• Nuclear Engineering and Technology
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• 제55권2호
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• pp.749-755
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• 2023

A new approximation method was proposed for treating the non-resonant spatial self-shielding effects of double heterogeneous region such as the double heterogeneous effect of VHTR fuel compact in the thermal energy range and that of BP compact with BISO. The method was developed based on the effective homogenization method and a spherical unit cell model with explicit coated layers and a matrix layer. The self-shielding factor was derived from the relation between the collision probabilities for a double heterogeneous compact and the effective cross section for the homogenized compact. First, the collision probabilities and transmission probabilities for all layers of the spherical model were calculated using conventional collision probability solver. Then, the effective cross section for the homogenized sphere cell representing the homogenized compact was obtained from the transmission probability calculated using the probability density function of a chord length. The verification calculations revealed that the proposed method can predict the self-shielding factor with a maximum error of 2.3% and the double heterogeneous effect with a maximum error of 200 pcm in the typical VHTR problems with various packing fractions and BP compact sizes.

• Nuclear Engineering and Technology
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• 제52권2호
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• pp.238-247
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• 2020

The Burnable Poison (BP) is very important for all Light Water Reactors in order to hold-down the initial excess reactivity and to control power peaking. The use of BP is even more essential as the excess reactivity increases significantly with a longer operation cycle. In this paper a feasibility study was conducted in order to investigate the benefits of a new combinational BP concept designed for 24-month cycle PWR core. The reference designs in this study are based on the two Korean fuel assemblies; 17 × 17 Westinghouse (WH) design and 16 × 16 Combustion Engineering (CE) design. A modification was done on these two designs to extend their cycle length from 18 months into 24 months. DeCART2D-MASTER code system was used to perform assembly and core calculations for both designs. A preliminary test was conducted in order to choose the best BP suitable for 24-month as a representative for single BP concept. The comparison between the results of two concepts (combinational BP concept and single BP concept) showed that the combinational BP concept can replace the single BP concept with better performance on holding down the initial excess reactivity without violating the design limitations.

• Nuclear Engineering and Technology
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• 제51권5호
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• pp.1195-1208
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• 2019

This paper introduces a new two-step procedure for PWR depletion analyses. This procedure adopts the albedo-corrected parameterized equivalence constants (APEC) method to correct the lattice-based raw cross sections (XSs) and discontinuity factors (DFs) by accounting for neutron leakage. The intrinsic limitations of the conventional two-step methods are discussed by analyzing a 2-dimensional SMR with the commercial DeCART2D/MASTER code system. For a full-scope development of the APEC correction, the MASTER nodal code was modified so that the group constants can be corrected in the middle of a microscopic core depletion. The basic APEC methodology is described and color-set problems are defined to determine the APEC functions for burnup-dependent XS and DF corrections. Then the new two-step method was applied to depletion analyses of the SMR without thermal feedback, and its validity was evaluated in terms of being able to predict accurately the reactor eigenvalue and nodal power profile. In addition, four variants of the original SMR core were also analyzed for a further evaluation of the APEC-assisted depletion. In this work, several combinations of the burnup-dependent and -independent XS and DF corrections were also considered. The results show that the APEC method could enhance the nodal equivalence significantly with inexpensive additional costs.

• Nuclear Engineering and Technology
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• 제48권3호
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• pp.624-634
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• 2016

Nuclear criticality safety analyses (NCSAs) considering burnup credit were performed for the GBC-32 cask. The used nuclear fuel assemblies (UNFAs) discharged from Hanbit Nuclear Power Plant Unit 3 Cycle 6 were loaded into the cask. Their axial burnup distributions and average discharge burnups were evaluated using the DeCART and Multi-purpose Analyzer for Static and Transient Effects of Reactors (MASTER) codes, and NCSAs were performed using SCALE 6.1/STandardized Analysis of Reactivity for Burnup Credit using SCALE (STARBUCS) and Monte Carlo N-Particle transport code, version 6 (MCNP 6). The axial burnup distributions were determined for 20 UNFAs with various initial enrichments and burnups, which were applied to the criticality analysis for the cask system. The UNFAs for 20- and 30-year cooling times were assumed to be stored in the cask. The criticality analyses indicated that $k_{eff}$ values for UNFAs with nonuniform axial burnup distributions were larger than those with a uniform distribution, that is, the end effects were positive but much smaller than those with the reference distribution. The axial burnup distributions for 20 UNFAs had shapes that were more symmetrical with a less steep gradient in the upper region than the reference ones of the United States Department of Energy. These differences in the axial burnup distributions resulted in a significant reduction in end effects compared with the reference.