ML19312C852

Proposed Replacement Pages to Tech Specs 2.1 & 2.3 Re Safety Limits for Reactor Thermal Power & Power Imbalance Combination & to flux-flow & power-imbalance Trip Setpoints

ML19312C852
Person / Time
Site:	Oconee
Issue date:	08/26/1975
From:	DUKE POWER CO.
To:
Shared Package
ML19312C850	List: ML19312C849 ML19312C852
References
NUDOCS 8001130007
Download: ML19312C852 (11)

Text

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O Power peaking is not a dir ectly 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

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2.1-2C to the expected minimum flow rates with four pumps, three pumps, one pump in each loop and two pumpt in one loop, respectively.

The curve of Figure 2.1-lD is the most restrictive of all possible reactor 2.1-1C 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 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 Fercent is justified on the basis of experimental data.(4)

The maximum thermal power for three pump opt v.lon 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 left of the curve would result in a DNBR greater than 1.3 or a local quality  :

at the point cf minimum DNBR less than 15 percent for that particular reactor ,

I coolant pump situstion. The 1.3 DNBR curve for four-pump operation is more restrictive than any other reactor coolant pumn situation because any pressure / i temperature point above and to the left of the four-pump curve will be above i 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

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8001130 007

Thermal Power Level, %

120 kw/ft Limit kw/ft Limit

. . 100

. 80 DNBR Limi

.. 60 DNBR Limit 40

.. 20 I I I t t f

-60 -40 -20 0 +20 +40 '+60 Reactor Power Imbalance, %

CURVE REACTOR COOLANT FLOW (LB/ll2) 6 1 131.3 x 10 2 98.1 x 10 6 3 64.4 x 10 6 4 60.1 x 10 PROTECTIVE SYSTEM MAXIMUM ALLOWABLE SETP0INTS UNIT 2 OCONEE NUCLEAR STATION )

Figure 2.3-2B l 2.1-8

THERMAL POWER LEVEL, %

. 120 kw/ft Limit m _ kw/ft Limit

" 100 0

- 80 DNBR Limit O

. . 60 DNBR Limit O

DNBR 40 Limit 20

. . . . . m

-60 -40 -20 0 +20 +40 +60 Reactor Power Imbalance, ".

CURVE REACTOR COOLANT FLOW (LB/HR) 1 131.3 x 10 6 2 98.1 x 10 6 3 64.4 x 10 4 60.1 x 10 6 CORE PR0TECTION SAFETY LIMITS

! UNIT 3 h OCONEE NUCLEAR STATION Figure 2.1-2C 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 automatic protective action to prevent any combination of process variables from exceeding a safety limit.

Specification The reactor protective system trip setting limits and the per:fssible bypasses for the instrument channels shall be as stated in Table 2.3-1A - Unit 1 and 2.3-1B - Unit 2 2.3-1C - Unit 3 Figure 2.3-2Al } Unit 1 2.3-2A2 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 protective system consists of four instrument channels to monitor each'of several selected plant conditions which will cause a reactor trip if any one of these conditions deviates from a pre-selected operating range to the degree that a safety limit may be reached.

The trip setting limits for 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 level (neutron flux) is provideu 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 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.)

Overpower Trip Based ,n Flow and Imbalance The power level *.ip set point produced by the reactor coolant system flow is based on a pov2r-to-flow ratio which has been established to accommodate the most severe .hermal transient considered in the design, the loss-of-coolant flow accide: t from high power. Analysis has demonstrated that the specified power-to-fl.w ratio is adequate to prevent a DNBR of less than 1.3 should a low flow :endition 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-flaw 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 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-lc the values are 52% power if the operating loop flow rate is 54.5% or flow rate is 48%

and power level is 46%.)

4. Trip would occur when one reactor coolant pump is operating in each loop l (total of two pumps operating; if the power is 53% and reactor flow rate is 49.0% or flow rate is 45% and the power level is 49%.

For safety calculations the maximum calibration 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 pr duced. The power-to-flow ratio reduces the power 2.3-2A2 2.3 Unit 2 2.3-2C - Unit 3 2.3-2 1

level trip and associated reactor power /recctor power-imbalcnc3 b:undtrics 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-to-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 Unit 2 2.3-1C - Unit 3 for high reactor coolant system pressure (2355 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 -

(1800) psig (16.25T$[7756) 7756) setpoints shown in Figure 2.3-1A have been established to maintain the DNB 2.3-1B 2.3-lc 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 Tout -

96)

(16.25 T g g -7796)

Coolant Outist 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-lc temperatures in the operating range. Due to calibration and instrumentation errors, the safety analysis used a trip set point of c20*F.

Reactor Building Pressure The high reactor 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.

2.3-3

x Shutdown Bypass in order to provide for control rod drive tests, zero power physics testing, and startup procedures, thereis provision for bypassing certain segments of the reactor protection system. The reactor protection system segments which can be bypassed are shown in Table 2.3-1A. Two conditions are imposed when 2.3-1B i 2.3-lc the bypass is used:

1. By administrative control the nuclear overpower trip set poine 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. After 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 af ter the reactor has been tripped.

Af ter 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 i 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 e

(1) FSAR, Section 14.1.2.2 (5) FSAR, Section 14.1.2.6 i

(2) FSAR, Section 14.1.2.7 (3) FSAR, Section 14.1.2.8 (4) FSAR, Section 14.1.2.3 I

2.3-4

Power Level , ". ,

-- 120 FOUR PUMP SETPOINTS

- - 100 THREE PUMP SETPOINTS 80 TWO PUMP

. 60 SETPOINTS

. 40 20 l I I I n 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 OCONSE NUCLEAR STATION Figure 2.1-2B 2.3-9

Power Level, %

. 120 FOUR PUMP SETPOINTS

. 100 THREE PUMP SETPOINTS A0 TWO PUMP - - 60 SETPOINTS

\

/

__ 40 3

- 20 ,

-60 -40 -20 0 20 40 60 Power Imbalance, %

l

• For two pumps in one loop, the I flux-flow setpoint must be 0.961 PROTECTIVE SYSTEM MAXIMUM ALLOWABLE SETPOINTS UNIT 3 I OCONEE NUCLEAR STAT!ON Figure 2.3-2C 2.3-10 I

o Table 2.3-1B Unit 2 kcactor Protec tive System Trip Set ting Limit s TVoIReactor 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% RateJ) ,

Bypass

1. Nuclear Power Max. 105.5 105.5 105.5 105.5 5.0 I3I (I Rated)

2. Nuclear Power Max. Based 1.07 times flow 1.07 times flow 0.961 times flow 1.07 times flow Bypassed on Flow (2) and labalance, minus reduction minus reduction minus reduction minus reduction (I Rated) due to Labalance due to imbalance due to imbalance due to imbalance

3. Nuclear Power Hax. Based NA NA 55% (5)(6) 551 Bypassed on Pump Monitors. (1, Rated)

4. High Reactor Coolant 2355 2355 2355 2355 1720I 'I ra System Pressure, psig, Max.

u i 5. Low Reactor Coolant 1800 1800 , 1800 Bypassed

[] System Pressure, psig, Min. 1800

6. Variable Low Reactor (16.25 T "

-7756)(I) (16.25 T "

-7756)( I (16.25 T" -7756)III (16.25 T" -7756)III Bypassed Coolant System Pressure psig. Min.

7. Reactor Coolant Temp. 619 619 619 (6) 619 619 F., Ham.

8. High Reactor Building 4 4 4 4 4 Pressure, psig Hax.

(1) T is in degrees Fahrenheit ( F). (5) Reactor puwer level trip ser point produced "E by pump contact monitor reset to 55.01.

(2) Resctor Coolant System Flow, %.

(6) Specification 3.1.8 applies. Trip one of the (3) Administratively controlled reduction set two protection channels receiving outlet temper-only during reactor shutdown. ature information from sensors in the idle loop.

(4) Automatically set when other segments of the RF3 are bypassed.

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9 Table 2.1-IC Unit 3 Reactor Protective System Trip Setting 1.imits bro 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) -46% Rated) -49% Rated) Bypass 5.0 I3)

l. Nuclear Power Max. 105.5 105.5 105.5 105.5

(% Rated)  ;

2. Nuclear Power Max. Based 1.07 times flow 1.07 times flow 0.961 times flow 1.07 times flow Bypassed on Flow (2) and imbalance, minus reduction minus reduction minus reduction minus reduction

(% Rated) due to imbalance due to imbalance due to imbalance du to imbalance

3. Nuclear Power Max. Based NA NA 55% (5)(6) 55% Bypassed on Pump Monitors. (%, Rated) w 4. High Reactor Coolant 2355 2355 2355 2355 1720 II y System Pressure, psig, Max.

1

[ 5. Low Reactor Coolant -

System Pressure, psig, Min. 1800 1800 1800 1800 Bypassed

6. Variable few Reactor (16.25 Tout -7756) (16.25 T -7756) (16.25 T -7756)III (16.25 Tout -7756)II) Bypassed ut ut Coolant System Pressure psig Min.

7. Reactor Coolant Temp. 619 619 619 (6) 619 619 f., Max.

8. High Reactor Building 4 4 4 4 4 Pressure, psig. Max.

(1) Tout 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) Administven.!vely controlled reduction set two protection cl.annels receiving outlet temper-only durog reactor shutdown, ature information from sensors in the idle loop.

9 (4) Automatic n11y set when other segrLents of the U.i are bypassed.

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