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Fuel economy and novel fuel cycles for nuclear industry is a wide subject that required a massive effort along the decades. The reprocessing methods has shown a great potential to reach this goal to reduce the environmental issues related to burnt fuel disposal in nuclear resources utilization frame which is the primary subject. In second to reduce the mining and milling disadvantages which has minor effect if compared with the first. Besides the advantages reprocessing wastes should be considered. Thus, advanced nuclear fuel cycles has been proposed and one of the interest for CANDU technology is the communication between CANDU and LWR fuel cycles. In this concept, for short DUPIC/OREOX method is one of the most promising due cost and lower residual products and retain non-proliferation features. Since the spent PWR fuel has a higher fissile content than natural uranium, the DUPIC fuel results in a greater amount of fissile nuclides in fuel matrix. The objective of this paper is to state a roadmap to keep neutronics of the reprocessed fuel bundle as closer as possible to natural uranium fuel bundle to achieve a reliable scenario for further CANDU 6 reactor core fueling scheme. An evaluation of neutron absorber materials (depleted uranium, thorium, dysprosium and hafnium) was presented in sense of reactivity reduction due reprocessed fuel attributes. In the systems assessment the use of hafnium had shown to be suitable relating to system criticality. The SCALE 6.0 code was used for modeling and analysis.

In this paper, the SiC reinforced with Hi-Nicalon type S fibers (SiC HNS) was used as fuel cladding, and its behavior from the neutronic viewpoint was compared with the Zircaloy. To evaluate the average temperature of the fuel, cladding, and moderator used in the neutronic studies, a simplified thermal-hydraulic simulation of the core was performed using the RELAP5 code for the two cladding materials. Then, after that, a standard PWR fuel element was simulated, considering normal conditions of operation at Beginning Of Life (BOL) and full power, and using the average temperatures coming from the RELAP5 code calculation. The effects of burnable poison rods and temperature feedback on the core neutron multiplication factor have been evaluated changing the clad material at criticality steady state condition. Finally, the fuel element is analyzed during burnup. The neutronic analysis was made using the SCALE 6.0 (Standardized Computer Analysis for Licensing Evaluation) code.

A typical pressurised water reactor (PWR) fuel element containing TRU–Th fuel cycle was simulated. The study analysed the behaviour of thorium insertion spiked with reprocessed fuel by considering different enrichments that varied from 5.5 to 7.0 %. The reprocessed fuels were obtained by using the ORIGEN 2.1 code from a burnt PWR standard fuel (33,000 MWd/tHM burned), with 3.1 % of initial enrichment, which has remained in the cooling pool for five years. The k eff, hardening spectrum, and the fuel evolution during burnup were evaluated. This study was performed by using the SCALE 6.0 code.

A micro heteregenous reprocessed fuel spiked with thorium in a PWR fuel element considering (TRU-Th) cycle was simulated using three different configurations and different fissile materials that varied from 6.0% to 7.0%. The reprocessed fuels were obtained using the ORIGEN 2.1 code from a burned PWR standard fuel (33,000 MWd/tHM burned), with 3.1% of initial enrichment, which was remained in the cooling pool for five years and then reprocessed using UREX+ technique. The keff and plutonium generation during the burnup were evaluated and compared with the standard fuel. This study was performed using the SCALE 6.0.

The insertion of reprocessed fuel spiked with thorium in a typical PWR fuel element considering (TRU-Th) cycle was simulated using different fissile materials that varied from 5.5% to 7.0%. The reprocessed fuels were obtained using the ORIGEN 2.1 code from a burned PWR standard fuel (33,000 MWd/tHM burned), with 3.1% of initial enrichment, which was remained in the cooling pool for five years and then reprocessed using UREX+ technique. The kinf, hardening spectrum and the fuel evolution during the burnup were evaluated. This study was performed using the SCALE 6.0

Silicon carbide (SiC) has a potential to replacement zircaloy as fuel cladding material due to its high temperature tolerance, chemical stability and low neutron affinity. These characteristics may improve the economic and safety of nuclear reactors. Previous work has examined the possible use of SiC as a fuel cladding material in a PWR (Pressurized Water Reactor) environment. However, the advantage thermo mechanical and neutronic analysis replacement zircaloy cladding is not clear. Literature reviews has been done to predict the thermo mechanical behavior of SiC in high temperatures. The neutronic analysis was made using the SCALE 6.0 (Standardized Computer Analysis for Licensing Evaluation) code. This codes system is widely accepted and used worldwide for safety analysis and criticality of nuclear reactors has been utilized to model a typical fuel element of a PWR. It was used the CSAS6 and TRITON modules. The goals are to evaluate the behavior of the infinite multiplication factor (kinf) and neutron flux using SiC as a fuel cladding material.