne ment, or such later time atisfy any applicable as the Commission may specify. the Licensee shall/agreenent with the Secretary requirement of P.L.97-425 related to pursuing ap of Energy for the disposal of high-level radio live waste and spent nuclear fuel.
4.
This license amendment is effective as of its date of issuance.
FOR E NUCLEAR REGULATORY COMMISSION Elinor G. Adensam, Chief Licensing Branch No. 4 Division of Licensing Attactnent:
Appendix A Technical Specification Changes Date of Issuance:
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s ATTACHMENT TO LICENSE NiENDMENT NO. 27 FACILITY OPERATING LICENSE NO. DPR-77 DOCKET NO. 50-327 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 area of change.
Amended Page 3/4 3-11 3/4 3-12 3/4 3-34 3/4 3-35 3/4 3-36 3/4 3-37 I
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M TABLE 4.3-1 "a
SE REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS E
CHANNEL MODES IN VHICH b
CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST Rf. QUIRED
]
l.
Manual Reactor Trip N.A.
N.A.
S/U(1) 1, 2, and
- 2.
Power Range, Neutron Flux S
0(2), M(3)
Q 1, 2 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 M
5.
Intermediate Rar.ge, S
R(6)
S/U(1) 1, 2, and *
?
Neutron Flux 6.
Source Range, Neutron Flux S(7)
R(6)
M and S/U(1) 2, 3, 4, 5, and
- 7.
6vertemperature Delta T S
R M
1, 2 8.
Overpower Delta T S
R M
1, 2 9.
Pressurizer Pressure--Low 5
R Q
1, 2
\\
?
10.
Pressurizer Pressure--High 5
R Q
\\1,2 S
11.
Pressurizer Water Level--High S
R Q
1, 2 5
12.
Loss of Flow - Single Loop S
R Q
l h
13.
Loss of Flow - Two Loops S
R N.A.
1 a--_
.._m.m,._
.4 p
TABLE 4.3-1 (Continued)
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' REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS i
SE CHANNEL MODES IN WHICH.
- i Ei CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE p
FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED 1
14.
Main Steam Generator Water S
R Q
1, 2 Level--Low-Low 15.
Steam /Feedwater Flow Mismatch and S
R Q
1, 2 -
l Low Steam' Generator Water Level i
l 16.
Undervoltage - Reactor Coolant N.A.
R M
1 i
Pumps t
l, 17.
Underfrequency - Reactor Coolant N.A.
R M
1 Pumps 18.
A.
Low Fluid Oil Pressure N.A.
N.A.
S/U(1) 1 o,
];
B.
Turbine Stop Valve Closure N.A.
N.A.
S/U(1) 1 l
y) 19.
Safety Injection Input from E3 N.A.
N.A.
M(4) 1, 2 N
20.
Reactor Trip Breaker N.A.
N.A.
M(5) and S/U('s) 1, 2, and
- 21.
Automatic Trip Logic N.A.
N.A.
M(5) 1, 2, and *
[
22.
Reactor Trip System Interlocks A.
Intermediate Range N.A.
R S/U(8) 2, and
- f.l Neutron Flux, P-6 B.
Power Range Neutron N.A.
R S/U(8) 1 i
Flux, P-7 i
C.
Power Range Neutron N.A.
R S/U(8) 1 y
Flux, P-8 g
D.
Power Range Neutron N.A.
R S/U(8) 1, 2 fp Flux, P-10 Q;
E.
Turbine Impulse Chamber N.A.
R S/U(8) 1 Pressure, P-13 re y
ll F.
Power Range Neutron N.A.
R S/U(8) 1 M
Flux, P-9
- "9]
G.
Reactor Trip, P-4 N.A.
R S/U(8) 1, 2, and
- q.:
I 3
t
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M TABLE 4.3-2 IS 53 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION E5 SURVEILLANCE REQUIREMENTS 8
e j
c
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5 CHANNEL MODESIN\\WHICH
-d CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED 1.
SAFETY INJECTION AND FEEDWATER ISOLATION a.
Manual Initiation N.A.
N.A.
M(1) 1, 2, 3, 4 b.
Automatic Actuation Logic N.A.
N.A.
M(2) 1, 2, 3, 4 c.
Containment Pressure-High S
R Q
1, 2, 3 d.
Pressurizer Pressure--Low S
R Q
1, 2, 3 w2 w
e.
Differential Pressure S
R Q
1, 2, 3 d,
Between Steam Lines--High a
f.
Steam Flow in Two Steam S
R Q
1,2,3 Lines--High Coincident with T
--L w-L w r Steam Line avg Pressure--Low 2.
a.
Manual Initiation N.A.
N.A.
M(1) 1, 2, 3, 4 s
i b.
Automatic Actuation Logic N.A.
N.A.
M(2) 1, 2, 3, 4 S"
c.
Containment Pressure--High-High 5 R
Q 1,2,3 5
l l
TABLE 4.3-2 (Continued) g g
ZNGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS Eq CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED 3.
CONTAINMENT ISOLATION a.
Phase "A" Isolation i) Manual N.A.
N.A.
M(1) 1, 2, 3, 4
- 2) From Safety Injection N.A.
N.A.
M(2) 1, 2, 3, 4 Automatic Actuation Logic b.
Phase "B" Isolacion
- 1) Manual N.A.
N.A.
M(1) 1, 2, 3, 4
[
- 2) Automatic Actuation N.A.
N.A.
M(2) 1, 2, 3, 4 Logic m
- 3) Containment Pressure--
S R
Q 1,2,3 High-High c.
Containment Ventilation Isolation
- 1) Manual N.A.
N.A.
M(1) 1, 2, 3, 4 S
R
- 2) Automatic Isolatior Logic N.A.
N.A.
M(2) 1, 2, 3, 4 M
,=,
- 3) Cnntainment Gar, Monitnr S
R M
1,2,3,4 Radioactivity-High
\\
=
4
jj TABLE 4.3-2 (Continued)
ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION E
SURVEILLANCE REQUIREMENTS i
5:
CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE
~~
FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED
- 4) Containment Purge Air S
R M
1, 2, 3, 4 Exhaust Monitor Radio-activity-High
- 5) Containment Particulate S
R M
1,2,3,4 Activity-High 4.
STEAM LINE ISOLATION a.
Manual N.A.
N.A.
M(1) 1, 2, 3 N
b.
Automatic Actuation Logic N.A.
N.A.
M(2) 1, F. 3 y
c.
Containment Pressure--
S R
Q 1,2,3 High-High d.
Steam Flow in Two Steam S
R Q
1, 2, 3 Lines--High Coincident with T
-- Low-Low or Steam Line 3yg Pressure--Low v
8 5.
TUR3INE TRIP AND FEEDWATER EL ISOLATION ij 5
a.
Steam Generator Water S
R Q
1, 2, 3 Level--iligh-liigh
?
U$
6.
AUXILIARY FEECWATER a.
Manual N.A.
N.A.
M(1) 1,2,3 b.
Automatic Actuation Logic N.A.
N.A.
M(2) 1, 2, 3
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TABLE 4.3-2 (Continued) i4 IS ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION l
f$
SURVEILLANCE REQUIREMENTS
,e z
E~'
CHANNEL MODES IN WHICH g.
EE CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE
- {
El FUNCTIONAL UNIT CHECK CALIBRATION TEST __
REQUIRED
,'I c.
Main Steam Generator Water S
R Q
1,2,3 Level-Low-Low d.
S.I.
See 1 above (all SI surveillance requirements) e.
Station Blackout N.A.
R N.A.
1,2,3 f.
Trip of Main Feedwater N.A.
N.A.
R 1, 2 F
Pumps g.
Auxiliary Feedwater Suction N.A.
R M
1,2,3 4
ki 7.
LOSS OF POWER a
w a.
6.9 kv Shutdown Board I
h Undervoltage 1.
Loss of Voltage S
R M
1,2,3,4
- 2. ' Load Shedding S
R N.A.
1, 2, 3, 4 8.
ENGINEERED SAFETY FEATURE
\\
ACTUATION SYSTEM INTERLOCKS
\\
1, a.
Pressurizer Pressure, N.A.
R (4)
N.A.
1,2,3 P-11 l
EI b.
T N.A.
R (4)
N.A.
1, 2, 3 avg, P-12 m
i El c.
Steam Generator N.A.
R (4)
N.A.
1, 2 i
S Level, P-14
=
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(
TENNESSEE VALLEY AUTHORITY DOCKET NO. S0-323 SE000YAH NUCLEAR PLANT, UNIT 2 AMEN 0 MENT TO FACILITY OPERATING LICEASE Amendment No. 16 License No. OPR-79 1.
The Nuclear Regulatory Comnission (the Commission) has found that:
A.
The application for amendment to the Sequoyah Nuclear Plant, Unit 2 (the facility) Facility Operating License No. DPR-79 filed by the Tennessee Valley Authority (licensee), dated September 17, 1982, and supplemented December 29, 1982, cmplies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's regulations as set forth in 10 CFR Chapter I; B.
The facility will operate in confomity with the license, as amended, the provisions of the Act, and the rules and regulations of the Com-mission; 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 conmon 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 Connission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is hereby anended by page changes to the Appendix A Technical Specifications as indicated in the attachments to this license amendnent and paragraph 2.C.(2) of Facility Operating License No. OPR-79 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment flo.16, are hereby incorporated into the license, i
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3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION f
/)
Elinor G. Adensam, Chief Licensing Branch No. 4 oivision of Licensing
Attachment:
Appendix A Technical Specification Changes oate of Issuance: March 16, 1983 1
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ATTACHMENT TO LICENSE AMENDMENT NO. 16 FACILITY OPERATING LICENSE NO. DPR-79 DOCKET NO. 50-328 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 area of change.
Amended Page 3/4 3-11 3/4 3-12 3/4 3-34 3/4 3-35 3/4 3-36 3/4 3-37 1
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TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODES FOR WHICH c
CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE IS y
FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED 1.
Manual Reactor Trip N.A.
N.A.
S/U(1) 1, 2, and
- 2.
Power Range, Neutron Flux S
D(2), M(3)
Q 1, 2 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 S.
Intermediate Range, S
R(6)
S/U(1) 1, 2, and
- y Neutron Flux 6.
Source Range, Neutron Flux S(7)
R(6)
M and S/U(1) 2, 3, 4, S, and
- 7.
Overtemperature AT S
R M
1, 2 8.
Overpower AT S
R M
1, 2 9.
Pressurizer Pressure--Low S
R Q
1, 2
{
10.
Pressurizer Pressure--High S
R Q
1, 2 a
11.
Pressurizer Water Level--High S
R Q
1, 2 \\
\\
c~
f 12.
Loss of Flow - Single Loop S
R Q
1 5
13.
Loss of Flow - Two Loops S
R N.A.
I 14.
Steam Generator Water Level--
S R
Q 1, 2 Low-Low
,.e ra-vemme
,w
M TABLE 4.3-1-(Continued)
'O S
REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS E
CHANNEL MODES FOR WHICH i.
c-CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE IS 5
FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED N
15.
Steam /Feedwater Flow Mismatch and S
R Q
1, 2 Low Steam Generator Water Level 16.
Undervoltage - Reactor Coolant N.A.
R M.
1 Pumps 17.
Underfrequency - Reactor Coolant N.A.
R M
1 Pumps 18.
Turbine Trip A.
Low Fluid Oil Pressure N.A.
N.A.
S/U(1) 1 Y
B.
Turbine Stop Valve Closure N.A.
N.A.
S/U(1) 1 M
19.
Safety Injection Input from ESF N.A.
N.A M(4) 1, 2 20.
Reactor Trip Breaker N.A.
N.A.
M(5) and S/U(1) 1, 2,_and
- 21.
Automatic Trip Logic N.A.
N.A.
M(5) 1,.2, and
- 22.
Reactor Trip System Interlocks l'
A.
Intermediate Range N.A.
R S/U (8) 2, and
- Neutron Flux, P-6 E7 B.
Power Range Neutron N.A.
R S/U (8)'
1 fj Flux, P-7 C.
Power Range Neutron N.A.
R S/U (8) 1 Flux, P-8 D.
Power Range Neutron N.A.
R S/U (8) 1, 2 Flux, P-10 E.
Turbine Impulse Chamber N.A.
R S/U (8) 1 Pressure, P-13 F.
Power Range Neutron Flux, P-9 N.A.
R S/U (8) 1 G.
Reactor Trip, P-4 N.A.
R S/U (8) 1, 2, and *
' TABLE 4.3-2 g
g ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION
=
SURVEILLANCE REQUIREMENTS CHANNEL MODES'f0R WHICH L
CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE IS FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED
}
y 1.
SAFETY IMECTION AND i
FEEDWATER ISOLATION a.
Manual Initiation N.A.
N.A M(1) 1, 2, 3, 4 b.
Automatic Actuation Logic N.A.
N.A.
M(2) 1, 2, 3, 4 c.
Containment Pressure-High S
R Q
1,2,3 d.
Pressurizer Pressure--Low S
R Q
1,2,3 u,2 e.
Differential Pressure S
R Q
1, 2, 3 g
Between Steam Lines--High f.
Steam Flow in Two Stea:a S
R Q
1,2,3 Lines--High Coincident with T
--L w-L w r Steam Line avg Pressure--Low 2.
??
i a.
Manual Initiation N.A.
N.A M(1) 1, 2, 3, 4 o
b.
Automatic Actuation Logic N.A.
N.A.
M(2) 1, 2, 3, 4
.g.
c.
Containment Pressure--High-High S R
Q 1,2,3 t
s
't
~
\\
M TABLE 4.3-2 (Continued)
S S
ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION i
E SURVEILLANCE REQUIREMENTS c-CHANNEL MODES FOR WHICH 5
CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE IS FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED L
N 3.
CONTAINMENT ISOLATION i >
a.
Phase "A" Isolation
- 1) Manual N.A.
N.A.
M(1) 1, 2, 3, 4
- 2) From Safety Injection N.A.
N.A.
M(2) 1, 2, 3, 4 Automatic Actuation Logic l
1 b.
Phase "B" Isolation.
U
- 1) Manual N.A.
N.A.
M(1) 1,2,3,4 T
g
- 2) Automatic Actuation Logic N.A.
N.A.
M(2) 1, 3, 3, 4
- 3) Containment Pressure--
S R
Q 1,2,3
't High-High c.
Containment Ventilation Isolation l
- 1) Manual N.A.
N.A.
M(1) 1, 2, 3, 4 If
- 2) Automatic Isolation Logic N.A.
N.A.
M(2) 1, 2, 3, 4 ft
- 3) Containment Gas Monitor S
R M
1,2,3,4 g
Radioactivity-liigh e
Z O
5 I
i i
s
i\\
X; TABLE 4.3-2 (Continued)
I iS 5!
ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION 3
3; SURVEILLANCE REQUIREMENTS e
CHANNEL MODES FOR WHICH~
5 CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE IS
]
FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED
- 4) Containment Purge Air S
R M
1,2,3,4 Exhaust Monitor Radio-activity-High S) Containment Particulate S
R M
1,2,3,4 Activity-High 4.
STEAM LINE ISOLATION gg a.
Manual N.A.
N.A.
M(1) 1, 2, 3 Y
b.
Automatic Actuation Logic N.A.
N.A.
M(2) 1, 2, 3 c.
Containment Pressure--
S R
Q 1,2,3 High-High d.
Steam Flow in Two Steam S
R Q
1,2,3 Lines--High Coincident with T
-- L w-Low r Steam Line avg Pressure--Low 3
g S.
TURBINE TRIP AND FEE 0 WATER R
ISOLATION
$r*
a.
Steam-Generator Water S
R Q
1,2,3 g:
Level--High-High
$t 6.
a.
Manual-N.A.
N.A.
M(1) 1, 2, 3 b.
Automatic Actuation Logic N.A.
N.A.
M(2) 1, 2, 3 1
s TABLE 4.3-2 (Continued) iS S
ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION 3!
SURVEILLANCE REQUIREMENTS c:
CHANNEL MODES FOR WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE IS 4
-4 FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED m
l c.
Main Steam Generator Wate.-
S.
R Q
1,2,3 Level-Low-Low d.
S.I.
See 1 above (all SI surveillance requirements) e.
Station Blackout N.A.
R N.A.
1,2,3
\\
f.
Trip of Main Feedwater N.A.
N.A.
R 1, 2 Pumps b
g.
Auxiliary Feedwater Suction N.A.
R M
1,2,3 Pressure-Low u,
03 7.
LOSS OF POWER a.
6.9 kv Shutdown Board Undervoltage 1.
Loss of Voltage S
R M
1,2,3,4 2.
Load Shedding S
R N.A.
1,2,3,4 3"
8.
ENGINEERED SAFETY FEATURE g
ACTUATION SYSTEM INTERLOCKS s
a.
Pressurizer Pressure, N.A.
R (4)
N.A.
1,2,3 P-11 l
b.
T 3yg, P-12 N.A R (4)
N.A.
1,2,3 c.
Steam Generator N.A.
R (4)
N.A.
1, 2 4
Level, P-14 k j