Mb. Carver et al., VALIDATION OF THE ASSERT SUBCHANNEL CODE - PREDICTION OF CRITICAL HEAT-FLUX IN STANDARD AND NONSTANDARD CANDU BUNDLE GEOMETRIES, Nuclear technology, 112(3), 1995, pp. 299-314
The ASSERT code has been developed to address the three-dimensional co
mputation of flow and phase distribution and fuel element surface temp
eratures within the horizontal subchannels of Canada uranium deuterium
(CANDU) pressurized heavy water reactor fuel channels and to provide
a detailed prediction of critical heat flux (CHF) distribution through
out the bundle. The ASSERT subchannel code has been validated extensiv
ely against a wide repertoire of experiments; its combination of three
-dimensional prediction of local flow conditions with a comprehensive
method of predicting CHF at these local conditions makes it a unique t
ool for predicting CHF for situations outside the existing experimenta
l database. In particular, ASSERT is an appropriate tool to systematic
ally investigate CHF under conditions of local geometric variations, s
uch as pressure tube creep and fuel element strain. The numerical meth
odology used in ASSERT, the constitutive relationships incorporated, a
nd the CHF assessment methodology are discussed. The evolutionary vali
dation plan is also discussed, and early validation exercises are summ
arized. More recent validation exercises in standard and nonstandard g
eometries are emphasized.