ML19312B969

Forwards Proposed Tech Specs 2.1,2.3 & 3.5 Re Reactor Core Safety Limits,Protective Instrumentation Limiting Safety Sys Settings & Control Rod Group Withdrawal Limits.Supports Operation of Unit 1,Cycle 2 at Rated Power

ML19312B969
Person / Time
Site:	Oconee
Issue date:	10/28/1974
From:	DUKE POWER CO.
To:
Shared Package
ML19312B963	List: ML19312B962 ML19312B969
References
NUDOCS 7911270751
Download: ML19312B969 (8)

Text

_

O REPLACEMENT PAGES FOR PROPOSED TECHNICAL SPECIFICATIONS i

5911270 7 9 /

Rev. 1.

10/2d/74

^i

~)

2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS, REACTOR CORE Applicability Applies to reactor thermal power, reactor power imbalance, reactor coolant system pressure, coolant temperature, and coolant flow during power operation of the plant.

Objective To maintain the integrity of the fuel cladding.

Specification The combination of the reactor systen pressure and coolant temperature shall not exceed the safety limit as defined by the locus of points established in Figure 2.1-lA-Unit 1.

If the actual pressure / temperature point is below 2.1-1B-Unit 2 2.1-lC-Unit 3 and to the right of the line, the safety limit is exceeded.

The combination of reactor thermal power and reactor power imbalance (power in the top half of the core minus the power in the bottom half of the core expressed as a percentage of the rated power) shall not exceed the safety limit as defined by the locus of points (solid line) for the specified flow set forth in Figure 2.1-2A-Unit 1.

if the actual reactor-thermal-power / power 2.1-2B-Unit 2 2.1-2C-Unit 3 imbalance point is above the line for the specified flow, the safety limit is exceeded.

Bases - Unit 1 The safety limits presented for Oconge Unit 1 have been generated using BAW-2 critical heat flux (CHF) correlation (l)and the actual measured flow rate at Oconee Unit 1 (2).

This development is discussed in the Oconee 1, Cycle 2-Reload Report, reference (2). The flow rate utilized is 107.6 percent of the design flow (131.32 x 106 lbs/hr) based on four-pump operation.(2)

To maintain the integrity of the fuel cladding and to prevent fission product release, it is necessary to prevent overheating of the cladding under normal operating conditions.

This is accomplished by operating within the nucleate boiling regime of heat transfer, wherein the heat transfer coefficient is large enough so that the clad surface temperature is only slightly greater than the coolant temperature. The upper boundary of the nucleate boiling regime is termed " departure from nucleate boiling" (DNB). At this point, there is a sharp reduction of the heat transfer coefficient, which would result in high cladding temperatures and the possibility of cladding faflure. Although DNB is not an observable parameter during reactor c;eration, the observable parameters of neutron power, reactor coolant flow, temperature, and pressure j

i 2.1-1

}

For Figure 2.1-3A, a pressure-temperature point above and to the left of the curve would result in a DNBR greater than 1.32.

The 1.32 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) Correlation of Critical Heat Flux in a Bundle Cooled by Pressurized Water, BAW-10000, March, 1970.

(2) Oconee 1, Cycle 2 - Reload Report - BAW-1409, Sepetmeber, 1974.

4 l

l 2.1-3

.s

..)

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 2.1-1C 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 maximue 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 qualit. 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 ta 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% - Unit 2 86% - 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 maxi =um calibration and instrument error.

The maximum thermal power for other coolant pump conditions are produced in a similar manner.

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.l.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 2.1-3b

s

)

2.3 LIMITING SAFETY SYSTEM SETTINGS, PROTECTIVE INSTRLEENTATION 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 permissible bypasses for the instrument channels shall be as stated in Table 2.3-1A - Unit 1 and 2.3-1B - Unit.2 2.3-lc - Unit 3 Figure 2.3-2A1 y 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. Power /RC pump trip setpoint is reset to 55% for all modes of 2 pump operation for Unit 1.)

Loss of one or two pumps during two-pump operation.

c.

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 Tiole 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 provided to prevent damage te the fuel-cladding from reactivity excursions too rapid to be detected by pressure and temperature measurements.

2.3-1

r 1.

Rod index is trie percentage sum of the withdra:al of the operating

~

groups.

2.

The withdrawal limits are modified after 250 1 5 full power days of operation.

173 209 100 154 213 945 Power Level Cutoff Restricted Region 80 122 230 80% b

' 75%

g y

2 v*+s,

60

'*/f ae

~

[

Permissinte

-(52% P)

E p

40 g[p Operating Region Y

20 0

0 50 100 150 200 250 300 Rod Index, 5 Witndrawal 25 50 75 100 t

f f

I O

25 50 75 100 Gp7 1

1 i

i I

O 25 50 75 100 Gp6 I

i i

I Gp5 CONTROL ROD GROUP WITHDRAWAL LIMITS FOR 4 PUMP OPERATION UNIT 1 3.5-12 ovu Ewin OCONEE NUCLEAR STATION Figure 3.5.2-1A1

?

l.

Rod index is the percentage sum of the withdrawal of the operating groups.

2.

T he withdrawal limits are in ef f ect af ter 25015 full power days of operation. (Tne applicable power level cutof f is 1005 power) 257.5 288. 1 100 1

2 G

80 Restricted

{

E Pcgion f

735 y

E 60 h'

/[

(525 P) t 4

PermissiDie j

40 Operating Region 20 0

t i

i i

0 50 100 150 200 250 300 Rod index, 5 withdrantl 0

25 50 75 100 i

f I

t 1

UPI 0

25 50 75 100 a

l I

I t

0 25 50 75 100 Gp6 I

I I

f l

Gp5 CONTROL ROD GROUP WITHDRAWAL LIMIT FOR 4 PUMP OPERATION UNIT 1 3.5-13 ountrental OCONEE NUCLEAR S TATION Figure 3.5.2-1A2

/

~

l.

Rod index is the porcentage sua of the withdra:al of the operating groups. (The applicable power level cutoff is 100% power) 2 Pump Withdrawal I.imit 100 150 275 C

Restricted j

E Region i

3 u

80 S

4 d

e E

D Permissible

[

Operating 6

4,'

2 2

Region 60 E

a 2

a 40 e

E 2

20 0

i i

i i

0 50 100 150 200 250 300 Rod index, t, withdrawal CO.IROL ROD GROUP WITHDRA$4AL LIMITS FOR 3 K D 2 PUMP OPERATION UNIT 1 OCONEE NUCLEAR STATION 3.5-18 Figure 3.5.2-2A

.__