Integrated analysis of VVER-1000 fuel assembly fueled with accident tolerant fuel (ATF) materials
Introduction
Nuclear fuel is one of the main and important components of the nuclear reactor, which has been undergoing continuous development for nearly 50 years. Researches in the fuel development were not related to economic considerations only, but safety considerations to face possible accidents that the reactor might be exposed to it. Uranium dioxide fuel pellets clad with zirconium alloy tubing was the fuel of choice for the vast majority of commercial nuclear power plants. Severe accidents, such as those that occurred at Three Mile Island and Fukushima Daiichi prompted the researchers to look for accident-tolerant fuels (ATF) that can be safely and reliably operated under harsh conditions inside the reactor instead of uranium dioxide (traditional fuel). The traditional fuel under such nuclear accidents will fail and high temperature reactions between zirconium alloys and water occur. This will lead to the generation of hydrogen that can cause a huge explosion in the plant (Baranov et al., 2013, Baranov et al., 2014; Zhou and Zhou, 2018, Ebrahimgol et al., 2019, Costa et al., 2020, Pourrostam et al., 2020). Due to, the current fuel designs are exposed to severe accident conditions, the international nuclear community is striving to design an appropriate fuel design instead of UO2-Zr fuel-cladding design, which has been shown to degrade rapidly in such a severe accident scenario. Such new fuel designs would need to be compatible with existing fuel and reactor systems to use in the current reactor. It stands to reason; all fuel designs could fail under imagined scenarios and ultimately melt. Therefore, the ATFs can be defined as it is the fuel that will remain intact under severe accident conditions as long as possible (2016, Costa et al., 2020, Terrani et al., 2020). The ATFs are fuels or fuel clads with improved features in comparison with standard UO2-Zircaloy in commercial light water reactors. The ATFs could tolerate in vessel loss of coolant accidents. Finding applicable solutions to reduce accident consequences such as hydrogen production and core degradation due to loss of coolant in core have been evaluated by researchers at nuclear laboratories and universities considerably (Ebrahimgol et al., 2019). Currently, many researches are being made on different operating reactors to reach the more accident-tolerant fuels. These fuel materials should provide an acceptable behavior in neutronic, thermal–hydraulic and solid mechanics. The most trouble for using the traditional fuel materials UO2 is its poor thermal–hydraulic and solid mechanics performance. UO2 has an extremely little thermal conductivity, leading to a high temperature and a high heat accumulation inside it, which causes partial melt-down for fuel material. So, the demand for new fuel materials becomes imperative to overcome the conventional fuel problems. The new fuel materials should accomplish certain requirements. The neutronic requirements are a high fissionable nuclei atom density relative to non-fissionable nuclei and a little absorption cross-section for thermal neutrons. For thermal–hydraulic requirements, these compounds should have high thermal conductivity, a high melting point, a high resistance to decomposition or phase change at high temperature and high corrosion resistance. Enhanced thermal conductivity accident tolerant fuels have become one of the most promising possible solutions. For solid mechanics requirements, these compounds should have little thermal expansion coefficient to avoid pellet swelling and cracking problems. They should have a high yield stress to avoid phase change (fracture or ductile) (Zhou and Zhou, 2018, Zhao et al., 2020). The accurate prediction of thermal-hydraulics performance of a nuclear reactor is a major concept in its design for both economic and safety reasons. An integrated module nuclear reactor core modular simulator (CMS) was proposed to simulate the transient thermo-neutronic behavior of a VVER-1000 reactor core during reactivity insertion accidents. The CMS contains three modules; a neutronic module based on the point kinetic equations, a thermal–hydraulic module based on the time-dependent single heated channel approximation and reactivity calculation one (Faghihi et al., 2016, Ajami et al., 2020). A COBRA-IV-I code for thermal–hydraulic analysis was used to determine the maximum capability of heat removal in the most powered channel (hot channel) of the Bushehr Nuclear Power Plant (VVER-1000) core (Faghihi et al., 2016). The use of uranium mono-nitride (UN) fuel in light water reactors has recently received an increasing interest due to its potential advantages (Al-Qasir et al., 2018). It has a high density, a high thermal conductivity, and a high melting point. These thermo-physical properties provide a respectable performance for Uranium Nitride in the thermal–hydraulic analysis, where its thermal conductivity is higher than the thermal conductivity of UO2. So, Uranium Nitride's use will reduce the high fuel temperature and heat accumulation problems that lead to partial fuel melt-down accidents. From the neutronic analysis point of view, 14N has an extremely high thermal neutrons absorption cross-section, which will affect the various neutronic parameters and will not produce an acceptable neutronic performance compared to the traditional fuel material. A costly enrichment process will be carried to increase of 15N, which has a low thermal neutron absorption cross-section, to overcome this problem. There are several techniques for Uranium Nitride production, such as carbo-thermic reduction for Uranium dioxide, Sol-gel and Ammonolysis of uranium tetrafluoride. Uranium Nitride (UN)-uranium dioxide (UO2) composite fuels are being considered as the most promising ATF option for light water reactors. However, the complexity related to the chemical interactions between UN and UO2 during sintering is still an open problem (Costa et al., 2020). Uranium Carbide (UC) has a higher density, thermal conductivity than Uranium Nitride and Uranium dioxide. So, the thermal–hydraulic performance of the UC will be the best one of them. For the neutronic analysis, 12C has a very low thermal-neutron absorption cross-section. UC applied for several nuclear aspects such as nuclear fuel in pellets in Pebble bed reactors or mixed with Plutonium carbide (U, Pu) C. Also, it can be applied to nuclear thermal rockets. Furthermore, it is utilized for particle accelerators (Pandya et al., 2020, Zhao et al., 2020). Uranium monocarbide (UC) is an attractive alternative to uranium dioxide (UO2) as a fuel material, because of several reasons. The first is its thermo-physical properties are greater than UO2. Secondly, UC exhibits good dimensional stability and fission gas retention during irradiation. Thirdly, it is chemically stable with the coolant and potential cladding materials. These properties provide operation at higher power density, which cause a reduction in reactor core size, and hence a reduction in capital cost. Using a high temperature coolant such as organic liquid and liquid metals with UC fuel material enhances the thermal to electrical conversion efficiency. In this study, the neutronic, thermal–hydraulic, and solid mechanics performance of the Uranium Carbide and Uranium Nitride are investigated in VVER-1000 and compared with the Uranium dioxide. These analyses play a critical role in evaluating the most promising accident tolerant fuel. In this study, the neutronic, thermal–hydraulic, and solid mechanics performance of the Uranium Carbide and Uranium Nitride are investigated in VVER-1000 and compared with the Uranium dioxide. These analyses play a critical role in evaluating the most promising accident tolerant fuel.
Section snippets
Reference VVER-1000 fuel assembly
The VVER-1000/UTVS constructed in Iran, India and china is used as a reference model in this study. The VVER-1000/UTVS thermal and electrical power outputs are about 3000 and 1000 , respectively, with a thermal efficiency of 32.05% for its Rankine cycle (Safaei Arshi et al., 2010). The reactor design is generation III nuclear power reactor. Its core design consists of 163 hexagonal fuel assemblies. Each fuel assembly contains 331 positions distributed in a hexagonal array. These
Investigated fuels
In this study, Uranium dioxide, Uranium Carbide, and Uranium Nitride were investigated as fuel materials with 235U enriched to 2.4%. Uranium Carbide (UC) and Uranium Nitride (UN) are potential fuels for future applications due to their high thermal conductivity, high melting points, and higher uranium content compared with UO2 (Pandya et al., 2020, Zhao et al., 2020). Table 2 illustrates the properties of the materials used in the VVER-1000 fuel assembly.
Calculation methods
The study steps and the applied codes used to do integrated analysis of the proposed fuel types are illustrated in Fig. 1, where the MCNPX 2.7 was used to investigate the main neutronic parameters such as criticality, reactivity, fuel burn-up, and power mapping calculations. The power mapping calculation was used to determine the hot channel at which the thermal–hydraulic and solid mechanics analyses were done. To simulate the thermal–hydraulic analysis using COMSOL-Multiphysics software and
Burn-up calculations
The VVER-1000 assembly fueled with the UN and UC in a separate fuel cycle was burnt and their results were compared with UO2. Fig. 3 illustrates the variation of the infinity multiplication factor (Kinf) as a function of reactor operating time for the investigated fuel types. At the beginning of the cycle (BOC), there is a slight difference between the Kinf values of UO2 and UC ) despite the absorption cross-section of 12C is greater than 16O as illustrated in
Conclusion
For applying Uranium Nitride (UN) and Uranium Carbide (UC) as accident-tolerant fuels, they must achieve the safe operational conditions from the three points of views (neutronic, thermal–hydraulic and solid mechanics). From the neutronic analysis, UC was more economic than UO2 and UN as the fuel cycle length for UC was 540 days while for UO2 and UN were 360 days and 270 days respectively. Also, the power generated per fuel assembly was 20.20 MWth for UC, 18.13 MWth for UO2 and 17.26 MWth for
CRediT authorship contribution statement
Mohamed Y.M. Mohsen: Conceptualization, Methodology, Software, Data curation, Writing - original draft, Visualization, Investigation, Validation, Writing - review & editing. Mohamed A.E. Abdel-Rahman: Conceptualization, Methodology, Software, Data curation, Writing - original draft, Visualization, Investigation, Validation, Writing - review & editing. A. Abdelghafar Galahom: Conceptualization, Methodology, Software, Data curation, Writing - original draft, Visualization, Investigation,
Declaration of Competing Interest
The authors declare that they have no known competing financial interests or personal relationships that could have appeared to influence the work reported in this paper.
Acknowledgments
The authors would like to express their deepest thanks to the Aerospace department in the military technical college for permitting them access to run their computational experiments on the Workstation Computer.
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