ML19294C266
| ML19294C266 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 02/28/1980 |
| From: | GEORGIA POWER CO. |
| To: | |
| Shared Package | |
| ML19294C265 | List: |
| References | |
| NUDOCS 8003070394 | |
| Download: ML19294C266 (10) | |
Text
TABLE 2.2.1-1
~
g REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS A
=
ALLOWABLE g
FUNCTIONAL UNIT TRIP SETPOINT VALUES M
1.
Intermediate Range Monitor, Neutron Flux-High -< 120/125 divisions of full scale -< 120/725 divisiens y
(2C51-K601 A,B,C,D,E,F,G,H) of full scale 2.
Average Power Range Monitor:
(2C51-K605 A,B,C,D,E,F)
Neutron Flux-High,15%(a)
-< 15/125 divisions of full scale
-< 20/125 divisions of a.
full scale b.
Flow Reference, Simulated
< (0.66 W + 51%) (113.5% fiax)
< (0.66 W + 54%)
Thermal Power T115.5% Max) c.
Neutron Flux - Upscale (Run Mode)
< 118%
<120%
3.
Reactor Vessel Steam Dome Pressure - High
-< 1045 psig
-< 1045 psig (2B21-f4023 A,B,C,D) 4.
Reactor Vessel Water Level - Low
>12.5lc)hesaboveinstrument
>12.5inchesabov{c) nc (2B21-N017 A,B,C,D) zero
_ instrument zero
[
5.
Main Steam Line Isolation Valve - Closure
-< 10% closed
< 10% closed (NA) 6.
Main Steam Line Radiation - High
-< 3 x full power background
-< 3 x full power (2Dll-K603A,B,C,D) background 7.
Drywell Pressure - High
< 2 psig
< 2 psig (2C71-N002A,B,C,D) 8.
Scram Discharge Volume Water Level - High
-< 57.15 gallons
< 57.15 callons O
(2Cll-N013A,B,C,D) 9.
Turbine Stop Valve - Closure
< 10% closed
< 10% closed O
(NA) d
- 10. Turbine Control Valve O
Fast Closure, Trip Oil Pressure-Low 1 600 psig 1 600 psig J
(2C71-N005A,B,C,D)
__D
.4
TABLE 2.2.1-1 (Continued)
REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS TABLE NOTATION (a) This Average Power Range Monitor scram function is a fixed point and is increased when the reactor mode switch is in the Run position.
(b) The Average Power Range Monitor ' flow referenced scram function is varied as a function of recirculation loop flow (W). The trip setting of this function must be maintained in accordance with Specification 3.2.2.
(c) See Bases Figure B 3/4 3-1.
HATCH - UNIT 2 2-6
2.2 LIMITING SAFETY SYSTEf1 SETTINGS _
4 Q
c BASES 2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS The Reactor Protection System Instrumentation Setpoints specified in Table 2.2.1-1 are the values at which the reactor trips are set for each parameter. The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their Safety Limits. Operation with a trip set less conservative than its Trip Setpoint, but within its specified Allowable Value, is acceptable nn the basis that each Allowable Value is equal to or less than the drif t allowance assumed for each trip in the safety analyses, l.
Intermediate Range Monitor, Neutron Flux The IRM system consists of 8 chambers, 4 in each of the eactor trip systems. The IRM is a 5 decade 10 range instrument.
The trip setpoint of 120 divisions of scale is active in each of the 10 ranges.
Thus, as the IRM is ranged up to accommodate the increase in power level, the trip setpoint is also ranged up. The IRM instruments provide for overlap with both the APRM and SRM systems.
The most significant source of reactivity changes during the power increase are due to control rod withdrawal.
In order to ensure that the IRM provides the required protection, a range of rod withdrawal accidents have been analyzed, Section 7.5 of the FSAR.
The most severe case involves an initial condition in which the reactor is just subcritical and the IRM's are not yet on scale. Additional conservatism was taken in this analysis by assuming the IRM channel closest to the rod being withdrawn is bypassed.
The results of this analysis show that the reactor is shutdown and peak power is limited to 1%
of RATED THERMAL POWER, thus maintaining MCPR above 1.06.
Based on this analysis, the IRM provides protection against local control rod errors and continuous withdrawal of control rods in sequence and provides backup protection for the APRM in the low power (<15%) range.
The design instrument drift allowance has been chosen to enable the trip setpoint to remain within the allowable value over the priod between calibration tests of the instrument electronics alone, and is based on cumulative field experience derived from similar applications and environmental conditions.
2.
Average Power Range Monitor For operation at low pressure and low flow during STARTUP, the APRM scram setting of 15/125 divisions of full scale neutron flux provides adequate thermal margin between the setpoint and the Safety Limits. The margin accommodates the anticipated maneuvers associated with power plant startup.
Effects of increasing pressure at zero or low void content are minor, cold water from sources available during startup, is not much colder than that already in the system, temperature coefficients are small and control rod patterns are constrained by the RSCS and RWM.
HATCH - UNIT 2 B 2-9
,, q g
2.2 LIMITING SAFETY SYSTEM SETTINGS BASES (Continued)
REACTOR PROTECTION SYSTEii INSTRUMENTATION SETP0INTS (Continued)
Average Power Range Monitor (Continued)
Of all the possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power increase.
Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks and because several rods must be moved to change power by a significant amount, the rate of power rise is very slow. Generally the heat flux is in near equilibrium with the fission rate.
In an assumed uniform rod withdrawal approach to the trip level the rate of power rise is not more than 5% of RATED THERMAL POWER per minute and the APRM system would be more than adequate to assure shut-down before the power could exceed the Safety Limit.
The 15% neutron flux trip remains active until the mode switch is placed in the RUN position.
The APRM flux scram trip in the run mode consists of a flow referenced scram setpoint and a fixed high neutron flux scram setpoint. The APRM flow referenced neutron flux signal is passed through a filtering network with a time constant which is representative of the fuel dynamics. This provides a flow referenced signal that approximates the average heat flux cr thermal power that is developed in the core during transient or steady-state conditions.
This prevents spurious scrams, which have an adverse effect on react.r safety because of the resulting thermal stresses.
Examples of events which can result in momentary neutron flux spikes are momentary flow changes in the recircu-lation system flow, and small pressure disturbances during turbine stop valve and turbine control valve testing. These flux spikes represent no hazard to the fuel since they are only of a few seconds duration and less than 118% of rated thermal power.
The APRM flow referenced scram trip setting at full recirculation flow is adjustable up to 113.5% of rated power. This reduced flow referenced trip setpoint will result in an earlier scram during slow thermal transients, such as the loss of 1000F feedwater heating event, than would result with the 118% fixed high neutron flux scram trip.
The lower flow referenced scram setpoint therefore decreases the severity (ACPR) of a slow thermal transient and allows lower Operating Limits if such a transient is the limiting abnormal operational transient during a certain exposure interval in the cycle.
The APRM fixed high neutron flux signal does not incorporate the time constant, but responds directly to instantaneous neutron flux.
This scram setpoint scrams the reactor during fast power increase transients if credit is not taken for a direct (position) scram, and also serves to scram the reactor if credit is not taken for the flow referenced scram.
HATCH - UNIT 2 B 2-10
2.2 LIMITINU SAFETY SYSTEM SETTINGS BASES (Continued)
REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)
Average Power Range Monitor (Continued)
The differential between the allowable value and the trip setpoint is equal to or greater than the design drif t' allowance. The drift value is only related to the electronic portion of the measurement, in particular to the reference setting signal, and is based on the design specification supported by cumulative field experience derived from virtually identical applications and er.vironmental conditions.
3.
Reactor Vessel Steam Dome Pressure-High High pressure in the nuclear system could cause a rupture to the nuclear system process barrier resulting in the release of fission products. A pressure increase while operating will also tend to in-crease the power of the reactor by compressing voids thus adding reactivity. The trip will quickly reduce the neutron flux, counter-acting the pressure increase by decreasing heat generation. The trip setting is slightly higher than the operating pressure to permit normal operation without spurious trips. The setting provides for a wide margin HATCH - UNIT 2 B 2-10a
POWER DISTRIBUTION LIMITS 3/4.2.2 APRM SETPOINTS jQ 0 9)
LIMITING CONDITION FOR OPERATION 3.2.2 The flow biased APRM scram trip setpoint (S) and rod block trip setpoint(SRB) shall be established according to the following relationships:
S < (0.66W + 51%) T SRB 5 (0.66W + 42%) T where: S and SRB are in percent of RATED THERMAL POWER, W = Loop recirculation flow in percent of rated flow, T = Lowest value of the ratio of design TPF divided by the MTPF obtained for any class of fuel in the core (T < l.C), and Design TPF for: 8 x 8 fuel = 2.48.
APPLICABILITY:
CONDITION 1, when THERMAL POWER > 25% of RATED THERMAL POWER.
ACTION:
With S or SRB exceeding the allowable value, initiate corrective action within 15 minutes and continue corrective action so that S and SRB are within the required limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.2.2 The MTPF for each clast of fuel shall be determined, the value of T calculated, and the flow referenced neutron trip setpoint adjusted, as required:
a.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.
Whenever THERMAL POWER has been increased by at least 15% of RATED THERMAL POWER and steady state operating conditons have been established, and c.
Initially 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 /> when the reactor is operating with a MTPF > 2.48.
HATCH - UNIT 2 3/4 2-5
TABLE 3.3.1.-l REACTOR PROTECTION SYSTEM INSTRUMENTATION APPLICABLE MINIMUM NUMBER z4 OPERATIONAL OPERABLE CHANNEL <
9 FUNCTIONAL UNIT C0NDITIONS PER TRIP SYSTEM (a)
ACTION g
1.
5 (2C51-K601 A, B, C, D, E, F, G, H)
- a. Neutron Flux - High 2(c),5(b) 3 1
2 2
3,4(b) 2, 5 3
1
- b. Inoperative 3, 4 2
2 2.
Average Power Range Monitor:
(2C51-K605 A, B, C, D, E, F)
- a. Neutron Flux - High,15%
2, 5 2
1
- b. Flow Refererence STP-Upscale
- c. Upscale (Run Mode) 1 2
3 a3
- d. Inoperative 1,2,5 2
4
- e. Downscale 1
2 3
[
- f. LPRM 1,2,5 (d)
NA 3.
Reactor Vessel Steam Dome Pressure -
High (2B21-N023 A, B, C, D) 1,2(e) 2(J, B2 N0 5-
)5 4.
2(j,2821-N024-5 Low (2B21-N017 A, B, C, D) 1, 2 A, B and 2B21-N025-A,B) 5.
Main Steam Line Isolation Valve -
Closure (NA) 1(f) 4 3
6.
Main Steam Line Radiation - High 1,2(*)
2 6
(2Dil-K603 A, B, C, D) 7.
Drywell Pressure - High 1,2(9) 2 5
(2C71-N002 A, B, C, D)
TABLE 3.3.1-2 REACTOR PROTECTION SYSTEM RESPONSE TIMES 5
FUNCTIONAL UNIT RESPONSE TIME y
(Seconds) l.
E a.
Neutron Flux - High*
NA U
b.
Inoperative NA m
2.
Average Power Range Monitor:*
a.
Neutron Flux - Upscale (Setdown) 1 0.09 b.
Flow Reference STP-Upscale
<0.09**
c.
Upscale (Run Mode) 7 0.09 d.
Inoperative iTA e.
Downscale NA f.
LPRM NA 3.
Reactor Vessel Steam Dome Pressure - High 1 0.55 4.
Reactor Vessel Water Level - Low 1 1.05 5.
Maic Steam Line Isolation Valve - Closure 1 0.06 6.
Main Steam Line Radiation - High NA 7.
Drywell Pressure - High NA 8.
Scram Discharge Volume Water Level - High NA 9.
Turbine Stop Valve - Closure 1 0.06
- 10. Turbine Control Valve Fast Closure, Trip Oil Pressure - Low 1 0.08 11.
Reactor Mode Switch in Shutdown Position NA
- 12. Manual Scram NA
- Neutron detectors are exempt from response time testing.
Response time shall be measured from detector output or input of first electronic component in channel.
TABLE 4.3.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL I
CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH FUNCTIONAL UNIT CHECK TEST CALIBRATION (a)
SURVEILLANCE REQUIRED r
1.
i q
a.
Neutron Flux - High D
S/U(b)(c)
R 2
D W
R 3,4,5 5
~ NA W
NA 2,3,4,5 b.
Inoperative m
2.
Average Power Range Monitor:
a.
Neutror Flux - Upscale (Setdown)
S S/U(b)(c), g(d) S/U(b),g(d) 2 S
W W
5 b.
Flow Reference SPT - Upscale S
S/U(b),y g(e)(f), Sg
)
c.
Upscale (Run Mode)
S W
W l
d.
Inoperative NA W
NA 1,2,5 e.
Downscale NA W
NA 1
f.
LPRM D
NA (g) 1, 2, 5
~
E 3.
Reactor Vessel Steam Dome
[
Pressure - High NA M
Q 1, 2 4.
Low D
M Q
1, 2 5.
Main Steam Line Isolation Valve -
Closure NA M
R(h) 1 6.
Main Steam Line Radiation - High D
W(i)
R( }
1, 2 7.
Drywell Pressure - High NA M
Q 1, 2 8.
Scram Discharge Volume Water Level - High NA M
R(h) 1, 2, 5
TABLE 4.3.1-1 (Continued)
REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH
,_!j FUNCTIONAL UNIT CHECK TEST CALIBRATION SURVEILLANCE REQUIRED o
7 9.
Turbine Stop Valve - Closure NA M
R(h) 1
- 10. Turbine Control Valve Fast Closure, Trip Oil Pressure -
-4 N
Low NA M
R 1
11.
Reactor Mode Switch in Shutdown NA R
NA 1,2,3,4,5 Posi tion
- 12. Manual Scram NA M
NA 1,2,3,4,5 a.
Neutron detectors may be excluded from CHANNEL CALIBRATION.
b.
Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prios to startup, if not performed within the previous 7 days.,
{
y c.
The APRM, IRM and SRM channels shall be compared for overlap during each startup, if not performed within the previous 7 days.
m d.
When changing from CONDITION 1 to CONDITION 2, perform the required surveillance within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering CONDITION 2.
e.
This calibration shall consist of the adjustment of the APRM channel to conform to the power values calculated by a heat balance during CONDITION 1 when THERMAL POWER > 25% of RATED THERMAL POWER.
Adjust the APRM channel if the absolute difference > 2%. Any APRM channel gain adjustment made in compliance with Specification 3.2.2 s:1all not be included in determining the absclute difference.
f.
This calibration shall consist of the adjustment of the APRM flow reference channel to conform to a calibrated flow signal.
g.
The LPRM's shall be calibrated at least once per 1000 effective full power hours (EFPH) using the TIP system.
h.
Physical inspection and actuation of switches.
i.
Instrument alignment using a standard current source.
j.
Ca'iibration using a standard radiation source.
,