ML19317E928
| ML19317E928 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 03/09/1973 |
| From: | Stello V US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Deyoung R US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 8001070727 | |
| Download: ML19317E928 (2) | |
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MAR 9 1973 R. C. DeYoung, Assistant Director for Pressurized Water Reactors, L STATUS OF THE OCCNEE 1 ECCS EVALUATION FOR THE CFT LINE BREAK ACCIDENT in the TR review of the core flooding tank (CFI) line break accident, i
i two major areas of concern were identified:
(1) the amount of water lef t in the vessel during this accident and (2) the resulting height of two phase mixture in the core (level swell). In an effort to j
better understand the phemenonology, we have conducted independent calculations using RELAP, TOODEE, and a newly developed code, SWELL.
In some cases we compared the results with B&W analyses.
%e amount of water remaining in the vessel during this accident has been calculated by B&V using their hydraulics code, CRAFT. We have also run several, more conservative, cases with RELAF. %e major areas of difficulty, encountered both by us and B&V, in this analysis involves the modelling and phase separation (bubble-rise) assumptions.
To eliminate " pancaking" (multiple layers of water and steam) in the vessel B&W uses only one node for the core and upper plenum volumes. This simplification produces an unrealistically high prediction of the mixture height for "nonnal" bubble rise velocities due to the method of heat addition to such a large node. %e re-suits of a calculation, using the bubble rise model approved in the B&W evaluation model, predicts a collapsed water level approxi-mately at the six foot core elevation, with swallen-level mixture as high as the vent valves ( 20 ft.).
A mixture height covering the core permits the calculation of sufficient heat transfer to cool the core te neer seturetton temoerature. However, a mixture height above the core seemed nonconservative to us, in consideration of low fluid velocities, near constant pressure, and no heat generation above the core. To investigate this alleged inconsis-tency (mixture height too high) B&W performed a scoping calculation by increasing the bubble-rise velocity to a large value (20 ft /sec) such that the mixture height would be below the vent valves. Since no 2-)hase mixture could new leave the vessel through the vent valves and out the break, the net vessel mass increased, and the collapsed water level thaa increased, to approximately the ten foot l
elevation in the core.
From these calculations, we consider that use of a collapsed (i.e., " quiet") level of somewhere between six and ten feet in the core would be reasonable as input to a fuel-pin heatup calculation.
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In our independent heat transfer analysis a series of TOODEE steam-cooling calculations were made. These calculations assumed cooltag by convection to steam, starting at the steam-two-phase boundary. The stesa mass flow rate was that supplied by one high-pressure injection pump. The temperature transient calculation was teminated at 900 seconde due to the assumed operator action providing additional water from the low pressure injection system.
%ese parametric TOODEE runs allowed construction of a curve of maximum peak power level (cladding temperature below the Interim Acceptance critaria) as a function of two-phase mixture height in the core. Based on these calculations, a mixture height of 8 feet would permit operation at a power level above 75 percent of nominal.
To determine the two-phase mixture height in the core, a code (SWELL) was developed using the Wilson bubble-rise madel and a calculational procedure developed by CE for the Quad-Cities review. Preliminary calculations from this code would indicate a level swell due to core heat generat. ion of approximately 2-3 feet. B&W has performed similar calculations. This level swell when added to the 6 to 10 feet of collapsed water would permit operation above 757. power for Oconee 1.
In this respect we are using as 1007. of power the kw/ft ratings in the FSAR.
We next step in our review of operation of Oconee 1 at full power is to verify the level swell model and associated heat transfer which would exist in the quasi-equilibrium portion of this accident.
Duke Power is obtaining experimentri data for this purpose frosa a Westinghouse facility. his facility has a 500 rod full length
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bundle and is capable of pressures up to at least 400 psia.
Based on our review to date, we see no reason to limit the power level of Oconee 1 to below 757. until the experimental program and our subsequent review are completed. Until our densification review is complete, there is still a 757. limit on power. Should our densiff. cation review be completed prior to the final experimental resolution of the heat transfer-level swell problem, we will re-examine the necessity for a power level restriction between 75 and 1007. power.
Driginal Signe<! by yletor Stello Victor Stello. Assistant Director for Reactor Safety Directorate of Licensing i
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