Accident tolerant fuel cladding development: Promise, status, and challenges

Characterization of SiC–SiC composites for accident tolerant fuel cladding

Microstructure, corrosion resistance and oxidation behavior of Cr-coatings on Zircaloy-4 prepared by vacuum arc plasma deposition

Recent progress in the development of SiC composites for nuclear fusion applications

Stress analysis and probabilistic assessment of multi-layer SiC-based accident tolerant nuclear fuel cladding

Advanced reactor concepts such as Generation-IV (Gen-IV) have been studied to fulfill the ambitious long term goals of developing a safe, sustainable, reliable, proliferation-resistant and economic nuclear energy system. The gas-cooled fast reactor (GFR) is a Gen-IV candidate for which a carbide fuel (UC, (U,Pu)C, ThC) has desirable properties in a fuel system using ceramic (SiC) cladding. This study reviews advanced fuel concepts and associated fabrication methods for a GFR, followed by available carbide fuel property data, update of fuel performance code, and assessment of the fuel performance analysis model based on irradiation test results in the Fast Flux Test Facility. The purpose of this study is to develop a fuel performance code for the design and analysis of carbide fuel and to verify the implementation of fuel property models using earlier fuel irradiation tests jointly conducted by US and Switzerland. The comparison of carbide fuel simulation results to the experimental data shows differences in the prediction of fuel temperature and swelling. Considering the limitation and uncertainty of the existing material property data, the results obtained from the updated fuel performance code indicate that more work is required to adjust and update some of the carbide fuel material property models. In the future, the coolant model will also be updated for the flexibility of code application to various advanced fuel analyses. In the long term, the code will be used for the evaluation of advanced carbide fuel with ceramic cladding and the simulation of carbide fuel irradiation tests.Copyright © 2014 by ASME

Changes to Sustainability of Uranium Carbide Fuel for a Long-Burn Fast Reactor Core

Commercial Light Water reactor (LWR) nuclear power plants currently utilize Zircaloy-4 as the fuel cladding material. As seen during the Fukushima accident, zirconium in the cladding reacts exothermically with water to form free hydrogen gas if cooling is not maintained. To solve this problem, General Atomics (GA) is developing Accident Tolerant Fuel (ATF) [1], where SiGA TM (silicon carbide technology developed by GA) is used as the cladding material, which has strong advantages in nuclear applications due to its stability under long-term irradiation [2], its ability to retain essential mechanical strength at very high temperatures; and its resistance to oxidation in steam and air environments. Iron-chromium-aluminum (FeCrAl) alloys are also being considered as advanced fuel cladding materials owing to their slower oxidation kinetics and higher strength when compared to zirconium-based alloys at high temperatures [3]. In order to quantitatively evaluate the merits of advanced cladding materials, accident analyses have been conducted for small break loss-of-coolantaccident (SBLOCA), large break loss-of-coolant-accident (LBLOCA), station blackout (SBO) and reactivity initiated accident (RIA). For modeling advanced cladding materials, GA has updated fuel performance codes FRAPCON [4] and FRAPTRAN [5]. The accident analyses were conducted by TRACE [6] which is an advanced best-estimate reactor system code developed by the U.S. Nuclear Regulatory Commission (NRC) for analyzing transient and steady-state neutronic-thermal-hydraulic behavior in LWRs. The Symbolic Nuclear Analysis Package (SNAP) [7] was used as a suite of integrating and processing calculation results obtained from TRACE, FRAPCON and FRAPTRAN. The accident analyses used a benchmarked version of the H. B. Robinson (Unit 2) plant in Hartsville, South Carolina, which is a 3-loop Westinghouse LWR.

Joshua Stone

Papers

Characterization of SiC–SiC composites for accident tolerant fuel cladding

Stress analysis and probabilistic assessment of multi-layer SiC-based accident tolerant nuclear fuel cladding

Changes to Sustainability of Uranium Carbide Fuel for a Long-Burn Fast Reactor Core

Preliminary Evaluation of Advanced Cladding Material Performance Under Accident Conditions