Comparisons of Deterministic Neutronic Calculations with Monte Carlo Results for an Advanced BWR Fuel Assembly with Hafnium Control Blades

Abstract

From the neutronic viewpoint, the optimization of BWR core designs is strongly related to the accurate determination of flux variations inside and around fuel assemblies. These fluctuations, which are mainly due to the high heterogeneity of the fuel and moderator regions, as additionally to the presence of cruciform absorber blades, have a direct impact on reactor safety and performance. Of particular importance is the pin power distribution, leading to the need of assessing the capabilities of design tools in a sufficiently rigorous manner. The basic configuration chosen for the code comparisons corresponds to a SVEA-96 fuel assembly under full-density water moderation conditions, with inserted hafnium absorber blades. The calculational schemes employed are the Monte Carlo code MCNPX2.5, in conjunction with various nuclear data libraries (ENDF/B-VI, JEF2.2, JEFF3.0, JENDL-3.2 and JENDL3.3), and the deterministic codes CASMO4 with JEF2.2, BOXER with JEF1.0 and HELIOS1.6 with ENDF/B-VI based libraries, respectively. The significant discrepancies observed in k 1 predictions (>500 pcm) are found to be mainly nuclear data related. On the other hand, data library effects have been found to be quite small for the prediction of pin-wise distributions of total fissions (F tot ), 238 U captures (C 8 ), as also of the C 8 =F tot ratio. Significant differences in these reaction rate distributions (up to several percent) have, however, been observed between the Monte Carlo and deterministic calculations, particularly in the vicinity of the hafnium blades and in the gadolinium pins.

Experimental Validation of Reaction Rate Distributions for a SVEA-96+ BWR Assembly with Hafnium Control Blades

“…As discussed in the purely calculational comparisons reported earlier 7) for the corresponding reflected-assembly configuration, these difficulties stem mainly from (a) the overall flux and power tilt across the assembly from the farthest rod (in the sense of rod distance to the absorber vertex) to the closest rod, and, (b) the strong neutron spectrum and flux variations superimposed upon this tilt, as a fine flux structure, due to the assembly's own inherent material and geometrical heterogeneities (wide range of fuel rod types, water channel, wings, etc. ).…”

“…The relatively low sensitivity to nuclear data libraries (<1:5% for the F tot distribution in individual pins) was discussed in some detail in the recently reported calculational benchmark exercise. 7) In order to assess the sensitivity of the results to a small displacement of the Hf control blades, an additional MCNPX2.5 simulation has been performed, in which a single modification was made in the model, viz., the displacement of the blade that is parallel to column 10 by 1 mm in the direction of the Hf-vertex. The results are found to be significantly affected for the prediction of the F tot distribution, but there is little impact on the C 8 distribution.…”

“…These factors effectively correspond to pinspecific ratios of the F tot (or C 8 ) values as obtained using a whole-core LWR-PROTEUS I-2A model with a given code, relative to the corresponding values obtained from a reflected assembly model (based on the same code) with a half-thickness of the Hf blades. 7 These FCCFs are thus an index of the representativity of the PROTEUS central test zone from a neutronics viewpoint. Based on the F tot MCNPX2.5 results, the main correction factor ($1:2, i.e., about 20% correction) corresponds to the J10 location (Hf-vertex).…”

“…The use of such a range of recently developed cross section files presents a significant advantage in assessing the influence of different types of data on the calculated results. 7) 2. Deterministic Codes PHOENIX4, BOXER, CASMO4, and HELIOS1.6 PHOENIX4 is a lattice code allowing detailed transport calculations at 2D level.…”

“…The contribution of uncertainty from evaluation and processing of basic nuclear data (Point 2) were investigated in a previous calculation benchmark exercise and was concluded to be of minor importance. 7) From the results of the present code validation exercise against PROTEUS experimental data, the combined accuracy of basic nuclear data (Point 1) and calculation methodologies (Point 4) will be assessed.…”