Volume 89, Issue 6

The floating nuclear power plant (FNPP) is a new type of reactor. Due to the unique marine environment, the reliability of its equipment differs significantly from that of land-based power plants, making conventional nuclear power plant databases unsuitable for direct application to FNPPs. Consequently, recognizing the need for reliability data specific to FNPP equipment, this paper proposes a framework for predicting reliability and analyzing associated uncertainties. This framework employs the parts count method and the parts stress method, as outlined in reliability prediction handbooks, to predict reliability based on different equipment categories. This paper studies the reliability of two typical types of equipment in nuclear power plants: valves and main coolant pumps. First, the reliability data for these two equipment types in land-based environments are calculated and compared with data from conventional nuclear power plant databases to validate the accuracy of the calculation results. Next, the reliability of the equipment is estimated for a marine environment, and the uncertainty distribution of these estimates is presented. The results indicate that the proposed framework effectively predicts the failure data for valves and main coolant pumps in FNPPs operating in marine conditions, serving as a reference for future reliability predictions of additional equipment.

The main aim of this study is the thermohydraulic and safety analysis of alternative UN and UC ceramic fuels in the next generation of VVER-1000/V446 reactor core. One of the advantages of these alternative fuels is their higher thermal conductivity and density relative to the UO 2 conventional fuels. Thermohydraulic analysis has been performed under full power conditions. Also, the safety evaluation of alternative ceramic fuels has been researched by the Rod Ejection Accident (REA) scenario. RELAP5 Mod3.2 code is used for thermo-hydraulic analysis in both steady state and transient state in VVER-1000/V446 reactor. In addition, calculations of the fuel center temperature in mean and hot channels for steady state, the temperature of the hottest rod in the fuel assembly, the radial distribution of fuel temperature for both channels, and the maximum clad temperature during the reactivity injection have been performed. The results showed that in the steady state with full power, the fuel centreline temperature in UO 2 is about twice the corresponding values for UN and UC alternative fuels which is very attractive from a safety point of view. In general, due to the large difference between the hot spot point and the melting point in alternative fuels, the thermal power and safety factors can be significantly increased in the next generation of VVER-1000/V446 reactors.

The divertor mockup design for the Pakistan Spherical Tokamak (PST) needs several improvements. These include reducing the total number of upper and lower divertors to 32 and enhancing the heat load capacity to 1 MW/m 2 . This study involves a thermal analysis of the upgraded divertor design for PST through simulation using MOOSE Framework. The simulation involves two steps – validation and thermal analysis. Validation is achieved by modeling 1D and 2D conduction problems in MOOSE Framework and comparing the results with the analytical solutions. The thermal analysis of a mockup of the upgraded PST divertor configurations with tungsten and graphite as armor materials under peak thermal heat load conditions of 1 MW/m 2 is performed. The results of the transient thermal analysis reveal that the mockup surface temperature rise for tungsten remains lower than that for graphite. The findings of the study indicate that both tungsten and graphite mockups can sustain a specific heat flux of 1 MW/m 2 for up to 3 s. However, even being a good conductor of heat, tungsten’s high atomic number (Z) poses challenges in plasma stability. On the other hand, graphite when paired with an efficient cooling system for controlling surface temperature rise is a superior armor material due to its affordability, ease of availability, and better compatibility with plasma environments. It is recommended as the preferred armor material for the upgraded divertor mockup in the PST system.

Heat removal from the core of nuclear reactors is a crucial safety requirement that needs much analysis and attention. Coolant channel flow blockage is one of the unintentional cases that may influence the required heat removal rate. As a result, unsustainable over heating of the fuel cladding could occur resulting eventually in affecting the fuel integrity. This paper deals with simulating the effect of total channel flow blockage in a single fuel assembly on its steady state thermal hydraulic parameters. Thermal hydraulic calculations are carried out for the much less studied IRT-4M ducted fuel assembly type in a 10 MWth WWR-S tank in pool research reactor. The simulated IRT-4M fuel assembly comprises six rounded rectangular fuel elements in which seven channels are used for cooling. A mathematical model is developed in which all steady state thermal hydraulic parameters and their distributions could be predicted for the blocked and unblocked flow channel cases. Calculations were performed for two separate hypothetical flow channel blockage cases in the innermost channel and the middle channel of the IRT-4M fuel assembly. The results showed that the maximum clad temperatures for the two-flow blockage cases are nearly 127.4 °C, and 132.2 °C, respectively. The temperature increases resulting from a single channel blockage are not high enough to cause fuel damage, although some local coolant boiling might occur. Model validation and results support the model outcomes and its applications in the safety of the research reactor during abnormal operation.

The purpose of the present work is to determine the neutronic characteristics of annular cylindrical prismatic Very High Temperature Gas-cooled Reactor (VHTR) with different advanced fuel designs. The study includes detailed neutronic investigations of the eight types of TRISO (containing SiC layer) and TRIZO (containing ZrC layer) fuel particles for annular VHTR core. All calculations were performed using continuous energy Monte Carlo code MVP2.0 and MVPBURN with JENDL4.0 nuclear data library. The obtained results indicated that SiC + PyC and PyC layers in TRISO fuel appeared to be not excellent choice for the annular VHTR core due to its shorter operation year and, lower fuel burnup. The annular cylindrical prismatic core fueled with TRIZO fuel particle shows good neutronic characteristics compared to those designed with other TRISO fuel particles.

Over the last two decades, extensive research has been conducted to enhance the control performance of reactor power. Various methodologies have been suggested and implemented to achieve optimal power control in nuclear reactors. However, due to the diverse characteristics and inherent uncertainties in the models, devising optimal controllers for nuclear systems remains a complex task. To address this, numerous approaches have been adopted to ensure controllability and resilience, aiming for an optimal nuclear power reactor controller. The Model Predictive Control (MPC) algorithm has garnered significant attention as a viable approach to boost operational efficiency and overall system utility. In this research, Particle Swarm Optimization (PSO) method and MPC controller are combined together to form a novel algorithm termed PSO-MPC, aiming to amplify the system’s performance and overcomes the local minima problem that basically happens when using MPC controller. First, MPC controller is applied to the nuclear reactor model, then the suggested technique PSO-MPC also is applied to the system and a comparison of the outcomes using both techniques is done. The results demonstrate enhanced system response using the innovative technique.

NaI (Tl) scintillation detectors are widely used in laboratories and in the field of gamma-ray spectrometry where detector efficiency is a key factor for successful quantitative γ ray analysis. In this work, Monte Carlo simulations using MCNPX code have been performed in order to evaluate energy resolution as a full width at half maximum (FWHM) of the peak and absolute efficiency of NaI detectors. Also, the detection efficiency and energy resolution for the NaI (Tl) scintillation detector of Tehran Research Reactor (TRR) pollution monitoring laboratory was measured. Simulation results were compared to experimental data and good agreement are obtained. In radioactive analyzing laboratory, precise determination of the detector efficiency according to any different gamma energy is a very important parameter whereas it uses in daily laboratory work.

In the light of radioactive material leakage (due to containment bypass or failure) to the auxiliary building and its service compartments under severe accident circumstances, key techniques related to accident source terms in the nuclear emergency building were researched. During the research, the source-term diffusion and migration patterns were simulated in the building compartments, including the simulation of the gradual diffusion and decay process (driven by circulating air) of fission-derived gases and aerosol toward each compartment. Besides that, by taking a M310 unit at Fangjiashan NPP as the application object, the scenarios of initial accidents of LB/MB/SB-LOCA, MSLB and SGTR were chosen to establish a real-time diffusion calculation model for in-building radioactive source terms under nuclear accident circumstances. While considering the circumstances before and after being intervened by accident mitigation measures, the concentration distribution of radionuclides in the auxiliary building and personnel total-exposure radiation dose rate were quickly assessed online, and a basis was provided for the visualization of radiation dosage and the assessment of received radiation dosage and injury & death hazards during personnel activities outside the main control room.