The Effect of Thermal Cycling on Mechanical Behavior of U-10Mo Monolithic Mini-Plate

Mini fuel plates used in Advanced Test Reactor (ATR) undergo five steps during the reator operation: startup, ATR Cycle 146A, transition from Cycle 146A to 146B, ATR Cycle 146B, and shutdown. Although the overall irradiation behavior of U-10Mo fuel is expected to be similar to the reactors operating at comparable power levels, there is a concern regarding how variation in operation schedules (that is, how many start-ups and shut-downs could be inserted during the Cycle A and Cycle B) may affect the mechanical behavior of the fuel plate during operation. To investigate any potential effect of number of start-stop activities, we simulated the thermo-mechanical behavior of L1P756 mini-plate, one of fuel plates inserted in ATR under RERTR-12 irradiation conditions with various numbers of start-stop activities artificially added at different points of time in irradiation cycle. This paper reviews four cases by varying the number of thermal cycles. Finite Element (FE) analyses were performed on the cases to investigate the effect of thermal cycling on the mechanical performance of L1P756 mini-plate. As a result, we observed that U-10Mo is in low stress level during the irradiation Cycle A and B due to creep behavior and that the maximum stress of aluminum cladding increases as the irradiation Cycle A and Cycle B proceeds. However, the number of thermal cycles did not affect the maximum stresses of U-10Mo, liner, and aluminum cladding.

FEASIBILITY STUDY FOR PRODUCTION OF IODINE-131 USING DIOXIDE OF TELLURIUM-130

Objective: Currently, nuclear medicine is becoming increasingly important, through the discovery of several medical radioisotopes, which are used in diagnosis, treatment, and medical imaging. Among the most important radionuclide which is commonly used is iodine-131, with a half-life of 8.02 d. Iodine-131 is one of the mainly essential elements in nuclear medicine. Since their first use, several studies have been conducted to meet the world need of hospital specialists in nuclear medicine. The purpose of this study was to participate in a lawsuit about the feasibility of producing 131I.Methods: using neutron activation of the dioxide of tellurium (TeO2) under a neutron flux which varies between 5 1011 and 1013 n/cm²s for 4, 6 and 8 hours** per irradiation cycle during 5 d, and used the Fortron90 Code to calculate the activity of iodine-131.Results: The result of the activity of iodine-131 found about 4,634 Curie with an irradiation of 4 hours** per day and 9.381 Curie with an activation of 8 hours** per day.Conclusion: Production of iodine-131 can be very effective if an acceptable capsule is used for different masses of tellurium and a neutron flux in a nuclear reactor.

Research and design of the moving system for the cobalt-60 industrial irradiator

On the way of localization of Cobalt-60 industrial irradiators, Research and Development Center for Radiation Technology (VINAGAMMA) has successfully designed and manufactured the first version of Co-60 industrial irradiator, VINAGA1. The second version of Co-60 industrial irradiator has been studied and designed by VINAGAMMA in the frame of the scientific project No. ĐTCB.02/15/TTNCTK. The nucleus of a Co-60 industrial irradiator is a mechanical system inside an irradiation room namely a tote box moving system. This report presents the tote box moving system designed by VINAGAMMA. The tote box moving system contains 52 tote boxes with the dimensions of 50 cm (w) × 70 cm (l) × 150 cm (h) that are moving around the source racks in the manner of 4 passes and 2 levels. The irradiator with this tote box moving system has good specifications: The minimum time of an irradiation cycle is 1h 20 min. and the dose uniformity ratio (DUR) at the product densities of 0.1 g/cm3 and 0.5 g/cm3 is 1.4 and 1.8, respectively. Radiation energy utilization efficiency at the product densities of 0.1 g/cm3 and 0.5 g/cm3 is 19.7% and 48.8%, respectively. These specifications meet the requirements for a multi-purpose Co-60 industrial irradiator and the present irradiation requirements in Vietnam.

Decomposition Analysis of Void Reactivity Coefficient for Innovative and Modified BWR Assemblies

The decomposition analysis of void reactivity coefficient for innovative BWR assemblies is presented in this paper. The innovative assemblies were loaded with high enrichment UO2and MOX fuels. Additionally the impact of the moderation enhancement on the void reactivity coefficient through a full fuel burnup discharge interval was investigated for the innovative assembly with MOX fuel. For the numerical analysis the TRITON functional module of SCALE code with ENDF/B-VI cross section library was applied. The obtained results indicate the influence of the most important isotopes to the void reactivity behaviour over a fuel burnup interval of 70 GWd/t for both UO2and MOX fuels. From the neutronic safety concern positive void reactivity coefficient values are observed for MOX fuel at the beginning of the fuel irradiation cycle. For extra-moderated assembly designs, implementing 8 and 12 water holes, the neutron spectrum softening is achieved and consequently the lower void reactivity values. Variations in void reactivity coefficient values are explained by fulfilled decomposition analysis based on neutrons absorption reactions for separate isotopes.

Low Burnup Inert Matrix Fuels Performance: TRANSURANUS Analysis of the Halden IFA-652 First Irradiation Cycle

The inert matrix fuels are a promising option to reduce-eliminate worldwide plutonium stockpiles by burning it in LWRs. These fuels, where plutonium is hosted in a U-free inert matrix phase, may reach high burning efficiency while preventing new plutonium build-up under irradiation. A specific investigation on CSZ and thoria inert matrices has been developed by ENEA since several years. In-pile testing on the ENEA-conceived innovative fuels is ongoing in the OECD Halden HBWR since June 2000 (IFA-652 experiment). The registered burnup at the end of 2005 is about 38 MWd·kgUeq−1 vs. 45 MWd·kgUeq−1 (40 MWd·kgUOXeq−1) target. Fuel pins are equipped with fuel temperature thermocouples, internal pressure transducers and fuel stack elongation sensors, with the task of studying thermal conductivity and its degradation with burnup, densification-swelling behaviour and the FGR. In this paper, the response at low burnup (< 7 MWd·kgUeq−1) of CSZ-based fuels loaded in IFA-652, is analysed by means of the TRANSURANUS code. To this purpose, a comprehensive modelling of the above mentioned un-irradiated fuels, mainly relying on the thermophysical characterisation performed at the JRC/ITU-Karlsruhe, has been implemented in a custom TRANSURANUS version (TU-IMF). A comparison of the code predictions vs. the experimental data, aimed at evaluating the early-stage under irradiation phenomena, particularly densification and relocation, has been performed.