ML20028H235
| ML20028H235 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 10/31/1990 |
| From: | Hebdon F Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20028H236 | List: |
| References | |
| NUDOCS 9011190173 | |
| Download: ML20028H235 (64) | |
Text
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION g
WASHINoToN, D. C. 20$$5
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TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-328 SEQUOYAH NUCLEAR PLANT UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.132 License No. DPR-79 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The applications for amendment by Tennessee Valley Authority (the
-licensee) dated January 24, April 25, May 15, and October 2, 1990, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regula-tions set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the applications, 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.
90l1190173 901031 i
PDR ADOCK 05000328 P
o
. 2.
Accordingly, the license is amended by ::hanges to the Technical Specifications as indicated in the attachmant to this lhense amer;dment and paragraph 2.C.(2) of Facility Operating License No. DPR-79 is ht;reby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.132, 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.
TGR THE NUCLEAR REGULATORY COMMISSION He Frederick J. Hebdon, Director Project Directorate II-4 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: October 31, 1990 I
ATTACHMENT TO LICENSE AMENOMENT NO.132 FACILITY OPERATING LICENSE NO. OPR-79 DOCKET 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 and contain marginal lines indicating the area of change.
Overleaf pages, marked with an "*",
are provided to maintain document completeness.
REMOVE.
INSERT 1-2 1-2 1-5 1-5*
1-6 1-6 2-4 2-4*
2-5 2-5 2-6 2-6
~~~
2-7 2-7 28 2-8 2-9 2-9 2-10 2-10 2-11
~~~
2-12 B 2-4 8 2-4 8 2-5 B 2-5 B 2-6 8 2-6 B 2-7 B 2-7 8 2-8 3/4 3-1 3/4 3-1*
3/4 3-2 3/4 3-2 3/4 3-3 3/4 3-3 3/4 3-3a 3/4 3-4 3/4 3-4*
3/4 3-5 3/4 3-5 3/4 3-6 3/4 3-6 3/4 3-7 3/4 3-7 3/4 3-8 3/4 3-8 3/4 3-9 3/4 3-9 3/4 3-10 3/4 3-10 3/4 3-11 3/4 3-11 3/4 3-12 3/4 3-12 3/4 3-13 3/4 3-13 3/4 3-14 3/4 3-14" 3/4 3-15 3/4 3-15 3/4 3-16 3/4 3-16 3/4 3-17 3/7 3-17*
3/4 3-18 3/4 3-18 3/4 3-19 3/4 3-19 3/4 3-19a 3/4 3-20 3/4 3-20*
i 3/4 3-21 3/4 3-21 3/4 3-21a 3/4 3-21a 3/4 3-22 3/4 3-22 3/4 3-23 3/4 3-23 3/4 3-23a
.- REMOVE INSERT 3/4 3-24 3/4 3-24 3/4 3-25 3/4 3-25 3/4 3-26 3/4 3-26 3/4 3 3/4 3-27 3/4 3-27a 3/4 3-27a 3/4 3-27b 3/4 3-28 3/4 3-28 3/4 3-30 3/4 3-30 3/4 3-31 3/4 3-31 3/4 3-33a 3/4 3-33a
. 3/4 3-34 3/4 3-34 3/4 3-35 3/4 3-35*
3/4 3-36 3/4 3-36 3/4 3-37 3/4 3-37 3/4 3-38 3/4 3-38 l
B 3/4 3-1 B 3/4 3-1 l
e r
o.
DEFINITIONS-CH!.NNELFUNCTIONALTEST 1.6 A CHANNEL FUNCTIONAL TEST shall be:
a.
Analog channels - the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions, b.
Bistable channels - the injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions.
c.
Digital-channels - the injection of a simulated signal into the channel as close to the sensor input to the process racks as practi-cable to verify OPERABILITY including alarm and/or trip functions.
CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:
a.
All penetrations. required to be closed during accident conditions are either:
.1)
Capable of being closed by an OPERABLE containment automatic isolation valve system, or 2)
Closed by manual valves, blind flanges, or deactivated auto-matic valves secured in their closed positions, except as provided in Table 3.6-2 of Specification 3.6.3.
b.
All-equipment hatches are closed and sealed, c.
. Each air' lock is in compliance with the requirements of i
Specification 3.6.1.3, d.
The containment leakage rates are within the limits of Specification 3.6.1.2, and
. e.
The sealing mechanism associated with each penetration (e.g., welds, bellows or 0-rings) is OPERABLE.
CONTROLLED LEAKAGE
- 1. 8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.
CORE ALTERATION 1.9 CORE ALTERATION shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel.
Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.
SEQUOYAH - UNIT 2 1-2 Amendment No. 63, 117, 132
u 1
DEFINITIONS OPERATIONAL MODE - MODE L
1.19 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive l
combination of core reactivity condition, power level and average reactor coolant temperature specified in Table 1.1.
PHYSICS TESTS 1.20 PHYSICS TESTS shall be those tests performed to measure the fundamental l
nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 14.0 of the FSAR, 2) authorized under the provisions of 10-CFR 50.59, or 3) otherwise approved by the Commission.
PRESSURE BOUNDARY LEAKAGE 1.21 PRES 5URE BOUNDARY LEAKAGE shall be leakage (except steam generator tube
~
leakage) through a non-isolable fault in a Reactor Coolant System component tody, pipe wall or vessel wall.
PROCESS CONTROL PROGRAM (PCP) 1.22 The PROCESS CONTROL PROGRAM shall contain the current formula sampling, l
analysis tests, and determinations to be made to-ensure that the processing and packaging of, solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with-10 CFR Part 20, 10 CFR Part 71, and federal and state regulations and other requirements governing the disposal of radioactive wastes.
PURGE'- PURGING
-1.23 PURGE or PURGING is the controlled process.of discharging air or gas l
from a confinement to maintain temperature, pressure, humidity, concentration
,or other operating condition, in such a manner that replacement air or gas is required to purify the confinement; QUADRANT. POWER TILT RATIO 1[24 QU50 RANT POWER TILT MTIO shall be the ratio of the maximum upper excore l
c.
detector calibrated output to the average of the upper excore detector cali-brated outputs, or'the ratio of the maximum lower excore detector calibrated.
output to the average of the lower excore detector calibrated outputs, which-ever is greater.
With one excore detector inoperable 'the remaining three s
detectors shall be used for computing the average.
SEQUOYAH - UNIT 2 1-5 Amendment No.63
- )
o DEFINITIONS RATED THERMAL POWER (RTP) 1.25 RATED THERMAL POWER (RTP) hall be a total reactor core heat transfer rate to the reactor r.oolant of 3411 MWt.
REACTOR TtIP SYSTEM RESPONSE TIME 1.26 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from k> hen the monitored parameter exceeds its trip setpoint at the channel sensor until loss of stationary gripper coil voltage.
REPORTABLE EVENT 1.27 A REPORTABLE EVENT shall be-any of those conditions specified in Section 50.73 to 10 CFR Part 50.
SHIELD BUILDING INTEGRITY L
l 1.28 SHIELD BUILDING INTEGRITY shall exSt when:
a.
The door in each access opening is closed except when the access opening is being used for normal transit entry and exit.
b.
The emergency gas treatment system is OPERABLE.
c.
The sealing mechanism associated with each penetration (e.g., welds, bellows or 0-rings) is OPERABLE.
i SHUTOOWN MARGIN SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which 1.29
'the' reactor is subcritical or would be suberitical from its present condition-assuming all full length rod cluster assemblies (shutdown and control) are
-fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.
SITE BOUNDARY 1.30 :The SITE BOUNDARY shall be that line beyond which the land is not owned,
-l 1 eased, or otherwise controlled by the licensee (see figure 5.1-1),
[-
1 g
SEQUOYAH - UNIT 2 1-6 Amendment No. 63,132 l-
O l
l l
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS
- 2. 2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP WsTEM INSTRUMENTATION SETPOINTS 2.1.1 The reactor trip system instrumentation and interlocks setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1.
APPLICABILITY:
As shown for each channel in Table 3.3-1.
nCTION:
With a reactor trip system instrumentation or interlock setpoint less conservative than the value shown in the Aliowable Values column of Table 2.21, declare the channel inoperable and ar. ply the applicable ACTION statement requirement of Specification 3.3.] until the channel is restored to OPERABLE status wi @ its trip setpoint adjusted cor.sistent with the Trip Setpoint value.
)
1
'SEQUOYAH - UNIT 2 2-4
TABLE 2.2-1 4
h REACTOR TRIP SYSTEM INSTRtMENTATION TRIP SETPOINTS 8
j FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES
- 1. Mar.ual Reactor Trip Not Applicable Not Applicable Z
- 2. Power Range, Neutron Flux Low Setpoint 1 25% of RATED Low Setpoint 1 27.4% of RATED l
m THERNAL POWER THERMAL POWER High Setpoint 1 109% of RATED High Setpoint 1111.4% of l
THERMAL POWER RATED THERMAL POWER 4
- 3. Power Range, Neutron Fluu, 1 5% of RATED THERMAL POWER wi "
$ 6.3% of RATED THERMAL POWER High Positive Rate a time constant 1 2 seconds with a time constant 1 2 seconds
- 4. Power Range, Neutron Flux, 5 5% of RATED THERMAL POWER with 5 6.3% of RATED THERMAL POWER High Negative Rate a time constant 1 2 seconds with a time constant 1 2 seconds
- 5. Intermediate Range, Neutron 5 25% of RATED THERMAL POWER
$ 30% of RATED THERMAL POWER m
d, Flux l
- 6. Source Range, Neutron Flux
$ 105 counts per second i 1.3 x 105 counts per s=cond
- 7. Overtemperature AT See Note 1 See Note 3 l
- 8. Overpower AT See Note 2 See Note 4
- 9. Pressurizer Pressure--Low 1 1970 psig 1 1964.8 psig
- 10. Pressurizer Pressure--High
< 2385 psig i 2390.2 psig
>g
- 11. Pressurizer Water Level-High 192% of instrument span i 92.7% of instrument span E
2
- 12. Loss of Flow
> 90% of design flow per loop *
> 89.4% of design flow per loop
- 5 z?
- Design flow is 91,400 gpm per loop.
l
~
(a)
N
. ~.
m
o.
TABLE 2.2-1 (Continued)
M REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS W$
FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES
- 13. Steam Generator Water c
]
Level--Low-Low 1
a.
RCS toop AT Equivalent RCS Loop AT variable input RCS Loop AT variable input to Power i 50% RTP
$ 50% RTP
$ trip setpoint +2.5% RTP Coincident with Steam Generator Water
> 15.0% of narrow range
> 14.4% of narrow range Level--Low-Low instrument span instrument span (Adverse) and Containment Pressure 1 0.5 psig 1 0.6 psig (EAM)
~
E or Steam Generator Water
> 10.7% of narrow range
> 10.1% of narrow range Level--Low-tow (EAM)
Instrument span Instrument span With A time delay (T ) if iTs (Note 5) i (1.01)T3 (Note 5) one Steam Gener3 tor is affected or 2
A time delay (T ) if
-<T" (Note 5)
- (1.01)T" (Note 5)
R twoormoreStek 2
Generators are a
affected E
~
~
~~
c.
i j
TABLE 2.2-1 (Continued) w E
REACTOR TRIP SYSTEM IMS RUMENTATION TRIP SETPOINTS 8s 7
FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES i
E b.
RCS Loop AT Equivalent Z
to Power > 50% RTP m
Coincident with i
Steam Generator Water
> 15.0% of narrow ra.wje
> 14.4% of narrow range Level--Low-Low instrument span instrument span i
(Adverse) and Containment Pressure
< 0.5 psig
< 0.6 psig (EAM) or m
Steam Generator Water
> 10.7% of narrow range
> 10.1% of narrow range Level--Low-Low (EAM) instrument span instrument span t
- 14. Deleted
- 15. Undervoltage-Reactor
> 5022 volts-each bus
> 4739 volts each bus Coolant Pumps
- 16. Underfrequency-Reactor
> 56 Hz - each bus
> 55.9 Hz - each bus Coolant Pumps
?
- 17. Turbine Trip g
A.
Low Trip System
> 45 psig
> 43 psig Pressure B.
Turbine Stop Valve
> 1% open
> 1% open Closure F
- 18. Safety Injection Input Not Applicable Not Applicable from ESF 5
M (A9 ro j
E TABLE 2.2-1'(Continued) g-REACTOR TRIP SYSTEM INSTRUNENTATION TRIP SETPOINTS 5
- r i
FUNCTIONAL UNIT TRIP SETPOINT ALLOWA8LE VALUES l
c 3
- 19. Intermediate Range Neutron
> 1 x 10 % of RATED
> 6 x'10 % ef RATED I
-5
-6 i
Flux, P-6, Enable Block THERMAL POWER THERMAL POWER y
}
Source Range Reactor Trip i
l
- 20. Power Range Neutron Flux
< 10% of RATED
< 12.4% of RATED l
(not P-10) Input to low THERMAL POWER THERMAL POWER l
Power Reactor Trips Block P-7 l
l
- 21. Turbine Impulse Chamber Pressure -
< 10% Tu4fne Impulse
< 12.4% Turbine Impulse l
(P-13) Input to Low Power Reactor Trips Pressure Equivalent Fressure Equivalent l
Block P-7 l
m 22.
Power Range Neutron Flux - (P-8) Low
< 35% of RATED
< 37.4% of RATED l
Reactor Coolant Loop Flow, and THERMAL POWER THERMAL POWER i
l
{
23.
Power Range Neutron Flux - (P-10) -
> 10% of RATED
> 7.6% of RATED Enable block of Source, Intermediate, THERMAL POWER THERMAL POWER and Power Range (low setpoint) Reactor
,g Trips o
[
24.
Reactor Trip P-4 Not Applicable Not Applicable a
25.
Power Range Neutron Flux - (P-9) -
< 50% of RATED
< 52.4% of RATED E
Blocks Reactor Trip for Turbine THERMAL POWER THERMAL POWER
(
Trip Below 50% Rated Power y
a to i
I m
i
=.
w
-4 w
.~
e
..,,_e
,w.
,... ~
4 TABLE 2.2-1 (Continued)
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS s
x NOTATION e
E 5
y (I * '1 )[T - T'] + K (P-P') - f (AI)}
)
Overtemperature aT (1 + T 5) $ AT, {K 4
-K NOTE 1:
3 y
y 1+15 1 + '25 5
1+T54 where:
= Lag c m nsa u r on measur d aT 1 * '55
= Time constants utilized in the lead-lag controller for AT, t4 = 12 secs, T4,15 5 = 3 secs 1
AT,
= Indicated AT at RATED THERMAL POWER ro E
K 5 1.15 y
K
= 0.011 2
1+rSy The function generated t,y the lead-lag controller for T,,, @namic cesation
=
1+r52 Time constants utilized in the lead-lag cc.: troller for Tyg, ty = 33 secs.,
r,&T
=
y 2
2 = 4 secs.
1 g
Average temperature 'F T
=
?+
T' 5 578.2*F (Nominal T,9 at RATED THERMAL POWER) z 0.00055 K
=
3 Pressurizer pressure, psig P
=
N P'
2235 psig (Nominal RCS operating pressure)
=
C ro
L a
N TABLE 2.2-1 (Continued)
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS g
r NOTATION (Continued)
E NOTE 1:
(Continued) l
-1 S
Laplace transform operator. sec
=
and f (AI) is a function of the indicated difference between top and bottom detectors 7
of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:
(i) for qt g
e-percen a
+ 5 p u nt f (aI) = 0 ( h qt*
7 O
are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt + 9b is total THERMAL POWER in percent of_ RATED THERMAL POWER).
l (ii) for each percent that the magnitude of (q 9 ) exceeds -29 percent, the AT trip set-l t
b point shall be automatically reduced by 1.50 percent of its value at RATED THERMAL POWER.
i (iii) for each percent that the magnitude of (q ~ 9 ) exceeds +5 percent, the AT trip set-t b
point shall be automatically reduced by 0.86 percent of its value at RATED THERMAL POWER.
Overpower AT (1 + T 5) < AT, {K 5
NOTE 2:
4
-K 5 (1 + r 5 3
) T -K U ~ D ~ # (OIII 4
6 2
g 1+ISS 3
1+r5 i
4 y
where:
= as defined in Note 1 1+rS m
5 w
e
~
m.
TABLE 2.2-1 (Continued) m5 g
REACTOR TRIP SYSTEM INSTRUNENTATION TRIP SETPOINTS 5
x NOTATION (Continued)
EZ NOTE 2:
(Continued) as defined in Note 1
=
AT, as defined in Note 1
=
K 5 1.087 4
K
=
5 0.02/*F for increasing average temperature and 0 for decreasing average temperature 7
T3 3
U 1+T5= The function generated by the rate-lag controller for T,,9 dynamic 3
compensation Time constant utilized in the rate-lag controller for T,,, t3 = 10 secs.
t
=
3 K
=
g 0.%11 for T > T* aM Kg = 0 for T S T*
T
=
as defined in Note 1 o.l T"
Indicated T,,9 at RATED THERMAL POWER (Calibration temperature for
=
[
AT instrumentation, 5 578.2*F) i l
?
S
=
as defined in Note 1 y
f (aI) 0 for all AI
=
2 y
J' l
N
.=
=.. -...
~.
TABLE 2.2-1 (Continued)
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS SE E
NOTATION (Continued)
Ea NOTE 3: The channel's maximum trip setpoint shall not exceed its computed trip point by more than 1.9 percent AT span.
l NOTE 4:
~
The channel's maximum trip setpoint shall not exceed its computed trip setpoint by more than 1.7 percent AT span.
i l
NOTE 5: Trip Time Delay - Steam Generator Water Level--Low-Low l
l T = {(-0.00583)(P)3 + (0.735)(P)2 - (33.560)(P) + 649.5}{0.99}
s T,= {(-0.00532)(P)3 + (0.678)(P)2 - (31.340)(P) + 589.5]{0.99]
l Where:
i l
P = RCS Loop AT Equivalent to Power (% RTP), P < 50% RTP T = Time delay for Steam Generator Water Level--Low-Low Reactor Trip, one Steam Generator 3
y affected.
E g-T, = Time delay for Steam Generator Water Level--Low-Low Heact.or Trip, two or more Steam g
Generators affected.
.O M
l
A 3/4.3 INSTRUMENTATION l
3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the reactor trip system instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE with RESPONSE TIMES as shown in Table 3.3-2.
APPLICABILITY:
As shown in Table 3.3-1.
ACTION:
As shown in Table 3.3-1.
SURVEILLANCE REQUIREMENTS 4.3.1.1.1 Each reactor trip system instrumentation channel and interlock 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 4.3-1.
4.3.1.1.2 The logic for the interlocks shall be demonstrated OPERABLE prior to each reactor startup unless performed during the )receeding 92 days.
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.
4.3.1.1.3 The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip function shall be demonstrated to be within its 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 leest once every N times 18 months where N is the total number of redundant channels in a specific reactor trip function as shown in the " Total No. of Channels" column of Table 3.3.1.
SEQUOYAH - UNIT 2 3/4 3-1
TABLE 3.3-1 M
REACTOR TRIP SYSTEM INSTRUMENTATION o8 3E I
MINIMUM TOTAL I40.
CHANNELS CHANNELS APPLICABLE FUNCTIONAL LHIT OF CHANNELS TO TRIP OPERABLE MODES ACTION E
Z 1.
1 2
1, 2, and
- 1 m
2.
Power Range, Neutron Flux 4
2 3
1, 2 2#
3.
Power Range, Neutron Flux 4
2 3
1, 2 2#
High Positive Rate 4.
Power Range, Neutron Flux, 4
2 3
1, 2 2#
High Negative Rate 5.
Intermediate Range, Neutron Flux 2
1 2
1, 2, and
- 3 w1 6.
Source Range, Neutron Flux A.
S'.artup 2
1 2
2", and
- 4 w
j c',
B.
Snutdown 2
0 1
3, 4 and 5 5
l 7.
Overt mperature AT Four Loop Operation 4
2 3
1, 2 6,
8.
Overpower aT l
Four Loop Operation 4
2 3
1, 2 6,
9.
Pressurizer Pressure-Low 4
2 3
1, 2 6#
10.
Pressurizer Pressure--High 4
2 3
1, 2 6
l 3
g 11.
Pressurizer Water Level--High 3
2 2
1, 2 6#
l 7
g
?
I",
D'uS M
N l
TABLE 3.3-1 (Continued)
M x>
REACTOR TRIP SYSTEM INSTRUMENTATION 85 MINIMUM TOTAL NO.
CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION EZ 12.
Loss of Flow - Single Loop 3/ loop 2/ loop in 2/ loop in 1
6#
m (Above P-8) l any oper-each oper-ating loop ating loop 13.
Loss of Flow - Two Loops 3/ loop 2/ loop in 2/ loop in 1
6#
l (Above P-7 and below P-8) two oper-each oper-ating loops ating loop 14.
Main Steam Generator Water Level--Low-Low w
A.
Steam Generator Water 3/Sta. Gen.
2/Stm. Gen.
2/Stm. Gen.
1,2 9#
1 Level--Low-Low in any in each (Adverse) operating operating w
J, Stm. Gen Sta. Gen.
B.
Steam Generator Water 3/Stm. Gen.
2/Stm. Gen.
2/Stm. Gen.
1,2 9#
Level--Low-Low in any in each (EAM) operating operating Sta. Gen.
Stm. Gen.
1 l
C.
RCS Loop AT 4 (1/ loop) 2 3
1,2 13#
D.
Containment Pressure 4
2 3
1,2 11#
y (EAM) s 15.
Deleted.
.e
~
l
~.
TABLE 3.3-1 (Continued) mE
.g REACTOR TRIP SYSTEM INSTRUMENTATION 5
z MINIMUM TOTAL NO.
CHANNELS CHANNELS APPLICABLE g
FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION
[
16.
Undervoltage-Reactor Coolant 4-1/ bus 2
3 1
6#
Pumps 17.
Underfrequency-Reactor Coolant 4-1/ bus 2
3 1
6#
Pumps 18.
Turbine Trip A.
Low Fluid Oil Pressure 3
2 2
1 6,
B.
Turbine Stop Valve Closure 4
4 4
1 6g 19.
Safety Injection Input from ESF 2
1 2
1, 2 12 S'
20.
Reactor Trip Breakers A.
Startup and Power Operation 2
1 2
1, 2 12, 15 B.
Shutdown 2
1 2
3*,4* and 5*
16
- 21. Automatic Trip Logic A.
Startup and Power Operation 2
1 2
1, 2 12 B.
Shutdown 2
1 2
3*,4* and 5*
16 22.
Reactor Trip System Interlocks E
A.
Intermediate Range
{
Neutron Flux, P-6 2
1 2
2, and*
8a 1
9 E
B.
Power Range Neutron Flux, P-7 4
2 3
1 8b E
L O
m I
TA8LE 3.3-1 (Continued)
-y REACTOR TRIP SYSTEM INSTRUNENTATION
-e E
MININUM TOTAL NO.
CHANNELS CHANNELS APPLICABLE e
g FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE NODES ACTION C.
Power Range Neutron Flux, P-8 4
2 3
1 8c D.
Power Range Neutron Flux, P-10 4
2 3
1, 2 Bd E.
Turbine Impulse Chamber Pressure, P-13 2
1 2
1 8b F.
Power Range Neutron 4
2 3
1 8e g
Flux, P-9
=
G.
Reactor Trip, P-4 2
1 2
1, 2, and
- 14 SI a
3-
.O
TABLE 3.3-1 (Continued)
TABLE NOTATION a
With the reactor trip system breakers in the closed position, the control rod drive system capable of rod withdrawal, and fuel in the
,, reactor vessel.
l The channel (s) issociated with the protective functions derived from the out of service Reactor Coolant Loop shall be placed in the tripped condition.
j
- The provisions of Specification 3.0.4 are not applicable.
- ource Range outputs may be disabled above the P-6 (Block of Source S
Range Reactor Trip) setpoint.
ACTION STATEMENTS ACTION 1 - With the number of OPERABLE channels one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and/or open the reactor trip breakers.
ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:
a.
The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, b.
The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing of other channels per Specification 4.3.1.1.1.
c.
The QUADRANT POWER TILT RATIO is monitored in accordance with Technical Specification 3.2.4.
l l
L SEQUOYAH - UNIT 2 3/4 3-5 Amendment No. 39, 122, 129, 132
4 i
TABLE 3.3-1 (Continued)
ACTION 3 - With the number of channels OPERABLE one less than required by t
the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:
Below the P-6 (Block of Source Range Reactor Trip) setpoint, a.
restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint.
b.
Above the P-6 (Block of Source Range Reactor Trip) setpoint, but below 5% of RATED THERMAL POWER, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 5% of RATED THERMAL POWER.
l c.
Above 5% of RATED THERMAL POWER, POWER OPERATION may continue.
d.
Above 10% of RATED THERMAL POWER, the provisions of Specification 3.0.3 are not applicable.
ACTION 4 - With the number of OPERABLE chocoels one less than required by I
the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:
Below the P-6 (Block of Source Range Reactor Trip) setpoint, a.
restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint.
I b.
Above the P-6 (Block of Source Range Reactor Trip) setpoint, l
operation may continue.
i ACTION 5 - With the number of OPERABLE channels one less than required by the Minimum Channels OPEP.ABLE requirement, verify compliance i
with the SHUTDOWN MARGIN requirements of Specification 3.1.1.1 or 3.1.1.2, as applicable, within I hour and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter.
ACTION 6 - With the number of OPERABLE channels one less than the Total i
Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditiens are satisfied:
a.
The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
b.
The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing of other channels per Specif1 cation 4.3.1.1.1.
i ACTION 7 - Deleted.
SEQUOYAH - UNIT 2 3/4 3-6 Amendment No. 39,132
-e--
p
--w w
TABLE 3.3-1 (Continued)
ACTION 8 - With less than the Minimum Number of Channels OPERABLE, declare 4
the interlock inoperable and verify that all affected channels of the functions listed below are OPERABLE or a priate ACTION statemer.t(s) for those functions.pply the appro-Functions to be evaluated are:
a.
Source Range Reactor Trip,
~
b.
Low Reactor Coolant Loop Flow (2 loops)
Undervoltage Underfrequency Pressurizer Low Pressure Pressurizer High Level c.
Reactor Trip Low Reactor Coolant Loop Flow (1 loop) d.
Retctor Trip Intermediate Range Low Power Range Source Range e.
Reactor Trip Turbine Trip ACTION 9 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed J
provided the following conditions are satisfied:
a.
The inoperable channel is placed in the tripped condition within G hours, b.
For the affected protection set, the Trip Time Delay for one affected steam generator (T ) is adjusted to 3
match the Trip Time Delay for multiple affected steam generators (T ) within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
g c.
The Minimum Channels OPERABLE requirement is met; i
however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing oi other channels per Specification 4.3.1.1.1.
SEQUOYAH - UNIT 2 3/4 3-7 Amendment No. 46, 99, 104, 132
s O
TABLE 3.3-1 (Continued)
ACTION 10 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided that within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, for the affected protection set, the Trip Time Delays (T3 and T ) threshold g
power level for zero seconds time delay is adjusted to 0%
RTP.
ACTION 11 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided that within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, for the affected protection set, the Steam Generator Water Level - Low-Low (EAM)channelstripsetpointisadjustedtothesamevalue as Steam Generator Water Level - Low-Low (Adverse).
ACTION 12 With the number of OPERABLE channels one less than required by the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; however, one channel may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing per Specification 4.3.1.1.1 provided the other channel is OPEPABLE.
ACTION 13 - With the number of OPERABLE channels one less than the Total Number of Channels and with the THERMAL POWER level above the P-7 (enable reactor trips) setpoint place the inoperable channel in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, operation may continue until performance of the next required CHANNEL FUNCTIONAL TEST.
ACTION 14 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, j
i l
ACTION 15 - With one of the divarse trip features (undervoltage or i
shunt trip attachment) inoperable, restore it to operable status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or declare the breaker inoperable and apply ACTION 12.
The breaker shall not be bypassed while one of the diverse trip featores is inoperable except for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for trforming maintenance to i
restore the breaker to OPERABt.E status.
ACTION 16 - With the number of OPERABLE channels one less than the minimum channels operable requirement, restore the inoperable 1
channel to operable status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or open the reactor trip breakers within the next hour.
J l-
)
SEQUOYAH - UNIT 2 3/4 3-8 Amendment No. 46, 132
~.
TABLE 3.3-2 REACTOR TRIP SYSTEN INSTRUNENTATION RESPONSE TINES Y
x FUNCTIONAL UNIT RESPONSE TINE 1.
Manual Reactor Trip Not Applicable 2.
Power Range, Neutron Flux
< 0.5 seconds
- 3.
Power Range, Neutron Flux, High Positive Rate Not Applicable 4.
Power Range, Neutron Flux, High Negative Rate
$ 0.5 seconds
- 5.
Intermediate Range, Neutron Flux Not Applicable 6.
Source Range, Neutron Flux Not Applicable 7.
Overtemperature AT i 8.0 seconds
- 8.
Overpower AT i 8.0 seconds 9.
Pressurizer Pressure--Low 1 2.0 seconds 10.
Pressurizer Pressure--High i 2.0 seconds 11.
Pressurizer Water Level--High Not Applicable 12.
Loss of Flow - Single Loop j
(Above P-8) i 1.0 second S#
r
- Neutron detectors are exempt from response time testing.
Response time of the neutron flux signal portion of the channel shall be measured from detector output or input of first electronic component in channel.
p U
~ ~ ' ~ ~ ' ~ ~ ^ ^ ^ ~ ' ^ ~ ~ ^ ^ ~ ' ' " '
TABLE 3.3-2 (Continued)
REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES a.
2
~
FUNCTIONAL UNIT RESPONSE TIME E
13.
Loss of Flow - Two Loops Q
(Above P-7 and below P-8) i 1.0 second to 14.
Main Steam Generator Water Level--Low-Low A.
RCS Loop AT II)
-< 8.0 seconds B.
Steam Generator Water II) i 2.0 seconds Level--Low-tow (Adverse, EAM)
C.
Containment Pressure (EAM) 1 2.0 seconds (1)
- 15. Deleted w
16.
Undervoltage-Reactor Coolant Pumps T
i 1.2 seconds g
17.
Underfrequency-Reactor Coolant Pumps 1 0.6 seconds 18.
Turbine Trip A.
Low Fluid Oil Pressure Not Applicable 8.
Turbine Stop Valve Not Applicable 19.
Safety Injection Input from ESF Not Applicable 20.
Reactor Trip Breakers Not Applicable g
21.
Automatic Trip Logic Not Applicable i
i g-22.
Reactor Trip System Interlocks Not Applicable
- e (1) Does not include Trip Time sMlays.
Response times noted include the transmitters, Eagle-21 2o process protection cabinets, solid state protection cabinets and actuation devices.
This reflects the response time r.ecessary for THERMAL POWER in excess of 50% RTP.
i
k N
TABLE 4.3-1 E
REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REOUIREMENTS SN CHANNEL MODES FOR ndHICH c
CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE IS
}
FUNCTIONAL UNIT-CHECK CALIBRATION TEST REQUIRED m
1.
Manual Reactor Trip N.A.
N.A.
S/U(1)and R(9) 1, 2, anc
- 2.
Power Range, Neutron Flux S
0(2), M(3)
Q 1, 2 and Q(6) 3.
Power Range, Neutron Flux, M.A.
R(S)
Q 1, 2 High Positive Rate 4.
Power Range, Neutron Flux, N. A.
R(6)
Q 1, 2 i
High Negative Rate w
S.
Intermediate Range, Neutron Flux S
R(6)
S/U(1) 1, 2, and
- s*
6.
Source Range, Neutron Flux S(7)
R(6)
M and S/U(1) 2, 3, 4, Y
S, and
- 7.
Overtemperature AT S
R Q
1, 2 8.
Overpower AT S
R Q
1, 2 l
9.
Pressurizer Pressure--Low S
R Q
1, 2 10.
Pressurizer Pressure--High 5
R Q
1, 2 11.
Pressurizer Water Level--High 5
R Q
1, 2 12.
Loss of Flow - Single Loop 5
R Q
1 13.
Loss of Flow - Two Loops S
R N.A.
1 F
14.
Steam Generator Water Level--
Low-Low A.
Steam Generator Water Level--
S R
Q 1, 2 g
Low-Low (Adverse)
B.
Steam Generator Water Level--
S R
Q 1, 2 Low-Low (EAM)
C.
RCS Loop AT S
R Q
1, 2 l
D.
Containment Pressure (EAM)
S R
Q 1, 2
~
w~
-c.
TABLE 4.3-1 (Continued)
ME REACTOR TRIP SYSTEM INSTRUNENTATION SURVEILLANCE REQUIREMENTS E
CHANNEL MODES FOR WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE IS g
FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED Z
15.
Deleted 16.
Undervoltage - Reactor Coolant N.A.
R Q
1 Pumps
- 17. Underfrequency - Reactor Coolant N.A.
R Q
1 Pumps
- 18. Turbine Trip A.
Low Fluid Oil Pressure M.A.
N.A.
S/U(1) 1 B.
Turbine Stop Valve Closure M.A.
M.A.
5/U(1)
I l
- 19. Safety Injection Input from ESF N.A.
N.A.
R 1, 2 g
[
20.
Reactor Trip Breaker M.A.
M.A.
M(5) and S/U(1) 1, 2, and
- d
- 21. Automatic Trip Logic N.A.
N.A.
M(5) 1, 2, and
- 22.
Reactor Trip System Interlocks A.
Intermediate Range N.A R
N.A.
2, and
- Neutron Flux, P-6 8.
Power Range Neutron N.A.
N.A.
N.A.
I Flux, P-7 C.
Power Range Neutron M.A.
R N.A.
1 Flux, P-8 E
D.
Power Range Neutron N.A.
R N.A.
1, 2 2
Flux, P-10
?+
E.
Turbine Impulse Chamber M.A.
R N.A.
1 z
Pressure, P-13
?
F.
Power Range Neutron Flux, P-9 N.A.
R N.A.
1 G.
Reactor Trip, P-4 N.A.
M.A.
R 1, 2, and
- 23.
Reactor Trip Bypass Breaker N.A.
N.A.
M(10)R(11) 1, 2, and
- m
Table 4.3-1 (Continued)
NOTATION With the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal.
(1)
If not performed in previous 31 days.
(2)
Heat balance only, above 15% of RATED THERMAL POWER.
Adjust channel if absolute difference greater than 2 percent.
(3)
Compare incore to excore AXIAL FLUX DIFFERENCE above 15% of RATED THERMAL POWER.
Recalibrate if the absolute difference greater than or equal to 3 percent.
(4)
Deleted.
(5)
Each train or logic channel shall be tested at least every 62 days on a STAGGERED TEST BASIS.
The test shall independently verify the OPERABILITY of the undervoltage and automatic shunt trip circuits.
(6)
Neutron detectors may be excluded from CHANNEL CALIBRATION.
(7)
Below P-6 (Block of Source Range Reactor Trip) setpoint.
(8)
Deleted.
(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.
(11) -
Automatic and manual undervoltage trip.
SEQUOYAH - UNIT 2 3/4 3-13 Amendment No. 46, 104, 132
o INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2 The Engineered Safety Feature Actuation System (ESFAS) instrumentation 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:
With an ESFAS instrumentation channel or interlock trip setpoint a.
less conservative 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 con-sistent with the Trip Setpoint value, b.
With an ESFAS instrumentation channel or interlock inoperable, take the ACTION shown in Table 3.3-3.
SURVEILLANCE REQUIREMENTS t
4.3.2.1.1 Each ESFAS instrumentation channel and interlock shall be demon-strated 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 legic 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 leas; 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 l.
the " Total No. of Channels" Colunn of Table 3.3-3.
l t
L SEQUOYAH - UNIT 2 3/4 3-14 L
l i ii.. In -
TABLE 3.3-3 Nj -
ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUNENTATION Y
x MININUM e
TOTAL NO.
CHANNELS CHANNELS APPLICA8tE c5 FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 1.
SAFETY INJECTION, TURBINE TRIP AND FEEDWATER ISOLATION a.
Manual Initiation 2
1 2
1,2,3,4 20 b.
Automatic Actuation 2
1 2
1,2,3,4 15 Logic c.
Containment 3
2 2
1,2,3 17*
{
Pressure-High Y
d.
Pressurizer 3
2 2
1, 2, 3#
17*
M Pressure - Low e.
Deleted
?
3e es
~.
M TABLE 3.3-3 (Continued)
ENGINEERED SAFETY FEATURE ACTUATIDM SYSTEM INSTRUMEN1dTION 2
I MINIMUM TOTAt NO.
CHAMMELS CHANNEL.S APPt.1 CABLE I.
g FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERA 8tE MODES ACTION M
f.
Steam Line Pressuce-3/ steam line 2/ steam line. 2/ steam line 1, 2, 3" 17*
m Low in any steam line 2.
CONTAIMMENT SPRAY a.
Manual ~
Z 1**
2 1,2,3,4 20 b.
Automatic Actuation 2
1 2
1,2,3,4 15 Logic i
c.
Containment Pressure--
4 2
3 1,2,3 18 w
}
High-High l
3.
CONTAINNENT ISOLATION I
a.
Phase "A" Isolation l
I). Manual 2
1 2
1, 2, 3, 4 20 y
2)
From Safety Injection 2
1 2
1,2,3,4 15 g
Automatic Actuation g
Logic Er E.
.U
- Two switches must be operated simultaneously for actuation.
t w
N t
l J
l l
I
O.-
TABLE 3.3-3 (Continued)
ENGINEERED SAEETY FEATURE ACTUATION SYSTEM INSTRUMENTATION s
x 4
MININtM c
TOTAL NO.
CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE M00ES ACTION 3.
CONTAINMENT ISOLATION b.
Phase "B" Isolation 1)
Manual 2
1**
2 1,2,3,4 20 2)
Automatic 2
1 2
1,2,3,4 15 Actuation Logic 3)
Containment 4
2 3
1,2,3 18 Pressure-High-High t
T c.
Containment Ventilation 0
Isolation 1)
Manual 2
1 2
1,2,3,4 19 2)
Automatic Isolation 2
1 2
1,2,3,4 15 Logic 3)
Containment Gac 2
1 1
1,2,3,4 19 Monitor Radioactivity-High 4)
Containment Purge 2
1 1
1,2,3,4 19 F
Air Exhaust Monitor E
Radioactivity-High E
s 5)
Containment Particu-- 2 1
1 1,2,3,4 19
{
1 ate Activity High o
- Two switches must be operated simultaneously for actuation.
r
~.
f TABLE 3.3-3 (Continued)
E g
ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION, z
MIdIMUM e
T3TAL NO.
CHANNELS CHANNELS MPLICABLE l
g FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE
_ MODES ACTION
[
4.
~ STEAM LINE ISOLATION a.
Manual 1/ steam line 1/ steam line 1/ operating 1, 2, 3 25 steam line b.
Automatic 2
1 2
1,2,3 23 Actuation Logic c.
Containment Pressure--
4 2
3 1,2,3 18 High-High d.
Steam Line Pressure-3/ steam line 2/ steam line 2/ steam line 1, 2, 3 17*
Low in any steam line oo e.
Negative Steam Line 3/ steam line 2/ steam line 2/ steam line 3"
17*
Pressure Rate-High in any steam r
lines e
.O W
e M
(Al f4 I
. =
=
=
~.
TABLE 3.3-3(bontinued)
ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION-8 5
MINIMUM
-TOTAL NO.
CHANNELS CHANNELS APPLICABLE E
FUNCTIONAL UNIT OF CHANNELS
.TO TRIP OPERABLE MODES ACTION
[
m 5.
FEEDWATER ISOLATION a.
Steam Generator 3/ loop 2/ loop in 2/ loop in 1, 2, 3 17*
l Water Level--
any oper-each oper-High-High ating loop ating loop b.
Automatic Actuation 2-1 2
1,2,3 23 Logic 6.
Manual Initiation 2
1 2
1,2,3 24 Y
b.
Automatic Actuation 2
1 2
1,2,3 23 G
Logic c.
Main Steam Generator.
Water Level--Low-Low i.
Start Motor-Driven Pumps 2,
2
-a.
Steam Gen.
3/Stm. Gen.
2/Stm. Gen.
2/Stm. Gen.
1,2,3 36*
R Water Level--
in any in each Low-Low operating operating 3
(Adverse)-
Stm. Gen.
Sta. Gen.
~E b.
Steam Gen.
3/Stm. Gen.
2/Stm. Gen.
2/Stm. Gen.
1, 2, 3 36*
Water Level--
in any in each O
Low-tow operating operating h
(EAM)
Stm. Gen.
Stm. Gen.
s w
^
7.
RCS Loop AT 4(1/ loop) 2 3
1, 2, 3 37*
U
- Two switches must be operated simultaneously for actuation.
C
~
~
TABLE 3.3-3 (Continued)
E g
ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION
-e?
MINIMUM TOTAL NO.
CHANNELS CHANNELS APPLICABLE g
FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION A
d.
Containment 4
2 3
1,2,3 38*
Pressure (EAM) 11.
Start Turbine-'
Driven Pump a.
Steam Gen.
3/Stm. Gen.
2/Stm. Gen.
2/Stm. Gen.
1,2,3 36*
L ter Level--
in any 2 in each Low-Low operating operating
,g (Adverse)
Stm. Gen.
Sta. Gen.
T
'b.
Stm. Gen.
. 3/Sta. Gen.
2/Sta. Gen.
2/Stm. Gen.
1, 2, 3 36*
I$
- Water _ Level--
in any 2 in each Low-Low operating operating (EAM)
Stm. Gen.
Stm. Gen.
c.
RCS Loop AT 4(1/ loop) 2 3
1,2,3 37*
d.
Containment 4
2 3
1,2,3 38*
Pressure (EAM) d.
S.I.
E Start Motor-Driven Pumps and Turbine g
Driven Pump See 1 above (all S.I. initiating functions and requirements)
?r E
N
(.A)
FO 4
.s n'
i s'
a
M TABLE 3.3-3 (Continued) og ENGINEERED SAFETY FEATURE AC!UATION SYSTEM INSTRUMENTATION s
z-MINIMUM i
TOTAL NO.
CllANNELS CHANNELS APPLICABLE g
FUNCTIONAL UNIT'
.0F CHANNELS TO TRIP OPERABLE MDDES ACTION "e
Station Blackout e.
y Start Motor-Driven Pump associated 2/ shutdown 1/ shutdown 2/ shutdown with the shutdown board-board board 1,2,3 20 board and Turbine Driven Pump f.
Trip of Main Feedwater Pumps Start Motor-Driven -
R Puaps and Turbine
[
Driven Pump 1/ pump 1/ pump 1/ pump 1, 2 20*
b g.
Auxiliary Feedwater Suction Pressure-Low 3/ pump 2/ pump 3/ pump 1, 2, 3 21*
l 1
l h.
Auxiliary Feedwater Suction Transfer Time Delays 1.
Motor-Driven Pump 1/ pump 1/ pump 1/ pump 1, 2, 3 21*
g 2.
Turbine-Driven g
Pump 2/ pump 1/ pump 2/ pump 1,2,3 21*
"s r
Z*
t.D o,
l s
w
, m s
~
=
. TABLE 3.3-3 (Continued) b ENGINEERED' SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION 8
Y MINIMUM TOTAL NO.-
CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE' MODES ACTION E
l Z
- 7. LOSS OF POWER l
m a.
6.9 kv Shutdown Board
--Loss of Voltage 1.
Start Diesel 2/ shutdown 1 loss of 2/ shutdown 1, 2, 3, 4 20*
Generators board voltage on board any shutdown board 2.
Load Shedding 2/ shutdown 1/ shutdown 2/ shutdown 1, 2, 3, 4 20*
R board board board s
T b.
.6.9 kv Shutdown Board N
Degraded Voltage 1.
Voltage Sensors 3/ shutdown 2/ shutdown 2/ shutdown 1, 2, 3, 4 20 board board.
board 2.
Diesel Generator 2/ shutdown 1/ shutdown 1/ shutdown 1, 2, 3, 4 20 Start and Load board board board Shedding Timer 3.
SI/ Degraded 2/ shutdown 1/ shutdown 1/ shutdown 1, 2, 3, 4 20 Voltage Enable board board board Timer 5
2 P.
F m
-TABLE 3.3-3 (Continued) m 5g ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION g
1 MINIMUM g
TOTAL'NO.
CHANNELS CHANNELS APPLICABLE q
FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 8.
ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INTERLOCKS-a.
Pressurizer Pressure -
3 2
2 1,2,3 22a P-11/Not P-11' b.
Deleted c.
Steam Generator 3/ loop 2/ loop 3/ loop 1, 2 22c
,g Level P-14 any loop T
9.
AUTOMATIC SWITCHOVER TO y
CONTAINMENT SUMP a.
RWST Level - Low 4
2 3
1,2,3,4 18 COINCIDENT WITH Containment Sump Level - High 4
2 3
1,2,3,4 18 AND Safety Injection
'(See 1 above for Safety Injection Requirements)
E b.
Automatic Actuation 2
I 2
1,2,3,4 15
- g Logic R
E F
s N
- - - - - - ~ - - -
- ^ - ~
i
z.:--------
.w TABLE 3.3-3 (Continued)
TABLE NOTATION
- rip function may be bypassed in this MODE below P-11 (Pressurizer T
Pressure Block of Safety Injection) setpoint.
,, Trip function automatically blocked above P-11 and may be blocked below P-11 when Safety Injection on Steam Line Pressure-Low is not blocked.
,,,The channel (s) associated with the protective functions derived from the
,out of service Reactor Coolant Loop shall be placed in the tripped mode.
The provisions of Specification 3.0.4 are not applicable.
ACTION STATEMENTS ACTION 15 -
With the number of OPERABLE Channels one less than the Total Number of Channels, be in HOT STANDBY within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in l
~ COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />; however, one channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance l
testing per Specification 4.3.2.1.1 provided the other channel is OPERABLE.
ACTION 16 -
Deleted.
ACTION 17 -
With the qumber of OPERABLE Channels one less than the Total Number of Channels, STFTUP and/or POWER OPERATION may proceed provi M tha followin
.onditions are satisfied:
The inoperable channel is placed in the tripped a.-
condition within'6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
b.
The Minimum Channels OPERABLE requirements is met;.
however, the. inoperable channel may be bypassed for
)
up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing of other i
channels'per Specification 4.3.2.1.1.-
ACTION 18 -
With the number of OPERABLE Channels one less than the Total Number of Channels, operation may proceed provided the inoperable channel is placed in the bypassed condition and the Minimum Channels OPERABLE requirement is met; one addi-l tional channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing per Specification 4.3.2.1.1.
l ACTION 19 -
With less than the Minimum Channels OPERABLE, operation may continue provided the containment ventilation isolation valves are maintained closed.
. ACTION 20 -
With the number of OPERABLE Channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE l
status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SEQUOYAH - UNIT 2 3/4 3-22 Amendment No. 55,132 l
i
(-
- s a-l Y TABLE 3.3-3 (Continued)
ACTION 21 -
With less than the Minimum Number of Channels OPERABLE, declare the associated auxiliary feedwater pump inoperable, and c.omply with the. ACTION requirements of Specification 3.7.1.2.
ACTION 22 With less than the Minimum Number of Channels OPERABLE, declare the interlock inoperable and verify that all affected channels 4
of the functions listed below are OPERABLE or apply the appropriate ACTION statement (s) for those functions.
Functions to be evaluated are:
a.
Safety Injection Pressurizer Pressure V
Steam Line Pressure Negative Steam Line Pressure Rate b.
Deleted c.
Turbine Trip Steam Generator Level High-High Feedwater-Isolation Steam Generator Level High-High ACTION 23 -
With-the number of OPERABLE channels one less than the Total Number of Channcis, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; however, one channel may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />'for surveillance testing per Specification 4.3.2.1.1.
ACTION 24 -
With the number of OPERABLE channels one less than the Total Number of Channels, restore the. inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HOT-STANDBY within i
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
y ACTION 25 -
With the number of OPERABLE channels one less than the Total Number of Channels, restore-the inoperable channel to OPERABLE status within 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />s-or-declare the associated valve inoperable and'take the ACTION required by Specification 3.7.1.5.
ACTION 36 -
With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may
- r proceed provided the following conditions are satisfied:
The inoperable channel is placed.in the tripped-a.
condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, b.
For the affected protection set, the Trip Time Delay for one affected steam generator (T ) is adjusted to 3
match the Trip Time Delay for multiple affected steam generators (T ) within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
g SEQUOYAH - UNIT 2 3/4 3-23 Amendment No. 55, 116, 132
A r.1
-c r
TABLE 3.3-3 (Continued)
The Minimum Channels OPERABLE requirement is met; c.
however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing of other channels per Specification 4.3.1.1.1.
ACTION 37 -
With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided that within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, for the affected protection set, the Trip Time Delays (T and T ) threshold 3
g power level for zero seconds time delay is adjusted to 0%
RTP.
ACTION 38 -
With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided that within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, for the affected pro-tection set, the Steam Generator Water Level - Low-Low (EAM) channels trip setpoint is adjusted to the same value as Steam Generator Water Level - Low-Low (Adverse).
i i
l+
l s
1 l'
l' i
O
'1 i
l SEQUOYAH - UNIT 2 3/4 3-23a Amendment No.132 l
l
k'Y TABLE 3.3-4 b
8 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS
-Y I
FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES E
1.
SAFETY INJECTION, TURBINE. TRIP AND
'Z FEEDWATER ISOLATION
~
a.
Manual Initiation Not Applicable Not Applicable b.
Automatic Actuation Logic Not Applicable Not Applicable Containment Pressure--High 11.54 psig 11.6 psig c.
d.
Pressurizer Pressure--Low 11870 psig 11864.8 psig e.
Deleted R
l f.
Steam Line Pressure--Low Y
1600 psig steam line 1592.2 psig steam line pressure (Note 1) pressure (Note 1) a 2e ta
~
l
TABLE 3.3-4 (Continued)
ENGINEERED SAFETY FEATURE ACTt,ATION SYSTEM INSTRUMENTATION TRIP SETPOINTS 85 FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES E
2.
m a.
Manual Initiation
-Not Applicable Not Applicable b.
Automatic Actuation Logic Not Applicable Not Applicable c.
Containment Pressure--High-High 12.81 psig 12.9 psig 3.
CONTAINMENT ISOLATION a.
Phase "A" Isolation 1.
Manual Not Applicable Not Applicable g
T 2.
From Safety Injection Not Applicable Not Applicable Automatic Actuation logic b.
Phase "B" Isolation 1.
Manual Not Applicable Not Applicable 2.
Automatic Actuation Logic Not Applicable Not Applicable 3.
Containment Pressure--High-High 12.81 psig 12.9 psig y
c.
Containment Ventilation Isolation Eg 1.
Manual Not Applicable Not Applicable a"
2.
Automatic Isolation Logic Not Applicable Not Applicable 5
m
I.
TABLE 3.3-4 (Continued)
ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS
- 8. j' FUMCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES 3.
Containment Gas Monitor
-3
~3 E
Radioactivity-High
--<8.5 x 10 Ci/cc
<8.5 x 10 pCi/cc
-4
-3
-3 ro 4.
Containment Purge Air Exhaust 18.5 x 10 Ci/cc 18.5 x 10 C1/cc Monitor Radioactivity-High
-5
-5 5.
Containment Particulate 11.5 x 10 Ci/cc 11.5 x 10 Ci/cc Activity-High~
4.
STEAM LINE ISOLATION a.
Manual Not Applicable.
Not Applicable R
b.
Automatic Actuation Logic Not Applicable Not Applicable s
Y c.
Containment Pressure--High-High
$2.81 psig 12.9 psig d.
Steam Line Pressure--Low
>600 psig steam line
>592.2 psig steam line
_pressure (Note 1) pressure (Note 1) e.
Negative Steam Line Pressure
$100.0 psi (Note 2) 1107.8 psi (Note 2)
Rate--High 5.
TURBINE TRIP AND FEEDWATER ISOLATION a.
Steam Generator Water level--
581% of narrow range 181.7% of narrow range 3,,
High-High instrument span each instrument span each steam generat.r steam generator S
b.
Automatic Actuation Logic N. A.
N.A.
8
.O W
U fb
~.-
.r.
~.
TABLE 3.3-4 (Continued)
.mE
_g ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES g
6.
[
a.
Manual Not Applicable Not Applicable b.
Automatic Actuation Logic Not Applicable Not Applicable c.
Main Steam Generator Water Level--Low-Low i.
RCS Loop AT Equivalent RCS Loop AT variable RCS Loop AT variable-.
to Power 150% RTP input 5 0% RTP input < trip setpoint 5
+2.5% RTP Coincident with Steam
>15.0% of narrow range
>14.4% of narrow
[
Generator Water Level--
Instrument span range instrument span Low-Low (Adverse) and u
Containment Pressure-EAM 10.5 psig 10.6 psig or Steam Generator Water
>10.7% of narrow range
>10.1% of narrow Level--Low-Low (EAM)
Instrument span Instrument sp'n with A time delay-(T ) if one
$T3 (Note 5, Table 2.2-1) 5 (1.01) T3 (Note 5, 3
Steam Generator is affected Table 2.2-1) 30 or 5
A time delay (T,) if two 1 T, (Note 5, Table 2.2-1) 1 (1.01) T,(Note 5, E
or more Steam Generators Table 2.2-1) are affected 5
T e
~
e' M
TABLE-3.3-4 (Continued)-
j ENGINEERED SAFETY FEATifRE ACTUATION SYSTEM' INSTRUMENTATION TRIP SETPOINTS b
FUNCTIONAL UNIT ~
TRIP SETPOINT ALLOWABLE VALUES c-ii.
RCS Loop AT Equivalent to y
Power > 50% RTP Coincident with Steam
>15.0% of narrow range
>14.4% of narrow.
Generator Water Level--
Instrument span range instrument span
-Low-Low (Adverse) and Containment Pressure (EAM) 10.5 psig 10.6 psig or Steam Generator Water
>10.7% of narrow range
>10.1% of narrow range Level--Low-tow (EAM)
Instrument span.
Instrument span
{
d.
S.I.
See 1 above (all SI Setpoints)
T e.
Station Blackout 0 volts with a 5.0 second 0 volts with a 5.0 i y
time delay 1.0 second time delay f.
Trip of Main Feedwater N.A.
N. A.
Pumps
,a 8
g.
Auxiliary Feedwater Suction
> 2 psig (motor driven pump)
> 1 psig (motor driven pump)
{
Pressure-Low
[13.9psig(turbinedriven i12psig(pumpturbinedriven) g pump)
I h.
Auxiliary Feedwater Suction 4 seconds'(motor driven pump) 4 seconds 10.4 seconds Transfer Time Delays (motor driven pump) 5.5 seconds (turbine driven pump) 5.5 seconds 0.55 seconds mf (turbine driven pump)
U 3
m
.. [,
TABLE 3.3-4 (Continued)
ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS 5
FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES 7.
LOSS OF POWER zZ a.
6.9 kv Shutdown Board Undervoltage Loss of Voltage m
1.
Start of Diesel Generators a.
Nominal Voltage Setpoint 4860 volts 4860 volts i97.2 volts b.
Relay Response' Time for 0 volts with a 1.5 second 0 volts with a 1.5 i0.5 Loss of Voltage time delay second time delay 2.
Load Shedding a.
Nominal Voltage Setpoint 4860 solts 4860 volts 197.2 volts b.
Relay Response Time for 0 volts with a 5.0 second 0 volts with a 5.0 il.0 Loss of Voltage time dt_ lay second time delay b.
6.9 kv Shutdown Board-Degraded Y
Voltage-1.
Voltage Sensors 6560 volts 6560 volts i 33 volts
. 2.
Diesel Generator Start and Load Shed Timer 300 seconds 300 seconds i 30 seconds 3.
SI/ Degraded Voltage Logic Enable Timer 10 seconds 10 seconds 1 0.5 seconds E
8.
ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INTERLOCKS a"
a.
Pressurizer Pressure 5
1.
Not P-11, Automatic Unblock
$1970 psig 11975.2 psig of Safety Injection on Increasing Pressure d
~
11956.8 psig 2.
P-11 Enable Manual Block of 11%2 psig Safety Injection on Decreasing Pressure w
N
s TABLE 3.3-4 (Continued) b ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS O
l 5
i f
FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES EE 8.
ENGINEERED SAFETY FEATURE ACTUATION l
SYSTEM INTERLOCKS (Continued) b.
Deleted c.
Deleted d.
Steam Generator Level Turbine Trip, Feedwater Isolation P-14 (See 5. above) 9.
AUTOMATIC SWITCHOVER TO R
CONTAINMENT SUMP T
a.
RWST Level - Low 130" from tank base 130" i 2.71" from tank base i
COINCIDENT WIlH Containment Sump Level - High 30" above elev. 680' 30" i 1.68" above elev. 680' AND Safety Injection (See 1 above for all Safety Injection Setpoints/ Allowable Valves) b.
Automatic Actuation Logic N.A.
N.A.
Note 1:
Time constants utilized in the lead-lag controller for Steam Pressure-Low are 11 ~> 50 g
seconds and T2 < 5 seconds.
j[
Note 2:
Time constant utilized in the rate-lag controller for Negative Steam Line Pressure Rate-High 2
is ti > 50 seconds.
5 W
M M
to
.(
TABLE 3.3-5 (Continued)
ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 3.
Pressurizer Pressure-Low a.
Safety Injection (ECCS) 132.0(1)/28.0(7) b.
Reactor Trip (from SI) 5 3.0 c.
Feedwater Isolation
< 8.0(2) d.
Containment Isolation-Phase "A"(3) 18.0(8) e.
Containment Ventilation Isolation 5.5(8)(13) f.
Auxiliary Feedwater Pumps 160(11) g.
Essential Raw Cooling Water System 565.0(8)/75.0(9) h.
Emergency Gas Treatment System 128.0(8) 4.
Deleted 5.
Negative Steam Line Pressure Rate-High a.
Steam Line Isolation 5 8.0 1
I SEQUOYAd - UNIT 2 3/4 3-30 Amendment No. 47, 68, 96, 132 l
n a
TABLE 3.3-5 (Continued)
ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 6.
Steam Line Pressure-Low a.
Safety Injection (ECCS) 5 28.0(7)/28.0(1) b.
Reactor Trip (from SI)
$ 3.0 c.
Feedwater Isolation
< 8.0(2) d.
Containment Isolation-Phase "A"(3)
}18.0(8)/28.0(9) e.
Containment Ventilation Isolation Not Applicable f.
Aur?liary Feedwater Pumps 160(11) j g.
Essential Raw Cooling Water System 5 65.0(9)/75.0(9) h.
Steam Line Isolation
< 8.0 1.
Emergency Gas Treatment System 38.0(9) 7.
Containment Pressure--High-High a.
Containment Spray 5 208(9) b.
Containment Isolation-Phase "8"(12) 1 65(8)/75(9) c.
Steam Line Isolation 5 7.0 d.
Containment Air Return Fan
> 540.0 and 1660 8.
Steam Generator Water Level--Hich-High a.
Turbine Trip 5'2.5 b.
Feedwater Isolation 5 11.0(2) 9.
_ Main Steam Generator Water LC? ell Low-Low a.
Motor-driven Auxiliary 1 60.0(14)
Feedwater Pumps (4) b.
Turbine-driven Auxiliary
< 60.0(14)
~
Feedwater Pumps (5)(11)
SEQUOYAH - UNIT 2 3/4 3-31 Amendment No. 47, 51, 55, 68, 73, 132
TABLE 3.3-5 (Continued)
TABLE NOTATION (10) The response time for loss of voltage is measured from the time voltage is lost until the time fell voltage is restored by the diesel.
The response time for degraded voltage is measured from the time the load shedding signal is generated, either from the degraded voltage or the SI enable timer, to the time full voltage is restored by the diesel.
The response time of the timers is covered by the requirements on their setpoints.
l (11) The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 for the turbine-driven Auxiliary feedwater Pump.
(12) The following valves are exceptions to the response times shown in the table and will have the values listed in seconds for the initiating signals and the function indicated:
Valves:
FCV-67-89, -90, -105, -106 Response times:
7.b.
75(8)/85(9)
Valve:
FCV-70-141 Response times:.7.b, 70(8)/80(9)
(13) Containment purge valves _only.
Containment radiation monitor valves have l
a response time of 6,5 seconds or less.
1 (14) Does not include Trip Time Delays.
Response times noted include the transmitters, Eagle-21 process protection cabinets, solid state 4
protection cabinets, and actuation devices (up to and including pumps).
This reflects the response times necessary for THERMAL POWER in excess of 50% RTP.
l t
SEQUOYAH - UNIT 2 3/4 3-33a Amendment No. 18, 68, 73, 96, 132-
TABLE 4.3-2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION 8
SURVEILLANCE REQUIREMENTS Y
1 CHANNEL MODES FOR WHICH E
CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE IS Q
FUNCTIONAL UNIT-CHECK CALIBRATION TEST _
REQUIRED to 1.
SAFETY INJECTION AND FEEDWATER ISOLATION a.
Manual Initiation M. A..
.N.A.
R 1,2,3,4 b.
Automatic Actuation Logic N.A.
N.A.
M(1) 1, 2, 3, 4 c.
Containment Pressure-High 5-R Q
1, 2, 3 g
d.
Pressurizer Pressure--Low S
R Q
1,2,3 l
T e.
Deleted to Jm f.
Steam Line Pressure--Low S
R Q
1,2,3 l
2.
Manual Initiation N.A.
N.A.
R 1,2,3,4 b.
Automatic Actuation Logic N.A.
N.A.
M(1) 1, 2, 3, 4 c.
Containment Pressure--High-High S R
Q 1,2,3 N
a 8e l'
M IV
~
a.
i TABLE 4.3-2 (Continued) b ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION l
l 8
SURVEILLANCE REQUIREMENTS l
5 t
x CHANNEL MODES FOR WHICH E
FUNCTIONAL UNIT
. CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE IS CHECK CALIBRATION TEST REQUIRED Q
j m
3.
CONTAINMENT ISOLATION l
a.
Phase "A" Isolation 1
j
- 1) Manual N.A.
N.A.
R 1,2,3,4
- 2) 'From Safety Injection..
N.A.
N. A.
M(1)-
1,2,3,4 Automatic Actuation Logic b.
Phase "B" Isolation
- 1) Manual N.A.
N.A.
R 1, 2, 3, 4 R
- 2) Automatic Actuation Logic N. A.
N.A.
M(1) 1, 2, 3, 4 Y
M
- 3) Containment Pressure--
S R
Q 1, 2, 3 High-High c.
Containment Ventilation Isolation
- 1) Manual N.A.
N.A.
R 1,2,3,4
- 2) Automatic Isolation Logic N. A.
N.A.
M(1) 1, 2, 3, 4
- 3) Containment Gas Monitor S
R M
1,2,3,4 l
Radioactivity-High.
N a
ao
.n_.
TABLE 4.3-2 (Continued)
M ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIC6MENTS 0>*
CHANNEL MODES FOR WHICH CHANNEL CFANNEL FUNCTIONAL SURVEILLANCE IS FUNCTIONAL UNIT-CHECK Cf.t.IBRATION TEST-REQUIRED i
Z
- 4) Containment Purge Air S
R M
1,2,3,4 m
Exhaust Monitor Radio-l activity-High
- 5) Containment Particulate S
R M
1,2,3,4 Activity-High 4.
STEAM LINE ISOLATION i
a.
Manual N.A.
N. A.
R 1,2,3 b.
Automatic Actuation Logic N.A.
N.A.
M(1) 1, 2, 3 w1 c.
Containment Pressure--
S R
Q 1,2,3 m
J, High-High cn d.
Steam Line Pressure--Low S
R Q
1, 2, 3 e.
Negative Steam Line Pressure S
R Q
3 Rate--High 5.
TURBINE TRIP AND FEEDWATER ISOLATION k
a.
' Steam Generator Water S
R Q
1,2,3 g
Level--High-High e"g b.
Automatic Actuation Logic N.A.
N. A.
M(1) 1, 2, 3 g
6.
g a.
Manual N.A.
N.A.
R 1, 2, 3 b.
Automatic-Actuation Logic N.A.
M.A.
M(1) 1, 2, 3 l
w (4
.fV
.s m
m.
TABLE 4.3-2 (Continued) b ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUNENTATION 8-SURVEILLANCE REQUIREMENTS h
CHANNEL MODES FOR WHICH CHANNEL CHANNEL-FUNCTIONAL SURVEILLANCE IS FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED c-5 c.
Main Steam Generator Water Level--Low-Low 1.
Steam Generator Water.
S R
Q 1, 2, 3 Level--Low-Low (Adverse) 2.
Steam Generator Water S
R Q
1,2,3 Level--Low-Low (EAM) i 3.
RCS Loop AT S
R Q
1, 2, 3 R
4.
Containment-Pressure S
R Q
1, 2, 3 (EAM)
Y 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 g.
Auxiliary Feedwater Suction M.A.
R M
1, 2, 3 y
Pressure-Low
'h.
Auxiliary Feedwater Suction N.A.
R N.A.
1,2,3 Transfer Time Delays-g 7.
LOSS OF POWER g
a.
6.9 kv Shutdown Board -
Loss of Voltage h
1.
IStart Diesel Generators S
R' M
1,2,3,4 2.
Load Shedding S
R N.A.
1,2,3,4
1
=.
TABLE 4.3-2 (Continued) b ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODES FOR WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE IS E
FUNCTIONAL UNIT CHECK CALIBRATION TEST-REQUIRED l
b.
6.9 kv Shutdown Board -
Degraded Voltage 1.
. Voltage sensors S
R M
1,2,3,4 2.
Diesel Generators N.A.
R N.A.
1,2,3,4 Start.and Load Shedding Timer 3.
SI/ Degraded Voltage-N.A.
R N.A.
1, 2, 3, 4 Logic Timer ws 8.
ENGINEERED SAFETY FEATURE Y
ACTUATION SYSTEM INTERLOCKS a.
Pressurizer Pressure, N.A.
R(2)
N.A.
1, 2, 3 P-11/Not P-11' b.
Deleted c.
Steam Generator N.A.
R(2)
N.A.
1, 2 g
Level, P-14 Eg 9.
AUTOMATIC SWITCHOVER TO 5
CONTAINMENT SUMP a.
RSWT Level - Low S'
R Q
1,2,3,4 2
COINCIDENT WITH Containment Sump Level - High S
R Q
1,2,3,4 AND Safety Injection (See 1 above for all Safety Injection Surveillance Requirements) b.
Automatic Actuation Logic N.A.
N.A.
M(1) 1, 2, 3, 4
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g IN
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LIMITING SAFETY SYSTEM SETTINGS BASES Intermediate and Source Rance, Nuclear Flux (Continued)
Range Channels will initiate a reactor trip at approximately 25 percent of
. RATED THERMAL POWER unless manually blocked when P-10 becomes active.
No j
credit was taken for operation of the trips associated with either the Inter-mediate or Source Range Channels in the accident analyses; however, their functional capability at the specified trip settings is required by this specification to enhance the overall reliability of the Reactor Protection
- System, t
Overtemperature AT
.The Overtemperature Delta T trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power dis-l tribution, provided that the transient is slow with respect to transit, thermo-twell, and RTD response-time delays from the core to the temperature detectors
>(about 8 seconds),-and pressure is within the range between the High and Low Pressure reactor trips.
This setpoint includes corrections for axial. power-distribution, changes in density and heat capacity of water with temperature and dynamic compensation for transport, thermowell, and RTD response time delays from the core-to the RTD output indication.
With normal axial power distribu-tion, this reactor trip limit is always below the core safety limit as shown in Figure 2.1-1.
If axial peaks'are greater than design, as indicated by the dif-L
-.ference between top and bottom power range nuclear detectors, the reactor trip E
is automatically. reduced according to the notations in Table 2.2-1.
l Operation.with a-reactor coolant loop out of service below the 4 loop P-8 setpoint does not require reactor protection system setpoint modification because the P-8 setpoint-and associated trip will prevent DNB during 3 loop operation ^ exclusive of.the Overtemperature Delta T setpoint.
Delta-T,, as used in the Overtemperature and Overpower AT trips, represents the -
100 percent RTP value as measured by the plant for each loop.
This normalizes m
~
each' loop's AT trips to the actual operating conditions existing at the time of l
measurement, thus forcing the trip to reflect the equivalent full power condi-tions as assumed in the accident analyses.
These differences in RCS loop AT can be due to several factors, e.g., measured RCS loop flows greater than ther-7' mal design flow, and slightly asymmetric power distributions between quadrants.
While;RCS loop flows are not expected to change-with cycle life, radial power h
' redistribution between quadrants may occur, resulting in small changes in_ loop
. specific AT values. Accurate determination of the loop. specific AT value should l?s be made when performing Incore/Excore quarterly recalibration and under steady L
state conditions (i.e., power distributions not affected by xenon or other transient conditions.).
l l
l
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SEQUOYAH - UNIT 2 B 2-4 Amendment No. 129, 132 i
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LIMITING SAFETY SYSTEM SETTINGS BASES Overpower AT The Overpower Delta T reactor trip provides assurance of fuel integrity, e.g.,
no melting, under all possible overpower conditions, limits the required range for Overtemperature Delta T protection, and provides a backup to the High Neu-tron Flux trip.
The setpoint includes corrections for changes in density and heat capacity of water with temperature, and dynamic compensation for transport, thermowell, and RTD response time delays from the core to the RTO output indication.
The Overpower Delta T trip provides protection to mitigate the consequences of various size steam breaks as reported in WCAP-9226, " Reactor Core Response to Excessive Secondary Steam Releases."
Delta-T, as used in the Overtemperature and Overpower AT trips, represents the g
100 percent RTP value as measured by the plant for each loop.
This normalizes each loop's AT trips to the actual operating conditions existing at the time of measurement, thus forcing the trip to reflect the equivalent full power condi-tions as assumed in the accident analyses.
These differences in RCS loop AT can be due to several factors, e.g., measured RCS loop flows greater than ther-mal design flow, and slightly asymmetric power distributions between quadrants.
While-RCS loop flows are not expected to change with cycle life, radial power redistribution between quadrants may occur, resulting in small changes in loop specific AT values.
Accurate determination of the loop specific AT value should be made when performing Incore/Excore quarterly recalibration and under steady _
state conditions (i.e., power distributions not affected by xenon or other transient conditions.).
Pressurizer Pressure The Pressurizer High and Low Pressure trips are provided to limit the pressure range in which reactor operation is permitted.
The High Pressure trip is backed-up by the pressurizer code. safety va'i.es for RCS overpressure protection, and is therefore set lower than the set pressure for these valves (2485 psig).
The
-Low Pressure trip provides protection by tripping the reactor in the event of a loss of reactor coolant pressure.
Pressurizer Water Level The Pressurizer High Water Level trip ensures protection against Reactor Coolant System overpressurization by limiting the water level to a volume sufficient to retain a steam bubble and prevent water relief through the pressurizer safety J
valves.
No credit was taken for operation of this trip in the accident analyses; however, its functional capability at the specified trip setting is required by this specification to enhance the overall reliability of the Reactor Protection System.
SEQUOYAH - UNIT 2 B 2-5 Amendment No.132
M IT_ING SAFETY SYSTEM SETTINGS BASES _
Loss of Flow The Loss of flow trips provide core protection to prevent DNB in the event of a loss of one or more reactor coolant pumps.
Above 11 percent of RATED THERMAL POWER, an automatic reactor trip will occur
)
if the flow in any two loops drops below 90 percent of nominal full loop flow.
Above the P-8 interlock, automatic reactor trip will occur if the flow in any-single loop drops below 90 percent of nominal full loop flow.
This latter trip will prevent the minimum value of the DNBR from going below 1.30 during normal operational transients and anticipated transients when 3 loops are in operation and the Overtemperature Delta T trip setpoint is adjusted to the value specified for all loops _ in operation.
J Steam _ Generator Water Level The. Steam Generator Water Level Low-Low trip protects the reactor from loss.of heat sink in the event of a sustained steam /feedwater flow mismatch resulting from loss of normal feedwater or a feedwater system pipe break, inside or out-side of containment.
This function also provides input to the steam generator level control system.
IEEE 279 requirements are satisfied by 2/3 logic for protection function actuation, thus allowing for a single failure of a channel and still performing the protection function.
Control / protection interaction is addressed by the use of the Median Signal felector which prevents a single failure of a channel providing input to the control system requiring protection i
function action.
That is, a single failure of a channel providing input to the control system does not result in the control system initiating a condition requiring protection function action.
The Median Signal Selector performs this
- by not selecting the channels indicating the highest or lowest steam generator levels as input to the control system.
With the transmitters located inside= containment and thus possibly experiencirJ.
adverse environmental conditions (due to a feedline break), the Environmental Allowance Modifier (EAM) was. devised.
The EAM function (Containment Pressure (EAM) with a setpoint of < 0.5 psig) senses the presence of adverse containment conditions (elevated pressure) and enables the Steam Generator Water Level--
Low-Low trip setpoint (Adverse) which reflects the increased transmitter uncer-tainties due to this environment. The EAM allows the use of a lower Steam Gen-erator Water Level - Low-Low (EAM) trip setpoint when these conditions are not present, thus allowing more margin to trip for normal operating conditions.
The Trip Time De'ay (TTD) creates additional operational margin when the plant needs it most, during early escalation to power, by allowing the operator time to recover level when the primary side load is sufficiently small to allow such action.
The TTD is based on' continuous monitoring of. primary side power through the use of RCS loop AT. Two time delays are calculated, based on the number of steam generators indicating less than the Low-Low Level trip setpoint and the primary side-power level.
The magnitude of the delays decreases with increasing SEQUOYAH - UNIT 2 B 2-6 Amendment No. 132
LIMITING SAFETY SYSTEM SETTINGS BASES Steam Generator Water Level (Cont'd)
L primary sid' power level, up to 50 percent RTP.
Above 50 percent RTP there are no time d81,sys for the low-Low level trips.
In the event of failure of a Steam Generator Water Level channel, it is placed in the trip condition as input to the Solid State Protection System and does not affect either the EAM or TTD setpoint calculations for the remaining oper-t able channels.
It is then necessary for the operator to force the use of the shorter TTD time delay by adjustment of the single steam generator time delay calculation (T ) to match the multiple steam generator time delay calculation 3
(T ) for the affected protection set, through the MMI.
Failure of the Contain-M l-ment Pressure (EAM) channel to a protection set also does not affect the EAM setpoint calculations.
This results in the requirement that the operator adjust the affected Steam Generator Water Level - Low-Low (EAM) t. rip setpoints to the same value as the Steam Generator Water Level - Low-Low (Adverse).
Failure of the RCS loop AT channel input (failure of more than one T RTD or failure of a H
T RTD) does not affect the TTD calculation for a protection set.
This results C
in the requirement that the operator adjust the threshold power level for zero seconds time delay from 50 percent RTP to 0 percent RTP, through the MMI.
v L
.Undervoltage and Underfrequency - Reactor Coolant Pump Busses The Undervoltage and Underfrequency Reactor Coolant Pump bus trips provide-reactor core protection against DNB as a result of loss.of voltage or underfre-quency to more than one reactor coolant pump.
The specified setpoints' assure a reactor trip. signal is generated before the low flow trip setpoint is reached, l
Time delays are incorporated in the underfrequency-and undervoltage trips to l
i; prevent spurious reactor trips from momentary electrical power transients.
For undervoltage, the delay is set so that the time required for a signal to reach the reactor trip ~ breakers following the simultaneous trip of two or more reactor i
coolant pump bus circuit breakers shall not exceed 1.2' seconds.
For underfre-quency, the delay is set so that the time required for a signal to reach the reactor trip breakers after the underfrequency trip setpoint is reached shall not exceed 0.6' seconds.
Turbine Trip A Turbine Trip causes a direct reactor trip when operating above P-9.
Each of the-turbine trips provide turbine protection and reduce the severity of the ensuing transient.
No credit was taken in the accident analyses for operation of these trips.
Their functional capability at the specified trip settings is required to enhance the overall reliability of the Reactor Protection System.
Amendment No. 132 SEQUOYAH - UNIT 2 B 2-7 Revised 08/18/87
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LIMITING SAFETY SYSTEM SETTINGS BASES Safety Injection Input from ESF If a reactor trip has not already been generated by the reactor protective instrumentation, the ESF automatic actuation logic channels will initiate a reactor trip upon any signal which initiates a safety injection.
This trip is provided to protect the core in the event of a LOCA.
The ESF instrumentation channels which initiate a safety injection signal are shown in Table 3.3-3.
Reactor Trip System Interlocks The Reactor Trip System Interlocks perform the following functions on increasing power:
P-6 Enables the manual block of toe source range reactor trip (i.e.,
prevents premature block of source range trip).
P-7 Defeats the automatic block of reactor trip on:
Low flow in more P-13 than one primary coolant loop, reactor coolant pump undervoltage and underfrequency, pressurizer low pressure, and pressurizer high level.
P-8 Defeats the automatic block of reactor trip on low RCS coolant flow in a single loop.
P-9 Defeats the automatic block of reactor trip on turbine trip.
P-10 Enables the manual block of reactor trip on power range (low setpoint),
intermediate range, as a backup block for source range, and intermedi-ate range rod stops (i.e., prevents premature block of the noted functions).
'On decreasing power, the opposite function is performed at reset setpoints..
P-4 Reactor-tripped - Actuates turbine trip, closes main feedwater valves on T,yg below setpoint, prevents the opening of the main feedwater valves which were closed by a safety injection or high steam generator-water level signal, allows manual block of the automatic reactuation of safety injection.
Reactor not tripped - defeats manual block preventing automatic reactuation of safety injection.
Amendment No. 132 SEQUOYAH - UNIT 2 B 2-8 Revised 08/18/87
s i
3/4.3 INSTRUMENTATION i
BASES
~
3/4.3.1 and 3/4.3.2 REACTOR TRIP AND ENGINEERED SAFETY FFATURE ACTUATION SYM EM INSTRUMENTATION Yhe OPERABP..ITY 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 i
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 l
- 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 feeJwater isolation valves and intiibits feedwatr' control valve moculation.
1 On decreasing the required parameter the oppos' te action is performed at l
reset setpoints.
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SEQUOYAH - UNI 1' 2 B 3/4 3-1 Amendment No.132 i
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