ML17053B617
| ML17053B617 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 05/02/1980 |
| From: | Ippolito T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17053B619 | List: |
| References | |
| NUDOCS 8005200323 | |
| Download: ML17053B617 (70) | |
Text
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UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C..20555 NIAGARA MOHAWK POWER CORPORATION
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8 0 052 00pg P DOCKET NO. 50-220 NINE MILE POINT NUCLEAR STATION UNIT NO.
1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.
37
'icense No.
DPR-63 1.
The Nuclear Regulatory Commission (the COIImIission) has found that:
A.
The application for amendment by Niagara Mohawk Power Corporation (the licensee) dated February 15, 1979, as supplemented March 27 and April 3, 1979, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the 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)'hat such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the coIImIon defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No.
DPR-63 is hereby amended to read as follows:
(2) Technical S ecifications The Technical'pecifications contained in Appendices A and B, as revised through Amendment No.
37
, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
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3.
This license amendment fs effective as of the date of tts issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
Attachment:
Changes to the Technical Specifications Date of Issuance:
Mal 2. lgBO
/
Thomas
. Ippolito, Chief Operating Reactors Branch P3 Division of Operating Reactors
ATTACHMENT TO LICENSE AMENDMENT NO. 37 FACILITY OPERATING LICENSE NO. DPR-63 DOCKET NO. 50-220 Revise Appendix A by removing the following pages and replacing with revised identically numbered pages.
Marginal lines indicate area of change.
6 13 188 192-201 203-215 231 232 232a 233 237a Add page 212a
SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING Ci d.
e.
The neutron flux shall not exceed its scram setting for longer than 1.5 seconds as indicated by the process computer.
When the process computer is out of service, a safety limit violation shall be assumed if the neutron flux exceeds the scram setting and contro1 rod scram does not occur.
To ensure that the Safety Limit established in Specifications 2.1.1a and 2.1.1b is not
The Safety Limit shall be assumed to be exceeded when scram is accomplished by a means other than the expected scram signal.
Whenever the reactor is in the shutdown condition with irradiated fuel in the reactor
- vessel, the water level shall not be more than 7 feet ll inches (3.88 inches indicator scale) below minimum normal water level (Elevation 302'9"), except as specified "e" below.
For the purpose of performing major maintenance (not to exceed 12 -weeks in duration) on the reactor vessel; the reactor water level may be lowered 9'elow the minimum normal water level (Elevation 302'9").
Whenever the reactor water level is to be lowered below the low-low-low level set point redundant instrumentation will be provided to monitor the reactor water level; e.-
The reactor water low-low level setting for core spray initiation shall be no less than -5 feet (5 inches indicator scale) relative to the minimum normal water level (Elevation 302'9").
The flow biased APRM rod block trip settings shall be less than or equal to that shown in Figure 2.1.1.
d.
The reactor water low level scram trip setting shall be no lower than -12 inches (53 inches indicator scale) relative to the minimum normal water level (302'9").
Amendment No.
, 37
I
8ASES FOR 2; I.I FUEL CLADDING - SAFETY LIMIT During periods when the reactor is shut down, consideration must also. be given to water level requirements, due to the effect of decay heat. If reactor water level should drop below the top of the active fuel during this time, the ability to cool the core is reduced.
This reduction in core cooling capability could lead to elevated cladding temperatures and clad perforation.
The core will be cooled sufficiently to prevent clad melting should the water level be reduced to two-thirds of the core height.
The lowest point at which the water level can normally be monitored is approximately 4 feet 8 inches above the top of the active fuel.
This is the low-low-low water level trip point, which is 7 feet 11 inches (3.88 inches indicator scale) below minimum normal water level (Elevation 302'9").
The safety limit has been established here to provide a point which can be monitored and also can provide adequate margin.
However, for performing major maintenance as specified in Specification 2.1.1.e, redundant instrumentation will be provided for monitoring reactor water level below the low-low-low water level set point.
(For example, by installing temporary instrument lines and reference pots to redundant level transmitters, so that the reactor water level may be monitored over the required range.)
In addition written procedures, which identify all the valves which have the potential of
.lowering the water level inadvertently, are established to prevent their operation during the major maintenance which requires the water level to be below the low-low level set point.
The therma) power transient resulting when a scram is accomplished other than by the expected scram signal (e.g.,
scram from neutron flux following closure'f the main turbine stop valves) does not necessarily cause fuel damage.
However, for this specification a safety limit violation will be assumed when a scram is only accomplished by means of a backup feature of the plant design.
The concept of not approaching a safety limit provided scram signals are operable is supported by the extensive plant safety analysis.
Amendment No.
37 13
L I
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6.2 PROTECTIVE INSTRUMENTATION
~II 1i bill Applies to the operability of the plant instrumentation that performs a safety function.
~0b ective:
To assure the operability of the instrumentation required for safe operation.
e if'.
The set points, minimum number of trip
.systems, and minimum number of instrument channels that must be operable for each position of the reactor mode switch shall be as given in Tables 3.6.2a to 3.6.2k.
If the requirements of a table are not met, the actions'isted below for the respective type of instrumentation shall be taken.
4.6.2 PROTECTIVE INSTRUMENTATION A~li el i Applies to the surveillance of the in-strumentation that performs a safety function.
~0b ective:
To verify the operability of protective instrumentation.
a.
Sensors and instrument channels shall be checked.
tested and calibrated at least as frequently as listed in Tables 4.6.2a to 4.6.2k.
(1)
Instrumentation that initiates scram - control rods shall be
- inserted, unless -there is no fuel in the reactor vessel..
Amendment No.
37 188
l
Table 3.6.2a (cont'd)
INSTRUMENTATION THAT INITIATES SCRAM Limitin Condition for 0 eration Parameter Minimum No.
of Tripped or Operable
~Ti i Minimum No. of Operable Instrument Channels per Operable Tri S stem Set Point Reactor Mode Switch Position in Which Function Must Be 0 erable (6)
Main-Steam-I ine Isolation Valve Position (7)
High Radiation
.Hain-Steam-Line (8)
Shutdown Position of Reactor Mode Switch (9)
Neutron Flux (a)
IRM (i)
Upscale O(h) 3(d)
<10 percent valve closure from.full open
<5 times normal background at rated power
<96 percent of full scale
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C1.
Q S
Cll A5 CL C/l C/l (c)
(c)
X X
X X
(k)
X X
(g)
(g) (g)
Amendment No.
37 192
s
Table 3.6.2a (cont'd)
INSTRUMENTATION THAT INITIATES SCRAM Limiting Condition for Operation Parameter Minimum No.
of Tripped or Operable
~Ti S t Minimum No. of Operable Instrument Channels per Operable T~iS Set Point Reactor Mode Switch Position in Mhich Function Must Be 0 erable oo Cll (ii) Inoperative (b)
APRM (i )
Upscale 2
(ii)
Inoperative 2
(iii )
Downs cal e 2
(10) Turbine Stop Valve Closure (ll) Generator Load Rejection 3(d) 3(e) 3(e) 3(e)
Figure 2.1.1
>5 percent of full scale
<10% valve closure X
X X
X X
X X
X (g)
(g)
(g)
Amendment No.
37 193
I
Table 4.6.2a INSTRUMENTATION THAT INITIATES SCRAM Surveillance Re uirement Parameter (I)
Manual Scram (2)
High Reactor Pressure (3)
High Orywell Pressure (4)
Low Reactor Mater Level (5)
High Mater Level Scram Oischarge Volume (6)
Main-Steam-Line Isolation Valve Position (7)
High Radiation Main-Steam Line Sensor Check None None None Once/day None None Once/shift Instrument Channel Test Once per 3
months Once per month Once per month Once per month'1)
Once per month Once per 3 months Once per week Instrument Channel Calibration None Once pqr=3 months(l)
Once pyr 3 months<1)
Once pqr 3
months(1)
Once per 3 months None Once per 3 months Amendment No.
37 194
Table 4.6w2a (cont'd)
INSTRUMENTATION THAT INITIATES SCRAM Surveillance Re uirement Parameter (8) Shutdown Position of Reactor Mode Switch t
(9) Neutron F)ux (a)
(i)
Upscale (ii)
Inoperative Sensor.
Check None (f)
(f)
- Instrument Channel Test Once during each major refueling outage (f)
(f)
Instrument Channel Calibration None (f)
(f)
(b)
APRM (i)
Upscale (ii)
Inoperative (iii)
Downscale
{10) Turbine Stop Valve Closure t (ll) Generator Load Rejection None None Hone None None Once/week Once/week Once/week Once per 3 months Once per month Once/week Once/week Once/week None Once per 3
months 195 Amendment No.
37
33
NOTES FOR TABLES 3.6.2a AND 4.6.2a (a)
May be bypassed when necessary for containment inerting.
(b)
May be bypassed in the refuel and shutdown positions of the reactor mode switch with a keylock switch.
(c)
May be bypassed in the refuel and startup positions -of the'reactor mode switch when reactor pressure is less than 600 psi.
(d)
No more than one of the four IRM inputs to each trip system shall be bypassed.
(e)
No more than two C or D level LPRM inputs to an APRH shall be bypassed and only four LPRM inputs to an APRH shall be bypassed in order fot the APRM to be considered operable.
No more than one of the four APRM inputs to each trip system shall be bypassed provided that the APRM in the other instrument channel in the same core quadrant is not bypassed.
A Travel)ing In-Core Probe (TIP) chamber may be used as a substitute APRH input if the TIP is positioned in Close proximity to the failed LPRM it is replacing.
(f)
Calibrate prior to starting and normal shutdown and thereafter check once per shift and test once per week until no longer required.
(g)
IRM's are bypassed when APRH's are onscale.
APRM downscale is bypassed when IRM's are onscale.
(h)
Each of the four isolation valves has two limit switches.
Each limit switch provides input to one of two instrument channels in a single trip system.
(i)
Hay be bypassed when reactor power level is below 45K.
(j) Trip upon loss of oil pressure to the acceleration relay.
(k)
May be bypassed when placing the reactor mode switch in the SHUTDOMN position and all control rods are fully inserted.
(1)
Only the trip circuit will be calibrated and tested at the frequencies specified in Table 4.6.2a, the primary sensor will be calibrated and tested once per operating cycle.
196 Amendment No.
37
I
Table 3.6.2b INSTRUMENTATION THAT INITIATES PRIMARY C OLANT SYSTEM OR CONTAINMENT ISOLATION Limitin Condition for 0 eration Parameter Minimum No. of Minimum No.
Oper able Instrument of Tripped or Channels per Operable Operable T
S Set Point Reactor Mode Switch'osition in Which Function Must Be 0 erable PRIMARY COOLANT ISOLATION
~iiain Steam,
- Cleanup, and Shutdown)
(1)
Low-Low Reactor Water Level (2)
Manual
>5 inches (Indicator Scale)
O Cl
'0 2
W C
4 S-8 aCI CY CA C/l X
X X
X X
X HAIN-STEAM-LINE ISOL TION (3)
High Steam Flow Main-Steam Line
<105 psid X
X Amendment No.
197
1
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Table 3.6.2b (cont'd)
INSTRUMENTATION THAT INITIATES PRIMARY COOLANT SYSTEM OR CONTAINMENT ISOLATION Limitin Condition for 0 eration Par ameter Minimum No.
of Tripped or Operable Tri S stems Minimum No. of Operable Instrument Channels per Operable Tri S stem Set Point Reactor Mode Switch Position in Which Function Must Be 0 erable C
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4 C
8 S
CY ns cC t/l C/l (4)
High Radiation Hain-Steam Line (5)
Low Reactor Pressure
{6)
Low-Low-Low Condenser Vacuum (7)
High Temperature Main-Steam-Line Tunnel
<5 times normal back-ground at rated power
>850 psig
>7 in. mercury vacuum
<200F X
X (a)
X X
X Amendment No.
37 198
Table 3.6.2b (cont'd)
INSTRUMENTATION THAT INITIATES PRIMARY C L NT S STEM CONTAIN ENT SOLATION Limitin Condition for 0 eration Parameter Minimum No."
of Tripped or Operable Tri S stems Minimum No. of Operable Instrument Channels per Operable Tri S stem Set Point Reactor Mode Switch Position in Which Function Must Be 0 erable 3
e 0
S Ql l/l CL C
S cg l/)
CLEANUP SYSTEM ISOL TI N
(8).High Area Temperature
<190 X
X X
X SHUTDOWN COOLING SYSTEM ISOLATION (9)
High Area Temperature
<170 X
X X
X CONTAINMENT ISLT N
{10) Low-Low Reactor Mater
)5 inches
{c)
(Indicator Scale)
X X
Amendment No. P~ 37 199
/g
Table 3.6.2b (cont'd)
INSTRUMENTATION THAT INITIATES PRIMARY C LANT SYSTEH OR CONTAINMENT ISOLATION Limitin Condition for 0 eration Parameter Minimum No.
of Tripped or Operable Tri S stems Hinimum No. of Operable Instrument Channels per Operable Tri S stem Set Point Reactor Mode Switch Position in Which
-Function Must Be 0 erable e
ap LL O0 4
CP Sl5 CY L9
{11) High Drywell Pressure (12) Manual
<3.5 psig (c)
(b)
(b)
X X
X X
200 Amendment No.
37
Table 4.6.2b INSTRUMENTATION THAT INITIATES PRIMARY COOLANT SYSTEM OR CONTAINMENT ISOLATION Surveillance Re uirement Parameter PRIHARY COOLANT ISOLATION
~Main Steam, Cleanup and Shutdown)
(1)
Low-Low Reactor Water Level (2)
Manual Sensor Check Once/day Instrument Channel Test (d)
Once pN'onth' Once during each major refueling outage Instrument Channel Calibration Once per 3 months (d)
HAIN-STEAM-LINE ISOLATION (3)
High Steam Flow Hain-Steam Line (4)
High Radiation Main-Steam Line (S)
Low Reactor Pressure Once/day Once/shift Once/day Once per month (d)
Once/week Once per month (d)
Once per 3 months (d)
Once per 3 months Once per 3 months (d)
Amendment No.
37 201
I
Table 4.6.2b {cont'd)
INSTROMENTATION THAT INITIATES PRIMARY COOLANT SYSTEM OR CONTAINMENT ISOLATION Surveillance Re uirement Parameter CONTAINMENT ISOLATION
{10) Low-Low Reactor lfater Level (11) High Drywell Pressure (12) Manual Sensor Check Once/day Once/day Instrument Channel Test Once per month Once per month Once during each operating cycle Instr ument Channel Calibration Once per 3 months (d)
Once per 3 months (d)
Amendment No.
37 203
NOTES FOR TABLES 3.6.2b AND 4.6.2b (a)
May be bypassed in the refuel and startup positions of the reactor mode switch when reactor pressure is less than 600 psi.
(b)
May be bypassed when necessary for containment inerting.
(c)
May be bypassed in the shutdown mode whenever the reactor coolant system temperature is less than 215 F.
(d)
Only the trip circuit will be calibrated and tested at the frequencies specified in Table 4.6.2b, the primary sensor will be calibrated and tested once per operating cycle.
Amendment No. fg, 37 204
Table 3.6;2c-INSTRUMENTATION THAT INITIATES OR ISOLATES EMERGENCY COOLING Limitin Condition for 0 eration Parameter Minimum No.
of Tripped or Operable Tri S stems Minimum No. of Operable Instrument Channels per Operable Tri S stem Set-Point Reactor Mode Switch Position in Which Function Must Be 0 erable EMERGENCY COOLING INITIATION CL O
Cl D
et-5 03 stl CY C/l f/l (1)
High-Reactor Pressure
<1080 psig
. (b)
X X
(2)
-Low-Low Reactor Mater Level EMERGENCY COOLING ISOLATION
~for each of two systems)
(3)
High Steam Flow Emergency Cooling System (4)
High Radiation Emergency Cooling System Vent 2 (a)
> 5 inches (Indicator Scale)
(b) 19 psid 25 mr/hr X
X X
X X
X amendment No.g Jg, 37 205
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Table 4.6.2c INSTRUMENTATION THAT INITIATES OR ISOLATES EMERGENCY COOLING
- Surveillance Re uir ement Parameter EMERGENCY COOLING INITIATION
~1 High Reactor Pressure (2)
Low-Low Reactor Water Level Sensor Check None Once/day Instrument Channel Test Once per month Once per month Instrument Channel Calibration Once per 3 months Once per 3 months EMERGENCY COOLING ISOLATION
~for each of teo systems)
(3)
High Steam Flow Emergency Cooling
'ystem (4)
High Radiation Emergency Cooling System Vent None Once/shi ft Once per 3 months Once during each major refueling outage Once per 3 months Once during each major refueling outage Amendment No.
37 206
r
NOTES FOR TABLES 3.6.2c AND 4.6.2c (a)
Each of two differential pressure switches provide inputs to one instrument channel in each trip system.
(b)
May be bypassed in the cold shutdown condition.
(c)
Only the trip-circuit will be calibrated.and tested at the frequencies specified in Table 4.6.2c, the primary sensor will be calibrated and tested once per operatihg cycle.
Amendment No. J4, 37 207
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Table 3.6.2d INSTRUMENTATION THAT INITIATES CORE SPRAY 0
Limitin Condition for 0 eration Parameter Minimum No. of Minimum No.
Operable Instrument of Tripped or Channels per Operable Operable
~li I Ti S t Set-Point Reactor Mode Switch Position in lAich Function Must Be 0 erable OO C/)
START CORE SPRAY PUMPS (I)
High Drywel 1 Pressure (2)
Low-Low Reactor Mater Level OPEN CORE SPRAY DISCHARGE VALVES (3)
Reactor Pressure and either (I) or (2) above.
<3.5 psig
>5 inches (Indicator Scale)
>365 psig
{d) x (a)
{a)
{b)
X X
X X
X X
X Amendment No. 7, 7g, 37 208
4
Table 4.6.2d INSTRUMENTATION THAT INITIATES CORE SPRAY Surveillance Re uirement Parameter START CORE SPRAY PUMPS (I)
High Drywel1 Pressure (2)
Low-Low Reactor Water Level OPEN CORE SPRAY DISCHARGE VALVES (3)
Reactor Pressure and either (I) or (2) above Sensor Check Once/day Once/day None Instrument Channel Test
/,r' Once per month Once per month Once per month Instr ument Channel Calibration Once per 3 months (c)
Once per 3 months (c)
Once per 3 months (c)
Amendment No.
37 209
NOTES FOR TABLES 3.6.2d ANO 4.6.2d (a)
May be bypassed when necessary for containment inerting.
(b)
May be'bypassed when necessary for performing major maintenance as specified in Specification 2.1.1.e.
(c)
Only the trip circuit will be calibrated'and tested at the frequencies specified in Table 4.6.2d, the primary sensor will be calibrated and tested once per operating cycle.
(d)
May be bypassed when necessary for integrated leak rate testing.
(e)
The instrumentation that initiates the Core Spray System is not required to be operable, if there is no fuel in the reactor vessel.
210 Amendment No. 7g, 37
Table 3.6.2e INSTRUMENTATION THAT INITIATES CONTAINMENT SPRAY Limitin Condition for 0 eration Parameter Minimum No.
of Tripped or Operable Tri S stems Minimum No. of Operabl e Ins trument Channels per Operable Tri S stem Set-Point Reactor Mode Switch Position in Which~
Function Must Be ~
0 erable c
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n.
a dl C)
D W
IY U
9-S-
CJ C5 CY'/)
(1) a.
High Drywell Pressure
< 3.5 psig (a)
X X
and b.
Low-Low Reactor Water Level
> 5 inches (Indicator Scale)
(a)
X X
Amendment No.
37 211
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~
Table 4.6.2e INSTRUMENTATION THAT INITIATES CONTAINMENT SPRAY Surveillance Re uirement Parameter Sensor Check Instrument Channel Test Instrument Channel Calibration (l)a. High Drywell Pressure
.b. Low-Low Reactor Water Level Once/day Once/day Once per month Once per 3 months Once per month Once per 3 months Amendment No.
37 212
ll
+
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NOTES FOR TABLES 3.6.2e AND 4.6.2e (a)
Hay be bypassed in the shutdown mode whenever the reactor coolant temperature is less than 215 F.
(b)
Only the trip circuit will be calibrated and tested at the frequencies specified in Table 4.6.2e,
'he primary sensor will be calibrated and tested once per operating cycle.
Amendment No.
212a
o C
aii Table 3.6.2f INSTRUMENTATION THAT INITIATES AUTO DEPRESSURIZATION Limitin Condition for 0 eration Parameter Minimum No.
of Tripped or Operable Tri S stems Minimum No. of Operable Instrument
~
Channels per Operable Tri S stem Set-Point Reactor Node Switch Position in which Function Rust Be 0 erable a
0 S
0 4-S-
OP 45 CL CX CA CA INITIATION (1) a.
Low-Low-Low Reactor Water Level 2 (a) 2 (a)
> 3 88 inches (Indicator scale)
(b) and b.
High Drywell Pressure 2 (a) 2 (a)
< 3.5 psig (b)
(b) x Amendment No.
37 213
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Table 4.6.2f INSTRUMENTATION THAT INITIATES AUTO OEPRESSURIZATION Surveillance Re uirement Parameter Sensor Check Instrument Channel Test Instrument Channel Calibration INITIATION (1) a.
Low-Low-Low Reactor Mater and b.
High Drywell Pressure None Once/day Once per month(
)
Once per month Once per 3 months (c)
Once per 3 months Amendment No.
37 214
l
NOTES FOR TABLES 3.6.2f AND 4.6.2f (a)
Both instrument channels in either trip system are required to be energized to initiate auto depressurization.
One trip system is powered from power board 102 and the other trip system from power board 103.
(b)'ay be bypassed when the reactor pressure is less than 110.psig and the reactor coolant temperature is less than the corresponding saturation temperature.
(c)
Only the trip circuit will be calibrated and tested at the frequencies specified in Table 4.6.2f, the primary sensor will be calibrated and tested once per operating cycle.
Amendment No.
37 215
C A
Table 3.6.2k HIGH PRESSURE COOLANT INJECTION Limitin Condition for 0 eration Parameter Minimum No.
of Tripped or Operable Tri S stems Minimum No. of Operable Instrument Channels per Operable Tri S stem Set-Point Reactor Mode Switch Position in Which Function Must Be 0 erable
~
CL O
CCJ D
4 S-Cll C/l C/l (1)
Low Reactor Water Level (2)
Automatic Turbine Trip
> 53 inches (Indicator scale)
(a)
(a)
(a)
X (a)
X Amendment No
, 37 231
Table 4.6.2k HIGH PRESSURE COOLANT INOECTION Surveillance Re uirement Parameter (1)
Low Reactor Water Level Sensor Check Once per day Instrument Channel Test Once per month (b)
Instrument Channel Calibration Once per 3.months (b)
(2)
Automatic Turbine Trip None Once during each operating cycle None Amendment No.
37 232
C I
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NOTES FOR TABLES 3.6.2k AND 4.6.2k (a')
Hay be bypassed when the reactor pressure is less than 110 psig and the reactor coolant temperature is less than the corresponding saturation temperature.
(b)
Only the trip circuit will be calibrated and tested at the frequencies specified-in Table 4.6.2k, the primary sensor will be calibrated and tested once per operating cycle.
Amendment No. Jg, 37 232a
f
BASES FOR 3.6.2 AND 4.6.2 PROTECTIVE INSTRUMENTATION The reactor protection system automatically initiates a reactor scram to prevent exceeding established limits.
In addition; other protective instrumentation is provided to initiate action which mitigates the consequences of accidents or terminates operator error.
The reactor protection system is a dual channel type (Table 3.6.2.a).
Each trip system except the manual scram has two independent instrument channels.
Operation of either channel will trip the trip system, i.e.,
the trip logic of the channel is one-out-of-two.
A simultaneous trip of both trip systems will cause a
reactor scram, i.e., the tripping logic of the trip systems is two-out-of-two.
The tripping logic of the total system is referred to as one-out-of-two taken twice.
This system will accommodate any single failure and still perform its intended function and in addition, provide protection against spurious scrams.
The reliability of the dual channel system or probability that it will perform its intended function is less than that of a one-out-of-two system and somewhat greater than that of a two-out-of-three system (Section VIII-A.1.0 of the FSAR).
The instrumentation used to initiate action other than scram is generally similar to the reactor protection system.
There are usually two trip systems required or available for each function.
There are usually two instrument channels for each trip system.
Either channel can trip the trip system but both trip systems are required to initiate the respective action.
Where only one trip system is provided only one instrument channel is required to trip the trip system.
All instrument channels except those for automatic depressurization are normally energized.
De-energizing causes a trip.
Power to the trip systems for each function is from reactor protection system buses ll and 12.
The signals for initiating automatic blowdown and rod block differ from other initiating signals in that only one of the two trip systems is required to start blowdown or initiate rod block.
Both instrument channels in the trip system must trip to initiate automatic blowdown.
This difference is due to the requirement that automatic depressurization be prevented unless A.C. power is available to the emergency core cooling systems.
The instrument channels in the trip system for automatic depressurization are normally de-energized.
In order to cause a trip both instrument channels must be energized.
Power to energize the instrument channels is from power boards 102 and 103. If A.C. power is lost to one power board, one trip system becomes inoperable Amendment No.
233
Lr
BASES FOR 3.6.3 AND 4.6.2 PROTECTIVE INSTRUMENTATION b.
The control rod block functions are provided to prevent excessive control rod withdrawal so that HCPR is maintained greater than 1.06.
The trip logic for this function is I out of n; e.g.,
any trip on one of the eight APRH's, eight IRM's or four SRM's will result in a rod block.
The minimum instrument channel requirements provide sufficient instrumentation to assure the single failure criteria is met.
The minimum instrument-channel requirements for the rod block may be reduced by one for a shor t period of time to allow maintenance,
- testing, or calibration.
This time period is only UX of the operating time in a month and does not significantly increase the risk of preventing an inadvertent control rod withdrawal.
The APRM rod block trip is flow biased and prevents a significant reduction in HCPR especially during operation at reduced flow.
The APRH provides gross core protection; i.e., limits the gross core power increase from withdrawal of control rods in the normal withdrawal sequence.
The trips are set so that MCPR is maintained greater than 1.06.
The APRH rod block also provides local protection of the core; i.e., the prevention of critical heat flux in a local region of the core, for a single rod withdrawal error from a limiting control rod pattern.
The trip point is flow biased.
The worst case single control rod withdrawal error has been analyzed and the results show that with the specified trip settings rod with-drawal is blocked before the HCPR reaches 1.06, thus allowing adequate margin.
Below %OX power the worst case withdrawal of a single control rod results in a MCPR > 1.06 without rod block action, thus below this level it is not required.
The IRM rod block function provides local as well as gross core protection.
The scaling arrange-ment is such that trip setting is less than a factor of 10 above the indicated level.
Analysis of the worst case accident results in rod block action before HCPR approaches 1.06.
A downscale indication on an APRM or IRM is an indication the instrument has failed or the instrument is not sensitive enough.
In either case the instrument will not respond to changes in control rod motion and the control rod motion is prevented.
The downscale rod blocks are set at 5 percent of full scale for-IRM and 2 percent of full scale for APRH (APRH signal is generated by averaging the output signals from eight LPRM flux monitors).
Amendment No. 5, g 237a
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