Volume 89, Issue 4

The use of thorium oxide in thermal reactors is currently being explored due to its promising outcomes. One primary concern is how to reduce the pollution of core components and nuclear waste. The inclusion of 232 Th in the reactor leads to the production of 233 U and decreases the formation of minor actinides. On the other hand, 238 U increases the production of 239 Pu, a toxic and strategically significant isotope, along with minor actinides. Therefore, the potential of thorium to reduce the production of 239 Pu, one of the most harmful isotopes, can be assessed in the Bushehr VVER-1000 reactor. This study involved replacing some uranium in the Bushehr VVER-1000 reactor with 232 Th. The research focused on examining the environmental impact of nuclear waste, including activity, chain reactions, and isotope levels, over a two-year period. The impact of the new fuel substitution was evaluated in the Bushehr reactor, which has a power of 3,000 MWth, in three scenarios: thorium addition to 1.6 % enriched assemblies, thorium addition to 2.4 % enriched assemblies, and thorium addition to 3.6 % enriched assemblies. These changes were analyzed in terms of nuclear waste contamination, plutonium production, fuel burn-up, and conversion ratio, and compared to the reactor using UO 2 fuel. The simulation was conducted using the MCNPX 2.6.0 computational code and heterogeneous geometry. The results indicate that nuclear waste pollution decreased when thorium was added to the 1.6 % and 2.4 % enriched assemblies, but increased when added to the 3.6 % assemblies. Additionally, fuel burn-up increased with the addition of thorium in the 1.6 % and 2.4 % assemblies, but decreased in the 3.6 % assemblies. However, the conversion ratio increased in all cases. The fuel temperature coefficient (FTC), moderator temperature coefficient (MTC), and void coefficient (VC) were calculated and evaluated.

Diffusion is the predominant mechanism governing the transport of 129 I through geosphere. Consequently, the assessment of the experimental findings on two-compartment diffusion reservoirs of IO 3 − within compacted bentonite involved the application of three distinct diffusion models: CC–CC (constant concentration), CC–VC (variable concentration), and VC–VC. To ensure the reliability of the obtained diffusion coefficients, multiple laboratory tests were performed for internal comparison. The experimental results revealed that IO 3 − diffusion coefficients were ranging from 3.83 × 10 −13 to 1.91 × 10 −11   m 2 /s. These techniques using three mathematical models could be conducted to estimate the diffusion coefficients of non- or weakly-sorbing radionuclides on compacted bentonite for safety assessment of radioactive waste final disposal.

HTTRs are high temperature testing reactors which are considered generation IV nuclear power reactors. The reactor core startup and fuel burnup time dependent are simulated and analyzed. For the present computer model core startup, fuel loading with approach to criticality, control rod critical positions, scram reactivity and control rod worth, different core excess reactivity are evaluated. The results are compared with both typical measurements and other previously published methods. Moreover, the time dependent core model and fuel burn up are simulated to calculate the core multiplication factor, cycle length, discharged burnup, core isotopic evolution with time and burnable poisons behavior.

The physical protection system (PPS) is one of the main systems used in nuclear facilities to protect these facilities from terrorist attacks, the possibility of bombing, sabotage, and theft. Installation of a security system for nuclear waste stores from the outside is necessary and very important for early detection and physiological deterrence greater than the internal security system. From this perspective, this work provides a solution for the protection of nuclear waste stores, which means detection and defense systems. This paper introduces an advanced security system and proposes a design for a detection and defense security system for nuclear waste storage against stealth terrorist attacks. The security system that is being suggested will have smart camera surveillance system, electric chock system, intense flashlight barrier, high-intensity sound barriers, and dual-side infrared rays. Three consecutive processes – detection, deterrence, and defense – will be operated around the building of the waste storage. The concept may be implemented by establishing an early detection system to detect any intrusion by terrorist attackers through a protected area surrounded by storage. The following step is deterring the attackers from completing their mission by installing advanced deterrence system and putting delay components along the path of their attack on the waste storage to complicate the mission. Then the defense system is operating to stop terrorist progress. Finally, the work make an evaluation process of the overall security system which has been designed by using EASI program to verify the accepted level of physical protection, maintain the validity and updates of security system components, and follow procedures, rules and requirements of the nuclear regulatory authority at the national level and meet the international IAEA concept and recommendations.

Electron cyclotron resonance (ECR) is a common pre-ionization technique to assist ohmic startup in tokamaks. In this research, ECR assisted pre-ionization in GLAST-III tokamak is comprehensively investigated via Multiphysics 3D COMSOL simulations. Optimal conditions for ECR microwave absorption at fundamental mode are determined to achieve reliable plasma startup. The effect of filled gas pressure on the ECR plasma parameters such as plasma density and plasma temperature is thoroughly investigated in the range of 0.08–1 Pa. The results reveal that the ECR absorption of the incident microwaves is significantly enhanced at low enough filled argon pressure. The behavior of absorbed microwave power is significantly changed for different filled gas pressures. In case of high gas pressure, the microwave power is deposited locally in front of the waveguide, whereas in case of low pressure (0.08 Pa), the microwave power is uniformly deposited along the resonance layer. Moreover, the experimental investigation extends to nearly identical operating conditions, confirming the ECR absorption. The experimental data aligns with the simulation results, collectively affirming the efficacy of optimized pre-ionization in the low gas pressure scenario (0.08 Pa) within the GLAST-III tokamak.

Probabilistic safety assessment (PSA) has been applied to evaluate the safety of nuclear reactor systems. The results of PSA are used by designers, utilities and regulatory body to confirm the reactor design, to change operation procedures, to enhance the reliability of safety systems, or to modify regulation. Fault tree analysis (FTA) has been used as a deductive means for PSA. System components are assumed to always have defined failure probabilities; however, in reality this assumption is questionable and sometimes there is insufficient data about the failure probability of some components. To solve this problem, fuzzy approaches have been proposed and implemented. The aim of this study is to apply a fuzzy reliability algorithm to estimate basic events failure probabilities for the shutdown system of a research reactor. The reactor is a 22 MW light water open pool type material test reactor and the shutdown system is responsible for the fast shutdown of the reactor whenever safety limits are exceeded. Then, to illustrate the ability of the algorithm, the results are compared with real basic event failure probabilities. The results verify the competence of the algorithm and confirm that it can be applied as an alternative method when quantitative data are not available. Also, a risk importance measure is done on the fault tree of the shutdown system to illustrate which components need to be focused for improvements to achieve the safe operation of shutdown system.

A secure and highly reliable communication network is a crucial component of the distributed control system (DCS) in nuclear power plants (NPPs), particularly for the protection system. However, there have been numerous reported cases of cyber-attacks in the field of nuclear power, and the cybersecurity threats faced by a DCS are increasing. Given that the DCS is a complex integrated system combining computers, communication, and control with a large number of interactions between components and external interferences, traditional linear safety analysis methods fail to meet its requirements. To minimize cybersecurity accidents caused by potential vulnerabilities, it is necessary to conduct systematic safety analysis on DCS cybersecurity accidents in NPPs. Therefore, this paper selected a specific case of cybersecurity accidents in a NPP for abstract modeling and reproduction. The causal analysis based on system theory (CAST) method was applied to conduct an analysis of the accident causes. It can identify existing vulnerabilities and potential factors impacting the safe operation of NPPs, and countermeasures and recommendations can be proposed. The reference for assessing potential risks and devising solutions can be provided for other DCS communication networks.

Pressure and heat transfer coefficient (HTC) are parameters used to measure the performance of microchannel evaporators (MCE). By keeping the same overall dimensions, optimised hydraulic diameters of three different port shapes, square, rectangle and trapezoid, are analysed using CFD software to study the effect of geometry on the performance characteristics of MCE. The number of square, rectangle, and trapezoid channels is 580, 986, and 812, respectively. CFD analysis of the evaporator using R134a refrigerant is performed using the Volume of Fluid (VOF) with the SST k-ω model and Lee model for interphase mass transfer trace. Variation of pressure and HTC concerning volume of fraction liquid refrigerant (α) is compared for these three port structures. The trapezoid port microchannel evaporator (MCE) exhibits the highest pressure drop compared to square and rectangular port MCEs. The rectangular port MCE demonstrates the highest heat transfer coefficient among the configurations studied. The rectangular port MCE outperforms both square and trapezoid port MCEs in overall performance.

The experience of Fukushima brings out the importance of the emergency response plan during or after a severe nuclear power plant (NPP) accident. To deal with the accident emergency response situation, the National Atomic Research Institute (NARI) in Taiwan developed an emergent radiological dose evaluation system with the Central Weather Administration (CWA) called “A2CDOSE”. A2CDOSE can evaluate the eight-day projected dose after a nuclear accident and provide the predicted dose consequences for decision-making. Nevertheless, a source term generation method was still needed during the accident. This study chose MELCOR as the comparison target for the source term simulation issue with RASCAL. A MELCOR calculation of Maanshan NPP was built and a 5 % containment leakage was set for both MELCOR and RASCAL source term calculations. It seems the calculations of noble gases were consistent between MELCOR and RASCAL. However, the Cs-137 and I-131 calculated by MELCOR were twice compared to the RASCAL results, which may have a chance to cause different calculations of public protective actions during an accident. After the source term analysis, several atmospheric dispersion cases were simulated by A2CDOSE and RASCAL for further comparison.

This paper presents a reliability model for digital reactor protection systems (RPSs) in floating nuclear power plants (FNPPs) that accounts for both the internal characteristics of RPS and the external environment. The internal characteristics of RPS include independent failures and common-cause failures (CCFs) of components, repair behavior, and actuation logic degradation. For the external environment, we incorporated a parts-pressure method and used the environmental factors to describe the impact of marine environment at component level. Detailed Monte Carlo simulation (MCS) algorithm was proposed to solve the reliability models with different environmental factors, and the results showed that the maximum value of the environmental factor was 3.2 under the requirements that the probability for RPS failing to generate the trip signal does not exceed 1 × 10 −5 and the spurious trip frequency does not exceed one time per year. Reliability indexes, such as failure probability and spurious trip frequency, were also derived. The 90 % confidence intervals of these two indexes were further calculated in the uncertainty analysis by using the kernel density estimation (KDE) approach.

Mental workload has an important impact on the human performance of ship operators. The purpose of this study is to develop an evaluation method to evaluate the ship operator’s mental workload. First, an evaluation index system was constructed, and the DEMATEL (decision-making trial and evaluation laboratory) method was used to determine the relative weight of each index. Then, the fuzzy evaluation method was used to calculate the mental workload of ship operator based on the evaluation value and relative weight of indexes. The evaluation result of the developed method showed that the ship operators’ mental workload were 3.78 and 2.99 under tasks 1 and 2, and the ship operators’ mental workloads were at the ‘very high’ and ‘high’ levels based on the maximum membership principle. Also, the evaluation result of the developed method was consistent with the evaluation results of subjective mental workload assessment technique. It proves that the developed method can effectively evaluate the ship operator’s mental workload level.

With the widespread application of Field Programmable Gate Array (FPGA) technology in the instrumentation and control (I&C) systems of nuclear power plants (NPPs), its design reliability has become the key to ensuring the subsequent safe operation of NPPs. However, the traditional document-based design method is no longer sufficient to meet the current design needs of complex systems, and there is a lack of research related to the reliability and security analysis of FPGA-based systems. Therefore, the overall goal of this study is to integrate reliability analysis into the model-based systems engineering (MBSE)-based design of FPGA-based I&C systems. Based on the typical application characteristics of the FPGA, a digital design framework for FPGA-based systems is proposed to realize a traceable design process from requirements to implementation. Subsequently, a model-based forward comprehensive reliability analysis method is proposed, which uses a SysML-based mapping mechanism to integrate Failure Mode and Effects Analysis (FMEA) and Fault Tree Analysis (FTA) methods into the design process, so as to comprehensively identify and analyze failure modes from systems to components and optimize and improve system design. Finally, the proposed method is applied to the analog output design of FPGA-based systems. The proposed method can provide guidance for improving the reliability of FPGA-based DI&C systems and provide theoretical basis for the further engineering application.