Experimental investigations of two-phase mixture level swell and axial void fraction distribution under high pressure, low heat flux conditions in rod bundle geometry Page: 6 of 22
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5
The significan-e of these results can be understood through a simple
parametric study. A statement of the problem is as follows:
Given that a nuclear reactor is undergoing a slow, small
break loss of coolant accident, what fraction of initial
active core coolant inventory is required to assure that
uncovering of active core does not occur. Assumptions
and conditions that apply are:
a) Mixture swell is characterized by Eq. (6)
b) System pressure is 6.9 MPa
c) Vapor generation by flashing is small compared
to that by heat addition
d) Heat losses are negligible
e) Entire active core is in saturated boiling
f) ANS decay curve characterizes decay heat levels
g) Calculations are based on a 0.95-cm-diam fuel
rod and a pitch-to-diameter ratio of 1.34
h) Decay heat based initial power of 18.4 kW/m
Results of the study are displayed in Fig. 5. Time is defined with respect
to reactor SCRAM. The study indicates that in the time period of primary
interest in small break accidents, 1000 to 10,000 s after SCRAM, roughly 80
to 90% of the active core liquid inventory is required to maintain core
coverage. Also plotted in Fig. 5 is the increase in mixture level due to
level swell.
Figure 4 indicates that the level swell for the 8.0 MPa data is some-
what higher than the 4.0 MPa and Small Break Test Series I data for condi-
tions of similar vapor flux density. It is reasonable to ask what effect
this difference might have on the parametric study results. A line passing
through the origin and bounding the 8.0 MPa data has a slope of 0.0132/
(cm/s). ihe differences in mass inventory needed to maintain core coverage
that result from differences in slope are quite small for times greater
than or equal to 1000 s. Figure 5 shows that 88% of initial active core
liquid inventory is needed to maintain core coverage at 10,000 s. If the
slope is changed to 0.0132/(cm/.) the result is 86%. So while pressure may
have affected the amount of level swell at a given vapor flux density, the
calculated differences in core behavior were insignificant for the time
period of interest.
Despite the attractive simplicity of these results, they should not be
taken too literally. Recall that the THTF has an axially uniform power
profile. It is not known what effect, if any, the nonuniform power profile
of an operating reactor might have on level swell. However, these calcula-
tions should approximate the dependency of mixture swell on decay heat level.
In addition, the calculations form a basis for more detailed studies.
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Anklam, T. M. & White, M. D. Experimental investigations of two-phase mixture level swell and axial void fraction distribution under high pressure, low heat flux conditions in rod bundle geometry, article, January 1, 1981; Tennessee. (https://digital.library.unt.edu/ark:/67531/metadc1113391/m1/6/: accessed July 16, 2024), University of North Texas Libraries, UNT Digital Library, https://digital.library.unt.edu; crediting UNT Libraries Government Documents Department.