Radial Power Research Articles

Neutronic assessment and optimization of ACP-100 reactor core models to achieve unit multiplication and radial power peaking factor

This study focuses on the neutronic and burnup assessment of the PWR-based Chinese ACP-100 SMR core. Four different core models with varying fuel compositions and assembly configurations were analyzed using the Monte Carlo code SERPENT 2. Three of these models shared uniformly enriched fuel assemblies of 3 wt.%, 4 wt.%, and 4.45 wt.% 235U respectively, while the fourth model (the proposed design for the ACP-100 in this study) featured a mix of differently enriched fuel assemblies along with the inclusion of burnable absorbers. The burnup analysis of the cores spanned 1570 EFPDs, evaluating several neutronic parameters (beta effective, reactivity coefficients) as well as fuel cycle characteristics (cycle burnup, discharge burnup and cycle length). Changes in the isotopic concentration of fissile, fertile, and poison materials were observed and the assembly-wise linear power distribution, neutron flux spectra, and radial power peaking factors of the models were determined. In the mixed core, a combination of borated water, IFBAs and control rods were employed to achieve exact criticality (unit keff) and uniform power distribution (unit radial power peaking factor). Insertion of control rods into 41 of the 57 assemblies, in conjunction with IFBAs placed strategically, attained a near unit radial power peaking factor (PPF) of 1.0586. Finally, the combination of IFBAs, sol-bor of 936 ppm concentration, and CRs with rod worth of 414.33 pcm per assembly brought down the keff to 1.000, but increased the PPF to 1.117. While the achievement of uniform power distribution came at the cost of fuel burnup, the proposed mixed core successfully remained critical for over 1.5 years and reached a discharge burnup of 23.37 GWd/t for a two-batch refueling scheme.

METAL: Methodology for liquid metal fast reactor core economic design and fuel loading pattern optimization

The design of Liquid Metal-cooled Fast Reactor (LMFR) cores encompass two levels of design. The first is the physical core design, which is concerned with the design of the fuel assembly geometry in the context of a full core. The second is the fuel loading design, which is an in-core fuel management problem concerned with the design of fuel enrichment and core loading pattern. In this paper, an optimized fuel design search is developed to improve core loading patterns. As part of the implementation, the multiphysics simulation suite LUPINE has been enhanced to allow for fuel shuffles and an in-core fuel management optimization methodology has been added with the objective of reducing the fuel cost. The design methodology is referred to as Methodology for Economical opTimization of Applied LMFR (METAL), and this methodology is demonstrated using the popular Super Power Reactor Innovative Small Modular (SPRISM) sodium-cooled fast reactor (SFR) as a reference core. One challenge currently facing the deployment of LMFRs is the availability of TRansUranium (TRU) and high assay low enriched uranium (HALEU) driver fuel. The two forms of fuel are expensive and not widely available, which poses a challenge on the startup of LMFR cores. To address the availability of driver fuel, and to lower fuel costs, low enriched uranium (LEU) blankets are investigated as replacements to the natural uranium or depleted uranium blankets typically used. The advantage of this solution is the wide availability of LEU and its ability to lower the amount of TRU or HALEU driver fuel needed. The design objectives, constraints, and optimization algorithm for a SFR are identified. Then, METAL is applied to design a SFR core fueled by TRU driver fuel assemblies and LEU blanket assemblies. To demonstrate the advantage of introducing LEU blankets, METAL is also used to design another SFR core following the same objectives and constraints, but using naturally enriched blankets. By comparing the two cores, it was found that the introduction of LEU blankets results in a ∼28% reduction in the driver fuel mass requirements. Upon development of a levelized fuel cycle cost (LFCC) model, the 28% reduction in driver fuel mass corresponds to a 10% decrease in the LFCC. Additional advantages of using LEU blankets include reducing the core conversion ratio (CR), and improving the assembly radial power peaking factor (RPPF), which enhances the safety and non-proliferation performances of the SFR core.

Multi-physics analysis of SBLOCA by coupled Neutronics/Thermal-Hydraulics full plant model

With the development of computational tools, the accident analysis would be performed by best estimate approaches. The employment of multi-physics codes would be considered as new calculation procedure to obtain high valuable results by improving calculations with specialized codes. The small break loss of coolant accident (SBLOCA) is an accident that has typically been analyzed using system codes with a limited number of Thermal-Hydraulics (TH) channels in the core region, neglecting the use of multi-physics calculations. This approach results in insufficient accuracy and resolution in both neutronics and TH calculations within the core zone. To address this issue, this study proposes the implementation of a Neutronics/Thermal-Hydraulics coupling with one-to-one mapping in the core zone, achieved through the development of a comprehensive plant TH model. This approach aims to enhance the accuracy and resolution of calculations, enabling the identification of local phenomena, parameters, and feedback during accident simulations, and the incorporation of these effects on the macroscopic cross-sections at each time step. The results of SBLOCA simulation reveals the effects of conservative/best estimate approach on the sequence of events, the initiation time of the safety systems, the critical controlled parameters of the power plant, and the precise understanding of the accident. Also, it was found that the radial power distribution becomes disrupted and loses its symmetry immediately after the SCRAM leading to different characteristics of steam generators (SGs) and natural circulation flow rate in each loop of primary side.

Multioptical bottles from the chirped circular Bessel Gaussian beams carrying annular spiral-zone power-exponent-phase

In this paper, we introduce a new structure of multioptical bottles formed by the chirped circular Bessel Gaussian beams carrying annular spiral-zone power-exponent-phase (ASP). The ASP is a general term for complex phases composed of the radial phase, equiphase, and power spiral phase. By combining the influence of autofocusing properties, chirp factor, and ASP mode, the propagation dynamics of the proposed multioptical bottles beam such as the sizes, lengths, number and locations can be flexibly manipulated. Intriguingly, the effect of the power order of ASP mode on the beam propagation dynamics is firstly introduced and examined. The proposed beams can enable diversification the optical tweezers system in multiparticle manipulation, which is verified by the radiation force analysis on a Rayleigh particle.

Radial Power Combiners—An Overview: A Comprehensive Analysis of Power Combiners, Their Structures and Evolution From Their Inception to the Present

Radial Power Combiners—An Overview: A Comprehensive Analysis of Power Combiners, Their Structures and Evolution From Their Inception to the Present

Comparative analysis of different FeCrAl alloys in pressurized water reactors

Discussing and researching issues related to their safety is essential to guarantee the continuity of fission systems in generating energy for existing and new electrical matrices. Commercial fission reactors currently in operation have undergone a series of redesigns. One aspect much discussed in recent years is related to research about new fuels and cladding materials. Therefore, in this work, the feasibility of replacing Zircaloy-4 with FeCrAl-based alloys (C06M, C35M, C36M, Kanthal APMT) was evaluated considering these materials' physical–chemical, thermal, and neutronic properties. Moreover, thermal–hydraulic and neutronic simulations have been presented to compare the cladding behavior in a PWR core in steady state and transient operation conditions. The results showed, in a general way, that the analyzed alloys have equivalent properties from a physical–chemical and thermal point of view. Despite the similarity, the Kanthal APMT alloy presented the worst performance, and the C35M alloy showed more advantageous characteristics. Considering the similarities in the composition of the C06M, C35M, and C36M alloys, which have the same elements in their constitution, the alloy with the best thermal performance has the lowest aluminum content. Considering the neutronic behavior, the Kanthal APMT alloy, with more Cr percentile, presents the lowest neutron effective multiplication factor, keff. Axial and radial power distribution did not change significantly after the cladding replacement. Nevertheless, modifications must be made to keep the criticality during the fuel cycle burnup for all FeCrAl alloys.

Optimization of physical design for small long life Lead Cooled fast Reactor based on SARAX

The Offshore Fixed Lead-Cooled Reactor is a Small Long Life Lead Cooled Fast Reactor. In order to verify the rationality of the physical design and applicability of SARAX and the core equivalent model method to the corresponding reactor type, the calculation of the key physical parameters of the Offshore Fixed Lead-Cooled Reactor and the physical design characteristics of the core were investigated and compared with the Monte-Carlo code OpenMC. The initial physical design has been shown to meet the design requirement. The reactivity coefficient of the core during life were calculated and verified, which proved the rationality of the initial core design. In order to solve the problem of high radial power peak factor at end-of-life, analysis of radial fuel zoning was conducted and the core optimization was carried out. The calculation of the key physical parameters of the optimization schemes was conducted, which proved the rationality of optimization schemes.

Whole-core analysis of Watts bar benchmark with three-dimensional MOC code STREAM3D

This paper presents a high-fidelity simulation of the Organization for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) 3D whole-core Watts Bar benchmark using the UNIST in-house STREAM3D (Steady State and Transient Reactor Analysis code with Method of Characteristics) neutronic code. The benchmark encompasses various whole-core exercises, including single physics problems, multi-physics simulations, and depletion problems. When comparing parameters during the zero-power physics tests, including ITC, DBW, CRW, and criticality tests, STREAM3D results indicate a strong agreement with the measured data and KENO-VI. The comparison with the MC21/CTF code in 3D HFP BOC condition demonstrated strong agreement, with only a 0.42% difference in the normalized radial power distribution, a 0.38 K difference in the RMS of the assembly coolant exit temperature, and a mere 4 ppm difference in CBC.

The influence of source locations of different multi-beam concept on the power density distribution and nuclear waste transmutation performance for ADS–NWT

The multi-beam concept for the accelerator driven subcritical reactor (ADS) has the advantages in reducing the requirements for the accelerator and can flatten the power distribution of the core. However, the layout of the multi-beam should meet the goal of balancing the relationship between the total power density, the neutron leakage, and the nuclear waste transmutation performance. For the further study of the multi-beam concept for the ADS, this paper focuses on the influence of the source locations of different multi-beam concept on the power density distribution, the neutron leakage, and the nuclear waste transmutation performance for the ADS for a nuclear waste transmutation reactor (ADS–NWT). The different variations on the radial power density distribution and the nuclear waste transmutation performance for the three-, four-, six-, and seven-beam concept for ADS–NWT were analyzed. The results show that the related neutronics performance for the multi-beam concept for ADS-NWT is not only affected by the external source locations and the core geometry, but also affected by the keff and the interference effect between the spallation targets. Moreover, the fission reaction ratio and the fission reaction / (n, gamma) absorbtion ratio for minor actinides (MA) and plutonium (Pu) for the few beams concept are both higher than that for the many beams concept. Based on the analysis of the influence of source locations on the neutron spectrum, the total neutron flux, and the nuclear waste transmutation performance for MA and Pu, the related conclusions about how to arrange the best beam position from the perspective of nuclear waste transmutation and production capacity are provided.

Multiphysics and multi-scale design and analysis of nuclear thermal propulsion pellet bed reactor based on spherical cermet fuel element

Due to the advantages of high thrust and high specific impulse, nuclear thermal propulsion (NTP) is a preferred near-term rocket engine technology for future Mars missions. This article introduces the reactor design based on spherical cermet fuel element, with the aim of studying the feasibility of applying this new fuel in NTP system. First, design requirements are determined based on the unique features of NTP. In terms of neutronics, the effect of rocket launch accidents and fuel loss caused by high-temperature hydrogen are considered. Thermal hydraulics refer to the requirements of NASA's Mars Design Reference Architecture 5.0 (DRA 5.0) for impulse, thrust, and thrust to weight ratio. Second, based on the requirements above, the optimal reactor configuration parameters are obtained with the goal of maximizing the thrust to weight ratio: reactor radius 51.1 cm, height 106 cm, aspect ratio 1.04, reflector 13 cm, and reactor mass 3940 kg. Finally, the performances of the optimal reactor design are analyzed. The neutron spectrum shows a characteristic of fast neutron spectrum, and fuel loss analysis shows that the core can withstand a fuel mass loss of about 4 wt%. In addition, the core axial and radial power distribution are calculated, with an overall power peak factor of 1.26. The thermal hydraulic calculation provides a range of core operating conditions that satisfy the design requirements of DRA 5.0 and temperature, with power of 620 ∼ 950 MW and flow rate of 16 ∼ 24 kg/s. Compared with the ANL200/2000 reactor using prismatic cermet, it has the advantage of higher specific impulse. The main disadvantage is the large pressure drop, which prevents it from being applied to the scenarios of high power and high thrust. Therefore, for NTP reactor using spherical cermet elements, it should be applied to low to medium power scenarios.

Optimal Design and Operation of Wind Turbines in Radial Distribution Power Grids for Power Loss Minimization

This research proposes a strategy to minimize the active power loss in the standard IEEE 85-node radial distribution power grid by optimizing the placement of wind turbines in the grid. The osprey optimization algorithm (OOA) and walrus optimization algorithm (WOA) are implemented to solve the problem. The two algorithms are validated in three study cases of placing two wind turbines (WTs) in the system for power loss reduction. Mainly, in Case 1, WTs can only produce active power, while in Case 2 and Case 3, WTs can supply both active and reactive power to the grid with different ranges of power factors. In Case 4, the best-applied methods between the two are reapplied to reach the minimum value of the total energy loss within one year. Notably, this case focuses on minimizing the total power loss for each hour in a day under load demand variations and dynamic power supply from WTs. On top of that, this case uses two different sets of actual wind power data acquired from the Global Wind Atlas for the two positions inherited from the previous case. Moreover, the utilization of wind power is also evaluated in the two scenarios: (1) wind power from WTs is fully used for all values of load demand, (2) and wind power from WTs is optimized for each load demand value. The results in the first three cases indicate that the WOA achieves better minimum, mean, and maximum power losses for the two cases than the OOA over fifty trial runs. Moreover, the WOA obtains an excellent loss reduction compared to the Base case without WTs. The loss of the base system is 224.3 kW, but that of Case 1, Case 2, and Case 3 is 115.6, 30.6 kW, and 0.097 kW. The placement of wind turbines in Case 1, Case 2, and Case 3 reached a loss reduction of 48.5%, 84.3%, and 99.96% compared to the Base case. The optimal placement of WTs in the selected distribution power grid has shown huge advantages in reducing active power loss, especially in Case 3. For the last study case, the energy loss in a year is calculated by WSO after reaching hourly power loss, the energy loss in a month, and the season. The results in this case also indicate that the optimization of wind power, as mentioned in Scenario 2, results in a better total energy loss value in a year than in Scenario 1. The total energy loss in Scenario 2 is reduced by approximately 95.98% compared to Scenario 1. So, WOA is an effective algorithm for optimizing the placement and determining the power output of wind turbines in distribution power grids to minimize the total energy loss in years.

Feasibility Study on the Application of Boron Carbide for Long Term Reactivity Control in the LOTUS Small Fast Reactor

Abstract 200 MWth lead-cooled fast reactor (LOTUS) reactor core is a small modular lead-cooled fast reactor with designed power of 200 MWth under development at Vietnam National University (VNU) University of Science, Hanoi for a floating nuclear power plant application. For that purpose, advanced passive safety features and no refueling requirement are the priorities in the core design process. To endure the continuous operation over a long lifetime, the startup core exhibits excess reactivity to cover the reactivity loss due to burnup. The reactivity control system includes burnable poison and absorber rods and layers made of B4C which are employed in the reactor to minimize the excess reactivity of the core to about 1 $ to enhance the safety features of the core. The burnable poison is fixed inside the reactor while absorber rods/absorber layers are withdrawn or inserted in sequence to achieve the required excess reactivity of about 700 pcm. The reactivity control was arranged into ten steps to achieve the operating time of 15 effective full-power years without refueling. Good neutronic behavior of the core was observed with negative fuel temperature coefficient and coolant void reactivity and maximum radial power peaking factor of 1.32. However, a quite large residual absorption caused by fixed burnable poison inside fuel assemblies was revealed. In further study, to increase the neutron absorption efficiency of burnable poison in the fast spectrum as well as the reactor lifetime, a neutron moderator will be considered to add to the burnable poison rods.

Development and validation of transient analysis module in nodal diffusion code RAST-V with Kalinin-3 coolant transient benchmark

This study introduces a transient analysis module developed for RAST-V and validates it using the Kalinin-3 benchmark problem. For the benchmark analysis, RAST-V standalone and STREAM/RAST-V calculations were performed. STREAM supplies the few-group constants and RAST-V conducts a 3D core simulation utilizing few-group cross-sectional data. To improve accuracy, the main solver was developed based on the advanced semi-analytic nodal method. To evaluate the computational capability of the transient analysis module in RAST-V, Kalinin-3 benchmark is employed. Kalinin-3 represents a coolant transient benchmark that offers experimental data during the deactivation of the Main Circulation Pumps. Consequently, the transient calculations reflected the changes in the reactor flow rate. Benchmark comprising steady-state and transient calculations. During the steady state, the STREAM/RAST-V combination demonstrated a 30 ppm root mean square difference from 0 to 128.50 EFPD. For the transient calculations, STREAM/RAST-V showed power differences within ±7 % over a range of 0–300 s. Axial offset differences were within ±3 %, and the RMS difference in radial power ranged within 2.596 % at both 0 and 300 s. Overall, this study effectively demonstrated the newly developed transient solver in RAST-V and validated it using the Kalinin-3 benchmark problem.

A proposed refueling scheme for the modular pebble bed reactor

In this paper, a suggested design of the fuel pebble and loading pattern in the MIT Modular Pebble Bed HTGR (MPBR) using thorium (232Th) is presented. Using the MCNP5 code model for the MPBR reactor, different compositions of Thorium Low Enriched Uranium (ThLEU) with different Heavy Metal (HM) loading and 235U enrichment were investigated to achieve the maximum thorium utilization with relatively minimum reactivity insertion in case of water ingress accident. Using pure Low Enriched Uranium (LEU) pebbles and ThLEU pebbles with the same HM loading and 235U enrichment, different pebbles loading patterns were studied. One loading pattern was selected to achieve the criticality with enough excess reactivity compensating for the poisoning effect of xenon build-up at the startup of the reactor operation and the negative effect of fuel and reflector temperature coefficients of reactivity. This loading pattern aims also to reduce the radial power peaking in the core and to minimize the axial Power Peaking Factor (PPF) in the central hot channel, with a minimum number of pure LEU fuel batches. A calculation procedure was developed that combines fuel pebble depletion and full core calculations to investigate the impact of a proposed online refueling scheme for the MPBR reactor and to simulate the multicycle refueling mechanism. Using SERPENT 1.1.7, the nuclide concentrations were calculated as a function of pebble burnup for both LEU and ThLEU pebbles. A multicycle online refueling scheme was proposed and the average number of passes required for LEU and ThLEU pebbles to reach their maximum burnup was estimated.

Core design optimization of a 200 MWt pressurized water SMR using evolutionary simulated annealing

The paper presents the design optimization of a pressurized small modular reactor (SMR) core using evolutionary simulated annealing (ESA). The core was designed for the power output of 200 MWt based on the typical 17 × 17 fuel assemblies of a traditional PWR. The core consists of 37 fuel assemblies of F235, F340 and F445 types corresponding to the 235U enrichment levels of 2.35, 3.40 and 4.45 wt%, respectively. The design target is to attain the fuel cycle length of about 900 effective full power days (EFPDs) with a minimum core average 235U enrichment, while flattening the radial power distribution. Lattice and core physics calculations were performed using the SRAC code system. The COREBN module of the SRAC code system was coupled with the ESA method for performing the core burnup calculations during the optimization process. A fitness function was chosen to combine the objectives of the cycle length, the core average enrichment and the radial power peaking factor (PPF). A survey was conducted for determining the optimal control parameters of the ESA method and the weighting factors in the fitness function associated to the design objectives. The optimal core was selected from the ESA search consisting of 21 assemblies of F235, 8 of F340 and 8 of F445 type, respectively. The core average 235U enrichment is 3.031 wt% with the cycle length of about 840 EFPDs. The PPF of the selected core is 1.428 which appears at the centered assembly.

Sensitivity studies of PWR MOX fuel management to the plutonium initial vector using Artificial Neural Networks

This paper presents new metamodels based on artificial neural networks trained on full core 3D depletion simulations performed with APOLLO2 and CRONOS2. They are used to estimate the irradiation cycle length, discharge burn-up of each fuel assembly type and radial power factor of a PWR loaded with 30% of MOX fuels, as a function of the initial plutonium composition. They allow to explore the impact of the plutonium isotopic vector on the reactor characteristics and can be used for scenarios studied for future fuel cycle. Some exclusion domains in the plutonium isotopic vector phase space are identified as a function of the cycle length. As an example, the potentialities of such fuel management for plutonium recycling from MOX spent fuel are studied.

Optimization of Arterial Stents for May-Thurner Syndrome Management in West Java: Experience and Outcome

Abstract Context: May-Thurner syndrome (MTS) is an anatomical defect where the right common iliac artery overrides and compresses the left iliofemoral vein. MTS causes deep vein thrombosis (DVT) but is often underdiagnosed. As anticoagulant management (warfarin) rarely achieves target INR, endovascular management is vital for MTS. Surgical and endovascular management of MTS has known to have similar good outcome. Endovascular management is generally facilitated using a venous stent. However, the unavailability of the venous stent in Indonesia made arterial stent be optimized during vein stenting procedure. Aims: We aim to report our experience. Settings and Design: Consecutive observational study within 26 months in a hospital in West Java including 6 MTS patients. Subjects and Methods: Diagnosis is established by the finding of proximal femoral DVT by ultrasonography further confirmed by computed tomography (CT) angiography. Endovascular management was done using arterial stent placement. After endovascular management, we reviewed the outcome on a follow-up including stent fracture, in-stent restenosis (ISR), and stent patency. These are also monitored by serial imaging by CT Angiography, with periods ranging from 1.5 to 10 months (mean: 4.08 months). Direct oral anticoagulants and antiplatelets are usually given along with compression stockings following the procedure. Statistical Analysis Used: None. Results: No stent fracture, no ISR, and the stent remained patent after follow-up. No patient had complications. Conclusions: Apart from the difference in radial power compared to the venous stent, the arterial stent is feasible to use during MTS endovascular management with good postprocedural results. Prompt management is needed to prevent complication and chronic total occlusion. Comprehensive management should be ensured to optimize patients’ quality of life. We do hope that venous stent will be available in the future.

Chance-Constrained Pre-Contingency Joint Self- Scheduling of Energy and Reserve in VPP

With the increasing penetration of primarily inertia-less distributed energy resources, allowable delay in the fast frequency reserve provision to ensure marginal stability condition can become comparable with the requisite dead-time for fast-switching in a radial power distribution system (PDS) connected to the transmission system. Consequently, the prevailing short-duration load-generation imbalance might propagate as system-wide frequency excursion in the future power system, which traditionally has not been observed. Furthermore, during the fast-switching, some of the fast-acting reserve (FAR) providing local generators remain inaccessible to the bulk power system (BPS). This work has answered the questions of requisite reserve requirements to cater to such events through efficient energy and FAR provision joint-scheduling. Stochasticity of fast-switching, with temporary faults as a use-case, requires modeling of load-generation imbalance and local resource unavailability, vis-á-vis FAR requirement problem, as a chance-constraint. Due to the limited visibility at the BPS level, PDS operators must ensure sufficient FAR availability for a given confidence level of the chance-constraint. Individual chance-constraints are used to ensure mathematical simplicity. Here, the PDS is operated as a virtual power plant (VPP), where the operator can procure resources locally or from the wholesale market with the aim of profit maximization before the contingency occurs. Results show that FAR requirements can influence the local energy schedule and raise energy costs. Furthermore, the confidence level of the chance-constraint can impact overall profitability, and FAR should be scheduled carefully. Comparative analysis with a modified benchmark IEEE 33-node radial test system and a 98-node test system shows the superior performance of the proposed approach. The impact of the limited available FAR is also demonstrated.

Prediction of critical heat flux in non-uniform heated rod bundle based on modified CHF empirical correlations

The accurate estimation of critical heat flux (CHF) continues to be a significant challenge in the study of radial and axial non-uniform heated rod bundles under NHR-200Ⅱ operating conditions. The present work develops the subchannel analysis code CTF algorithm that incorporates modified CHF empirical correlations to forecast CHF of non-uniform heated rod bundles preliminarily. The article selected experimental data from HTRF, which includes both uniform heated and non-uniform heated rod bundle CHF experiments, for preliminary validation and analysis of this method. The results demonstrated that the predicted results of modified Barnett correlation have the good agreement with the experimental values. Additionally, the modified Barnett correlation is used to predict the CHF of rod bundles under various power distribution. In different radial power distribution, the critical power of rod bundles decreases with the increase of radial power factor ratio. In different axial power distribution, the critical power in inlet and outlet peak distribution is lower than uniform heated rod bundle. Regarding the case of radial uniform heating and axial inlet peak distribution, CHF locations often occur near the central section of the heated rod height. In the axial intermediate peak distribution, CHF positions often emerge in the intermediate and downstream sections of the heated rod to the outlet. In a typical operating scenario of the NHR-200II, the CHF will occur towards the exit of the hot central subchannel, particularly near the radial heated rod exhibiting the maximum thermal power. The occurrence of CHF can be attributed to factors such as lower mass flux, higher local equilibrium quality, and heat flux.

Feasibility of Using Mixed-Oxide Fuel for a Pressurized Water Reactor Instead of Traditional UO2 Fuel Material

This study examines the feasibility of utilizing mixed-oxide fuel [(U0.9, rgPu0.1) O2] instead of traditional UO2 in nuclear reactors. The utilization of (U0.9, rgPu0.1) O2 is particularly significant as it represents an effective approach to nuclear fuel recycling by combining reactor-grade plutonium extracted from partially used nuclear fuel and depleted uranium obtained through the enrichment process. The fundamental neutronic characteristics, such as the radial power distribution, were investigated using the MCNPX 2.7 algorithm to identify the specific channel for subsequent thermal-hydraulic (TH) analysis. The TH analysis was conducted using COMSOL-Multiphysics, allowing for the estimation of the fuel rod’s axial and radial temperature profiles, as well as the determination of the departure from the nucleate boiling ratio. Furthermore, the coupling between heat transfer and solid structure (SS) was achieved using the Multiphysics tool in COMSOL-Multiphysics. This coupling facilitated the simulation of key SS parameters, including von Mises stress, volumetric strain, and displacement, while considering the influence of heat transfer. The results demonstrate significant improvements and enhanced safety margins when utilizing (U0.9, rgPu0.1) O2.