Uncertainty quantification in decay heat calculation of spent nuclear fuel by STREAM/RAST-K

This paper addresses the uncertainty quantification and sensitivity analysis of a depleted light-water fuel assembly of the Turkey Point-3 benchmark. The uncertainty of the fuel assembly decay heat and isotopic densities is quantified with respect to three different groups of diverse parameters: nuclear data, assembly design, and reactor core operation. The uncertainty propagation is conducted using a two-step analysis code system comprising the lattice code STREAM, nodal code RAST-K, and spent nuclear fuel module SNF through the random sampling of microscopic cross-sections, fuel rod sizes, number densities, reactor core total power, and temperature distributions. Overall, the statistical analysis of the calculated samples demonstrates that the decay heat uncertainty decreases with the cooling time. The nuclear data and assembly design parameters are proven to be the largest contributors to the decay heat uncertainty, whereas the reactor core power and inlet coolant temperature have a minor effect. The majority of the decay heat uncertainties are delivered by a small number of isotopes such as 241Am, 137Ba, 244Cm, 238Pu, and 90Y. (c) 2021 Korean Nuclear Society, Published by Elsevier Korea LLC. All rights reserved. This is an open access article under the CC BY-NC-ND license (http://creativecommons.org/licenses/by-nc-nd/4.0/)

Overcoming the challenges of Monte Carlo depletion: Application to a material-testing reactor with the MCS code

The theoretical aspects behind the reactor depletion capability of the Monte Carlo code MCS developed at the Ulsan National Institute of Science and Technology (UNIST) and practical results of this depletion feature for a Material-Testing Reactor (MTR) with plate-type fuel are described in this paper. A verification of MCS results is first performed against MCNP6 to confirm the suitability of MCS for the criticality and depletion analysis of the MTR. Then, the dependence of the effective neutron multiplication factor to the number of axial and radial depletion cells adopted in the fuel plates is performed with MCS in order to determine the minimum spatial segmentation of the fuel plates. Monte Carlo depletion results with 37,800 depletion cells are provided by MCS within acceptable calculation time and memory usage. The results show that at least 7 axial meshes per fuel plate are required to reach the same precision as the reference calculation whereas no significant differences are observed when modeling 1 or 10 radial meshes per fuel plate. This study demonstrates that MCS can address the need for Monte Carlo codes capable of providing reference solutions to complex reactor depletion problems with refined meshes for fuel management and research reactor applications. (c) 2020 Korean Nuclear Society, Published by Elsevier Korea LLC. This is an open access article under the CC BY-NC-ND license (http://creativecommons.org/licenses/by-nc-nd/4.0/)

Verification of adjoint-weighted tally calculation capability in MCS

A Monte Carlo method developed by B. C. Kiedrowski, performs adjoint-weighted tallies in continuous energy k-eigenvalue calculations. Each tally contribution, Tp, is weighted by estimating the iterated fission probability, Rp, which is proportional to the adjoint function having the progenitor index, p, in the asymptotic generation at the phase space, r. There is no need for any additional random walk, which means there is only small amount of increase in the computation time. The adjoint weighted tallies are used to calculate adjoint-weighted flux and point reactor kinetics parameters, such as the effective delayed neutron fraction, ??eff, and the prompt neutron generation time, A. This Monte Carlo method tallying adjoint-weighted parameters is implemented in MCS which is an in-house Monte Carlo code that has been developed at UNIST. The multi-energy group test problems are used to verify the adjoint-weighted tally calculation capability in MCS, and the results are compared with the result from the Monte Carlo N-Particle (MCNP). The continuous energy test problems are also used to verify the adjoint-weighted tally calculation capability, and the results are compared with the results from the McCARD. The comparison shows good agreement between MCS and the results from the other codes

Sensitivity and Uncertainty Analysis Capability in MCS for UAM Benchmark

S/U analysis has been performed with a Monte Carlo code MCS for TMI-1 PWR pin-cell problem in the UAM benchmark. The sensitivity coefficients of k-inf and one group microscopic cross sections are compared between MCS and SERPENT2. The sensitivity results show a good agreement. The uncertainties are calculated with 44 group covariance data libraries from ENDF/B-VII.1 and SCALE 6.1. The top contributors of uncertainties are also calculated

Monte Carlo Adjoint Eigenmode Calculation with the Application of Modified Power Method

This paper discusses about the possibility of the adjoint eigenmode calculation for the Monte Carlo neutron transport problem with the application of the modified power method. The modified power method has been studied to simulate several forward eigenmodes at the same time by introducing a weight vector to one neutron. The adjoint eigenmodes are introduced in this paper by adopting the bi-orthogonal property of the forward and adjoint eigenmodes. The preliminary results based on the BEAVRS 3D whole core model are very encouraging. Further study about the theory and implementation issues will be done to improve the performance of the method