Water tests for determining post voiding behavior in the LMFB : final report
Author(s)Hinkle, William D.
The most serious of the postulated accidents considered in the design of the Liquid Metal Cooled Fast Breeder Reactor (LMFBR) is the Loss of Pipe Integrity (LOPI) accident. Analysis models used to calculate the consequences of this accident assume that once boiling is initiated film dryout occurs in the hot assembly as a result of rapid vapor bubble growth and consequent flow stoppage or reversal. However, this assumption has not been put to any real test. Once boiling is initiated in the hot assembly during an LMFBR LOPI accident, a substantial gravity pressure difference would exist between this assembly and other colder assemblies in the core. This condition would give rise to natural circulation flow boiling accompanied by pressure and flow oscillations. It is possible that such oscillations could prevent or delay dryout and provide substantial post-voiding heat removal. The tests described in this report were conceived with the objective of obtaining basic information and data relating to this possibility. To accomplish this objective a natural circulation test loop was designed to simulate LMFBR geometry and flow conditions predicted to exist at the time boiling is initiated in a LOPI accident. The test loop included: (l) a vertical tube test section, (2) upper and lower plenum tanks, (3) an external down-commer, (4) sight flow indicators and (5) instrumentation. The test section was an electrically heated tube designed with a hydraulic diameter and length similar to current LMFBR (FTR) design. The upper and lower plenum tanks were provided with means for controlling liquid subcooling above and below the test section. The down-commer was large enough to eliminate down-commer hydraulics. Water at a pressure of 1 atmosphere was used to simulate sodium. Sight flow indicators were provided to observe flow conditions at the test section inlet and exit. Instrumentation was provided to measure test section pressures, inlet and exit temperatures, tube wall temperatures, heat flux and oscillation frequencies. Steady state tests were conducted for subcooled flow boiling, saturated flow boiling, CHF and post CHF conditions. Subcooled flow boiling was observed for heat fluxes below 1 x 104 BTU/hr ft2. For this condition, both pressure oscillations and temperature oscillations at the heated surface were observed; but the pressure oscillations were not observed continuously. Saturated flow boiling was observed for heat fluxes between 3 x 104 BTU/hr ft2 and CHF. For this condition, pressure oscillations were observed continuously. As the CHF condition was approached, a periodic downward expansion of vapor from the heated section was observed at the bottom sight flow indicator and the flow regime appeared to be annular at the top sight flow indicator. CHF was observed at the top of the heated section when the heat flux reached 6.4 x 104 BTU/hr ft2, but rewetting occurred after a few seconds. As the heat flux was increased further, the maximum surface temperature reached before rewetting increased; until, at a heat flux of 7.15 x 104 BTU/hr ft2, the maximum temperature exceeded 9000 F and rewetting no longer occurred. A transient test was conducted for a post CHF condition. The heat flux was 7.3 x 104 BTU/hr ft2. The oscillations observed under steady state conditions developed within a few seconds after the power was turned on. The equilibrium tube wall temperature upstream of the CHF location was reached in 10 seconds. The equilibrium tube wall temperature at the CHF location was reached in about 135 seconds. A similarity analysis was done in order to scale the test results to LMFBR LOPI conditions. The results of this analysis indicate that the CHF for the LMFBR (FTR) would be at least 6 x 104 BTU/hr ft2. This corresponds to a critical average linear power for the hot assembly of 1.06 kw/ft compared to an estimated 2.55 to 5.1 kw/ft being transferred to the coolant at the time boiling begins during a LOPI accident. On the basis of this analysis, the results of the water tests indicate that CHF would occur. But, this conclusion is conservative for a number of reasons and further experimental work on a more prototypical system is suggested.
MIT Energy Lab
Liquid metal cooled reactors, Fast reactors, Breeder reactors, Nuclear reactors -- Safety measures
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