The fuel form developed for the Transformational Challenge Reactor demonstration program leverages recent advances in manufacturing, materials, and computational sciences, delivering a new architecture for production of high-performance microencapsulated nuclear fuels. The fuel consists of conventionally manufactured uranium nitride tristructural isotropic fuel particles embedded inside a 3D-printed silicon carbide matrix. This paper describes the overall architecture and manufacturing process for this fuel form, its properties and behavior, and the ongoing development activities.

•Impact of control blade position on the deformation behavior of SiC-SiC channel box was analyzed.•Large gradients in fast neutron flux up to 35–40% will develop across the channel box.•The bending behavior of the channel box depends strongly on the control blade position.•The stress development in the channel box for the three positions of control blade was analyzed.

The Transformational Challenge Reactor (TCR) is a reactor technology development project led by Oak Ridge National Laboratory. The project is focused on adoption of technology advancements, particularly advanced manufacturing, into nuclear application. The seminal goal of this project is a brief operation of a micro-scale reactor made with novel core materials. A rendering of the TCR core, vessel, and shielding structures can be seen in Figure 1 [1]. The TCR core is cooled by inert gas and comprised of various novel materials such as yttrium hydride moderator and additively manufactured structural components. The focus of this paper, however, is upon TCR’s unique fuel form.

This research presents BISON simulations of stress, temperature, and failure probability of tristructural isotropic (TRISO) fuel particles subjected to transient power pulse conditions in the Nuclear Safety Research Reactor (NSRR). By modifying the default elastic properties of the PyC and the UO2 coefficient of thermal expansion correlation, BISON was found to produce suitable agreement with the observed and independently simulated results when appropriate shape and scale Weibull parameters for SiC failure were chosen. The Weibull parameter effects were also explored within the range of SiC failure and were applied using experimental data from hemispherical crush testing.

INTRODUCTION The Transformational Challenge Reactor (TCR) is a helium-cooled, yttrium-hydride moderated reactor being designed for the US Department of Energy Office of Nuclear Energy to demonstrate and deploy advanced manufacturing technologies in nuclear design and manufacturing processes [1]. To inform the TCR safety analysis, sample fuel elements will be placed in Idaho National Laboratory’s (INL) Transient Reactor Test Facility (TREAT) and exposed to pulses representative of TCR reactivity-insertion accident (RIA) conditions. Using TRACE and RELAP5-3D, we calculated fuel temperatures and average energy deposition in TCR fuel at both hot full power (HFP) and hot zero power (HZP) for both sub-prompt and prompt reactivity insertions. Though RELAP has been used for gas-cooled reactor safety analysis in the past [2], the prompt generation time in TCR is an order of magnitude shorter than for other gas-cooled reactors due to the use of yttrium-hydride instead of graphite as the moderator [3]. The TREAT facility at INL is an air-cooled, graphitemoderated reactor fueled with uranium oxide dispersed into the graphite blocks. It has been used for fuel testing in both fast and thermal reactors, and following its nearly 25-year shutdown, the facility has resumed operation and demonstrated an ability to match historical tests. This provides confidence in the TREAT team’s ability to design tests for a wide range of transients including pulses representative of reactivity insertions terminated by either temperature feedback or negative reactivity insertion and pulses providing a power profile representative of a loss of coolant accident in a light-water reactor [4]. The demonstrated capabilities of TREAT for both fast and thermal reactors give us confidence that TREAT can create pulses that mimic anticipated TCR transients. To gain a better understanding of the impact of our transients on the heterogeneous fuel form, the RELAP boundary conditions were also used to inform a thermomechanical analysis using the fuel performance code

This article presents calculated radionuclide release fractions from TRISO fuel during post-irradiation annealing and comparison with AGR-1 irradiated TRISO compacts using the BISON fuel performance code. The BISON code was found to produce good agreement in an analytical comparison and with legacy German TRISO compact irradiation data. However, the recommended IAEA fission product diffusion coefficients did not produce good agreement with the 2014 AGR-1 data (Demkowicz et al., 2015). Thus, the diffusion coefficients prefactors and activation energies were modified to obtain best agreement with the recent AGR data. Future work involves investigating the release behavior of silver from silicon carbide and implementing methods to provide predictive capability for the silver release behavior.

Silicon carbide fiber-reinforced silicon carbide matrix (SiC-SiC) composites are being considered as components in light water reactor cores to improve accident tolerance, including channel boxes and fuel cladding. In the nuclear reactor environment, core components like a channel box will be exposed to neutron and other radiation damage and temperature gradients. To ensure reliable and safe operation of a SiC-SiC channel box, it is important to assess its deformation behavior under in-reactor conditions including the expected neutron flux and temperature distributions. In particular, this work has evaluated the effect of non-uniform dimensional changes caused by spatially varying neutron flux and temperatures on the deformation behavior of the channel box over the course of one year. These analyses have been performed using the fuel performance modeling code BISON and the commercial finite element analysis code Abaqus, based on fast flux and temperature boundary conditions that have been calculated using the neutronics and thermal-hydraulics codes Serpent and CTF, respectively. The dependence of dimensions and thermophysical properties on fast flux and temperature has been incorporated into the material models. These initial results indicate significant bowing of the channel box with a lateral displacement greater than 6.5 mm. The channel box bowing behavior is time dependent and driven by the temperature dependence of the SiC irradiation-induced swelling and the neutron flux/fluence gradients. The bowing behavior gradually recovers during the course of the operating cycle as the swelling of the SiC-SiC material saturates. However, the bending relaxation due to temperature gradients does not fully recover and residual bending remains after the swelling saturates in the entire channel box. (C) 2018 Elsevier B.V. All rights reserved.

Fully ceramic microencapsulated (FCM) fuel is a proposed fuel type for improved accident performance in LWRs (Light Water Reactors) that involves TRISO (TRistructural-ISOtropic) particles embedded in a nano-powder sintered silicon carbide (SiC) matrix. The TRISO particles contain a spherical fuel kernel ranging from 500 to 800 mu m in diameter. The kernel and buffer layer are then coated with three layers, each of which is 30-40 mu m thick, composed of dense inner pyrolytic carbon (IPyC), chemically vapor deposited silicon carbide (SiC) layer, and an outer pyrolytic carbon (OPyC) layer. These TRISO particles are then embedded in a fully dense sintered SiC matrix with an expected particle packing fraction of about 35-40% by volume. As is the case for gas reactor applications, the release of radioactivity into the coolant is dependent on the integrity of the silicon carbide layer of the TRISO particles, in addition to the SiC matrix. In this work, we report on fuel performance modeling of TRISO-bearing FCM fuel using the BISON code to simulate the thermo-mechanical behavior of this fuel in a prototypic LWR environment. This paper considers the effects of embedding a TRISO particle in the SiC pellet matrix and includes a discussion of the irradiation-induced dimensional change in the pyrolytic carbon (PyC) layers of the TRISO particle. Additionally, methods were developed to simulate a FCM pellet containing a large number of discrete and independent particles. Future work will report on developing an interface debonding model, a fracture model, and a radionuclide transport model.