Science and Technology of Nuclear Installations

Volume 2017, Article ID 3146985, 12 pages

https://doi.org/10.1155/2017/3146985

## Neutronic Analysis on Potential Accident Tolerant Fuel-Cladding Combination U_{3}Si_{2}-FeCrAl

Sino-French Institute of Nuclear Engineering and Technology, Sun Yat-sen University, Zhuhai, Guangdong 519082, China

Correspondence should be addressed to Cenxi Yuan; nc.ude.usys.liam@xcnauy

Received 23 September 2016; Revised 9 December 2016; Accepted 25 December 2016; Published 18 January 2017

Academic Editor: Tomasz Kozlowski

Copyright © 2017 Shengli Chen and Cenxi Yuan. This is an open access article distributed under the Creative Commons Attribution License, which permits unrestricted use, distribution, and reproduction in any medium, provided the original work is properly cited.

#### Abstract

Neutronic performance is investigated for a potential accident tolerant fuel (ATF), which consists of U_{3}Si_{2} fuel and FeCrAl cladding. In comparison with current UO_{2}-Zr system, FeCrAl has a better oxidation resistance but a larger thermal neutron absorption cross section. U_{3}Si_{2} has a higher thermal conductivity and a higher uranium density, which can compensate the reactivity suppressed by FeCrAl. Based on neutronic investigations, a possible U_{3}Si_{2}-FeCrAl fuel-cladding system is taken into consideration. Fundamental properties of the suggested fuel-cladding combination are investigated in a fuel assembly. These properties include moderator and fuel temperature coefficients, control rods worth, radial power distribution (in a fuel rod), and different void reactivity coefficients. The present work proves that the new combination has less reactivity variation during its service lifetime. Although, compared with the current system, it has a little larger deviation on power distribution and a little less negative temperature coefficient and void reactivity coefficient and its control rods worth is less important, variations of these parameters are less important during the service lifetime of fuel. Hence, U_{3}Si_{2}-FeCrAl system is a potential ATF candidate from a neutronic view.

#### 1. Introduction

After the Fukushima nuclear disaster in 2011, extensive focuses on the accident tolerant fuel (ATF) have been developed into seeking advanced nuclear fuel and cladding options. Many materials for fuel and cladding have been investigated over the past years. Three potential approaches have been proposed for the development of the fuel and cladding with enhanced accident tolerance [1]:(1)Modifications of current zircaloy alloy in order to improve the oxidation resistance, including the coating layer design(2)Replacement of zircaloy cladding by an alternative high-performance oxidation resistant cladding(3)Improvement or replacement of the ceramic oxide fuel

The cladding material should have good oxidation resistance, proper delay of the ballooning and burst [2], stress resistance, and small thermal neutron absorption cross section. Among a large number of candidates, stainless steels have better mechanic performance than the current zircaloy-4 alloy. FeCrAl is a potentially promising excellent cladding material [1, 3, 4]. For example, its oxidation rate is at least two orders of magnitude lower than that of zircaloy [1]. On the other hand, FeCrAl is better to be applied as monolithic cladding than coating, in consideration of matching the thermal expansion coefficient [2], the diametrical compression [2], the volumetric and microstructural evolution [2], the high temperature oxidation protection [5], and the interdiffusion with zirconium [6], which is the reason that the material, Cr coated zircaloy, which is of better mechanic performance, has not been selected [7]. In the present work, monolithic FeCrAl is chosen as the potential ATF cladding material.

Many properties have to be considered for the investigation of fuel materials, such as the heavy metal density, the melting point, and the thermal conductivity. Uranium mononitride (UN) based composite fuels may have potential benefits when applied in light water reactor because of its enhanced thermal conductivity and large fuel density. However, UN chemically reacts with water [8], especially at high temperature. Additional shielding material UN/U_{3}Si_{5} has been studied to overcome the defect [9]. But a problem still exists, which is the determination of the percentage of U_{3}Si_{5} to prevent the fuel from reacting with water in an accident condition. One of the possible solutions is the tristructural-isotropic (TRISO) fuel design [10, 11].

Uranium-silicon binary system, which is thermodynamically stable, is another potential fuel [12]. Among the multiple compounds, U_{3}Si and U_{3}Si_{2} are the best candidates due to their high uranium densities. However, U_{3}Si swells considerably under irradiation and dissociates into U_{3}Si_{2} and solid solution U above 900°C, which is below some possible temperatures in uranium silicide fueled pins [13]. U_{3}Si_{2} has promising records under irradiation in research reactor fuels and maintains several advantageous properties compared with UO_{2}.

In consideration of neutronic performance, the thermal neutron absorption cross section of FeCrAl is 2.43 barns, while that of zircaloy is 0.20 barns [14]. In order to compensate the larger cross section, one needs to decrease the thickness of cladding and/or increase the quantity of fissile nuclides in the fuel. Under economical and safety considerations, the present work chooses nuclear fuel U_{3}Si_{2} which has higher uranium density than the current UO_{2} fuel.

Since the neutronic performance of FeCrAl is different from current zircaloy and U_{3}Si_{2} has different uranium percentage from UO_{2}, analysis on neutronic performance is necessary for the new fuel-cladding system. Neutronic analyses on U_{3}Si_{2}-FeCrAl, U_{3}Si_{2}-SiC, and UO_{2}-Zr are performed based on I^{2}S-LWR [15]. In addition, in order to study the core performance under normal and accident conditions, it is also necessary to investigate the fundamental properties of nuclear reactor core besides the neutron economy analysis, such as moderator and fuel temperature coefficients, control rods worth, radial power distribution, and different void reactivity coefficients.

#### 2. Methodology and Input Parameters

The neutronic behavior is performed using TRITON and KENO-VI modules from SCALE 6.1 [16–18]. SCALE provides a “plug-and-play” framework with 89 computational modules, including three Monte Carlo and three deterministic radiation transport solvers. It includes state-of-the-art nuclear data libraries and problem-dependent processing tools for both continuous-energy and multigroup neutronic calculations and multigroup coupled neutron-gamma calculations, as well as activation and decay calculations. TRITION is a module for isotopic depletion calculation. KENO-VI is used for critical calculation. The nuclear data used in our simulation is based on ENDF/B-VII.0 238-group neutron library [19].

Monte Carlo codes are more reliable for radial distribution calculation than the deterministic codes due to the self-shielding. The advantage of Monte Carlo codes is especially evident in such calculation for a new fuel-cladding combination considering the less correction formula existing in deterministic codes. The radial power distribution is calculated by the Monte Carlo based code RMC [20]. RMC is a 3D Monte Carlo neutron transport code developed by Tsinghua University. The code RMC intends to solve comprehensive problems in reactor, especially the problems on reactor physics. It is able to deal with complex geometry, using continuous-energy pointwise cross sections ENFF/B-VII.0 for different materials and at different temperatures. It can carry out both criticality and burnup calculations, which help to obtain the effective multiplication factor and the isotopic concentrations at different burnup level. Monte Carlo method is also used in fundamental nuclear physics. Nuclear structure problem can be solved with traditional nuclear shell model [21, 22] or Monte Carlo shell model [23].

##### 2.1. Geometric Parameters and Model Description

The Westinghouse 17 × 17 assembly design is used, as shown in Figure 1. Larger rings placed within the lattice represent the guide tubes and instrumentation tube. Control rods are inserted in the 24 tubes except the center tube for instrumentation. When no control rods or instruments are inserted, these tubes are filled with moderator. All the simulations are based on one assembly unit, except the calculation on radial power distribution. The infinite lattice cell is used to calculate the effective resonant cross section.