Research Article |
Corresponding author: Ouadie Kabach ( ouadie.kabach10@gmail.com ) Academic editor: Oleg Tashlykov
© 2024 Fadi El Banni, Bogbe L. H. Gogon, Ouadie Kabach, El Mahjoub Chakir.
This is an open access article distributed under the terms of the Creative Commons Attribution License (CC BY 4.0), which permits unrestricted use, distribution, and reproduction in any medium, provided the original author and source are credited.
Citation:
El Banni F, Gogon BLH, Kabach O, Chakir EM (2024) Analyzing (Th-233U-235U)O2 fuel performance in various assembly configurations: A comparative neutronic study. Nuclear Energy and Technology 10(3): 169-178. https://doi.org/10.3897/nucet.10.125376
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This article investigates alternate fuel options for Pressurized Water Reactors (PWRs), focusing on thorium use to address safety, efficiency, and waste issues associated with standard UO2 fuel. Challenges in thorium utilization, such as the lack of a fissile isotope, are handled using approaches such as homogeneous mixtures and heterogeneous arrangements, promoting the exploration of (Th-233U-235U)O2 fuel in various assembly configurations. According to recent research, the annular dual-cooled assembly design has promising results in terms of fuel efficiency and safety while lowering the requirement for higher fissile enrichment levels. Studies additionally demonstrate that annular dual-cooled duplex fuel configurations can produce higher discharge burnup and lower power peaking factors than traditional UO2 fuel. The purpose of this work is to analyze and compare the performance of (Th-233U-235U)O2 fuel in various configurations against conventional UO2 fuel, focusing on key characteristics such as reactivity change, criticality, discharge burnups, and reactivity feedback coefficients.
Thorium-based nuclear fuel, PWR, Dual-cooled annular assembly, Neutronic evaluation, Criticality, Safety coefficients
PWRs predominantly utilize low-enriched uranium dioxide (LEU) fuels; however, numerous alternative fuel options have been proposed over decades to address various concerns. These include enhancing safety parameters, developing fuels resilient to severe operating and irradiation conditions, minimizing nuclear waste, and increasing burnup efficiency. Among these alternatives, there’s a growing interest in exploring the potential of thorium utilization. However, utilizing thorium poses challenges. For instance, pure ThO2 lacks a fissile isotope, rendering it incapable of initiating a fission chain reaction in thermal reactors. Consequently, the initial utilization of thorium necessitates the incorporation of fissile material from the uranium cycle. In response to this challenge, numerous studies have investigated methods to incorporate thorium, either through different fuel mixtures or innovative fuel assembly arrangements (
In general, the approaches for introducing fissile material to produce a thorium-based fuel are well recognized, including homogeneous mixtures and heterogeneous arrangements (
Recent research indicates that using (Th-233U-235U)O2 fuel in an annular dual-cooled 13 × 13 assembly design can be a feasible alternative to traditional solid UO2 in homogeneous mixes (
Previous studies have also looked into the advantages of annular dual-cooled fuel and duplex fuel configurations in a Westinghouse SMR assembly, which is a scaled-down version of the AP1000 reactor. These investigations intended to improve the present performance criteria for such reactor types. To account for the dual-cooled feature, the study investigates both duplex configurations: ThO2 in the inner region and UO2 in the outer region, as well as UO2 in the inner region and ThO2 in the outer region (
Given these considerations, one of the main objectives of this paper is to analyze and comprehend the performance of (Th-233U-235U)O2 fuel in three different configurations: simple annular dual-cooled, dual-cooled macro-heterogeneous, and dual-cooled micro-heterogeneous. This examination will be carried out by comparing them to conventional LEU-UO2 fuel. Key characteristics to assess include reactivity change, criticality time, discharge burnups, power distribution, delayed neutron fractions, and reactivity feedback coefficients.
In this study, we examine two distinct fuel types, both possessing the same cumulative enrichment level of 4.95 wt. %, across five proposed fuel assembly (FA) configurations. The first fuel option is UO2, while the second comprises a composite of 232Th, 233U, and 235U, forming (Th-233U-235U)O2 fuel. Table
The main properties of the investigated fuel types (
Fuel | Density | Fertile | Fissile | Enrichment |
---|---|---|---|---|
UO2 | 10.53 (g/cm3) | 238U | 235U | 4.95 wt.% (235U) |
(Th-233U-235U)O2 | 9.54 (g/cm3) | 232Th | 233U & 235U | 2.475 wt.% (233U) 2.475 wt.% (235U) |
A single fuel assembly based on the Westinghouse-designed AP1000 was taken as the reference design (FA-1). This assembly contains a 17 × 17 grid, accommodating 264 fuel rod positions, 24 guide thimbles for control rods, and one central guide thimble for core instrumentation. The fuel rods are with UO2 fuel with 95.5% theoretical density, each with a radius of 0.409575 cm. The fuel pellets are separated from the cladding by a 0.008 cm gap. The cladding is made of 0.06 cm thick ZirloTM, and the fuel rods are pitched at 1.26 cm (
Given the objectives outlined in the introduction, the primary focus of this paper is to analyze and understand the performance of (Th-233U-235U)O2 fuel across various configurations. Prior research suggests that utilizing (Th-233U-235U)O2 fuel in dual-cooled assembly designs enhances fuel rod cooling, thereby mitigating the risk of overheating and cladding damage (
Geometrical specification of rod and guide in study assemblies (
Parameter | Solid | Dual-cooled | Dual-cooled duplex |
---|---|---|---|
Rod lattice pitch (cm) | 1.260 | 1.648 | 1.648 |
Inner clad inner radius (cm) | - | 0.431650 | 0.431650 |
Inner clad outer radius (cm) | - | 0.488650 | 0.488650 |
Inner helium gap outer radius (cm) | - | 0.494850 | 0.494850 |
Fuel 1 outer radius (cm) | 0.409575 | 0.705150 | 0.598697 |
Fuel 2 outer radius (cm) | - | - | 0.705150 |
Outer helium gap outer radius (cm) | 0.417750 | 0.711350 | 0.711350 |
Outer clad outer radius (cm) | 0.474750 | 0.768350 | 0.768350 |
Total fuel area per rod (cm2) | 0.527007 | 0.792812 | 0.792812 |
Guide tube | |||
Guide tube inner radius (cm) | 0.56 | 0.71 | 0.71 |
Guide tube outer radius (cm) | 0.60 | 0.77 | 0.77 |
Zone | Parameter | Value |
---|---|---|
Cladding | Material | ZirloTM |
Density | 6.50 g/cm3 | |
Gap | Gap material | Helium |
Gap density at 600 K | 0.012 g/cm3 | |
Moderator | Material | Light water |
Density at 573 K | 0.721 g/cm3 | |
Guide tube cladding | Material | Stainless Steel type 304 |
Density | 8.03 g/cm3 |
Case Name | Configuration and compositions |
---|---|
FA-1 | Solid (17 × 17) with 264 UO2 fuel rods per assembly. |
FA-2 | Dual-cooled (13 × 13) with 160 UO2 fuel rods per assembly. |
FA-3 | Dual-cooled (13 × 13) with 160 (Th- 233U, 235U)O2 fuel rods per assembly. |
FA-4 | Dual-cooled (13 × 13) with 72 UO2 and 88 (Th- 233U, 235U)O2 fuel rods per assembly in macro-heterogeneous configuration. |
FA-5 | Dual-cooled (13 × 13) with 160 fuel rods per assembly in micro-heterogeneous configuration (UO2 & (Th- 233U, 235U)O2). |
Case Name | Nuclide | Mass (kg) |
---|---|---|
FA-1 | 235U | 27.25 |
238U | 523.30 | |
FA-2 | 235U | 24.85 |
238U | 477.20 | |
FA-3 | 232Th | 431.40 |
233U | 11.23 | |
235U | 11.23 | |
FA-4 | 232Th | 237.30 |
233U | 6.18 | |
235U | 17.36 | |
238U | 214.70 | |
FA-5 | 232Th | 237.30 |
233U | 6.18 | |
235U | 17.36 | |
238U | 214.70 |
The methodology employed in this study builds upon the approaches used in prior research (
Additionally, MCNP, a Monte Carlo code, was employed in this study to verify and validate the DRAGON depletion results, particularly in terms of the eigenvalue (kinf). MCNP, utilizing the Monte Carlo probabilistic method, is designed to simulate the transport of large particles such as neutrons, photons, and electrons in complex geometric models (
As per available literature, large PWRs typically exhibit neutron leakage levels ranging from 3% to 4%, with the initial reactivity showing an approximately linear variation. In the present study, the constructed models include reflective boundary conditions on outer surfaces, hence neglecting neutron leakage in the eigenvalue calculation. To address neutron leakage, a 3% reactivity correction is applied. As a result, the single-batch discharge burnup is adjusted to yield a kinf value of ≈1.03. The Linear Reactivity Model (LRM) is then used to convert the single-batch discharge burnup into that of a three-batch core (
(1)
The corresponding cycle lengths were calculated using the formula:
(2)
where Tcycle represents the cycle length in Effective Full Power Days (EFPDs), and ρ is the specific power density of 40 kW/kg. In our cases, the discharge burnup and cycle length for each fuel model were modeled as calculated from the LRM under the assumption of a 3-batch core with 3% neutron leakage (
The dynamics of a nuclear reactor during normal operation can be approximated by considering the number of delayed neutron emissions resulting from the decay of fission products. Safety parameters for reactors predominantly fueled by UO2 include a higher ratio of effective delayed neutrons (βeff), which is crucial for licensing operations (
Fuel and moderator temperature coefficients are two critical reactivity coefficients. The fuel temperature coefficient (FTC) is defined as the fractional change in kinf per unit change in fuel temperature, while the moderator temperature coefficient (MTC) is defined similarly for moderator temperature changes. In this study, FTC and MTC were evaluated at the beginning of cycle (BOC) and end of cycle (EOC) steps. The evaluated models are compared for the following temperature perturbation scenarios:
The reactivity coefficients are then determined using Eq. (3), where α represents the evaluated coefficient, X1 and X2 are the changes in the considered parameter, and k1 and k1 are the multiplication factors arising from the states X1 and X2. These factors are multiplied by 105 to yield results in pcm (
(3)
The neutronic calculations for the analyzed fuel type assemblies were conducted at a boron concentration of 0 ppm, with all control rods withdrawn. Fig.
As shown in Fig.
Comparison of kinf at BOC and EOC, and the single-batch discharge burnup (BSC).
Case Name | kinf (BOC) | kinf (EOC) | BSC (MWd/kg) | |||
---|---|---|---|---|---|---|
DRAGON | MCNP | DRAGON | MCNP | DRAGON | MCNP | |
FA-1 | 1.41672 | 1.41245 | 0.89381 | 0.90146 | 38.38 | 39.39 |
FA-2 | 1.37707 | 1.37474 | 0.89087 | 0.89803 | 34.73 | 35.82 |
FA-3 | 1.41836 | 1.41753 | 0.91448 | 0.92239 | 40.67 | 42.00 |
FA-4 | 1.39207 | 1.39239 | 0.90579 | 0.91291 | 38.48 | 39.81 |
FA-5 | 1.31661 | 1.31482 | 0.91461 | 0.92149 | 33.93 | 35.14 |
Transitioning from an all (Th-233U-235U)O2-fueled assembly to the macro-heterogeneous (FA-4) and micro-heterogeneous (FA-5) configurations reveals distinct patterns in terms of initial reactivities and criticality, even though the total volumes occupied by UO2 and (Th-233U-235U)O2 are the same for FA-4 and FA-5 (see Tables
As previously mentioned, the LRM was utilized to accurately represent key cycle parameters, including cycle burnup, discharge burnup, and cycle length. The results are visually depicted in bar chart representations in Figs
The fissile inventory ratio (FIR) is a crucial metric for reactor operation and is an effective means to evaluate a reactor’s breeding capability. This ratio represents the proportion of the fuel’s fissile content to its initial fissile content, serving as an indicator of the conversion ratio. The fissile nuclei considered in this study to measure the conversion ratio include 233U, 235U, 239Pu, and 241Pu (
Fig.
The neutron spectrum is influenced by the balance between neutron moderation and absorption. Fig.
Fig.
Changes in βeff values with fuel burnup are depicted in Fig.
Fig.
Previous research studies have shown that using 233U as fissile material in LWRs is a problem because of its potential to produce lower negative or even positive MTC values. This behavior is essentially related to the epithermal fission resonance cross-section of 233U at higher neutron temperatures. However, as shown in Fig.
The increasing need for thorium-based fuel in PWRs highlights the significance of exploring novel approaches for enhancing nuclear energy performance. One such approach includes combining 233U and 235U with 232Th to create (Th-233U-235U)O2 fuel. This study aimed to compare the neutronic safety properties of (Th-233U-235U)O2 fuel in various configurations within a dual-cooled annular 13 × 13 assembly. The study examined criticality, cycle length parameters, power distributions, delayed neutron fractions, and reactivity feedback coefficients for various assembly configurations. The study focused on the simple annular dual-cooled (FA-3), dual-cooled macro-heterogeneous (FA-4), and dual-cooled micro-heterogeneous (FA-4) configurations of (Th-233U-235U)O2 with UO2.
The findings show that the micro-heterogeneous configuration exhibits lower criticality and cycle length compared to all-UO2 assemblies. In contrast, the macro-heterogeneous design has higher criticality and cycle parameters than all-UO2 assemblies. However, radial pin power data showed minor increases in the macro-heterogeneous assembly, resulting in power imbalances between the seed and blanket regions. On the other hand, the micro-heterogeneous assembly resulted in lower and more consistent pin power than UO2. However, to fully comprehend the impact of the fuel and assembly change, a thermal-hydraulic analysis is required. The delayed neutron fractions were also estimated, and it was found that using macro-heterogeneous or micro-heterogeneous configurations improved the net delayed neutron fraction.
In terms of moderator and fuel temperature coefficients, simulations indicated that both macro-heterogeneous and micro-heterogeneous assemblies exhibit more negative coefficients compared to the all-(Th-233U-235U)O2 assembly. Moreover, these designs demonstrated performance equivalent to traditional UO2 fuel. Therefore, the results of this paper can serve as essential data for safety analyses of PWRs with different fuel options.
Future work will focus on detailed full-core analyses to further evaluate the safety and performance of thorium-based fuel assemblies in PWRs. Investigating future safety parameters such as reactivity control will be essential. Additionally, studies could address the thermal-hydraulic behavior and burnup characteristics of these assemblies to provide a more comprehensive understanding of their viability.
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.