Depletion chain optimization of lattice code STREAM for LWR fuel assembly burnup analysis. (January 2019)

• Record Type: Journal Article
• Authors: Nguyen, Khang Hoang Nhat; Choi, Sooyoung; Lemaire, Matthieu; Lee, Deokjung
• Is Part Of: Annals of nuclear energy. Volume 123(2019)
• Page Start: 18
• Page End: 45
• Publication Date: 2019-01
• Subjects: Depletion chain -- Optimization -- STREAM code -- Generalized perturbation theory; Nuclear energy -- Periodicals; Nuclear engineering -- Periodicals; 621.4805
• DOI: 10.1016/j.anucene.2018.08.030
• Languages: English
• ISSNs: 0306-4549

Abstract

Highlights: A method of burnup chain optimization is proposed for STREAM code. Verification is conducted on 17 LWR fuel assemblies (VERA benchmark and OPR-1000 type). Excellent agreement for reactivity and pin-by-pin power calculations. The optimized chain divides the total simulation time by 3–4 compared to the detailed chain. 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, B4 C and IFBA, and one OPR-1000 fuel assembly with gadolinium bearing fuel depletion problem. For burnup up to 80 MWd/kg and 235 U 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.