Modeling spent TRISO fuel for geological disposal: corrosion and failure under oxidizing conditions in the presence of water

High temperature gas reactors (HTGRs) are being considered for near term deployment in the United States under the GNEP program and farther term deployment under the Gen IV reactor design (U.S. DOE Nuclear Energy Research Advisory Committee, 2002). A common factor among current HTGR (prismatic or pebble) designs is the use of TRISO coated particle fuel. TRISO refers to the three types of coating layers (pyrolytic carbon, porous carbon, and silicon carbide) around the fuel kernel, which is both protected and contained by the layers. While there have been a number of reactors operated with coated particle fuel, and extensive amount of research has gone into designing new HTGRs, little work has been done on modeling and analysing the degradation rates of spent TRISO fuel for permanent geological disposal. An integral part of developing a spent fuel degradation modeling was to analyze the waste form without taking any consideration for engineering barriers. A basic model was developed to simulate the time to failure of spent TRISO fuel in a repository environment. Preliminary verification of the model was performed with comparison to output from a proprietary model called GARGOYLE that was also used to model degradation rates of TRISO fuel. A sensitivity study was performed to determine which fuel and repository parameters had the most significant effect on the predicted time to fuel particle failure. Results of the analysis indicate corrosion rates and thicknesses of the outer pyrolytic carbon and silicon carbide layers, along with the time dependent temperature of the spent fuel in the repository environment, have a significant effect on the time to particle failure. The thicknesses of the kernel, buffer, and IPyC layers along with the strength of the SiC layer and the pressure in the TRISO particle did not significantly alter the results from the model. It can be concluded that a better understanding of the corrosion rates of the OPyC and SiC layers, along with increasing the quality control of the OPyC and SiC layer thicknesses, can significantly reduce uncertainty in estimates of the time to failure of spent TRISO fuel in a repository environment.     

乏燃料后处理厂退役用远距离操作技术探索

随着我国大型遗留核燃料后处理设施退役治理工作的按序推进,现已进入退役关键阶段,为使其中强放射性区域安全、顺利实施退役,研究、摸索和掌握远距离操作应用技术,良好的退役设计与策划,是推进退役事业、使之具备工作条件和能力的先决条件。由于我国尚未建立乏燃料后处理厂退役用远距离操作的相关标准体系,本文首次依据对我国遗留后处理厂现状特点,深入剖析典型退役难点,并参照国外同类型工程远距离操作经验提出了退役用远距离操作的总体设计要求,可以作为设计远距离操作技术决策的重要依据。     

Prediction of temperature and water level in a spent fuel pit during loss of all AC power supplies

A prediction method for water temperature in a spent fuel pit of a pressurized water reactor (PWR) has been developed to calculate the increase in water temperature during the shutdown of cooling systems. In this study, the prediction method was extended to calculate the water level in a spent fuel pit during loss of all AC power supplies, and predicted results were compared with measured values of spent fuel pools in the Fukushima Daiichi Nuclear Power Station. The calculations gave reasonable results, but overestimated the decreasing rate of the water level and the water temperature. This indicated that decay heat was overestimated and evaporation heat transfer from the water surface was underestimated. Results of calculations with 80% decay heat and 155% (Unit 4 pool) or 230% (Unit 2 pool) evaporation heat flux were in good agreement with measured values. The data-fitted evaporation heat fluxes agreed rather well with the evaporation heat transfer correlation proposed by Fujii et al.     

The adsorption and desorption of water from a carbonaceous deposit layer on the surface of stainless steel representing spent AGR nuclear fuel cladding

Spent fuel discharged from advanced gas-cooled reactor power stations carries a deposit of carbon firmly attached to the cladding surface. The fuel route involves contact with water, for cooling and transport. Long-term storage potentially includes dry storage, however, the carry-over of water entrained within the carbon deposit needs to be considered regarding the storage environment. Drying of the fuel is possible, but little is known concerning the drying characteristics of such deposits. This work reports preparation of a laboratory simulant of a carbon deposit on a fuel pin surface and measurement of its adsorption and desorption properties regarding liquid and vapour phase water. This work found that water vapour equilibration is rapid and reversible. Liquid water uptake is appreciable (up to 5.7 times the mass of carbon) and most (up to 88%) is removed on standing for 12 h. Heating removes little more. The implications for spent fuel management are discussed.     

The analysis of the insertion of UO2 microspheres in the gap between pellet and clad as a new design solution to increase the burnup of the nuclear fuel

The insertion of UO₂ microspheres (eventually graphite coated) in the gap between pellet-clad is observed to decrease substantially the clad hoop plastic strain concomitantly with the elimination of the rim effect at high burnup if low enrichment is used for the microspheres. Taking into account the special features of the specialized finite element code ELFIN'90 for the behavior of fuel elements, it was possible to introduce this new type of material viewed as a granular media. The results of the new code version ELFIN'MS applications to a PHWR fuel for a power ramp irradiation history show that the hoop plastic strain is reduced by about three times in comparison to standard fuel, and that the ridge phenomenon disappears. To establish critical plastic strain limit for irradiated clad failure onset, quantitative evaluations of iodine chemisorbtion on graphite and at the surface of the irradiated zircaloy, are presented. The indications on technology procedure are also discussed. Therefore, the insertion of 2–3 layers of UO₂ microspheres of 100 μm diameter, graphite coated to retain corrosive fission products for clad and with the diameter greater than the design gap, can be considered a design solution to increase the burnup of nuclear fuel.