Burnup estimation for plate type fuel assembly of research reactors through the least square fitting method Luay M. Alawneh b , Chang Je Park a,⇑ , Mustafa K. Jaradat b , Byungchul Lee b a Sejong University, Department of Nuclear Engineering, 209 Neungdong-ro, Gwangjin-gu, Seoul 143-747, Republic of Korea b Korea Atomic Energy Research Institute, 1045 Daeduk-Daero, Dukjin-Dong, Yuseong-Gu, Daejeon 305-353, Republic of Korea article info Article history: Received 28 January 2014 Received in revised form 21 March 2014 Accepted 22 March 2014 Keywords: Burnup Plate type fuel Research reactor Least square fitting Monte Carlo SCALE abstract This work is focused on estimation of burnup for a plate type fuel assembly of research reactors with the SCALE6 code sequences such as TRITON/NEWT and ORIGEN-ARP. And a simple and accurate model is pro- posed to calculate burnup based on the least square fitting method without additional depletion analyses. One fuel assembly is modeled and its burnup is obtained for different power densities, enrichments, and fuel densities. Linear and non-linear polynomial fitting methods are used to provide a suitable formula for the burnup of the plate type fuel assembly with a function of different parameters. And the proposed approach is applied to three configurations of research reactors, such as a 5 MW core of U 3 Si 2 fuel, 3 MW cores with U 3 Si 2 and UMo fuels. The results are also validated by comparing those of Monte Carlo codes. Ó 2014 Elsevier Ltd. All rights reserved. 1. Introduction In order to evaluate the fuel performance and characteristics in the reactor, it is very important to estimate accurately the fuel dis- charge burnup. Fuel burnup is defined as the amount of energy (usually heat) generated per metric ton of all uranium and pluto- nium isotopes contained in the fuel charged into a reactor (Kulikowska and Szezesna, 1984). And it is an important quantity for design and operation of reactors from a standpoint of safety as well as operability (Inoue et al., 1969). It is widely known that the uranium oxide fuel in normal com- mercial light water reactors approaches the 40–60 GWD/MTU. However, in the research reactor, the burnup is changeable due to different types of fuels such as U 3 Si 2 ,U 3 Si, U-Al, and U-Mo fuels. Generally, the metal fuels of research reactors provide higher bur- nup than the existing uranium oxide fuel of LWR. The discharge burnup is nearly 100 GWD/MTU, which is mainly resulting from the higher power density in research reactors. Therefore, it is important work to estimate accurately the fuel burnup for the safety analysis and the fuel performance analysis. (Ravnik, 1992; Hussein et al., 2011) In this study, linear and non-linear formulae for burnups of plate type fuel assemblies are suggested with the least square fitting method. This approach enables us to estimate burnup di- rectly without following the detail fuel history including depletion analysis. Accurate burnup estimation is not an easy job due to sev- eral reasons such as the effect of fission products and the power change caused by refueling and depletion. In general, power den- sity, uranium enrichment, and fuel density are key factors of the fuel burnup. The sensitivity of each factor has been investigated, and then their effects are combined into one fitted formula for each burnup step. Several code systems are used to estimate the discharge burnup such as SCALE6 (Bowman, 2007) code system including TRITON/ NEWT (Dehart, 2009), ORIGEN-ARP (Gauld et al., 2009a) and Monte Carlo codes such as McCARD (Shim et al., 2012) and MCNPX (Brat- ton, 2012; Pelowitz, 2010). The ORIGEN-ARP is the SCALE6 depletion analysis sequence used to perform point-depletion calculations with the ORIGEN-S (Gauld et al., 2009b) code by using problem-dependent cross sections. The NEWT (Dehart, 2005) computer code is a multigroup discrete-ordinates radiation transport code with flexible meshing capabilities that allow two-dimensional (2-D) neutron transport calculations using complex geometric models. The TRITON (Dehart, 2009) is a SCALE control module that enables depletion calcula- tions to be performed by coordinating iterative calls between cross-section processing codes, NEWT, and the ORIGEN-S point- depletion code. The McCARD (Shim et al., 2012) is a Monte Carlo (MC) neutron-photon transport simulation code. It is capable of performing the whole core neutronics calculations, the reactor fuel http://dx.doi.org/10.1016/j.anucene.2014.03.029 0306-4549/Ó 2014 Elsevier Ltd. All rights reserved. ⇑ Corresponding author. Tel.: +82 2 3408 4432. E-mail address: parkcj@sejong.ac.kr (C.J. Park). Annals of Nuclear Energy 71 (2014) 37–45 Contents lists available at ScienceDirect Annals of Nuclear Energy journal homepage: www.elsevier.com/locate/anucene