Volume 90, Issue 2

Various heat pipe-cooled reactors have been conceptualized and designed in recent years, ranging from kilowatts to megawatts, suitable for remote and extreme environments. However, existing designs employ monolith cores, which are saddled with heat transfer deteriorations due to thermal expansion and subsequent deformations are some of the key challenges facing monolith heat pipe reactors. Addressing this, a novel 300 kW heat pipe has been designed by introducing lead liquid instead of the monolith. A comprehensive Monte Carlo neutronics analysis of the designed core has been conducted including its thermal analysis. The core design comes with the following parameters: fresh core excess reactivity of 2.9914 $ that can support a 10-year life cycle, and a negative temperature coefficient of −0.1648 pcm/K. The axial and radial peak power factors at critical states were calculated to be 1.67 and 1.35 respectively. Preliminary thermal hydraulics analysis shows maximum temperatures of 1044 K under normal operation and worst-case accident scenarios (3 heat-pipe failures) of 1,149.9 K. These temperatures are lower than the melting point of uranium nitride fuel and lower than the possible heat pipe boiling failure limit at 1,300 K. Thus, the possibility of fuel melting/heat pipe boiling accidents is avoided. A hypothetical criticality accident was simulated, assuming the core was flooded with water, leading to an increased fission rate. However, the implementation of three emergency rods in the core ensures the subcritical of the reactor.

This study presents developing and validating a nonlinear dynamic model for simulating thermal processes in the VVER-1200 reactor core, a third-generation pressurized water reactor with enhanced safety features and improved thermal efficiency. The model employs a point kinetic mathematical approach, incorporating key neutronic parameters extracted from previous analyses of the VVER-1200 under normal operating conditions. These parameters, including neutron flux distribution, power peaking factors, and reactivity coefficients, were integrated into the reactor kinetic model to capture complex feedback mechanisms. In this study, the variation of reactor power with control rod insertion is examined under normal operating conditions. The analysis focuses on how different levels of control rod insertion influence reactor performance, providing insights into optimizing operational efficiency and safety. Monte Carlo code MCNP6 was used to estimate neutronic and kinetic parameters for the proposed model. The model’s validation involved comparing simulation results against detailed Monte Carlo N-Particle calculations, specifically focusing on power changes due to reactivity variations induced by control rod movements. Results showed strong agreement between the nonlinear dynamic model and MCNP calculations. The validated model provides valuable insights into the VVER-1200 core’s thermal behavior under steady-state and transient conditions. It can investigate critical hot spots, identify thermal limits, and assess system perturbation responses. The model’s computational efficiency makes it a practical tool for further optimization and safety analyses of the VVER-1200 design.

The sliding reflector control system is utilized for core power level regulation through neutron leakage, offering a simpler structure in comparison to the control drum and control rod systems, thus minimizing the mass of the space reactor. This paper presents a conceptual design of a small – scale homogeneous lithium-cooled space fast reactor based on the sliding reflector control system. In this study, accurate modeling of the radial subregion of the reflector block is achieved, and the influence mechanism of the leakage rate on the key parameters of core neutronics is quantitatively analyzed. By pulling down the sliding reflector layer until the core reaches the critical state, the designed core shows a negative reactivity temperature coefficient of −0.3083 pcm/K. Meanwhile, the peak value of the inhomogeneity coefficients for axial and radial neutron flux density distributions are 1.414 and 1.268, respectively, satisfying safety and economic requirements. The analysis indicates that changes in the neutron leakage rate directly impact the reactivity. Building upon these findings, three reflector arrangement schemes summarizing all possible sliding scenarios are proposed: the uplift sliding reflector, the down-sliding reflector, and the intermediate separating sliding reflector. The results of neutron calculations show that the down-sliding reflector has the largest adjustment accuracy, with reaching criticality when the axial leakage length reaches 2.85 cm. The uplift reflector produces the most uniform axial neutron flux density distribution, peaking at 1.341.

This work analyzes advancements in nuclear energy, focusing on spent fuel reprocessing and waste disposal. Key technologies such as pyroprocessing, aqueous reprocessing, and the DUPIC (Direct Use of Spent PWR Fuel in CANDU) process are examined for their effectiveness in reducing environmental and safety risks. Pyroprocessing is highlighted for its ability to enhance actinide recovery and reduce waste volume, while the PUREX process demonstrates high recovery efficiencies for uranium and plutonium. The integration of Computational Fluid Dynamics (CFD) and chemical kinetics modeling optimizes process parameters, improving reprocessing outcomes. The DUPIC method shows potential in improving reactor performance and minimizing waste. Long-term waste management strategies, such as dry cask storage and geological disposal, are also explored, with advancements in safety and heat dissipation technologies. Innovations like autonomous passive cooling and antineutrino monitoring are promising for spent fuel management security. Additionally, the study highlights Generation IV reactors and Small Modular Reactors (SMRs) as critical to minimizing nuclear energy’s environmental impact. The work emphasizes continued progress in reprocessing and waste management to address technical, economic, and environmental challenges, ensuring the viability of nuclear energy.

Understanding the movement of radionuclides (RN) in the subsurface environment is of paramount importance, particularly when it comes to the planning and assessment of facilities devoted to the disposal of radioactive waste. Comprehensive mathematical models serve as indispensable tools in this regard, demanding a profound and thorough understanding of the intricate mechanisms underlying radionuclide transport. The effective application of these models is contingent upon accurately determining the required input parameters. This is a critical aspect to consider given the inherent physical and chemical variations exhibited by the subsurface environment. These variations can induce significant effects on the movement of RNs below ground, potentially altering the predicted outcomes of radionuclide transport. This paper presents the findings of a comprehensive investigation that was conducted utilizing both advective-dispersive experiments (ADE) and multiple processes nonequilibrium (MPNE) inversion. These methodologies were employed using the advanced HYDRUS code, which is highly regarded in the field. The research specifically focuses on the transportation mechanisms of Cesium (Cs), a common radionuclide, in a medium of crushed granite. The study considers varying conditions, including different flow rates and column lengths, to provide a broad understanding of the behavior of Cs. The findings reveal that the transport behavior of Cs is not only influenced by the different flow rates and column lengths but is also significantly affected by the diffusive mass transfer and nonequilibrium sorption. These factors collectively contribute to our understanding of the complex processes involved in radionuclide transport.

The Kingdom of Saudi Arabia (KSA) is now among the few nations poised to benefit from nuclear reactor technology. For building nuclear reactors in a desert region like the Kingdom, one of the effective approach is to construct small nuclear power plants that can be swiftly deployed in remote areas. Key factors include the performance of these nuclear reactors, which encompasses their innovative design, safe operation, and economic viability. The research begins with a comprehensive analysis of the latest literature on very small modular reactors (vSMRs), including relevant policies and regulations. The reactors’ suitability for Saudi Arabia’s climate is also assessed. Considering the country’s extensive sustainable energy plan, the future energy market in Saudi Arabia is examined. The study covers the gasoline supply chain system, addressing both the front and back ends, safety features, development strategies, and implementation challenges. Additionally, technological advancements relevant to Saudi Arabia, such as control systems, computer platforms, and fuel tolerance during accidents, are explored. The research supports considering vSMRs as a viable addition to the nuclear fleet of KSA.

In order to investigate the transient safety characteristics, Unprotected Transient Overpower (UTOP) and Protected Transient Overpower (PTOP) of small modular natural circulation lead-cooled fast reactor SNCLFR-100 were performed using MPC_LBE, a multi-physics coupled safety analysis code for lead-based reactor. The simulation results indicated that the negative reactivity feedback effect of the reactor during transient slowed down the accident process, and the natural circulation capability of the reactor ensured the heat removal from the core after accident. Under both UTOP and PTOP transient conditions, the peak temperatures of coolant and fuel pellet in the hottest channel of the core did not exceed the safety design limits. The reactor could re-establish a new steady state through natural circulation in UTOP transient, and it was shut down safely after reactor scram triggered in PTOP transient.

Pellet cladding mechanical interaction (PCMI) is a potential failure mechanism for fuel claddings during a Reactivity Initiated Accident (RIA) type event, thus preventing failure of the cladding material during transient conditions is considered one of the primary safety criteria to ensure fuel element integrity. Eight irradiation experiments aimed to validate the experimental setup and test methodology developed in the frame of the MAXSIMA (Methodology, Analysis and eXperiments for the Safety In MYRRHA Assessment) project have been carried out in the Annular Core Pulse Reactor (ACPR) TRIGA core at Technologies for Nuclear Energy State Owned Company (RATEN) – Institute for Nuclear Research (ICN), Pitesti, Romania. Cladding damage due to PCMI in terms of cladding mechanical strain was measured. In Partitioning And Transmuter Research Initiative in a Collaborative Innovation Action (PATRICIA) Project, a work package, is dedicated to obtaining and processing information from fuel segments exposed to power transient during the abovementioned experiments. This paper includes the results of TRANSURANUS and ANSYS thermo-mechanical modelling of one experimental fuel element, consisting in UO 2 with a higher enrichment of about 18 % 235 U and 15-15Ti clad, surrounded by lead bismuth eutectic (LBE). The selected fuel element was exposed in a double pulse test. The results are discussed concerning both fuel and cladding performance and compared with available experimental data and gave positive indications about Multi-purpose hYbrid Research Reactor for High-tech Applications (MYRRHA) type fuel element behavior under RIA conditions.

The Fukushima disaster highlighted the critical importance of a comprehensive emergency response plan for severe nuclear power plant (NPP) accidents. In response to this need, Taiwan’s National Atomic Research Institute (NARI) collaborated with the Central Weather Administration (CWA) to develop an urgent radiological dose evaluation system known as “A2CDOSE.” This system is designed to assess projected radiation doses over eight days following a nuclear incident, aiding in informed decision-making. Considering enhancing the multi-consequence analysis via nuclear material transport, this study connects the settings of ocean radiation contribution in series to develop a multi-sequence, multi-batch method called the ocean-atmosphere coupled analysis of radioactive substances with ocean dispersion code SCHISM. This Year’s (2024) hypothetical cesium-137 release case from the nuclear power plant was used to conduct a technical feasibility test. This method can serve as a reference for evaluating the optimal radiation detection routes for the affected maritime areas during future nuclear safety events, simulating the short-to mid-term effects of radioactive substance releases.

Nowadays most of research in nuclear engineering is on safety of nuclear reactor. Passive cooling attracts most of researchers to use low melting point and high boiling point fluid. Molten salts exhibit low melting and high boiling points at normal atmospheric pressure condition. These qualities make molten salts useful in a variety of applications as a coolant or heat transfer fluid. In this study, Three-dimensional (3D) Computational Fluid Dynamics (CFD) simulations have been used to examine the hydraulic properties and heat transfer characteristics of molten salt NaNO 3 +KNO 3 (60:40 wt ratio) in a horizontal pipe. Analytical data has been used to validate computed results. A comparative analysis has been conducted between the computational results and several correlations. This helps to understand the heat transfer characteristics. Using computational data, the local Nusselt number in the heater region is determined and compared to different correlations. The thermal-hydraulic properties of molten salt obtained from CFD results are compared with existing heat transfer correlations. The most appropriate correlations for obtaining the molten salt’s thermal-hydraulic properties have been discussed. Shah and London’s correlation to determine Nusselt number predicts very well in thermally developing region compared to Churchill and Oze’s correlation. CFD results are helpful to understand the hydraulic behavior of the fluid.

The purpose of this paper is to investigate the optimal parameters for the design and use of eddy current (EC) sensors for non-destructive testing (NDT), aimed at crack characterization by data inversion. A methodology has been established to evaluate the performance of these sensors in the detection of opening cracks and to compare their performances in various measurement conditions. Detection and inversion algorithms were developed and implemented in order to use the most efficient and optimal structure, to quantify the detection quality of defects and to estimate their geometric characteristics. This study has produced very encouraging results in terms of detection performance and data inversion.

Physical protection systems (PPSs) are designed to counter and neutralize any theft or sabotage attempts carried out against a given site. PPS designs are crucial, requiring rigorous analysis and evaluation to ensure their reliability, sustainability and optimal security. This paper assumes a Hypothetical Nuclear Research Center (HNRC) for PPS analysis and evaluation. This work will determine the PPS’s most vulnerable path elements, by demonstrating expected intrusion attack scenarios by taking advantage of the HNRC’s design weaknesses. Also, this work determines the threats resulting from adversary intrusion through the weakness path elements located at the nuclear site. Additionally, this paper develops an analytical methodology for this evaluation and risk assessment, by using the Systematic Analysis of Vulnerability to Intrusion (SAVI) computer program. Through constructing adversary sequence diagrams (ASDs), SAVI can easily identify the HNRC’s most vulnerable paths, pinpointing the areas which require the most attention. Alongside SAVI is the quantitative risk analysis method - the Dai et al. Model – which can calculate the effectiveness and risk of the PPS itself by measuring its degree of uncertainty using effectiveness entropy. To comply with effectiveness and risk standards, this work also introduces security system upgrade processes and determines the optimal security solution for the HNRC.

The Chinshan Nuclear Power Plant is a BWR/4 plant and is in the long-term shutdown period with the fuels still residing in the reactor core. This study presents the methodology for immersion dose analysis of loss of coolant accident (LOCA) with the above conditions and assuming some of the safety systems, facilities and radiation shielding fail. The RADTRAD and MicroShield codes are used in the methodology. This methodology can be used to calculate whole body dose, thyroid dose, and beta-skin dose received by control room personnel for the condition with full core damaged and calculate direct gamma ray radiation dose of control room personnel for the condition with no fuels damaged. The analysis results show that even failures in the safety systems, facilities or radiation shielding during the LOCA, the whole body dose, thyroid dose, and beta-skin dose are well below the acceptance limits specified in the Standard Review Plan.