ML19312C899
| ML19312C899 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 08/26/1975 |
| From: | DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML19312C895 | List: |
| References | |
| NUDOCS 8001130072 | |
| Download: ML19312C899 (11) | |
Text
_
c.3 iL,F.F.- c M
t$gulator) Dcchet E.ilg
&L 3
M.
Power peaking is not a directly observable quantity and therefore limits have been established on the bases of the reactor power imbalance produced by the power peaking.
The specified flow rates for Curves 1, 2, 3, and 4 of Figure 2.1-2B correspond 2.1-2C to the expected minimum flow rates with four pumps, three pumps, one pump in each loop and two pumps in one loop, respectively.
The curve of Figure 2.1-1B is the most restrictive of all possible reactor 2.1-lC coolant pump-maximum thermal power combinations shown in Figure 2.1-3B.
2.1-3C The curves of Figure 2.1-3B represent the conditions at which a minimum DNBR
., v-2.1-3C of 1.3 is predicted at the maximum possible thermal power for the number of reactor coolant pumps in operation or the local quality at the point of minimum DNBR is equal to 15%,(3) whichever condition is more restrictive.
Using a local quality limit of 15 percent at the point of minimum DNBR as a basis for Curves 2 and 4 of Figure 2.1-3B is a conservative criterion even 2.1-3C though the quality of the exit is higher than the quality at the point of minimum DNBR.
The DNBR as calculated by the W-3 correlation continually increases from point of minimum DNBR, so that the exit DNBR is 1.7 or higher, depending on the pressure.
Extrapolation of the W-3 correlation beyond its published quality range of +15 percent is justified on the basis of experimental data.(4)
The maximum thermal power for three pump operation is 86.5% - Unit 2 86.5% - Unit 3 due to a power level trip produced by the flux-flow ratio 75% flow x 1.07.= 80%
1.07 = 80%
power plus the maximum calibration and instrument error.
The maximum thermal power for other coolant pump conditions are produced in a similar manner.
A flux-flow l ratio of 0.961 is used for single loop conditions.
I For each curve of Figure 2.1-3B, a pressure-temperature point above and to the 2.1-3C lef t of the curve would result in a DNBR greater than 1.3 or a local quality at the point of minimum DNBR less than 15 percent for that particular reactor coolant pump situation.
The 1.3 DNBR curve for four-pump operation is more restrictive than any other reactor coolant pump situation because any pressure /
temperature point above and to the lef t of the four-pump curve will be above and to the lef t of the other curves.
REFERENCES (1) FSAR, Section 3.2.3.1.1 (2) FSAR, Section 3.2.3.1.1.c (3) FSAR, Section 3.2.3.1.1.k 8001130 07A 2.1-3b
Thermal Power Level, %
-- 120 kw/ft Limit kw/ft Limit
. 100 80 DNBR Limit
.. 60 O
DNBR Limit 40
.. 20 I
I I
I I
1
-60
-40
-20 0
+20
+40
+60 Reactor Power Imbalance, %
CURVE REACTOR COOLANT FLOW (LB/HR) 6 1
131.3 x 10 2
98.1 x 106 3
64.4 x 106 4
60.1 x 10 PROTECTIVE SYSTEM MAXIMUM ALLOWABLE SETPOINTS UNIT 2 f OCONEE NUCLEAR STATIO i
Figure 2.3-2B 2.1-8
THERMAL POWER LEVEL, %
. 120 kw/ f t Limi t
\\\\,
_ kw/ft Limit
-- 100
- 80 DNBR Limit O
. 60 DNBR Limit DNBR Limit 40 20
-60
-40
-20 0
+20
+40
+60 Reactor Power Imbalance, %
CURVE REACTOR COOLANT FLOW (LB/HR) 6 1
131.3 x 106 2
98.1 x 106 3
64.4 x 106 4
60.1 x 10 CORE PROTECTION SAFETY LIMITS UNIT 3 Ot OCONEE NUCLEAR STATION i
i Figure 2.1-2C l
l 2.1-9
2.3 LIMITING SAFETY SYSTEM SETTINGS, PROTECTIVE INSTRUMENTATION Applicability Applies to instruments monitoring reactor power, reactor power imbalance, reactor coolant system pressure, reactor coolant outlet temperature, flow, number of pumps in operation, and high reactor building pressure.
Objective To provide c stomatic protective action to prevent any combination of process variables from exceeding a safety limit.
Specificatior The reactor p otective system trip setting limits and the permissible bypasses for the instr. ment channels shall be as stated in Table 2.3-1A - Unit 1 and 2.3-1B - Unit 2 Figure 2.3-2A:
2.3-2A' } Unit 1
~
2.3-2B Unit 2 2.3-2C - Unit 3 The pump monitors shall produce a reactor trip for the following conditions:
a.
Loss of two pumps and reactor power level is greater than 55% (0.0% for Unit 1) of rated power.
b.
Loss of two pumps in one reactor coolant loop and reactor power level is greater than 0.0% of rated power.
(Power /RC pump trip setpoint is reset to 55% of rated power for single loop operation and for Units 2 and 3, the flux-flow setpoint must be set at 0.961 prior to single loop operation.
Power /RC pump trip setpoint is reset to 55% for all modes of 2 pump operation for Unit 1.)
c.
Loss of one or two pumps during two-pump operation.
Bases The reactor rotective system consists of four instrument channels to monitor each of several sclected plant conditions which will cause a reactor trip if any one of these etnditions deviates from a pre-selected operating range to the degree that a safety limit may be reached.
The trip setting limits ' ar protective system instrumentation are listed in Table 2.3-1A - Unit 1.
The safety analysis has been based upon these protective 2.3-1B - Unit 2 2.3-1C - Unit 3 system instrumentation trip set points plus calibration and instrumentation errors.
Nuclear Overpower A reactor trip at high power leve) (neutron flux) is provided to prevent damage to the fuel cladding from reactivity excursions too rapid to be detected by pressure and temperature measurements.
2.3-1
During normal plant operation with all reactor coolant pumps operating, reactor trip is initiated when the reactor power level reaches 105.5% of rated power.
Adding to this the possible variation in trip setpoints due to calibration l
and instrument errors, the maximum actual power at which a trip would be actu-ated could be 112%, which is more conservative than the value used in the safety analysis.(4.)
l Overpower Trip Based on Flow and Imbalance The power 1svel trip set point produced by the reactor coolant system flow is based on a power-to-flow ratio which has been established to accommodate the most severe thermal transient considered in the design, the loss-of-coolant l
flow accident from high power. Analysis has demonstrated that the specified power-to-flow ratio is adequate to prevent a DNBR of less than 1.3 should a low flow condition exist due to any electrical malfunction.
The power level trip set point produced by the power-to-flow ratio provides both high power level and low flow protection in the event the reactor power level increases or the reactor coolant flow rate decreases.
The power level trip set point produced bythe power-to-flow ratio provides overpower DNB pro-tection for all modes of pump operation.
For every flow rate there is a maxi-mum permissible power level, and for every power level there is a minimum permissible low flow rate.
Typical power level and low flow rate combinations for the pump situations of Table 2.3-1A are as follows:
1.
Trip would occur when four reactor coolant pumps are operating if power is 108% and reactor flow rate is 100%, or flow rate is 93% and power level is 100%.
2.
Trip would occur when three reactor coolant pumps are operating if power is 81.0% and reactor flow rate is 74.7% or flow rate is 69% and power j
level is 75%.
3.
Trip would occur when two reactor coolant pumps are operating in a single loop if power is 59% and the operating loop flow rate is 54.5% or flow rate is 43% and power level is 46%.
(For Tables 2.3-1B and 2.3-1C the values are 52% power if the operating loop flow rate is 54.5% or flow rate is 48%
and power level is 46%.)
l 4.
Trip would occur when one reactor coolant pump is operating in each loop I
(total of two pumps operating) if the power is 53% and reactor flow rate l
is 49.0% or flow rate is 45% and the power level is 49%.
For safety calculations the maximum eslibration and instrumentation errors for the power level trip were used.
The power-imbalance boundaries are established in order to prevent reactor thermal limits from being exceeded.
These thermal limits are either power peaking kw/ft limits or DNBR limits.
The reactor power imbalance (power in the top half of core minus power in the bottom half of core) reduces the power level trip produced by the power-to-flow ratio such that the boundaries of Figure 2.3-2Al } Unit 1 are produced.
The power-to-flow ratio reduces the power 2.3-2A2 2.3-2B - Unit 2 2.3-2C - Unit 3 2.3-2
level trip and associated reactor power / reactor power-imbalance boundaries by 1.08% - Unit 1 for a 1% flow reduction.
1.07% - Unit 2 1.07% - Unit 3 For Units 2 and 3, the power-> 3-flow reduction factor is 0.961 during single loop operation.
Pump Monitors The pump monitors prevent the minimum core DNBR from decreasing below 1.3 by tripping the reactor due to the loss of reactor coolant pump (s).
The circuitry monitoring pump operational status provides redundant trip protection for DNB by tripping the reactor on a signal diverse from that of the power-to-flow ratio.
The pump monitors also restrict the power level for the number of pumps in operation.
Reactor Coolant System Pressure During a startup accident from low power or a slow rod withdrawal from high power, the system high pressure set point is reached before the nuclear overpower trip set point.
The trip setting limit shown in Figure 2.3-1A - Unit 1 2.3-1B - Unit 2 2.3-lc - Unit 3 for high reactor coolant system pressure (23,s psig) has been established to maintain the system pressure below the safety limit (2750 psig) for any design transient.(1)
The low pressure (1985) psig and variable low pressure (13.77 Tout-618D trip (1800) psig (16,25 T
-7756)
(1800) psig (16.25T$'t 7756) setpoints shown in Figure 2.3-1A have been established to mainta$n the DNB 2.3-1B 2.3-1C ratio greater than or equal to 1.3 for those design accidents that result in a pressure reduction.(2,3)
Due to the calibration and instrumentation errors the safety analysis used a variable low reactor coolant system pressure trip value of (13.77 Tout - 6221)
(16.25 T
-7796)
(16.25 T
-7796) g Coolant Outlet Temperature The high reactor coolant outlet temperature trip setting limit (619 F) shown in Figure 2.3-1A has been established to prevent excessive core coolant 2.3-1B 2.3-1C temperatures in the operating range.
Due to calibration and instrumentation errors, the safety analysis used a trip set point of 620*F.
l Reactor Building Pressure The high react'or building pressure trip setting limit (4 psig) provides positive assurance that a reactor trip will occur in the unlikely event of a loss-of-coolant accident, even in the absence of a low reactor coolant system pressure trip.
l 2.3-3 1
Shutdown Bypass j
In order to provide for control rod drive tests, zero power physics testing, and startup procedures, there is provision for bypassing certain segments of the reactor protection system.
The reactor protection systen segments which can be bypassed are shown in Table 2.3-1A.
Two conditions are imposed when 2.3-1B 2.3-lc the bypass is used:
1.
By administrative control the nuclear overpower trip set point must be reduced to a value < 5.0% of rated power during reactor shutdown.
2.
A high reactor coolant system pressure trip setpoint of 1720 psig is automatically imposed.
The purpose of the 1720 psig high pressure trip set point is to prevent normal operation with part of the reactor protection system bypassed.
This high pressure trip set point is lower than the normal low pressure trip set point so that the reactor must be tripped before the bypass is initiated. The over power trip set point of < 5.0% prevents any significant reactor power from being produced when performing the physics tests.
Sufficient natural circulation (5) would be available to remove 5.0% of rated power if none of the reactor coolant pumps were operating.
Two Pump Operation A.
Two Loop Operation Operation with one pump in each loop will be allowed only following reactor shutdown. Af ter shutdown has occurred, the following actions will permit operation with one pump in each loop:
1.
Reset the pump contact monitor power level trip setpoint to 55.0%.
2.
(Unit 1) Reset the protective system maximum allowable setpoint,as shown in Figure 2.3-2A2.
B.
Single Loop Operation Single loop operation is permitted only after the reactor has been tripped.
After the pump contact monitor trip has occurred, the following actions will permit single loop operation:
1.
Reset the pump contact monitor power level trip setpoint to 55.0%.
2.
Trip one of the two protective channels receiving outlet temperature information from sensors in the Idle Loop.
3.
(Unit 1) Reset the protective system maximum allowable setpoints as shown in Figure 2.3-2A2.
Tripping one of the two protective channels receiving outlet temperature information from the idle loop assures a protective system trip logic of one out of two.
4.
(Units 2 and 3) Reset flux-flow setpoint to 0.961.
REFERENCES (1) FSAR, Section 14.1.2.2 (5) FSAR, Section 14.1.2.6 (2) FSAR, Section 14.1.2.7 (3) FSAR, Section 14.1.2.8 (4) FSAR, Section 14.1.2.3 2.3-4
Power Level, %
-- 120 FOUR PUMP SETP0INTS 100 THREE PUMP SETPOINTS 80 TWO PUMP SETPOINTS 60 40 20 L_
i i
i i
i
-60
-40
-20 0
+20
+40
+60 Power Imbalance, %
- For two pumps in one loop, the flux-flow setpoint must be 0.961.
CORE PROTECTION SAFETY LIMITS UNIT 2 OCONEE NUCLEAR STATION Figure 2.1-2B 2.3-9
Pow:r Level, % '
- 120 FOUR PUMP SETPOINTS 100 THREE PUMP SETPOINTS R0 TWO PUMP
-- 60 SETPOINTS
__ 40
- 20
-60
-40
-20 0
20 40 60 Power Imbalance, %
- For two pumps in one loop, the flux-flow setpoint must be 0.961 PROTECTIVE SYSTEM MAXIMUM ALLOWABLE SETP0INTS UNIT 3 OCONEE NUCLEAR STATION Figure 2.3-2C l
2.3-10 l
Table 2.3-18 Unit 2 Reactor Protective System Trip Setting Limits Two Reactor One Reactor Four Reactor Three Reactor Coolant Pumps Coolant Pump Coolant Pumps Coolant Pumps Operating in A Operating in Operating Operating Single Loop Each Loop (Operating Power (Operating Power (Operating Power (Operating Power Shutdown RPS Segment
-100% Rated)
-75% Rated)
-462 Rated)
-49% Rated)
Bypass I3) 1.
Nuclear Power Max.
105.5 105.5 105.5 105.5 5.0 (1 Rated)
[1 2.
Nuclear Power Max. Based 1.07 times flow 1.07 times flow 0.961 times flow 1.07 times flav sypassed on Flow (2) and Imbalance, minus reduction minus reduction minus reduction minus reduction (I Rated) due to Labalance due to imbalance due to imbalance due to imbalance 3.
Nuclear Puwer Max. Based NA NA 55% (5)(6) 551 Bypassed on Pump Honitors. (%, Rated) 4.
High Reactor Coolant 2355 2355 2355 2355 1720 ')
I na System Pressure, psig, Max.
ta e
5.
Low Reactor Coolant
[]
System Pressure, psig, Min.
1800 1800 1800,
1800 typassed 6.
Variable Low Reactor (16.25 T
-7756)I )
(16.25 7
-7756)I )
(16.25 T
-7756)(I)
(16.25 7 -7756)II)
Bypassed Coolant System Pressure pois, Min.
7.
Beactor Coolant Temp.
619 619 619 (6) 619 619 F., Max.
8.
High Reactor Building 4
4 4
4 4
Pressure, psig, Max.
)
(1) T,,g is in degrees Fahrenheit ( F).
(5) Reactor power level trip set point produced by pump contact uonttor reset to 55.02.
(2) Reactor Coolant System Flow, %.
(6) Specification 3.1.8 applies. Trip one of the (3) Administratively controlled reduction set two protection channels receiving outlet temper-unty during reactor shutdown, ature information from sensors in the idle loop.
(4) Automatics 11y set when other segments of the Rf3 are bypassed.
Table 2. 'l-IC Unit 3 Reactor Prote:tive Svstem Trip Setting Limits two Reactor One Reactor Four Reactor Three Reactor Coolant Pumps Coolant Pump Coolant Pumps Coolant Pumps Operating in A Operating in 7erating Operating Single loop Each Loop aperating Power (Operating Power (Operating Power (Operating Power Shutdown RPS Segment
-100% Rated)
-75% Rated)
-46% Rated) _
-49% Rated)
Bypass I3) 1.
Nuclear Power tinx.
105.5 105.5 105.5 105.5 5.0
(% Rated) l 2.
Nuclear Power Max. based 1.07 times flow 1.07 tincs flow 0.961 times flow 1.07 times flow sypassedi on Flow (2) and imbalance,'
minus reduction minus reduction minus reduction minus reduction
(% Rated) due to imbalance due to imbalance due to imbalance due to imbalance 3.
Nuclear Power Max. Based NA NA 55% (5)(6) 55%
Bypassed on Pump Honitors. (%, Rated)
IN 4.
tilgh Reactor coulant 2355 2355 2355 2355 1720 p
System Pressure, psig, Max.
I C
5.
Low Reactor Coolant System Pressure, psig, Min.
1800 1800 1800 1800 Bypassed 6.
Variable Low Reactor (16.25 T
-7756)II)
(16.25 T
-7756)
(16.25 T
-7756)(1)
(16.25 T
-7756)fII Bypassed ou a
at at Coolant System Pressure psig, Min.
7.
Reactor Coolant Temp.
619 619 619 (6) 619 619 F., Max.
8.
tilgh Reactor Building 4
4 4
4 4
Pressure, psig, tbx.
(1) T is in degrees Fahrenheit ( F).
(5) Reactor power level trip set point produced by pump contact monitor reset to 55.0%.
(2) Reactor Coolant System Flow, %.
(6) Specification 3.1.8 applies. Trip one of the (3) Administratively controlled reduction set two protection ct.annels receiving outlet temper-only during reactor shutdown, ature information from sensors in the idle loop.
(4) Automatic 111y set when other segments of the RE.i are bypassed.