Behavior of thorium plutonium fuel on light water reactors (original) (raw)
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Feasibility Study of Thorium-Plutonium Mixed Oxide Assembly In Light Water Reactors
Scientific Reports, 2019
Thorium-plutonium mixed oxide, (Th,Pu)OX, is currently used as an alternative fuel in the light water reactors in the world. The main objective of this paper is not only to show the benefits of using the thorium, but mainly to study how the way thorium is introduced in the fuel affects the neutron parameters. Among these benefits is the possibility of extending the operating cycle length and the reduction of the increasing stockpiles of plutonium. The first investigated method is introducing thorium as (Th,Pu)OX. The second one is a homogeneous model of thorium plutonium oxide. It is carried out by adding an amount of plutonium separated from the uranium oxide cycle at 50 GWd/ton of heavy metal to the same amount of thorium. Thus, we studied three assemblies; the reference assembly is uranium oxide of 4.2% enrichment containing borated water as a moderator of concentration 500 ppm (part per million) of B-10. The second is a (Th,Pu)OX and the third one is an assembly with homogenized...
This work investigates the possibility of using thorium-based fuels as an alternative fuel for advanced power reactor APR-1400. MCNPX code version 2.7 with crosssection library ENDF.VII has been used to design an APR-1400 fuel assembly. This code has been used to study the neutronic performance of the proposed thoriumbased Fuel types (0.944 Th, U)O 2 , (0.955 Th, 233 U)O 2 , (0.934 Th, rgPu)O 2. The fuel burn-up parameters such as infinity multiplication factor (k inf), initial heavy metals concentrations, Minor actinides concentration and fission products concentration have been analyzed during 1500 effective full power days (FPDSs) for thoriumbased Fuel types and compared with the common fuel. The analysis of the horizontal thermal power and neutron flux distribution provides valuable insights into the behavior and performance of the suggested Fuel types in the APR-1400 assembly. The analysis of the neutronic results ensures the viability of using the proposed thorium-based fuel types as an alternative fuel to UO 2 because they achieved acceptable safety parameter values and provided a good power distribution through the fuel assembly compared to UO 2 .
Advanced Nuclear Fuels Based on Thorium Mixed Oxides
Journal of Engineering Research, 2023
of advanced fuel systems formed ThO 2 75 wt.% and UO 2 25 wt.%, which worked with 19.5% enrichment of U 235. It analyzed the physical properties of mixed fuels using the composition of mixtures, such as the lattice parameters, thermal conductivity, specific heat, mechanical strength, and fission gas release. The codes FRAPCON-4.0 and FRAPTRAN-2.0 adapted can calculate the composite fuel response compared with uranium dioxide fuel used for light water reactors. In addition, the increased diffusion coefficient produced lower fuel swelling compared with UO 2. Thorium fuels had included an extensive range of applications, such as pressure-tube heavy water reactors (HWRs), light water reactors (LWRs), and thorium molten salt reactors. Advanced reactors, such as sodiumcooled fast reactors, can support the thorium mixed oxide fuel [1]. Early the Shippingport reactor, in Pennsylvania, USA, was a light water breeding reactor that operated with thorium as fuel during 1977-1982 [2]. Thorium is at least three times more abundant than uranium [3]. The natural isotopic distribution of thorium is 100% of Th 232 and is not fissile, but it is a fertile material, like U 238. Thorium requires fissile materials, such as U 235 and Pu 239 , to begin the reaction. Today, exist a few mixed fuel cycles based on thorium use (Th 232 +Pu 239), (Th 232 +U 233), (Th 232 +U 235), and other formulations, including dopant additions [4]. On the other hand, it researches innovative fuels, such as uranium nitride (UN) and uranium carbide (UC), which have several advantages over UO 2 , such as increased burnup capabilities and higher
Thorium Fuel Options for Sustained Transuranic Burning in Pressurized Water Reactors
As described in companion papers, Westinghouse is proposing the adoption of a thorium-based fuel cycle to burn the transuranics (TRU) contained in the current Used Nuclear Fuel (UNF) and transition towards a less radiotoxic high level waste. A combination of both light water reactors (LWR) and fast reactors (FR) is envisaged for the task, with the emphasis initially posed on their TRU burning capability and eventually to their self-sufficiency. Given the many technical challenges and development times related to the deployment of TRU burners fast reactors, an interim solution making best use of the current resources to initiate burning the legacy TRU inventory while developing and testing some technologies of later use is desirable. In this perspective, a portion of the LWR fleet can be used to start burning the legacy TRUs using Thbased fuels compatible with the current plants and operational features. This analysis focuses on a typical 4-loop PWR, with 17x17 fuel assembly design and TRUs (or Pu) admixed with Th (similar to U-MOX fuel, but with Th instead of U). Global calculations of the core were represented with unit assembly simulations using the Linear Reactivity Model (LRM). Several assembly configurations have been developed to offer two options that can be attractive during the TRU transmutation campaign: maximization of the TRU transmutation rate and capability for TRU multi-recycling, to extend the option of TRU recycling in LWR until the FR is available. Homogeneous as well as heterogeneous assembly configurations have been developed with various recycling schemes (Pu recycle, TRU recycle, TRU and in-bred U recycle etc.). Oxide as well as nitride fuels have been examined. This enabled an assessment of the potential for burning and multi-recycling TRU in a Th-based fuel PWR to compare against other more typical alternatives (U-MOX and variations thereof). Results will be shown indicating that Th-based PWR fuel is a promising option to multi-recycle and burn TRU in a thermal spectrum, while satisfying top-level operational and safety constraints.
Breeding Capability of Uranium and Thorium Fuel Cycles For Water Cooled Reactors
BREEDING CAPABILITY OF URANIUM AND THORIUM FUEL CYCLES FOR WATER COOLED REACTORS. Nuclear energy has contributed to fulfill the world energy demand especially in relation to the sustainable development of the world without any greenhouse effect to the environment for more than 50 years. The breeder reactors seem to have a similar trend with the renewable energies as a sustainable energy source. A fuel breeding is very essential for extending the sustainability of nuclear fuel resource and furthermore, it can be used to perform the sustainable development of the world. The present study intends to find the feasible region of design parameters for light or heavy water cooled reactors using thorium and uranium fuels which fulfill the required design characteristics such as breeding, negative void reactivity coefficient, comparable burn up with standard PWR, homogeneous core and large pin gap. The basic reactor design parameters of investigated systems are basically based on the water coolant reactor technology. The required enrichment, breeding capability and void coefficient are evaluated for light and heavy water coolants with U-Pu and Th-233 U fuel systems. A breeding condition is feasible for all investigated cases which are mainly require very tight lattice pitch for light water coolant cases and relatively larger lattice pitch for heavy water coolant. Regarding a negative void coefficient, only Th-233 U fuel system for both water coolants obtains a negative void coefficient in the breeding regions. The required design characteristics such as breeding, negative void reactivity coefficient, comparable burnup (PWR) and large pin gap can be achieved easier by heavy water cooled Th-233 U fuel system.
Annals of Nuclear Energy, 2012
The production of long-lived transuranic (TRU) waste is a major disadvantage of fission-based nuclear power. Previous work has indicated that TRU waste can be virtually eliminated in a pressurised water reactor (PWR) fuelled with a mixture of thorium and TRU waste, when all actinides are returned to the reactor after reprocessing. However, the optimal configuration for a fuel assembly operating this fuel cycle is likely to differ from the current configuration. In this paper, the differences in performance obtained in a reduced-moderation PWR operating this fuel cycle were investigated using WIMS. The chosen configuration allowed an increase of at least 20% in attainable burn-up for a given TRU enrichment. This will be especially important if the practical limit on TRU enrichment is low. The moderator reactivity coefficients limit the enrichment possible in the reactor, and this limit is particularly severe if a negative void coefficient is required for a fully voided core. Several strategies have been identified to mitigate this. Specifically, the control system should be designed to avoid a detrimental effect on moderator reactivity coefficients. The economic viability of this concept is likely to be dependent on the achievable thermalhydraulic operating conditions.
Use of Thorium for Transmutation of Plutonium and Minor Actinides in PWRs
The objective of this work was to assess the potential of thorium based fuel to minimise Pu and MA production in Pressurised Water Reactors (PWRs). The assessment was carried out by examining destruction rates and residual amounts of Pu and MA in the fuel used for transmutation. In particular, sensitivity of these two parameters to the fuel lattice Hydrogen to Heavy Metal (H/HM) ratio and to the fuel composition was systematically investigated. All burn-up calculations were performed using CASMO4 -the fuel assembly burn-up code. The results indicate that up to 1 000 kg of reactor grade Pu can potentially be burned in thorium based fuel assemblies per GW e Year. Up to 75% of initial Pu can be destroyed per path. Addition of MA to the fuel mixture degrades the burning efficiency. The theoretically achievable limit for total TRU destruction per path is 50%. Efficient MA and Pu destruction in thorium based fuel generally requires a higher degree of neutron moderation and, therefore, higher fuel lattice H/HM ratio than typically used in the current generation of PWRs. Reactivity coefficients evaluation demonstrated the feasibility of designing a Th-Pu-MA fuel with negative Doppler and moderator temperature coefficients.
Reactor performance and safety characteristics of ThN-UN fuel concepts in a PWR
Nuclear Engineering and Design, 2019
The reactor performance and safety characteristics of mixed thorium mononitride (ThN) and uranium mononitride (UN) fuels in a pressurized water reactor (PWR) are investigated to discern the potential nonproliferation, waste, and accident tolerance benefits provided by this fuel form. This paper presents results from an initial screening of mixed ThN-UN fuels in normal PWR operating conditions and compares their reactor performance to UO 2 in terms of fuel cycle length, reactivity coefficients, and thermal safety margin. ThN has been shown to have a significantly greater thermal conductivity than UO 2 and UN. Admixture with a UN phase is required because thorium initially contains no fissile isotopes. Results from this study show that ThN-UN mixtures exist that can match the cycle length of a UO 2-fueled reactor by using 235 U enrichments greater than 5% but less than 20% in the UN phase. Reactivity coefficients were calculated for UO 2 , UN, and ThN-UN mixtures, and it was found that the fuel temperature and moderator temperature coefficients of the nitride-based fuels fall within the acceptable limits specified by the AP1000 Design Control Document. Reduced soluble boron and control rod worth for these fuel forms indicates that the shutdown margin may not be sufficient, and design changes to the control systems may need to be considered. The neutronic impact of 15 N enrichment on reactivity coefficients is also included. Due to the greatly enhanced thermal conductivity of the nitride-based fuels, the UN and ThN-UN fuels provide additional margin to fuel melting temperature relative to UO 2 .
Effect of the thorium oxide content on the leaching of a mixed thorium-uranium oxide fuel
Journal of Radioanalytical and Nuclear Chemistry, 2022
substitution for UO 2 from a chemical point of view. ThO 2 is a highly insoluble compound, mainly due to the fact that ionic thorium only exists as the strong Lewis acid in the tetravalent oxidation state and, therefore, could potentially be able to protect the more redox sensitive and thus more soluble UO 2 from oxidation from the tetravalent to the hexavalent state and, ultimately, its dissolution in water. UO 2 is not per se more soluble in its tetravalent state than ThO 2 but can be readily oxidised to the hexavalent state and then dissolve. Dissolution of UO 2 may be a problem under different scenarios, for example, during interim storage of spent fuel, in a final repository, and in the case of fuel cladding failure during normal reactor operations. It is assumed that groundwater could intrude in the repository, and one of the scenarios that have been evaluated by SKB (The Swedish Nuclear Fuel and Waste Management Co) for the proposed KBS-3 concept for spent fuel management, is that the assumed reducing conditions in the water-filled repository would change to oxidizing conditions [5]. Together with a scenario with defective fuel canisters, this would increase the dissolution rate of the deposited fuel, which would then be in direct contact with groundwater. The radioactivity release from such a worst-case scenario would possibly be mitigated by partial substitution of UO 2 in the fuel