Sürdürülebilir Kalkınma Amaçları
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- Conference ObjectAçık Erişim2026Neutronic and thermal-hydraulic performance analysis of a novel thorium-based fuel in dual-cooled annular assemblies for SMR applications
Rifai, Hanan; Kabach, Ouadie; Uzun, Sinem; Genç, Yasin; Uzun, Bedrettin; Amsil, Hamid; Chakir, El Mahjoub; Sadoune, Zouhair; Didi, A.; Baatout, S.
A coupled neutronic and thermal-hydraulic assessment is performed for a dual-cooled annular fuel concept applied to light-water SMR conditions. A mixed thorium-based oxide, (Th-233U-235U)O2, is evaluated against conventional UO2 and solid-pin benchmarks while sweeping the inner moderator radius (0.05-0.43165 cm) at constant fuel volume. The thorium annular cases sustain an infinite multiplication factor (kinf) of 1.002-1.003 at EOC and achieve 3-batch discharge burnups of 166.3-173.5 GWd/t, compared with 142.4 GWd/t for solid UO2. Radial peaking remains low (maximum PPF ≈ 1.056, well below the 1.65 Westinghouse-types PWR design limit). All configurations exhibit negative FTC and MTC, with stronger negative feedback at larger inner radii. Thermal-hydraulic results show smooth coolant heating and robust DNBR margins (minimum ∼ 2.3) with negligible sensitivity to inner radius. These outcomes support the feasibility of thorium-based annular fuel for SMR deployment with improved fuel utilization and preserved safety margins.
RDL Kalıcı Kaydıİncele - ArticleYalnız künye2026Coupled Neutronic and Thermal-Hydraulic Analysis of Inner Moderator Radius Effects in Dual-Cooled Annular UO2 and Thorium-Based Fuels for a LW-SMR
Kabach, Ouadie; Rifai, Hanan; Uzun, Sinem; Genç, Yasin; Amsil, Hamid; El Banni, Fadi; Uzun, Bedrettin; Chakir, El Mahjoub
This study evaluated the effect of the inner moderator radius on the neutronic and thermal-hydraulic performance of dual-cooled annular UO2 and (Th-233U-235U)O2 fuels for a light water (LW) small modular reactor (SMR) derived from the AP300™ concept. The analysis combined lattice neutronic calculations and fuel channel thermal-hydraulic modeling to assess reactivity behavior, discharge burnup, cycle length, isotopic evolution, power peaking factor, pressure drop, coolant temperature rise, and departure from nucleate boiling ratio (DNBR). The results showed that the thorium-based fuel provided better fuel utilization than UO2 over the entire investigated range of inner moderator radii. Under single-batch operation, the discharge burnup of the thorium-based fuel reached 36.96 to 38.57 gigawatt-days per metric tonne (GWd/t), whereas the UO2 cases achieved 27.14 to 35.41 GWd/t. For a three-batch strategy, the thorium-based option attained 166.3 to 170.8 GWd/t and yielded longer operating cycles. It also significantly suppressed plutonium production and higher actinide accumulations, while reducing 135Xe and 149Sm poisoning. Increasing the inner moderator radius improved fertile-to-fissile conversion in both fuel systems; however, the thorium-based fuel exhibited much smaller performance degradation across the examined range. From the thermal-hydraulic standpoint, all the cases exhibited very similar behavior, with low power peaking factors, DNBRs near 2.3, and nearly unchanged pressure drop trends. Overall, the results demonstrated that thorium-based dual-cooled annular fuel is a promising option for improving fuel cycle performance while maintaining favorable thermal-hydraulic safety margins in LW SMRs.
RDL Kalıcı Kaydıİncele - ArticleYalnız künye2026A comprehensive assessment of neutronic and thermal–hydraulic behavior of some advanced accident-tolerant fuels in solid and dual-cooled annular PWR assemblies
Kabach, Ouadie; Uzun, Sinem; Genç, Yasin; Uzun, Bedrettin; Chakir, El Mahjoub; Amsil, Hamid
Accident-tolerant fuel (ATF) concepts are being evaluated to strengthen margins without sacrificing core economics. This study compares five ATF matrices, UN (with natural and enriched nitrogen), UC, UCO, U3Si2, and UAl2, implemented in 17 × 17 solid and 13 × 13 dual-cooled annular PWR lattices. Results show that the multiplication factor decreases monotonically with burnup in both geometries, which allows the use of the linear reactivity model to predict the reactor cycle length; UAl2 and UO2 provide the longest cycles. Assembly-level radial peaking remains low (≈1.05–1.09). All fuel and moderator temperature coefficients are negative, with the strongest feedback in annular UN (Natural N). Solid pins show a single outer-rim power/Pu-239 peak; annular pins split this into two milder rims and flatten the in-pin profile. Prompt (neutron-induced) power is essentially burnup-invariant by normalization, while decay heat grows with matrix density and conversion. Thermal-hydraulic results show that the annular geometry lowers coolant heating and pressure drop and raises the minimum DNBR by ∼ 55–65% relative to solid pins under identical boundary conditions.
RDL Kalıcı Kaydıİncele - ArticleAçık Erişim2025Thermal-hydraulic investigation of novel thoriumuranium fuel mixtures in advanced SMR assembly configurations
Uzun, Sinem; Genç, Yasin; Rifai, Hanan; Kabach, Ouadie; Uzun, Bedrettin; Sadoune, Zouhair; Chakir, El Mahjoub; Amsil, Hamid; El Banni, Fadi
Nuclear energy sustainability and deployment flexibility can be significantly enhanced through Small Modular Reactors (SMRs) technology. Critical to their operational success is the thorough assessment of thermal-hydraulic characteristics, especially when incorporating advanced fuel design concepts. This research conducts an extensive thermal-hydraulic analysis examining various thorium-based fuel formulations, including (Th-235 U)O2, (Th-233 U) O2, and an innovative (Th-233 U-235 U)O2 composition, benchmarked against standard UO2 fuel. The investigation encompasses both solid fuel arrangements and dual-cooled annular assembly designs, focusing on safety optimization and operational efficiency enhancement. The analysis focuses on key safety parameters, including pressure drop, coolant enthalpy, fuel centerline temperature, and Departure from Nucleate Boiling Ratio (DNBR). Results for solid fuel configurations reveal that thorium-based fuels exhibit reduced pressure drop, more efficient enthalpy distribution, lower peak fuel temperatures, and higher DNBR values compared to conventional UO2, highlighting improved thermal stability and safety margins. The (Th-233 U-235 U)O2 mixture demonstrates a balanced performance by mitigating the limitations of other thorium compositions. In annular configurations, all fuel types benefit from enhanced heat removal due to the dual cooling surfaces, resulting in further reductions in pressure drop and peak temperatures, as well as a significant increase in DNBR values. The highest DNBR, reaching up to 3.051, confirms the annular geometry’s superior safety performance against boiling crises.
RDL Kalıcı Kaydıİncele - ArticleAçık Erişim2024The Effect of Nanofluid Coolant and Thorium-added Fuel on Burnup Dependent Isotopic Compositions in VVER-1000 Reactor
Acır, Adem; Genç, Yasin; Uzun, Sinem
Recently, the use of nanofluids has gained importance in studies aimed at increasing reactor efficiency while also addressing safety concerns in nuclear technology. Various studies have investigated the effects of adding nanoparticles of different types and proportions to the coolant water on the thermal and neutronic characteristics of power reactors using conventional UO2 fuels. Given the abundance of thorium compared to uranium, research on thorium-based fuels has become increasingly significant. In this study, the criticality and isotope changes in a VVER-1000 reactor loaded with 5% ThO2 and 95% UO2 by mass as fuel, and 0.1% by volume of Al2O3, CuO, and TiO2 nanoparticles as the coolant, were investigated. Neutronic analysis was performed using the MCNP5 and MONTEBURNS2.0 codes. The analysis results indicated that the operational lifespan of the reactor with only water coolant and thorium-based fuel was shortened due to the presence of nanoparticles. Furthermore, it was observed that while there was no significant change in the amount of fissile 235U and fertile 238U isotopes, the consumption of fertile 232Th isotope in the reactor increased with the insertion of nanoparticles into the coolant.
RDL Kalıcı Kaydıİncele - ArticleAçık Erişim2024Exploring thermal flow dynamics in pressurized water reactors using hybrid graphene nanoplatelet coolants
Uzun, Sinem; Genç, Yasin; Acır, Adem
This study investigates the impact of hybrid nanoparticles on the temperature of nuclear reactor coolant, with a focus on graphene nanoplatelet (GNP)-based hybrid nanoparticles. Sixteen different hybrid nanofluids were analyzed, and their performance was compared with a standard water-based coolant. The criticality values were obtained through MCNP modeling, revealing that higher nanoparticle ratios led to increased criticality, with the highest value of 1.3239 observed in GNP-Fe3O4 + Al2O3 nanofluids (0.05 wt%) and the lowest value of 1.2935 in GNP–Fe3O4 + SiO2 nanofluids (0.001 wt%). Temperature variations showed that increasing nanoparticle concentrations resulted in slightly higher temperatures, with a maximum of 611.97 K for 0.05 vol.% GNP nanoparticles. Additionally, the departure from nucleate boiling ratio values were consistently above the safety threshold of 2.08, with the lowest value of 3.657 for GNP–Fe3O4 + SiO2 nanofluids (0.05 vol.%). These findings suggest that hybrid nanofluids, particularly those with higher nanoparticle ratios, can enhance the thermal performance and safety margins of nuclear reactor coolants, offering a promising avenue for future research and application.
RDL Kalıcı Kaydıİncele - ArticleAçık Erişim2023INVESTIGATION OF THE USE OF CALCIUM CARBONATE AS NANOPARTICLES IN NUCLEAR REACTORS
Uzun, Sinem; Genç, Yasin; Acir, Adem
In this study, calcium carbonate (CaCO3) 3 ) was considered as nanoparticle. In the first part of the study, half-value layer (HVL) and mean free path (MFP) values, which are radiation shielding parameters, were investigated in determined energy ranges by Phy-X open access software. At increasing energy levels, the HVL value reached approximately 10 cm, while the MFP value reached approximately 17 cm. In the second part of the study, the reactor core geometry was modeled with the MCNP code and then the relative power distribution values were determined. COBRA code input was prepared with the determined relative power distribution values and thermal analyzes were made. These analyzes were performed for three different nanoparticle ratios. As a result of the analysis, the temperature value at the end of the channel was 613 K when only water was used as a coolant, while the temperature value at the end of the channel was 611.19 K when 0.03% nanoparticles were used. Although the coolant temperature increased with the nanoparticle ratio, it was observed that the temperature decreased when only water was used.
RDL Kalıcı Kaydıİncele - ArticleYalnız künye2022Investigation of hybrid nanofluids effects on heat transfer characteristics in VVER-1000 nuclear reactor
Uzun, Sinem; Genç, Yasin; Acır, Adem
In this study, the analysis of heat transfer characteristics of VVER-1000 nuclear reactor by using hybrid nanofluids as a coolants in subchannels in hot fuel assembly have been performed. Firstly, the criticality (keff) and fuel assembly power distributions of the VVER-1000 reactor by using Monte Carlo particle transport (MCNP) code to determine of the heat transfer characteristics with COBRA IV PC code have been examined. Secondly, the axial coolant temperature distributions and Departure from Nucleate Boiling Ratio (DNBR) values in hottest channel for Cu, CuO and TiO2 nanoparticles of the different volume fractions (0.05%, 0.01% and 0.1%) mixed with 0.1 vol % Al2O3 hybrid nanofluids have been analyzed. The obtained results by using hybrid nanofluids have been compared with the pure water results. The keff decreases with the increase of volumetric fractions of hybrid nanoparticles Cu, CuO and TiO2 mixed with 0.1% Al2O3 in the coolant. The maximum temperature has been reached by the püre water coolant at the end of the channel as 612.68 K whereas, the highest temperature change has been obtained in the case of 0.1% Al2O3 + 0.1% TiO2 as 617.35 K when evaluated among nanoparticle themselves. The results of analysis show that the heat transfer enhances with increasing hybrid nanoparticles volume fractions in the subchannel geometry. Also, the obtained minimum DNBR values using hybrid nanofluid haven't been exceed the safe operational limits; therefore the reactor operation would be safer.
RDL Kalıcı Kaydıİncele - ArticleAçık Erişim2021Investigation of Neutronic and Thermal Performance Using UGD and MOX Fuel in VVER-1000 Nuclear Power Reactor
Uzun, Sinem; Genç, Yasin; Acir, Adem
When examining the safety and design features of nuclear power reactors, its thermal performance in addition to neutronic characteristics is important. In this study, the neutronic and thermal performances of VVER-1000 reactor with two different fuel assembly arrangements were examined. Those fuel assemblies named as YD1 and YD2 are composed of 3.7% enriched LEU and 3.6% enriched LEU with 4% Gd2O3 uranium-gadolinium (UGD) and 2%, 3%, 4.2% Pu and 3.6% enriched LEU with 4% Gd2O3 (MOXGD), respectively. The effects of using UGD and MOXGD fuel assembly arrangements on criticality and isotope transformations according to burnup rate were investigated by means of MCNP5 and MONTEBURNS2.0 nuclear code, correspondingly the temperature and enthalpy changes of the coolant along the hot channel were examined with the help of the COBRA-IV PC thermal analysis code. The results obtained from this study were compared with similar studies in the literature and it was observed that the obtained findings were in accordance with the literature.
RDL Kalıcı Kaydıİncele - ArticleAçık Erişim2021Investigation of thermal performance of aluminum nanoparticle coolant in vver-1000 reactor
Uzun, Sinem; Genç, Yasin; Acir, Adem
Thermal analysis is of great importance when examining the safety and design features of nuclear power reactors. The aim of this study is to examine the effect of pure water and coolant containing Al2O3 nanoparticles on thermal performance of VVER-1000 nuclear reactor. In this study, the VVER-1000 reactor is in steady state and BOC (begining of cycle) condition, relative power density distributions for UO2 fuel with 3.7% enrichment were obtained with the MCNP code and then, thermal analysis were made with the COBRA code and DNBR value, which is very important for the safety criteria of the reactor, was examined. In addition, with the COBRA code, thermal values such as temperature, enthalpy, density, and temperature values of the fuel and clad regions, which are of great importance for the reactor, were calculated. As a result of the analysis, it was determined that the most efficient coolant is 0.1% Al2O3 fluid in the heat transfer of the reactor.
RDL Kalıcı Kaydıİncele - ArticleYalnız künye2021THERMAL ANALYSIS OF THE VVER-1000 REACTOR WITH THORIUM FUEL AND COOLANT CONTAINING Al2O3, CuO, AND TiO2 NANOPARTICLES
Acır, Adem; Uzun, Sinem; Genç, Yasin; Asal, Şulenur
The use of nanofluids has recently gained importance in studies carried out on increasing reactor efficiency with attention paid on safety factors in nuclear technology. Nanofluids used in reactor cooling water at certain properties and ratios generally use enriched UO2 as fuel. Through the discovery of thorium reserves and its use in nuclear technology, thorium has also become a candidate for being used as a fuel material in reactors. In this study, using ThO2 fuel and Al2O3, CuO, and TiO2 nanoparticles of the coolant in different ratios, the neutronic and thermal properties of the reactor, especially the criticality and temperature of coolant and DNBR as three leading parameters, were investigated. For the analyses, 5%, 10%, 15%, and 20% ThO2 was added to the reactor core containing 3.7% enriched UO2 fuel, and the analyses were performed adding 0.01%, 0.05%, and 0.1% volume of nanoparticles to the coolant water. Neutronic calculations were carried out with the help of MCNP code and COBRA-IV PC code was used for thermal analysis. The obtained results showed that by adding nanoparticles to coolant and using thorium as an additive to fuel the rise of coolant temperature can be augmented by 4.67 K compared to traditional combination of UO2 fuel and water while reactor remained within safety limits.
RDL Kalıcı Kaydıİncele - ArticleAçık Erişim2020Investigation of Criticality and Relative Power Density Distribution in VVER-1000 Nuclear Power Reactor
Genç, Yasin; Uzun, Sinem; Acir, Adem
One of the most important elements of the design and safe operation of nuclear power reactors is the examination of the criticality and relative power density distribution. In this study, the neutronic performance analysis of the VVER-1000 containing the critical and relative power density distribution was calculated with the help of the MCNP neutronic code. Neutronic calculations were performed for four different fuel types. As a result of the calculations for each fuel assembly, criticality and relative power density distributions were obtained in the reactor core. The results obtained in this study were compared with the literature and it was seen that the critical and relative power density distribution values were similar.
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