ML20052C639

Proposed Tech Specs,Modifying Reactor Protection Sys & ESF Actuation Sys Equipment Setpoints Per Revised B&W Calculations.Pages 9 & 11-15 Encl

ML20052C639
Person / Time
Site:	Arkansas Nuclear
Issue date:	04/29/1982
From:	ARKANSAS POWER & LIGHT CO.
To:
Shared Package
ML20052C635	List: ML20052C634 ML20052C639
References
NUDOCS 8205050320
Download: ML20052C639 (6)

Text

. _ _ .

The DNBR as calculated by the BAW-2 correlation continually increases from the point of minimum DNBR, so that the exit DNBR is always higher and is a function of the pressure.

The maximum thermal powe, for three pump operation is 88.92 percent due to a power level trip produced by the flux-flow ratic (74.7 percent flow x 1.054 = 78.73 percent power) plus the maximum calibration and instrumentation error. The maximum thermal power for other reactor coolant pump conditions is produced in a similar manner.

For each curve of Figure 2.1-3, a pressure-temperature point above and to the left of the curve would result in a DNBR greater than 1.3 or a local quality at the point of minimum DNBR less than 22 percent for that particular reactor coolant pump situation. Curves 1 and 2 of Figure 2.1-3 are the most restrictive because any pressure / temperature point above and to the left of this curve will be above and to the left of the other curve.

REFERENCES (1) Correlation of Critical Heat Flux in a Bundle Cooled by Pressitrized Water, BAW-10000A, May 1976.

(2) FSAR, Section 3.2.3.1.1.c.

i 9

8205050320 820429

~

PDR ADOCK 05000313 P PDR

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 Ta provide automatic protection action to prevent any combination of process variables from exceeding a safety limit.

Specification 2.3.1 The reactor protection system trip setting limits and the permissible bypasses for the instrument channels shall be as stated in Table 2.3-1 and Figure 2.3-2.

Bases-The reactor protection 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 preselected operating range to the degree that a safety limit may be reached.

The trip setting limits for protection system instrumentation are listed in Table 2.3-1. The safety analysis has been based on these protection ,

system instrumentation trip setpoints plus calibration and instrumentation errors.

Nuclear Overpower A reactor trip at high power level (neutron flux) is provided to prevent damage to the fuel cladding from reactivity excursions too rapid to be detected by pressure and temperature measurements.

During normal plant operation with all reactor coolant pumps operating, reactor trip is initiated when the reactor power level reaches 104.9 percent 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 actuated could be 112%, which is the value used in the safety analysis.

A. Overpower Trip Based on Flow and Imbalance The power level trip setpoint 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-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.

11

The power level trip setpoint 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 setpoint '

produced by the power-to-flow ratio provides overpower DNB protection for all modes of pump operation. For every flow rate there is a maximum 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 2.3-1 are as follows:

1. Trip would occur when four reactor coolant pumps are operating if power is 105.4 percent and reactor flow rate t 100 percent or flow rate is 94.88 percent and power level is 100 percent.

2. Trip would occur when three reactor coolant pumps are operating ir power is 78.73 percent and reactor flow rate is 74.7 percent or flow rate is 71.16 percent and power level is 75 percent.

3. Trip would occur when one reactor coolant pump is operating in each loop (total of two pumps operating) if the power is 51.85 percent and reactor flow rate is 49.2 percent or flow rate is 46.49 percent and the power level is 49.0 percent.

The flux / flow ratios account for the maximum calibration and instrumentation errors and the maximum variation from the average value of the RC flow signal in such a manner that the reactor protective system receives a conservative indication of the RC flow.

No penalty in reactor coolant flow through the core was taken for an open core tent valve because of the core vent valve surveillance program during each refueling outage. For safety analysis calculations the maximum calibration and instrumentation errors for the power level 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 top half of core minus power in bottom half of core) reduces the power level trip produced by the power-to-flow ratio so that the boundaries of Figure 2.3-2 are produced. The power-to-flow ratio reduces ths power level trip associated with reactor power-to-reactor power imbalance boundaries by 1.054 percent for a 1 percent flow reduction.

B. Pump Monitors In conjunction with the power imbalance / flow trip, the pump l

monitors prevent the minimum core DNBR from decreasing below 1.3 by tripping the reactor due to the loss of reactor coolant 12

pump (s). The pump monitors also restrict the power level for the number of pumps in operation.

C. RCS Pressure During a startup accident from low power or a slow rod withdrawal from high power, the system high pressure trip setpoint is reached before the nuclear overpower trip setpoint.

The trip setting limit shown in Figure 2.3-1 for high RCS pressure (2300 psig) has been established to maintain.the system pressure below the safety limit (2750 psig) for any design transient.(2)

The low pressure (1800 psig) and variable low pressure (11.75 T - 5103) trip setpoints shown in Figure 2.3-1 have beenestNkishedtomaintaintheDNBratiogreaterthanor 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 (11.75 T out - 5143).

D. Coolant Outlet Temperature Thehighreactorcoolantoutlettemperaturetripsettinglimitl (618F) shown in Figure 2.3-1 has been established to prevent excessive core coolant temperatures in the operating range.

Due to calibration and instrumentation errors, the safety analysis used a trip setpoint of 620F.

E. 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 steam line failure in the reactor building or a loss-of-coolant accident, even in the absence of a low reactor coolant system pressure trip.

F. Shutdown Bypass 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 system segments that can be bypassed are shown in Table 2.3-1. Two conditions are imposed when the bypass is used:

1. A nuclear overpourtrip setpoint of 55.0 percent of rated power is automatically imposed during reactor shutdown.

l 2. A high reactor coolant system pressure trip setpoint of l 1720 psig is automatically imposed.

13

1. O The purpose of the 1720 psig high pressure trip setpoint is to prevent normal operation with part of the reactor protection system bypassed. This high pressure trip setpoint is lower than the normal low pressure trip setpoint so that the reactor must be tripped before the bypass is initiated. The overpower trip setpoint of 55.0 prevents any significant reactor power from being produced when performing the physics tests.

Sufficient natural circulation (5) would be available to remove 5.0 percent of rated power if none of the reactor coolant pumps were operating.

References (1) FSAR, Section 14.1.2.3 (2) FSAR, Section 14.1.2.2 (3) FSAR, Section 14.1.2.7 (4) FSAR, Section 14.1.2.8 (5) FSAR, Section 14.1.2.6 14 l

Tablo 2.3-1 Reactor Protection System Trip Setting Units

.Four RC Pumps Three.RC Pumps One RC Pump Operating Operating (Nominal) Operating (Nominal in each loop (Nominal Operating Power Operating Power Operating Power. . Shutdown 100%) 75%) -

49%) Bypass

' Nuclear power, %'of 104.9 104.9 104.9 5.0" rated, max ,

Nuclearbp wer based 1.054 times flow minus 1.054 times' flow minus 1.054 times flow minus Bypassed on flow and imbal- reduction due to imbal- reduction due to imbal- reduction due to imbal- '

ance, % of rated, :alance(s). ance(s). ance(s).

. max.

Nuclear power based NA NA 55 Bypassed onpumpmonitgrs,%

of rated; max a

High RC system 2300 2300 2300 1720 pressure, psig, max.

Low RC system ,

1800 1800 1800 Bypassed pressure, psig, min.

11.75 T out -5103 d d d M Variable low RC 11.75 Tout -5103 11.75 T out -5103 Bypassed system pressure, '

psig, min.

RC temp, F, max 618 618 618 '618 High reactor bldg. 4(18.7 psia) 4(18.7 psia) 4(18.7 psia) 4(18.7 psia) pressure, psig,. max.

a Automatically set when other segments of the:RPS (as specified) are bypassed.

b Reactor coolant system flow.

c The pump monitors also produce a trip on (a) loss of two RC pumps in one RC loop, and (b) loss of one or two RC pumps during two pump operation.

d T

out is given indegrees Fahrenheit (F).

V i

t

. _ . . . - _ . _ ~_ ~ .._ . . _ - . . _ , . - . .- . _ _ _ _ _ _ _ _