Volume 89, Issue 5

The safety and reliability of nuclear facilities hinge critically on addressing metallic surface contamination in nuclear waste waters. This contamination poses significant hazards to the environment, human health, and the structural integrity of equipment. Key contaminants include heavy metals such as lead, cadmium, and mercury from industrial processes, and radioactive isotopes like uranium, plutonium, and cesium, which present severe radiological risks due to their formation during nuclear reactions and fuel cycles. Corrosive chemicals further exacerbate the problem by promoting the accumulation of rust and other metallic compounds. Additionally, organic contaminants from equipment leaks and microbiological elements, including fungi and bacteria, can form biofilms that accelerate the corrosion process. The objective of this review is to evaluate the various techniques used to identify and quantify these contaminants on metal surfaces, such as surface sampling and microbiological analysis. By implementing appropriate mitigation measures based on these findings, it is possible to reduce risks and ensure the safety and operational integrity of nuclear plants. This comprehensive assessment aims to provide a framework for enhancing contamination management practices in nuclear facilities.

The paper is focused on how to apply the Probabilistic Safety Assessment (PSA) methodology to assess the nuclear safety risks associated with the transportation of radioactive material. The methodology is comprehensive and covers the different mode of transports (for example, by road, by train, by ship and by air) used for the transportation of radioactive material. This methodology can be applied to any specific mode of transportation of radioactive material as well as to any specific country. The systematic assessment starts with identification of risk associated with transportation mode and its initiating events to analyze what can go wrong in transportation of radioactive material and how likely it can be. The accident involving the radioactive material has sensitive information and this information is not publicly available, therefore the generic accident data can be utilized to develop the probabilistic models and quantify the risk associated with radioactive material transportation. Another fact is the transportation of radioactive material accidents are extremely rare, consequently the associated statistics are insufficient for the development of PSA models. The lack of statistical data induced the need to use the generic accident data for predicting accident frequency estimations. The paper explores the feasibility of using probabilistic methods to assess the safety risks associated with radioactive material transport. The paper describes how to perform the comprehensive probabilistic assessment and creating a generic accident event tree that is based on road traffic accident data. The focus of the event tree is to outline a range of different traffic accident scenarios and their respective probability of occurrence and its consequences.

A novel methodology to analyze the trend of radioactive effluent data from nuclear power plants (NPPs) has been proposed to estimate the possibility of abnormal releases of radioactive materials, and to identify sensitive radionuclides with high correlation to specific events and abnormal releases. Actual discharge data including abnormal releases for 21 years (1,610 reactor-years) from 62 US and 22 Korean pressurized water reactors were collected. A release correlation factor (RCF) was newly proposed as a quantitative indicator of the abnormality of radioactive discharges based on annual discharge data from each NPP. The probability of correct estimation of abnormality using the RCF for a test case ranges 27–36 %, and such a modest estimation power was ascribed to the relatively low contribution (i.e., about 12 %) of a specific causal event to the annual radioactivity release. Through this, the feasibility of the RCF for estimating the abnormality of radioactive discharges was demonstrated. The RCF was derived for three types of causal events, and subsequently the sensitive radionuclides to each specific event were derived as 88 Kr, 137 Xe, and 138 Xe for leakage from gas decay tank, 3 H for leakage from steam generator’s power operated relief valve, and 85 Kr, 88 Kr, 133 Xe, 133m Xe, 3 H and  14 C for leakage from defected spent fuel during handling. The new approach proposed in this study is to be used to suppose the significant unreported abnormality in radioactive effluent data and the potential causal event, which may contribute to improving the safety and performance of radioactive effluent control as necessary.

Controlling and managing nuclear waste is a significant challenge due to the harmful effects of radioactive materials on human health. To address this, long-term storage solutions are essential. Artificial Intelligence (AI) and Machine Learning (ML) are being utilized to make nuclear waste management safer, more effective, and efficient. This paper evaluates various applications of AI and ML in the field of nuclear waste, covering aspects such as predictive maintenance, waste sorting, and classification. AI and ML enhance real-time monitoring of storage conditions and optimize waste handling procedures through advanced data processing capabilities. Implementing cutting-edge solutions is crucial to protect public health and the environment from radioactive waste. The purpose of this evaluation is to examine how AI and ML improve nuclear waste management processes. These technologies can reduce human exposure to harmful materials and increase the safety and efficiency of managing nuclear waste through advanced predictive capabilities. The introduction of AI and ML in nuclear waste management is driving significant changes and innovations, addressing current issues, and establishing new guidelines for future policies.

This study investigated gas-cooled fast reactor performance utilizing natural uranium as the fuel employs modified constant axial shape of neutron flux, nuclide densities, and power shape during life of energy production (CANDLE) shuffling strategy in the axial direction. The performance has been carried out on a reactor with various power levels in the range of 2,700–3,100 MWt and refueling every 10 years of burnup. The main importance of the modified CANDLE (MCANDLE) is that it can utilize natural uranium as fuel without needs reprocessing or enrichment plant. During this work, the core has been partitioned into 10 regions of similar volume along the axial direction. The fuel was filled into the first region, then transported to the second region after 10 years of burnup, and then to the third region 10 years later. This technique was practiced to all 10 regions, and the fuel in the tenth region was discharged in the reactor core. 10 % of natural thorium, natural uranium and enriched nitride (15 N) were used as fuel input. The calculations were performed employing a standard reactor analysis code (SRAC). The collision probability method (PIJ) module of SRAC was used to calculate cell burnup, and the reactor core design calculations were done with the CITATION module of SRAC, which used JENDL-4.0 as a nuclear data library. The result shows that a high power level causes an increasing rate of burnup level and effective multiplication factor. The highest average discharge burnup level is about 31.3 % HM for case E and the lowest average discharge burnup is about 27.2 % HM for case A.

Magnetite (Fe 3 O 4 ) is increasingly valued for its excellent performance on radioactive wastes disposal. Here, Fe 3− x Nd x O 4 (0 ≤ x  ≤ 0.9) components are synthesized using a co-precipitation method to simulate the solidification capacity for trivalent actinides in Fe 3 O 4 . These components are characterized by X-ray diffraction, X-ray photoelectron spectroscopy, Fourier transform infrared spectroscopy and Raman, demonstrating that with the increasement of Nd(III) content, the samples gradually transformed from inverse spinel to orthorhombic phase. And x  ≤ 0.6, Nd(III) can be solidified to Fe 3 O 4 lattice via replacing octahedron Fe(III). Furthermore, the vibrating sample magnetometer tests manifest that Fe 3− x Nd x O 4 (0 ≤ x  ≤ 0.9) components hold the capable of responding to external magnetic field, which makes magnetic separation feasible. Our findings are expected to provide a promising idea for radioactive wastes disposal.

To reduce the potential and volume of radiotoxicity of radioactive wastes, minor actinides (MAs) elements recovered from discharged fuel are to be recycled. In this context, the main aim of this work is to examine the possibility of using thorium-based fuel for burning extra MAs in the fourth generation European Sodium cooled Fast Reactor (ESFR). For that MCNPX transport code was used to design a representative fuel assembly of this concept. The reference uranium-based fuel was uniformly replaced by thorium mixed with different fractions of MAs and burned for 2050 effective full power days. The influence of MAs on the evolution of reactivity and fuel transmutation is being investigated to ensure the performance of the fuel. Also, different neutronic and safety parameters such as neutron yield, neutron spectrum, neutron flux, effective delayed neutron fraction, Doppler constant, sodium void reactivity coefficient, were computed and their change with MAs content was investigated. Results indicate that recycling of MAs in ESFR using thorium-based fuel has a positive effect on reactivity and transmutation evolution but on other hand has a negative effect on safety parameters.

Graphite nuclear properties such as moderating power and absorption cross-section, are not as good as those of heavy water. But its pure form can be prepared. Its structural and thermal properties are good and it has a high thermal conductivity. The thermal neutron in graphite performs an average of 1,200 scattering collisions before it is absorbed. This very low absorption cross section makes graphite as an ideal material for applications in nuclear reactors. In the current research, graphite is assumed as a diffusive medium due to its low absorption cross-section (0.0035 b) and having a low mass close to the neutron mass. In this medium: Boron ( 10 B), Cadmium ( 113 Cd), Samarium ( 149 Sm), Europium ( 151 Eu), Hafnium ( 177 Hf) and Gadolinium ( 157 Gd), separately are also considered as neutron absorbers. The aim of this paper is obtaining the solitary wave form of reaction rate in graphite diffusive medium using these neutron absorbers.

Probabilistic Safety Assessment is a tool for improving a nuclear installation and its safety in all its function phases. PSA provides a methodical way of identifying sequences resulting from a wide range of initiating events concerning an accident, and includes systematic and realistic determination of accident frequencies and their consequences. This paper analyzes the unavailability of the TRIGA SSR reactor primary circuit, obtaining the system unavailability value, identifying main contributors to system failure, and calculating risk importance factors. Two cases were considered: one without considering Common Cause Failure (CCF) in the developed fault tree, and the second considering CCF on pumps and heat exchangers lines. The aim was to assess their influence on the primary system’s unavailability. Two importance factors, Fussell–Vesely and Risk Achievement Worth, were employed to identify components with a strong impact on the risk (probability of system failure). These factors aid in enhancing maintenance activities and testing. The analysis utilized the EDFT code, part of the PSAMAN code package developed in SCN-Pitesti.

A thermal-hydraulic model has been developed by RELAP5 code to simulate the thermal-hydraulic behavior of a WWR-S research reactor under loss of flow accident (LOFA). The reactor power is 2 MW with downward flow direction and different types of fuel bundles of different power densities and different coolant flow-rates. The simulation is performed for two scenarios; protected and unprotected LOFA. In the protected LOFA scenario; Scram is triggered as soon as the coolant flow rate reaches 85 % of its nominal value while in the unprotected LOFA scenario; the reactor continues operation during pump cost-down and natural circulation flow. Once the flow is low enough, the buoyancy force increases in the core leading to flow inversion phenomenon and establishment of a natural circulation mechanism within the core coolant channels in both scenarios. In the protected LOFA scenario, the coolant remains subcooled by a vast margin during transient while bulk boiling is predicted at the upper part of the core for the unprotected LOFA scenario. The heat fluxes leading to the onset of nucleate boiling and the critical heat flux are predicted and the safety margins are determined. The results for both scenarios are analyzed and discussed.

A quick tool capable of processing output files generated by the SCHISM particle tracking module has been developed to handle complex data processing procedures and conduct parameter sensitivity testing. This tool performs calculations, considering the influence of half-lives and partition coefficients, for the conversion of activity concentrations, plots particle diffusion trajectories, conducts statistical analyses of regional particle amounts, and presents activity concentration within the defined area. Using the IAEA MODARIA program as an example, this study investigates the transport of radioactive 137 Cs in the ocean following the Fukushima accident on March 11, 2011. A three-year diffusion simulation was conducted by incorporating data such as ocean currents, wind fields, and tides. The calculated results demonstrate a high level of consistency with measurements, and the calculated results in the low activity concentration fall within the variability range reported in the literature.

NCABWR (Natural-Circulation-Advanced-Boiling-Water-Reactor) is a pressure-tube-type BWR with multiple inter-connected Main-Heat-Transport (MHT) system loops and steam drums (SDs). SD-Level-Control is related to, the MHT coolant inventory and sustained heat removal, hence the overall safety of the Nuclear-Power-Plant (NPP). SD-Level is vital for ensuring presence of adequate coolant-inventory with sufficient and prolonged cooling of fuel, for normal-operation and accidents. In the initial-MHT-configuration of NCABWR the four MHT-loops are, inter-connected at the Common-Reactor-Inlet-Header (CRIH), which influences the SDLC adversely; such occurrences are, not reported in the open literature. Open loop (without SD-level-controllers) step-change response showed that initial MHT-configuration response is stable, non-oscillatory and non-diverging even for a high-magnitude step-change in the feed-flows. Closed-loop (with SD-level-controllers) response, even to minor disturbances leads to diverging SD-Levels due to CRIH-MHT-configuration (Gaikwad, A.J., Vijayan, P., Iyer, K., Bhartiya, S., Kumar, R., Lele, H.G., Ghosh, A.K., Kushwaha, H.S., and Sinha, R.K. (2009). Effect of loop configuration on steam drum level control for a multiple drum interconnected loops pressure tube type boiling water reactor. IEEE Trans. Nucl. Sci. 56: 3712–3725). CRIH is required to overcome Postulated-Initiating-Event (PIE) for single-accumulator-failure-LOCA. A hypothetical MHT-configuration with partitioning of CRIH will isolate each loop from one another and lead to a stable SDLC, but then for the above-mentioned PIE safety is not met. LOCA in one of the four ECCS header-compartments require a mandatory interconnection among, the four MHT-Loops at CRIH. Solution to SDLC issue were explored with changes in MHT-configuration and with innovative solution-schemes for feed-water-system and SDLCs. In the present study the pioneering solution-schemes with individual SD pressure control for SD level control and the Split/Double baffle are proposed, implemented and the performance was tested with simulation are discussed in details with other solution schemes.

The essential service water system (ESWS) in nuclear power plants circulates the water that cools fan coolers, cooling water heat exchangers, and condensers and other components before dissipating the heat into the environment. Because this includes cooling the systems that remove decay heat from both the primary system and the spent fuel rod cooling ponds, the ESWS is a safety-critical system. In locations without a large body of water in which to dissipate the heat, water is recirculated via a cooling tower. Cooling towers fall into two main categories: Natural draft and Mechanical draft. Since, the mechanical draft cooling towers are much more widely used; the focus is on them in this work to measure its performance. Therefore; the present work addresses the thermal acceptance test for mechanical draft water cooling tower in order to determine its thermal capability in accordance with the cooling technology institute (CTI) code recommendations. Both the Characteristic Curve and the Performance Curve Methods are used. A detailed description of the test procedure according to the CTI ATC-105 test code is presented, and the test code is applied on the Egypt second research reactor (ETRR-2) cooling tower as a case study. All the instruments used in the reactor plant are calibrated prior the test and the measured data is collected at regular intervals according to the code recommendations. The test result shows a tower capability of about 105 % and so the tower is thermally accepted.

Proportional–Integral–Derivative (PID) controllers have been optimized and used to overcome many types of problems in nuclear reactor systems. The high performance of PID controllers depend on optimizing their gains. In this research, an optimized robust PID controller is proposed to control power perturbations in a pressurized water reactor (PWR). The optimization process of robust PID using particle swarm optimization (PSO) algorithm aims to adapt PID gains then after that, H-infinity controller is used. The results show a good performance when that suggested hybrid controller is applied to the nuclear power system since the suggested design makes the system robust due to applying H-infinity method, in addition to get the benefits of the optimized PID controller.