ML20235Z542
| ML20235Z542 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 07/17/1987 |
| From: | FLORIDA POWER CORP. |
| To: | |
| Shared Package | |
| ML20235Z494 | List: |
| References | |
| NUDOCS 8707270325 | |
| Download: ML20235Z542 (4) | |
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J 2.1 SAFETY LIMITS BASES 2,1.1 and 2.1.2 REACTOR CORE The restrictions of thin safety limit prevent overheating of the fuel cladding and possible cladding perforation which would result in the release of fission products to the reactor coolant.
Overheating of the fuel cladding i.s prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.
Operation above the upper boundary of the nuc1cate boiling regime would result in excessive cladding temperature because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient.
DNB is not a directly measurable parameter during operation but THERMAL POWER and Reactor Coolant Temperature and Pressure can be related to DNB using a Critical Heat Flux (CHF) correlation.
The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux is indicative of the martin to DNB.
The B&W-2 and BWC CHF correlations have been developed to predict DND for axially uniforn and non-uniform heat flux distributions.
The B&W-2 correlation applies to Mark-B fuel and the BWC correlation applies to Mark-BZ fuel.
The minimum value of the DNBR during staady state operation, normal operational transients, and anticipated transients is limited to 1.30 (B&W-2) and 1.18 (BWC).
A DNBR of 1.30 (B&W-2) or 1.18 (BWC) corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occur.
The curve presented in Figure 2.1-1 represents the conditions at which the minimum allowable DNBR or greater is predicted for the thermal power and number of operating reactor coolant pumps.
This curve is based on the following nuclear power peaking factors with potential fuel densification effects:
N N
N F
= 2.82 F = 1.71 F = 1.65 0
H Z
i The design limit power peaking factors are the most restrictive calculated at full power for the range from all control rods fully withdrawn to minimum allowable control rod withdrawal, and form the core DNBR design basis, i
l l
1 i
l 8707270325 870717 PDR ADOCK 05000302 p
PDR CRYSTAL RIVER - UNIT 3 B 2-1 L_1___ _ __ _
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SAFETY LIMILS 1.
I BASES l
l For each curve of BASES Figure 2.1, a pressure-temperature point above and to the left of the curve would result in a DNBR greater than 1.30 (B&W-2) or 1.18 (BWC) or a local quality at the point of minimum DNBR less than 22% (B&W-2) or 26% (BWC) for that particular reactor coolant pump situation.
The curve for j
three 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 three pump curve will be above and to the left of the oth3r curves.
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2.1.3 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.
The reactor pressure vessel and pressurizer are designed to Section III of the ASME Boiler and Pressure Vessel Code which permits a maximum transient pressure of 110%, 2750 psig, of design pressure.
The Reactor Coolant System piping, valves and fittings, are designed to USAS B 31.7, February, 1968 Draft Edition, which permits a maximum transient pressure of 110%, 2750 psig, of component design pressure.
The Safety Limit of 2750 psig is therefore consistent with the design criteria and associated code requirements.
The entire Reactor Coolant system is hydrotested at 3125 psig, 125% of design pressure, to demonstrate integrity prior to initial operation.
1 Lt CRYSTAL RIVER - UNIT 3 B 2-3
TABLE 14-10 sp.
M Moderator Dilution Accident Parameters Flow Rates Considered:
Dilution Flow Rate Condition Normal _ Dilution, gpn 70 Normal Dilution with Low Reactor Cholant System Pressure, gpn 100 Marimum Considered, gpn 500 L
Initial Boron Concentration in Reactor 2270-Coolant.cystem, ppu (hot, clean,BOL)
Boron Reactivity Worth, ppn/1%Ak/k 140 (HFP,0 EIPD,NO XENON)
Mc6erator coefficient, (Ak/k)/F
+0.5 x 10-4 Doppler Coefficient, (Ak/k)/F
-1.17 x 10-5 Dilution valve Interlock Setpoints:
i Dilution Valve May Be Opened When:
(1) Control rod is withdrawn to 95% out positior., and (2) Integrated flow timing device is set.
Dilution Valves Autanatically Close When:
(1) Control rod is inserted to 75% full in, or (2) Timing device has reached a preset value corresponding to a gien integrated flow, or (3) Reactor trips O
TABLE 14-14 Surary of Moderator Dilution Accident Analysis A. Dilution At Power Rate of Change in Core Average Dilution hbter Beactivity Rate F1uid Temp.,
Condition Flow, exm (ok/k)/sec F/sec NDBMAL 70 3.413 x 10-6 0.0966 LOf RCS 100 4.842 x 10-6 0.1095 IRESSURE MAXIMJM CDNSIDERED 500 2.183 x 10-6 0.2082 B. Dilution to Trip Dilution Water Peak Thermal Deak Pressure, Time To Trip, Flow,can Power, %FP psia sec 70 104.99 2455.7 72, ';0 500 109.45 2477.7 34.27 C. Dilution at Shutdown Initial Boron -
Initial Total Final Boron Final Total Concentration, Subcriticality, Concentratica, Subcriticality, pan pan l
t 2498
-5.0 2176
-1. 81 3 i
I The results are based on the following assuaptLons (a) Cold (70F), DOL, Clean
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(b) Emptying i full makeup tank volune of 600 f:3 (c) Asstmed inverso boron worth of 101 ppvuk/L I
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