• Energy Research

Abstract This study presents the first application of the advanced Rossi-alpha method (theoretically introduced by Kong et al., 2014) on the reactivity measurements in a research reactor: detector count signals at the Kyoto University Critical Assembly (KUCA) facility. The detector signals in the KUCA A-type core are analyzed by three subcriticality measurement methods: (1) Feynman-alpha (F-α) method, (2) Rossi-alpha (R-α) method, and (3) advanced Rossi-alpha (advanced R-α) method. Four cases are analyzed for two different subcritical states of the core and two different neutron source locations. Two different negative reactivity ρ values are obtained by the measurements of control rod worth and regarded as the reference reactivity values, comparing the results by the four methods. The F-α shows reactivity errors ranging between 7.1 and 7.3% due to its use of variance-to-mean ratios of detector count signals, which are not very sensitive to neutron background noise. However, the fitting uncertainties associated to the F-α results are large, ranging between 5.4 and 12.8% at one standard deviation. The R-α shows small fitting uncertainties ranging between 2.8 and 3.8%, although reactivity errors are in the range of 3.5–26.5% due to the neutron background noise. Finally, the advanced R-α that explicitly models the neutron background noise contrary to the previous methods shows the reactivity errors in the range of 1.0–11.8%, and provides the lowest uncertainties of the measured ρ in the range of 0.4–0.9%. In conclusion, among the four methods applied to the reactivity measurements at KUCA, the advanced R-α reveals the best accuracy with the lowest uncertainties.

Abstract A depletion chain simplification method is applied to UNIST lattice code STREAM (Steady state and Transient REactor Analysis code with Method of Characteristics) in this paper to alleviate the computational burden of depletion calculation associated with a large depletion matrix. A simplified burnup matrix (burnup chain) of 464 nuclides and 10,638 transitions is thus created from a large detailed burnup matrix containing 3,837 nuclides and 43,416 transitions from ENDF/B-VII.0 nuclear decay data. The simplified burnup matrix is optimized for the purpose of calculating effective neutron multiplication factors (keff) and power distributions with reduced computation time and memory usage. The nuclide selection method relies on the Generalized Perturbation Theory (GPT): by exploiting the adjoint function of nuclide number densities as derived with GPT, a set of nuclides which must be included in the simplified chain (nuclides with important contribution to reactivity) is determined. Numerical verification of the simplified burnup chain is conducted using the deterministic neutron transport analysis code STREAM, developed to perform whole light water reactor (LWR) core calculations with the direct transport analysis method and the two-step method. The simplified burnup chain is tested on the 16 LWR fuel assembly depletion problems from the Virtual Environment for Reactor Application (VERA) benchmarks, including burnable poison (BP) pin cells commonly used in nuclear reactor design such as gadolinia, Pyrex, AIC, B4C and IFBA, and one OPR-1000 fuel assembly with gadolinium bearing fuel depletion problem. For burnup up to 80 MWd/kg and 235U enrichment ranging from 2.1 w/o to 4.6 w/o, the calculations with the simplified burnup chain predict the keff variations within 10 pcm and identical power distributions, with a speed-up factor from 20 to 40 in depletion calculation and a reduction factor by 3–4 of the total simulation time compared to the ones with the detailed burnup chain.