ML20069D120
| ML20069D120 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 05/19/1994 |
| From: | Beckner W Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20069D123 | List: |
| References | |
| GL-93-08, GL-93-8, NUDOCS 9406020279 | |
| Download: ML20069D120 (22) | |
Text
.
g km Rf(ig O
Kg;j h[ Yp ' f UNITED STATES tb OLEAR REGULATORY COMMISSION
?
WASHINGTON, D.C. 20555-0001 GULF STATES UTILITIES COMPANY **
CAJUN ELECTRIC POWER COOPERATIVE AND ENTERGY OPERATIONS. INC.
DOCKET NO. 50-458 RIVER BEND STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 73 License No. NPF-47 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Gulf States Utilities * (the licensee) dated March 3, 1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, as amended, the provisions of the Act, ar.c M.9 nles 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 De conducted in compliance with the Commiss'on's regulations; D.
The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and
- E0I is authorized to act as agent for Gulf States Utilities Company, which has been authorized to act as agent for Cajun Electric Power Cooperative, and has exclusive responsibility and control over the physical construction, operation and maintenance of the facility.
- Gulf States Utilities Company, which owns a 70 percent undivided interest in River Bend, has merged with a wholly owned subsidiary of Entergy Corporation. Gulf States Utilities Company was the surviving company in the merger.
9406020279 940519 PDR ADOCK 05000458 P
,, E.
The issuance of this amendment is in accordance with 10 CFR Part 51' of the Commission's regulations and all applicable requirements have
'been satisfied.
2.
Accordingly, the license is amended by changes to the Technical.
Specifications as indicated in the attachment to this license amendment; and Paragraph 2.C.(2) of Facility Operating License No. NPF-47 is hereby amended to read as follows:
l (2) Technical Specifications and Environmental Protection Plan l
The Technical Specifications ' contained in Appendix A, as revised-through Amendment No. 73 and the Environmental Protection Plan i
contained in Appendix B, are hereby. incorporated in the license.
E01 shall. operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
The license amendment is effective as of its date of issuance.
I FOR THE NUCLEAR REGULATORY COMMISSION t'n D. he
~~
j William D. Beckner, Director l
Project Directorate IV-1 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: May 19, 1994 l
l-
~
we
.g
.~
ATTACHMENT TO LICENSE AMENDMENT NO. 73 FACILITY OPERATING LICENSE NO. NPF-47 DOCKET NO. 50-458 Replace the following pages of the Appendix A Technical Specifications'with the attached pages.
The revised pages are identified by Amendment number and contain marginal lines indicating the areas of change. The corresponding overleaf pages are also provided to maintain document completeness.
REMOVE INSERT xxiii xxiii 3/4 3-1 3/4 3-1 3/4 3-6 3/4 3-6 3/4 3-10 3/4 3-10 3/4 3-11 3/4 3-11 l
3/4 3-24 3/4 3-24 3/4 3-25 3/4 3-25 3/4 3-30 3/4 3-30 l
3/4 3-30a 3/4 3-30a 3/4 3-40 3/4 3-40 i
W I
w d'
,-r=-
~v+-
e n
p e,
wwwv.
v m
ev--*
e-~w e
~
~-,*m-eew-a-
INDEX LIST OF TABLES TABLE TITLE EAqE 1.1 Surveillance Frequency Notation.....................
1 10' l.2 Operation al Condi ti ons..............................
1-11 l
2.2.1-1 Reactor Protection System Instrumentation Setpoints.
2-4 3.3.1-1 Reactor Protection System Instrumentation...........
3/4 3-2 3.3.1-2 Deleted.............................................
3/4 3-6 l
4.3.1.1-1 Reactor Protection System Instrumentation-Surveillance Requirements...........................
3/4 3-7 l
3.3.2-1 Isol ation Actuation Instrumentation.................
3/4 3-12 l
3.3.2-2 Isolation Actuation Instrumentation Setpoints.......
3/4 3-19 3.3.2-3 Deleted.............................................
3/4 3-24 l
4.3.2.1-1 Isolation Actuation Instrumentation Surveillance Requirements........................................
3/4 3-26 j
3.3.3-1 Emergency Core Cooling System Actuation Instrumentation.....................................
3/4 3-31 l
3.3.3-2 Emergency Core Cooling System Actuation Instrumentation Setpoints...........................
3/4 3-36 3.3.3-3 Deleted.............................................
3/4 3-40 l
4.3.3.1-1 Emergency Core Cooling System Actuation j
Instrumentation Surveillance Requirements........
3/4 3-41 j
3.3.4.1-1 ATWS Recirculation Pump Trip System Instrumentation.
3/4 3-45 3.3.4.1-2 ATWS Recirculation Pump Trip System Instrumentation 1
Setpoints...........................................
3/4 3-46 RIVER BEND - UNIT 1 xxiii Amendment No.14,73
l INDEX LIST OF TABLES (Continued)
TABLE TITLE PAGE 4.3.4.1-1 ATWS Recirculation Pump Trip Actuation Instrumentation Surveillance Requirements...........
3/4 3-47
- 3. 3. 4. 2-1 End-of-Cycle Recirculation Pump Trip System l
Instrumentation.....................................
3/4 3-50 3.3.4.2-2 End-of-Cycle Recirculation Pump Trip System Setpoints...........................................
3/4 3-51
(
3.3.a.2-3 End-of-Cycle Recirculation Pump Trip System Response l
Time................................................
3/4 3-52 4.3.4.2.1-1 End-of-Cycle Recirculation Pump Trip System
)
l Surveillance Pequ1rements...........................
3/4 3-53 3.3.5-1 Reactor Core Isolation Cooling System Actuating I
Instrumentation.....................................
3/4 3-55 3.3.5-2 Reactor Core Isolation Cooling System Actuation Instrumentation Setpoints...........................
3/4 3-57 f
4.3.5.1-1 Reactor Core Isolation Cooling System Actuation Instrumentation Surveillance Requirements...........
3/4 3-58 3.3.6-1 Control Rod Block Instrumentation...................
3/4 3-60 1.3.G=2 Control Rod Block Instrumentation 5etpoints.........
3/4 3-62 4.3.6-1 Control Rod Block Instrumentation Surveillance l
Requirements........................................
3/4 3-63 3.3.7.1-1 Radiation Monitoring Instrumentation....s...........
3/4 3-66 i
l 4.3.7.1 Radiation Monitoring Instrumentation Surveillance Requirements........................................
3/4 3-69 3.3.7.2-1 Seismic Monitoring Instrumentation..................
3/4 3-71 4.3.7.2-1 Seismic Monitoring Instrumentation Surveillance Requirement2........................................
3/4 3-72 3.3.7.3-1 Meteorological Monitoring Instrumentation..........T 3/4 3-74
~
.A.3.7.3-1 Meteorological Monitoring Instrumentation Surveillance Requirements...........................
3/4 3-75 l
l RIVER BEND - UNIT 1-xxiv i
i
3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the reactor protection system instrumentation channels shown in Table 3.3.1-1 shall be OPERABLE.
APPLICABILITY: As shown in Table 3.3.1-1.
ACTION:
a.
With the number of OPERABLE channels less than required by the Minimun OPERABLE Channels per Trip System requirement for one trip syster., place the inoperable channel (s) and/or that trip system in the tripped condition
- within one hour.
b.
With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for both trip systems, place at least one trip system ** in the tripped condition within one hour and take the ACTION required by Table 3.3.1-1.
SURVEILLANCE RE0VIREMENTS 4.3.1.1 Each reactor protection system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.1.1-1.#
4.3.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months.***
4.3.1.3 THE REACTOR PROTECTION SYSTEM RESPONSE TIME of each required reactor tirp functional unit shall be demonstrated to be within its limit at least once per 18 months.
Neutron detectors are exempt from response time testing.
Each test shall include at least one channel per trip system such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip system.
An inoperable channel need not be placed in the tripped condition where this would cause the Trip Function to occur.
In these cases, the inoperable channel shall be restored to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or the ACTION required by Table 3.3.1-1 for that Trip Function shall be taken.
- The trip system need not be placed in the tripped condition if this would cause the Trip Function to occur.
When a trip system can be placed in the l
tripped condition without causing the Trip Function to occur, place the trip system with the most inoperable channels in the tripped condition, if
~both systems have the same number of inoperable channels, place either trip systen in the tripped condition.
The requirement to place a trip system in the tripped cuhdition does not apply to Functional Units 6 and 10 of Table 3.3.1-1.
- Logic System Functional Test period may be extended as identified by note
'p' on Table 4.3.1.1-1.
Channel Calibration period may be extended as identified by notes 'o' and 7
- q' on Table 4.3.1.1-1.
I RIVER BEND - UNIT 1 3/4 3-1 Amendment No. 8,47,72,73
iAHLt 3.3.1-1 g
REACTOR PROTECTION SYSTEM INSTRUMENTATION g
APPLICARLE MINIMUM g
OPERATIONAL OPERABLE CHANNELS FUNCTIONAL UNIT CONDITIONS _
PER TRIP SYSTEM (a)
ACTION 5
1.
-4 a.
Neutron Flux - High 2
3 1
w 3(b)4 3
2 5
3 3
b.
Inoperative 2
3 1
3, 4 3
2 5
3 3
2 2.
Average Power Range Monitor IC):
a.
Neutron Flux - High, Setdown 2
3 1
,.g 3(b)4 3
2 5
3 3
'b.
Flow Biased Simulated Thermal Power - High 1
3 4
c.
Neutron Flux - High 1
3 4
d.
Inoperative 1, 2 3
1 4
3, 4 3
2 5
3 3
3.
Reactor Vessel Steam Dome Pressure - High 1, 2(d) 2 1
i-4.
Reactor Vessel Water Level - Low, l
l Level 3 1, 2 2
1 5.
Reactor Vessel Water Level-High, l
Level 8 1(,)
2 4
l 6.
Main Steam Line Isolation Valve -
II ')
[.
Closure 4
10 7.
Main Steam Line Radiation -
High 1,2(d) 2 5
III 8.
Dry *11 Pressure - High 1, 2 2
1
e f
4 i
TABLE 3.3.1-1 (Continued)
REACTOR PROTECTION SYSTEM INSTRUMENTATION TABLE NOTATIONS (a)
T A channel may be placed in an inoperable status 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 required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter.
(b)
Unless adequate shutdown margin has been demonstrated per Specification 3.1.1, the " shorting links" shall be removed from the RPS circuitry prior to and during the time any control rod is withdrawn".
(c)
An APRM channel is inoperable if there are less than 2 LPRM inputs per level or less than 11 LPRM inputs to an APRM channel.
-(d)
This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1.
(e)
This function shall be automatically bypassed when the reactor mode switch is not in the Run position.
i (f)
This function is not required to be OPERABLE when DRYWELL INTEGRITY is not required.
(g) With any control rod withdrawn.
Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
(h)
This function shall be automatically bypassed when turbine first stage pressure is < 187 psig,** equivalent to THERMAL. POWER less than 40% of RATED THERMAL POWER.
~
~
"Not required for control rods removed
- To allow for instrumentation accuracy,per Specification 3.9.10.1 or 3.9.10.2.
< 177 psig turbine first stage pressure shall be used. calibration and drift, a setpo RIVER BEND - UNIT I 3/4 3-5
.~
Table 3.3.1-2 has been deleted, i
I l
I l
l l
l l
~
RIVER BEND - UNIT 1 3/4 3-6 Amendment No. 42, 72.73
5 lABLE 4.3.1.1-1 (Continued) i3 REACTOR PR01EC110N SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS E5 (f) The iPRMs shall be calibrated at least cnce per 1000 effective full power hours (EFPH) using the TIP system.
g (g) Calibrate Rosemount trip unit setpoint at least once per 31 days.
p (h) ~ Verify measured drive flow to be less than or equal to established drive flow at the existing flow control valve position.
y (i) This calibration shall consist of verifying the simulated thermal power time constant is within the limits specified in the COLR.
(j) This function is not required to be OPERABLE when the reactor pressure vessel head is removed i
per Specification 3.10.1.
(k) With any control rod withdrawn.
Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
(1) This function is not required to be OPERABLE when DRYWELL INTEGRITY is not required per Specifica-t.fon 3.10.1 y
(m) Verify the Turbine Bypass Valves are closed when THERMAL POWER is greater than or. equal to 40% RATED THERMAL POWER.
y (n) The CHANNEL FUNCTIONAL TEST and EHANNEL CALIBRATION shall include the turbine first stage pressure instruments..
~
e (o) - The CHANNEL CAllBRATION'shall exclude the flow reference transmitters; these transmitters shall tie calibrated at least once per 18 months, except that this test may be performed during the fifth refueling outage scheduled to begin April-16, 1994.
(p),This period may be extended to the completion of the fifth refueling outage scheduled to begin April 16, 1994.
(q) CHANNEL CALIpRATION may be performed during the fifth refueling outage scheduled to begin April N
16, 1994.
9 O
i f
L i
\\
i INSTRUMENTATION i
3!4.3.2 ISOLATION ACTVATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION l
3.3.2 The isolation actuation instrumentation channels shown in Table 3.3.2-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.2-2.
l APPLICABILITY:
As shown in Table 3.3.2-1.
ACTION:
a.
With an isolation actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.2-2, declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.
b.
With the number of OPERABLE channels less than required by the Minimum OPERABLE channels per Trip System requirement for one trip syste.m place the inoperable channel (s) and/or that trip system in the tripped condition
- within one hour.
c.
With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for both trip systems, place at least one trip system ** in the tripped condition within one hour and take the ACTION required by Table 3.3.2-1.
- An inoperable channel need not be placed in the tripped condition where this would cause the Trip Function to occur.
In these cases, the inoperable channel shall be restored to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or the ACTION required by Table 3.3.2-1 for that Trip Funciton shall be taken.
- The trip system need not be placed in the tripped condition if this would cause the Trip Function to occur. When a trip system can be placed in the tripped condition without causing the Trip Function to occur, place the trip system with the most inoperable channels in the tri,pped condition; if both systems have the same number of inoperable channels, place either trip system in the tripped condition.
RIVER BEND - UNIT 1 3/4 3-10 Amendment No. 47,73
. INSTRUMENTATION SURVEILLANCE RE0VIREMENTS 4.3.2.1 Each isolation actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS 'and at the frequencies shown in Table 4.3.2.1-1.**
4.3.2.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months.*
4.3.2.3 The ISOLATION SYSTEM RESPONSE TIME of each required isolation trip function shall be demonstrated to be within its limit at least once per 18 months.
Each test shall include at least one channel per trip system such that all channels are tested at least once every N times 18 months, where N is the total number of redundant channels in a specific isolation trip system.
. eds
+
- Logic System Functional Testing period may be extended as identified by note i
c on Table 4.3.2.1-1.
j
- Channel Calibration period may be extended as identified by note 'd' on Table 4.3.2.1-1.
I RIVER BEND - UNIT 1 3/4 3-11 Amendment No. 87 2,73 7
I
3 TABLE 3.3.2-1 h
ISOLATION ACTUATION INSTRUMENTATION to E
VALVE GROUPS MINIMUM APPLICABLE OPERATED BY OPERABLE CHANNELS OPERATIONAL TRIP FUNCTION SIGNAL *** PER TRIP SYSTEM (a)
CONDITION ACTION 1.
. PRIMARY CONTAINMENT ISOLATION
[
a.
Reactor Vessel Water Level-1,7,8,9ID)ICIII),
low Low, level 2 15, 16 2
1,2,3 20 b.
Drywell Pressure - High 1,3,8(b)(c)(j) 2 1,2,3 20 Containment Purge 8
1 1,2,3 21 c.
Isolation Radiation -
High 2.
MAIN STEAM LINE ISOLATION R
a.
Reactor Vessel Water Level-Low Low Low, level 1 6
2 1,2,3 20 Y
b.
Radiation - High 6,9(d) 2 1,2,3 23 c.
Main Steam Line Pressure - Low 6
2 1
24
.d.
Main Steam Line Flow - High 6
2/MSL 1,2,3 23
- e.
- Condenser Vacuum - Low 6
2 1, 2**, 3**
23 f.
Main Steam Line Tunnel Temperature - High 6
2 1,2,3 23 g.
Main Steam Line Tunnel g
A Temperature - High 6
2 1,2,3 23 S
h.
' Main Steam Line frea Temperature f
H:gh (Turbine Building) 6 2/ area 1, 2, 3 23 v.
m
~
~
s k
t l
\\
s l
i TABLE 3.3.2-2 (Continued) h
_ ISOLATION ACTUA110N INSTRUMENTATION SETPOINTS E
TRIP FUNCTION ALLOWABLE TRIP SETPOINT
[
VALUE 6.
RHR SYSTEM ISOLATION (Cont'd)
-4 e.
Reactor Vessel (RNR Cut-in w
Permissive) Pressure - High 5 135 psig i 150 psig f.
Drywell Pressure - High i 1.68 psig i 1.88 psig 7.
MANUAL INITIATION NA NA
{
- See Bases Ffgure B 3/4 3-1.
T U
i e
1 i
9 E
m i
0 4
i
\\
^
~~
4 4
i i
l Table 3.3.2-3 has been deleted.
I a
1 4
i 1
RIVER BEND - UNIT 1 3/4 3-24 Amendment No. M,13
'l Table 3.3.2-3 has been deleted.
RIVER BEND - UNIT 1 3/4 3-25 Amendment No. M,73 1
-~
~- -
TABLE 4.3.2.1-1
_ ISOLATION ACTUATION INSTRUNENTATION SURVEILLANCE REQUIREMENTS en CHANNEL OPERATIONAL E
CHANNEL FUNCTIONAL CilANNEL CONDITIONS IN WHICH TRIP FUNCTION CllECK TEST CALIBRATION SURVEILLANCE REQUIRED y
1.
PRIMARY CONTAINMENT ISOLATION a.
Low Low Level 2 S
M R(b) 1, 2, 3 l
br Drywell Pressure - High
'S M
R(b) 1, 2, 3 c.
Containment Purge Isalation Radiation - High S
M R
1,2,3 2.
MAIN STEAM LINE ISOLATION
~
a.
{
Low Low Low level 1 5
M R(b) 1, 2, 3 l
b.
Main Steam Line Radiation -
w 4,
High S
M R
1,2,3 l
o e
c.
Main Steam Line Pressure -
Low 5
M R(b) y d.
Main Steam Line Flow - High S
M R(b) 1, 2, 3
{
e.
Condenser Vacuum - Low 5
M R(b) 1, 2**. 3**
f.
Main Steam Line Tunnel Temperature - High S
M R
1, 2, 3 g.
Main Steam Line Tunnel a Temperature - High S
M R
1, 2, 3
[
h.
Main Steam Line Area S
M R(b) 1, 2, 3 g
Temperature-High 2
(Turbine Building) g e-
\\
4 JABLE 4.3.2.1-1 (Continued)
E ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS I
a, t
CHANNEL OPERATIONAL Q
CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WilCH o
TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE kEQUIRED 6.'
RHR SYSTEM ISOLATION 5
a.
RHR Equipment Area Ambient Temperature - High 5
M R
1,2,3 i
w b.
RHR Equipment Area A Temperature - High 5
M R
1, 2, 3 c.
Low Level 3 M
R(b) 1, 2, 3 d.
Low Low Low Level 1 S
M R(b) 1, 2, 3 l
e.
Reactor Vessel (RHR Cut-in Permissive) Pressure - High 5
M R
ICI(d) 1, 2, 3 g
f.
Drywell Pressure - High 5
M.-
R 1,2,3 y
7.
MANUAL INITIATION NA M
MA 1, 2' 3 4
i
- When handling irradiated fuel in the Fuel Building.
i
- When the reactor mode switch is in Run and/or any turbine stop valve is open.
(a) Each train or logic channel shall be tested at least every ether 31 days.
I I
(b) Calibrate trip unit setpoint at least once per 31 days.
(c) May be performed during the fif th refueling outage scheduled to begin April 16, 1994.
i (d) CHANNEL CALIBRATION may be perforwed durina the fifth refueling outaae scheduled to i
g begin April 16, 1994.
i 8
i gr w
- J l
l
.~.
i INSTRUMENTATION 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION llMITING CONDITION FOR OPERATION 3.3.3 The emergency core cooling system (ECCS) actuation instrumentation channels shown in Table 3.3.3-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.3-2.
APPLIC,BILITY: As shown in Table 3.3.3-1.
ACTION:
a.
With an ECCS actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.3-2, declare the channel ir. operable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value, b.
With one or more ECCS actuation instrumentation channels inoperable, take the ACTION required by Table 3.3.3-1.
i c.
With either ADS trip system "A" or "B" inoperable, restore the inoperable trip system to OPERABLE status:
1.
Within 7 days, provided that the HPCS and RCIC systems are OPERABLE, or 2.
Within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, provided either the HPCS or the RCIC system is inoperable.
Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reduce reactor steam dome pressure to less than or equal to 100 psig within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
SURVEl' LANCE REQUIREMENTS L
4.3.3.1 Each ECCS actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCIT0NAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.3.1-1.##
4.3.3.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all-channels shall be performed at least once per 18 months.##
I l
- Channel Calibration and Logic System Functional testing period may be extended as identified by note b on Table 4.3.3.1-1.
RIVER BEND - UNIT 1 3/4 3-30 Amendment No. 9 72,73 7
l l
l INSTRUMENTATION 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE RE0VIREMENTS 4.3.3.3 At least once per 18 months ##, the ECCS RESPONSE TIME of each required ECCS trip function shall be demonstrated to be within the limit.
l Each test shall include at least one channel per trip system such that all channels are tested at least once every N times 18 months ##, where N is the total number of redundant channels in a specific ECCS trip system.
l i
i l
l l
l arECCS Response time testing period may be extended to the completion of the j
fif th refueling outage scheduled to begin April 16, 1994.
RIVER BEND - UNIT 1 3/4 3-30a Amendment No. 9-74,73 i
. - ~
~
,1 t-TABLE 3.3.3-2 (Continued)
EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS 9
a, ALLOWABLE 5"
TRIP FUNCTION TRIP SETPOINT VALUE o
D.
LOSSOFPdWER (continued)
C 25 2.
Division III
-d 4.16 kv Standby Bus Undervoltage a.
4.16 kv Basis -
a.
6' (Sustained Undervoltage) 3045 1 153 volts 3045 1 214 volts b.
3 1 0.3 sec. time 3 1 0.33 sec. time delay delay b.
4.16 kv Standby Bus Undervoltage a.
4.16 kw Basis -
(Degraded Voltage) 3777 1 30 volts 3777 1 75 volts b.
60 1 6 sec. time 60 i 6.6 sec. time delay delay (w/o LOCA) c.
3 1 0.3 sec. time 3 1 0.33 sec. time delay
):
delay (w/LOCA)
Y M
- See Bases Figure 8 3/4 3-1.
- (Bottom of CST is at EL 95'1".)
The levels are measured from the instrument zero level of EL 98'6".
- (Bottom of suppression pool is at EL 70'.)
The levels are peasured from the instrument zero level of El 89'9".
l
- These are inverse time delay voltage relays or instantaneous voltage relays with a time delay.
The voltages shown are the maximum that will not result in a trip.
Lower voltage conditions will result in decreased trip times.
s
\\
\\
Table 3.3.3-3 has been deleted.
l l
[
t l
t
\\
G-RIVER BEND - UNIT 1 3/4 3-40 Amendment No M,73 f
. - -,..