Saudi International Conference On Nuclear Power Engineering (SCOPE)

The Paul Scherrer Institute (PSI) is the largest research institute for natural and engineering sciences in Switzerland, focusing on cutting-edge research in four fields, namely, Future Technologies, Energy and Climate, Health Innovation and Fundamentals of Nature. PSI develops, builds and operates complex large research facilities, in particular, one of the most powerful proton accelerators worldwide.
An important component of innovative radiopharmaceuticals, especially in oncology, is the availability of various radionuclides with optimal decay properties for the improvement of diagnostic or therapeutic efficacy. The Laboratory of Radiochemistry (LRC) at PSI, in collaboration with the Center of Radiopharmaceutical Sciences (CRS), produces and further develops a variety of accelerator, reactor (or Spallation Neutron Source) and spallation-induced radionuclides via its vast networks. Medical radionuclides must be available with high-specific activity and purity. Here, the choice of nuclear reaction and subsequent radiochemical isolation strategy play a key role.
To support the Swiss research community, PSI develops, builds and operates complex large research facilities (Fig. 1). The site hosts facilities such as the Swiss Light Source (SLS), the Swiss X-ray Free-Electron Laser (SwissFEL), the High Intensity Proton Accelerator (HIPA) facility - which also feeds the spallation neutron source (Swiss Neutron Source, or SINQ), the muon source (SμS) and the Swiss Research Infrastructure for Particle physics (CHRISP). More than 2500 scientists from all over the world use these large facilities for research and development.
A novel irradiation station with high-energy protons at PSI, in the view of enlarging the radionuclide production portfolio in Switzerland as well as Europe, referred to as TATTOOS (Targeted Alpha Tumor Therapy and Other Oncological Solutions), is proposed. The spallation process induced by high-energy protons, utilizing various target material, will provide access to a plethora of exotic radionuclides not otherwise accessible with great scientific potential for nuclear physics, astrophysics and fundamental radiochemistry.

The OECD Nuclear Energy Agency (NEA) Working Group on Analysis and Management of Accidents (WGAMA) addresses activities in the three technical fields of thermal-hydraulics (T/Hs), computational fluid dynamics (CFD) and severe accidents (SAs) related to safety aspects of potential accidental situations in nuclear power plants (NPPs). WGAMA covers existing nuclear reactors and related technologies as well as emerging challenges in evolutionary and innovative reactor designs and nuclear technologies, including small modular reactors (SMRs). WGAMA’s objective is to assess and, where necessary, strengthen the technical basis needed for the prevention, mitigation and management of potential accidents in nuclear reactors and related technologies, and to facilitate international convergence on safety issues, safety assessments and accident management (AM) measures and strategies.
The WGAMA's achievements have been outstanding in providing technical reports and position papers which are reference publications, and in organizing workshops and conferences to discuss innovative methods, materials and technologies in the fields of T/Hs, CFD and SAs.
The paper aims to review and summarize the recent WGAMA activities and outcomes by focusing on T/H analysis of water-cooled nuclear reactors and possible applications to advanced designs, including CFD applications to nuclear reactor safety.
The paper will also describe benefits from, and opportunities for, NEA Joint Safety Research Experimental Projects (JPs) which aim to enhance the technical bases on reactor accidents and CFD and T/H codes’ validation ,with performing series of SETs or IETs to support the analysis of outstanding safety issues. Such projects also contribute to the development and preservation of key technical capabilities, research infrastructure and expertise in participating organizations/countries. Joint Safety Research project also contribute to education and training of the future generation of nuclear safety experts.

KEYWORDS
WGAMA, thermal hydraulics analysis, joint safety research projects

This study evaluates the reactor degradation progression in VVER-1000 due to loss of coolant accident (LOCA) and total station blackout (SBO). The simulation evaluation was carried out using the adaptive SYStem Thermal-hydraulics (ASYST) program. The low-pressure scenario was simulated by modelling 80 mm small break LOCA (SBLOCA) in the cold pressurizer leg in during SBO. This double ended break size was selected to cause a significate faster depressurization during low-pressure simulation and a longer borated cold water injection from the passive hydro accumulators (HAs). This enabled evaluating the times required to attain critical set points during the transient progression. The investigation looked into loss of coolant circulation, fuel and clad heating up, commencement of hydrogen generation, the activation of the passive safety system, changes in pressure, and primary circulation. The results were compared to those obtained using ASTEC 2.1.1.0 for the same postulated events. Low pressure ASYST simulation results showed that the modelled break area was enough to allow primary loop depressurization. These kinds of analyses assist in estimating the time available to perform operator safety actions. This in turn aids in emergency planning and severe accident management. The results revealed that the fuel damage decreases after the introduction of HAs. Actuation of HAs at their actuation set-points provided core cooling by injecting water into reactor core.

In the coupled neutronics and thermal-hydraulics analysis, the void fraction plays a significant role in determining various reactor parameters such as reactor coolant mixture density, neutron moderation, local power distribution, two-phase pressure drop, two-phase flow regimes, and heat transfer coefficient. Consequently, accurate prediction of the void fraction is crucial for nuclear reactor simulations in both steady-state and transient conditions. This research aims to evaluate a range of drift flux correlations frequently employed in the nuclear and oil industries. Initially, a simple and robust one-dimensional two-phase flow code, based on the drift flux model, was developed and validated to assess the performance of the selected drift-flux correlations. Subsequently, a comprehensive statistical evaluation was performed using over 1,600 experimental tests drawn from open literature, which encompassed various vertical and horizontal flow regimes and geometries, including pipes, annulus channels, and fuel assemblies. The evaluation results identified the Hibiki & Ishii correlation as the most accurate, with a mean absolute error of 16.2%, followed by Toshiba and Antonio correlations, with mean absolute errors of 17.45% and 17.69%, respectively. In addition, the same experimental dataset was utilized to derive a new drift-flux correlation for various vertical and horizontal flow regimes. The performance assessment of the newly developed correlation showed an overall improvement, with a mean absolute error of 14.6%, offering a more precise correlation for future nuclear reactor simulations

The KAUST Supercomputing Laboratory (KSL) has been serving the engineering community of Saudi Arabia for many years. This paper aims to cover a few selected highlights of KSL's educational, training, and consulting services. This outreach program has two streams: modelling & simulation and deep learning for engineers. Since 2017, KSL has conducted nine training workshops in collaboration with ANSYS. The workshops were attended by researchers and students from KAUST, Saudi universities and industry. In total, 583 people attended the workshops. Based on the feedback from the participants, the training themes evolved with time from basic CFD/CSM (Computational Fluid Dynamics / Computational Structural Mechanics) to multiphase flows, electromagnetism and multiphysics simulations. Hands-on sessions on workstations and Shaheen II are the key components of the workshops. In 2019, KSL started a certification program that became quite popular among Saudi students. Since the start of this certification program, 260 students have been certified in CFD and CSM. These engineering services of KSL have a high potential to cater to the upcoming nuclear human capacity building within the kingdom and will be extensively discussed in the full-length paper.

The global electricity demand, volatile fossil fuel prices, and the need to reduce greenhouse gas emissions push power producers towards clean energy sources, including nuclear energy. Civil nuclear power generation is seen as a reliable contestant in the clean energy race, with about 437 nuclear power plants currently operating worldwide. This paper discusses the Kingdom of Saudi Arabia's (KSA) efforts to meet its growing electricity demand and reduce its dependency on fossil fuels through the introduction of nuclear energy into its national energy mix. The paper describes KSA's progress in developing its peaceful nuclear power energy infrastructure, including the establishment of the King Abdullah City for Atomic and Renewable Energy (KA-CARE) and the Nuclear and Radiological Regulatory Commission (NRRC). To achieve its goals, Saudi Arabia has followed the IAEA "Milestones Approach" and conducted the first Integrated Nuclear Infrastructure Review (INIR) at KA-CARE headquarters in 2018. This showed that the country had made noteworthy progress in the development of its nuclear power energy infrastructure and is ready to invite bids or negotiate a contract for its first nuclear power plant(s). The paper emphasizes KSA's commitment to fulfilling international conventions and treaties related to nuclear power generation and its desire to achieve its Vision 2030 nuclear power generation targets.

The availability of refined and efficient energy resources has played a decisive role in the advancement of societies, especially since the industrial revolution of the eighteenth century. In the twenty-first century, the international energy scenario is experiencing a profound transition in terms of energy resources and their utilisation. The energy transition is in response to the challenges the global energy landscape faces such as rapidly growing demand, depleting fossil fuel reserves, surging energy prices, risks associated with the security of supplies, and above all climate change. Nuclear power is an important form of energy making a significant contrition to the electricity mix around the world, especially in developed countries. One of the major advantages of nuclear power is its minimal greenhouse emissions as compared to fossil fuels. The paper examines the key technological and policy dynamics of the unfolding energy transition. It also explores the prospects of nuclear power in the energy transition taking into account both the trend of phase-out the technology has experienced in some of the developed countries over the last couple of decades as well as the growing interest it has received more recently as a low carbon energy solution towards addressing climate change. It also examines nuclear power concerning the broader opportunities and challenges. The findings of this study are supported by the results of a survey carried out with the energy sector stakeholders from over 40 countries around the world.

Small Modular Reactor (SMR) technology is increasingly popular for its potential to provide clean, affordable, and dependable energy. However, implementing and adopting SMRs pose unique challenges and opportunities for nuclear safeguards. This paper provides an overview of these issues, including proliferation risk, safeguards implementation, resource constraints, novel technologies/designs, and adapting existing frameworks. SMRs' challenges include increased proliferation risk due to their concealability, complex designs, and the need to adapt existing safeguards frameworks to new reactor types. Resource constraints further exacerbate these challenges.

Nevertheless, SMR technology offers opportunities for enhancing nuclear safeguards through intrinsic features such as the use of proliferation-resistant fuel cycles and self-contained fuel designs, standardizing designs, and remote monitoring technologies. International cooperation is crucial in addressing the challenges and harnessing the opportunities presented by SMR technology by sharing best practices, technology, and information. Innovation in safeguards technologies, such as advanced sensors and data analytics, is necessary to address the unique challenges posed by SMRs. These technologies can overcome resource constraints, streamline safeguards implementation, and adapt existing frameworks to accommodate novel SMR designs.

In conclusion, while SMR technology presents several challenges for nuclear safeguards, it also offers numerous opportunities for enhancing proliferation resistance, streamlining implementation, and fostering international cooperation. By leveraging these opportunities and addressing the challenges, the global community can ensure the safe and responsible deployment of SMRs as a key component of the future energy mix.

Worldwide, the demand of hydrogen has grown by threefold since 1975. Hydrogen is mainly produced from fossil fuels, due to lower costs compared to other production alternatives. Consequently, this industry resulted in the emission of about 830 million tons of carbon dioxide per year. In its effort to tackle this global issue, Saudi Arabia is considering the utilization of nuclear energy to power the production of hydrogen as part of a push by the world's largest exporter of crude to diversify its economy away from hydrocarbons and fulfill its net-zero target. Hence, this paper presents a Techno-Economic model for hydrogen production using System-Integrated Modular Advanced Reactor (SMART) as a clean alternative to reduce dependence on fossil fuels and mitigate its carbon footprint in Saudi Arabia. SMART is an advanced pressurized water reactor that has been developed on the basis of proven technologies, with a 365 MW thermal capacity co-owned by the Kingdom. The reactor has integral steam generators and advanced safety features designed for electricity generation and thermal applications. The paper identifies the key factors that could affect the competitiveness of SMART for hydrogen production, by estimating the associated capital and operational expenses while considering the impact of various factors such as electricity prices, fuel costs, and carbon prices. The tool used for this analysis is “Hydrogen Economic Evaluation Program (HEEP)” developed by International Atomic Energy Agency (IAEA), which includes numerous options for hydrogen generation. The Discounted Cash Flow (DCF) approach is the foundation of the mathematical modeling used in HEEP to determine the levelized cost of hydrogen at a certain discount rate. The findings of this paper identify the key factors that could affect the competitiveness of small nuclear reactors for hydrogen production, providing valuable insights for future research and development in this area.

This paper discusses the technology and usage of SMR technology in the Kingdom of Saudi Arabia (KSA). Implementing such technology is helpful and can facilitate meeting the 2030 vision, which states net zero carbon emissions by 2060. SMRs with around 300 MWe contain advanced passive safety to eliminate any possible risk. With their small size compared to large reactors, SMRs can be multi-unit to increase the power and reduce the cost. Also, KSA is the largest producer of desalinated water and diesel generators used as a power source. Therefore, using SMR rather than diesel generators can reduce costs and CO2 emissions. In KSA, SMRs can be used as the primary source of electricity production for far-distanced areas, for example, or as an addition to existing plants. One unit of SMR can generate electricity for 0.93% of the population and produce 0.78% of the total energy produced in KSA. Furthermore, SMRs can integrate with renewable sources to compensate for the drawbacks since the vision states that renewables and natural gas will reach 50% of the energy mix in 2030. Saudi Arabia is taking action towards nuclear technology by constructing the uranium extracting facility and for SMRs by signing the contract for the SMART reactor with South Korea.

In the complex field of nuclear reactor design and analysis, there is a continuous need for sophisticated computational models that can accurately capture the diverse and challenging thermal hydraulic phenomena during steady state and transient conditions. This research sets the stage for the development of a comprehensive system analysis code for nuclear reactor thermal hydraulic design, starting with an isentropic two-fluid model with four governing equations. The computational methodology for this model incorporated the Advection Upstream Splitting Method (AUSM) scheme with a staggered grid arrangement. The nonlinear system of governing equations was solved implicitly by employing Newton’s method while a numerical Jacobian matrix was calculated for the derivative terms, enhancing the stability and efficiency of the solution process. The performance of the model was assessed through a code-to-code comparison with the Safety and Performance Analysis Code (SPACE), a licensed tool for design and safety analysis of Pressurized Water Reactors (PWRs). Further validation of the model was also performed using three classical two-phase benchmarks: water faucet problem, oscillating manometer problem, and air-water phase separation problem. The validation results indicate a reliable and accurate prediction of the model. Consequently, the successful development and validation of current two-fluid isentropic model provide a solid foundation for the future development of a comprehensive nuclear system analysis code based on the two-fluid model.

The quality of TRISO coated nuclear fuel particles is crucial to the successful operation and safety of Very-High-Temperature Nuclear Reactors (VHTR). For this reason, the TRISO particles should be free of defects and uniform in size and shape, as well as have a uniform coating. In this respect, the gas-solid spouted bed coating technology of TRISO particles is important. Coating layers around the fuel kernel are delicate processes impacted mainly by the hydrodynamics of the spouted beds. It becomes even more complex when considering that the success of the preceding coating affects the probability of coating the next layer successfully. As the current spouted bed coaters are relatively small, large-scale spouted beds are also essential to fabricate high-quality and large quantities of TRISO particles for the VHTRs. It is imperative to develop large-scale spouted beds in order to meet the growing demand for TRISO particles for the VHTRs. Thus, the scaling up of spouted beds is considered one of the major challenges in manufacturing TRISO nuclear fuel particles. Although the insights provided by the literature on spouted beds, scale-up of gas-solid spouted beds is still far from satisfactory. Correspondingly, having a fundamental understanding of hydrodynamics and scale-up of spouted beds dynamics is essential towards the development and commercialization of the VHTRs. In this work, a comprehensive overview of our newly developed and validated mechanistic scale-up methodology for gas-solids spouted beds based on matching the radial profile of gas holdup is presented. The overview aims at pointing out the present findings and challenges about the mechanistic scale-up methodology of gas-solids spouted beds. In addition, the overview includes a comparison between this mechanistic scale-up methodology and traditional scale-up methodology based on matching dimensionless groups to demonstrate its improved accuracy in predicting the performance of spouted bed systems.

Demonstrations of the capabilities of nuclear system analysis codes are required to obtain a license for their use in various applications of nuclear power plants. The Safety and Performance Analysis CodE (SPACE) has been developed and approved to be used for licensing applications of Pressurized Water Reactors (PWRs). However, since new innovative designs such as SMART100, a 100 MWe system-integrated modular advanced reactor, incorporate inherent and passive safety design features that are not used in conventional loop-type reactors, especial models should be developed and validated to reflect the characteristics of the SMART-100 and obtain reliable predictions. A thermal-hydraulic integral effect test facility, SMART-ITL, was constructed to examine the system performance of SMART-100 and to Investigate the thermal hydraulic phenomena occurring in the reactor systems and components during the normal, abnormal, and emergency conditions. The experimental data also serves to validate the related thermal-hydraulic models of the safety analysis codes. This study presents a validation of the SPACE code, using the SMART-ITL facility, to evaluate its applicability for analyzing thermal hydraulics in integral reactors. Simulations were performed for two experimental test scenarios: pressurizer safety valve break and safety injection line break concurrent with total loss of secondary heat removal (TLOSHR). The validation results indicate that the SPACE code accurately predicted key thermal hydraulic behaviors, such as primary and secondary system pressures and temperatures. However, a slight underestimation of the reactor pressure vessel's water level was observed, attributed mainly to the overestimation of the accumulated break flow due to inaccuracies in the two-phase critical flow models.

Abstract:
Nuclear power is identified as a reliable solution to generate electricity and desalinate water for the base load without intermittence and non-controlled variations. The recourse to nuclear power in Gulf countries started a few years ago. Few plants have already been constructed, in particular in UAE. One of the numerous aspects to be addressed of nuclear power performance is the condenser cooling process which requires large quantities of cooling water resulting in important environmental impacts and energy requirements. This work aims to evaluate the impact of seawater cooling temperature on the first and second law efficiencies of a typical nuclear power plant. Energy and exergy analyses will be developed to quantify the thermodynamic performance of the nuclear power plant and its components. The methodology consists of developing a mathematical model based on energy and exergy balances on each of the components and the entire plant using updated technical specifications and accurate fluid properties. The study includes three different Saudi locations with different seawater temperature profiles. The variation in the electric production, thermal efficiency, and exergy efficiency for the three locations will be particularly investigated.

Keywords: Nuclear power, Cooling water, Condenser, Saudi Arabia, thermal efficiency, exergy efficiency.

The Digital Twin (DT) is the most advanced paradigm to
create a virtual representation of a real world multi-physical system
for the sake of designing, monitoring, and supporting the decision
making throughout its whole lifecycle. Therefore, DT requires a
flexible ecosystem made of a variety of engineering tools and services
that need to be highly interoperable and enable knowledge
management and traceability. However, the major challenges that
slows down the emergence of DT environment are the inherent
incompatibility of engineering tools with the absence of a unified
standard for interoperability, and the human change aversion to new
technologies, tools and paradigms. For this purpose, we have
developed an evolutive environment that answers the main
requirements of Interoperability, Traceability, Flexibility, and
Knowledge management. In this paper, we introduce a general
overview of DT ecosystem with the limitation of some state-of-art
existing implementations with respect to four metrics. Then we will
introduce our solution applied to a nuclear case study to demonstrate a
highly evolutive implementation of a DT ecosystem.

Energy Transition acceleration coupled to geopolitical stakes has translated into a major worldwide attention for nuclear sector mainly thanks to its capacity to produce stable electricity in a souverain way. Hence, in order to keep up with the high demand of nuclear power plants, it is fundamental to have an efficient project management in order to successfully deliver nuclear projects while ensuring high regulatory requirements.
Nuclear sector faces a major challenge, which paradoxically lies in the fact that despite all the accumulated experience in product design, it struggles with managing the associated organizations, which is becoming as complex as designing the product itself.
The aim of this paper is to present a new methodology called PDM, which stands for Project Delivery Model. PDM provides a digital framework that harnesses the power of model-driven engineering capabilities to enhance and optimize project delivery. Unlike traditional project tools that are restricted to having interfaces with KPIs, PDM offers a digital framework that enables the dynamic association and modelling of various dimensions of a project with each other. This enables project managers to make well-informed decisions regarding their projects throughout the entire lifecycle, from planning to implementation. By leveraging digital technology and artificial intelligence, the methodology offers several benefits, including the ability to accelerate the initialization of projects, to conduct simulations on the project organization, on the supply chain optimization, and to perform impact analysis of a specific decision.

Fossil-based fuels have been powering our economies for years. They account for about 80% of the global energy mix. However, despite their dominance, fossil fuels are finite, and has negative impacts of the environment and climate-related changes. Saudi Arabia under Vision 2030 sees this issue and build their Vision to reduce the indecency on oil and achieve net zero emission by introducing renewable and alternative source of energy in their energy mix such as nuclear power. Saudi Arabia is following the IAEA milestones in the development of national infrastructure for nuclear power. Because nuclear technology requires knowledgeable and highly skilled personnel to ensure its safe deployment and sustainability, human resource development (HRD) is 1 of the 19 key infrastructure requirement that IAEA mentioned. It needs to be implemented during the design, construct and subsequent operationalization of nuclear power plants. Saudi Arabia is constructing the Low Power Research Reactor (LPRR) as a tool to transfer the nuclear technology and train future operators for nuclear power plants. This study will focus on the potential role of research reactors that can play in building the nuclear human capacity for nuclear energy generation. Findings show that research reactors can help nuclear human resource capacity, especially with regard to education and training that can be used to not only develop but also maintain the human resources necessary for supporting the safe and sustainable operation of nuclear power programs.

Probabilistic safety assessment (PSA) is a powerful tool that can be used to identify and assess the risks associated with nuclear power plants (NPPs). Nevertheless, PSA is a complex and data-intensive process, and it requires a significant level of human expertise to be conducted reliably. As a result, there is a need to build a human capacity for PSA in order to ensure that it is used effectively in licensing NPPs.
There are a number of steps that can be taken to build a human capacity for PSA. First, it is important to develop a strong understanding of the principles and methods of PSA. This can be done through training programs, workshops, and other educational initiatives. Second, it is important to develop a database of relevant data on the reliability of NPP components and systems. This data can be collected from a variety of sources, including plant operating experience, industry databases, and research studies. Finally, it is important to develop a team of experienced PSA practitioners who can conduct and interpret PSA studies. By taking these steps, it is possible to build a human capacity for PSA that will enable NPPs to be licensed in a safe and informed manner. There are also a number of other factors that can contribute to the success of PSA in the licensing of NPPs.
There is also a need to form a robust R&D group for PSA that includes experts from a variety of supporting fields, such as: nuclear, mechanical, industrial, reliability and human factors engineering, computer science, statistics and probability, mathematics and computer sciences. This paper will discuss in detail the human capacity requirements to excel in PSA that is essential in licensing NPPs, and outlines the supporting fields to PSA in order to form a robust R&D group for performing PSA.

Evaluation of Nuclear Power Plants (NPPs) performances during accident conditions has been the main issue of research in nuclear fields during the last 40 years. Therefore, several complex system thermal-hydraulic codes have been developed for simulating the transient behavior of NPPs. Safety and performance analysis codes validation is required and important work that should be performed to obtain reliable results for simulating the NPPs behaviors during the steady state or transients.

SMART100 is System Integrated Modular Advanced Reactor with 100 MWe and fully Passive Safety Systems (PSSs). The design of SMART100 was upgraded from the standard design of SMART and developed by Korean Atomic Energy Institute (KAERI). Unlike loop-type commercial reactors, the SMART100 plan adopts a helically coiled steam generator, an internal pressurizer, inside the Reactor Pressure Vessel (RPV).In addition, SMART-ITL is an integral test loop facility that has been constructed by KAERI and finished its commissioning tests in 2012, to observe and understand the thermal hydraulic phenomena that occur in the systems of SMART100 during normal operation or transients. To simulate the thermal-hydraulic behavior well at SMART100 plant under various conditions including Steam Generator Rapture (SGTR), it is necessary to develop and validate safety and performance system analysis codes that reflect the characteristic of SMART100. In general, developing physical models and validation work for separate effect and integral effect tests are required to enhance the reliability of the simulation results of a system analysis code.

The purpose of this paper is to validate the Safety and Performance Analysis CodE (SPACE) based on the steam generator tube rapture experiment with SMART-ITL in order to predict and identify the capability of SPACE for analyzing thermal hydraulics in integral reactors

This is a preliminary conceptual design configurations study for SMART Plus reactor vessel (RV) and steam generators (SGVs) for the purpose of enhancing its economic efficiency and safety by introducing innovative element technologies. To enhance the competitiveness of SMART, several element technologies have been suggested such as printed circuit steam generator (PCSG), in-vessel control element drive mechanism (IV-CEDM), an improved RV module, and so on. The most important and promising one of those is to utilize a printed circuit heat exchanger (PCHE) as a steam generator, and to develop an improved RV module adopting the PCSG, which can enhance the economic efficiency of SMART Plus. Several types of conceptual design candidates of SMART Plus were investigated thoroughly in the present study to find out the most effective reactor configuration for economic enhancement without drastic degradation of safety. There are five conceptual designs for arrangement of an RV and SGVs, i.e., integral, 1-SGV modular, 2-SGV modular, 3-SGV modular, and 4-SGV modular arrangement.

The Phenomena Identification and Ranking Table (PIRT) of Thermal –Hydraulic (T-H) phenomena is used to identify the key phenomena associated with the intended application, then rank the relative importance and current state of knowledge for each identified phenomenon by the experts in the related field. This ranking provides guidance for code development and improvement for the specific simulation of the plant behaviors.

The Safety and Performance Analysis CodE for nuclear power plants (SPACE) has been developed for the safety analysis of operating PWRs and the design of advanced water reactors. The SPACE adopts advanced physical modeling of two-phase flows, mainly two-fluid three-field models that consist of gas, continuous liquid, and droplet fields. Based on that the Nuclear Safety and Security Commission (NSSC) approved the use of the SPACE for licensing applications of Korean PWRs in 2017. In addition, the SPACE has been improved continuously to extend its application for the Design Extension Conditions (DECs).

SMART100 is System Integrated Modular Advanced Reactor with 100 MWe and fully Passive Safety Systems (PSSs). The design of SMART100 was upgraded from the standard design of SMART and developed by Korean Atomic Energy Institute (KAERI). Unlike loop-type commercial reactors, the SMART100 plan adopts a helically coiled steam generator, and internal pressurizer inside the Reactor Pressure Vessel (RPV).

The main objectives of this paper are to develop and generate PIRT of important T-H phenomena for expected DECs of SMART100, and to implement T-H models and validation items in SPACE for the reference reactor and scenarios.

“Thermal Hydraulic Performance Analysis of a Printed Circuit Steam Generator (PCSG) in Innovative Smart-Plus Reactor: A Proof of Concept Study”

The Printed Circuit Steam generator (PCSG) is a kind of Printed Circuit Heat Exchanger (PCHE) designed for steam generator application. The PCSG manufacturing process is similar to the PCHE. The PCSG was suggested to be used in the innovative SMART-Plus reactor. Two PCSG designs were considered, one incorporating monitoring channel plates and the other without. The design with monitoring channels was introduced due to the inherent challenges in conducting in-service inspections with the design lacking these channels. To validate the PCSG concept and verify its thermal hydraulic performance, several experimental tests were conducted using two mockups in the SMART Integral Test Loop (ITL) facility. The tests simulated the operating conditions of SMART Plus reactor and its thermal hydraulic parameters such as temperature and pressure. The test facility comprised a primary circulation loop representing the hot side, and a secondary circulation loop representing the cold side. The two mockups were designed and manufactured to mirror the two PCSG concepts: one with monitoring channel plates and the other without. The results demonstrated the viability of the PCSG design as a superheated steam generator. Hence, it was concluded that introducing the PCSG to the SMART-Plus Design is a viable option.

In the event of a severe nuclear accident resulting in a core meltdown, the molten corium which is a mixture of the molten fuel, cladding, and structural elements, originating in the reactor core could penetrate through the reactor pressure vessel to interact with concrete structure underneath. This research paper presents the numerical modeling of molten corium concrete interaction. The complex phenomena of molten corium concrete interaction and melting of concrete are simplified by considering multi-region with the change in phases taking place only within each predefined region with the assumption that corium is a homogenous mixture of molten nuclear fuel, cladding, thermo-hydraulic and structural element. This study presents the use of the OpenFoam a Computational Fluid Dynamics (CFD) simulator where a new solver is developed to model the molten corium concrete interaction, its melting, solidification, and concrete ablation for the first time. Two sets of experimental data are used to validate the developed solver and demonstrate the thermal modeling and heat transfer capabilities of the developed solver for concrete ablation under severe conditions. We analyzed different boundary conditions and found that they had a pronounced effect on mitigating ablation and reactor integrity in case of a nuclear accident. In addition, the water-cooled boundary condition was found to be the controlling boundary condition to mitigate concrete ablation. The concrete ablation mechanisms during MCCI are very case-dependent on the concrete solidus, liquidus, and ablation temperatures.

The goal of the IAEA Coordinated Research Project “Benchmark of Transition from Forced to Natural Circulation Experiment with Heavy Liquid Metal Loop” (CRP - I31038) is to develop Member State advanced fast reactor analytical capabilities for simulation and design using system, CFD, and subchannel analysis codes. Here we present CFD validation employing the commercial CFD code Star CCM+ applied to the fuel pin simulator for forced and natural convection cases in the open phase where experimental data is provided in the benchmark specification provided by ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development) for the NACIE-Up facility (NAtural CIrculation Experiment-UPgrade). Considered is the fuel pin simulator with 19 pins, each consisting of a preheated lower section and heated upper sections, respectively. Three configurations (i) all pins heated, (ii) inner 7 pins heated and (iii) asymmetric heating are studied. For each heating configuration data for forced and natural convection are provided. Here case (i) is studied. Temperatures at three planes are measured near the inlet, in the middle and near the end of the heated section, respectively. In addition, the axial temperature along the wall of one fuel pin simulator (in second row) is measured so that in total 67 thermocouples measure fluid and wall temperatures for validation purposes.
Our validation confirms that the thermos hydraulics inside the fuel pin simulator can be simulated with a good accuracy. Applied is a polyhedral mesh with 2 prism layers, the k-omega SST model with all all-wall treatment and order unity y+ values. Moreover, we performed a grid-sensitivity study and analysed the importance of conjugate heat transfer inside the fuel-pin simulators and the wrapper. Our studies indicate that it is possible to implement further simplifications without corrupting the accuracy of the simulation to reduce computational effort.

Supercritical Water-Cooled Reactors are among the selected proposed designs for the upcoming GEN IV of nuclear power plants. They represent the natural evolution of presently available LWRs: borrowing the experiences and competences developed during the last decades for both PWRs and BWRs, it indeed aims at increasing the efficiency of the thermodynamic cycle while reducing the construction capital costs. During the last decades the European Union and IAEA launched projects in support of the development of such a technology: the University of Pisa joined the common efforts providing numerical analyses addressing the thermal-hydraulics aspects of SCWRs.
Among the most challenging issues related to the development of SCWRs, the prediction of heat transfer to supercritical fluids probably represents one of the toughest ones. In fact, the interesting phenomenon of heat transfer deterioration can hardly be predicted by both CFD approaches and heat transfer correlations. Being the consequence of dramatic thermophysical properties changes, buoyancy phenomena or flow acceleration, generally available approaches cannot provide suitable predictions of this complex phenomenon: the development of advanced modelling techniques is thus required.
At the University of Pisa, during the last years, a modified k-ε model adopting the algebraic heat flux model (AHFM) for the prediction of the turbulent heat fluxes and buoyancy production terms was implemented in the commercial code STAR-CCM+ reporting promising capabilities in the prediction of several supercritical fluids operating conditions. A fluid-to-fluid similarity theory for heat transfer to supercritical fluids was proposed as well achieving a preliminary validation on the basis of DNS, LES and RANS calculations. Limitations and capabilities of available heat transfer correlations were investigated as well.
The present paper reports on the recent numerical activities performed at the University of Pisa also providing information about the foreseen next steps and possible developments in modelling.

The nuclear fuel performance during accidents became a critical issue after the Fukushima Daiichi nuclear accident in 2011. Currently, various research and development programs are being carried out to enhance the fuel's reliability and durability under such conditions. These programs are collectively known as the Accident Tolerant Fuel (ATF) R&D program, which involves multiple countries, research institutes, and fuel vendors. ATF is an enhanced fuel that can tolerate longer periods of active cooling system failure, without significant fuel/cladding system degradation. Moreover, it can improve fuel performance in normal operations, transients, as well as design-basis accident (DBA) and beyond design-basis (BDBA) scenarios. This paper presents a preliminary neutronics analysis for Accident Tolerant Fuel (ATF) cladding materials for a standard PWR fuel rods. The candidate cladding materials were compared with the original Zircaloy-4 cladding material. To confirm the necessary geometry requirements for achieving end-of-cycle fuel reactivity, a parametric evaluation was conducted on fuel and cladding materials. The findings were then compared with the standard PWR reference fuel-cladding system. Several reactor safety parameters such as reactivity, radial power distribution of fuel pellet, reactivity coefficients, and spectral hardening are assessed for the candidate cladding materials. The neutronic and depletion calculations of ATF cladding materials was performed in this study by using the OpenMC code.

Accurate and reliable computer codes for neutronics are essential for nuclear reactor design and operation since they permit the simulation of neutron behavior inside the reactor core. These simulations are crucial for anticipating the reactor's performance, safety, and efficiency. In order to ensure the safe operation of nuclear reactors and the advancement of new reactor technologies, the accuracy and reliability of these codes are of the utmost importance. The modeling and simulation codes' accuracy and reliability are tested using the Virtual Environment for Reactor Analysis VERA Core Physics Benchmark's problems. MASTER's neutronics model solves the space-time dependent neutron diffusion equations with the advanced nodal methods and DeCART2D a Deterministic Core Analysis based on Ray Tracing, have been developed in Korea Atomic Energy Research Institute (KAERI) to design and analyze the Pressurized Water Reactor (PWR). This study aims to validate the DeCART2D - MASTER code system for the successful completion of the calculations related to the Zero Power Physics Tests (ZPPT) of VERA pressurized water reactor that are carried out at the start of each fuel cycle startup. Several crucial configuration estimates, RCCA bank reactivity worths, the isothermal temperature reactivity coefficient (ITC), and differential soluble boron worth (DBW) are among them. The calculation results are compard to ZPPT OF VERA banchmark.

During daily load-follow operation, the xenon concentration in the reactor changes over time due to reactor power variation. To compensate for this change, the soluble boron concentration in the core is adjusted through dilution or boration. However, since xenon is not directly measurable in the reactor, the sliding mode observer is utilized to estimate the xenon concentration prior to the daily load-follow operation. This estimation provides the operator with valuable information to determine the appropriate boron concentration adjustment scenario. The robustness of the sliding mode observer relies on accurate estimation of the observer gains used in the model, ensuring the sliding surface is reached within a finite time.
Mode-K+ was employed to perform the load-follow operation in the APR1400 reactor. Soluble boron scenarios were chosen with linear variation over time, and no dilution or boration occurred before the completion of power ramp-down and ramp-up to prevent any reactivity divergence in the core. The results indicate good agreement between the observed xenon concentration and the actual xenon concentration variation in the core, calculated using time-dependent xenon concentration simulation. This analysis was conducted using a two-step procedure: cross-section evaluation using the SERPENT continuous energy Monte Carlo code, and whole core calculations performed using an in-house diffusion code called KANT.

Severe accident management (SAM) measures play a significant role in mitigating severe accidents in
nuclear power plants (NPPs). Their proper design requires a thorough understanding of processes
leading to severe accidents and the related phenomena. As SAM measures are being increasingly
considered in regulation of NPPs operation, it is of utmost importance that suitable database is
available for validation of safety analyses codes to facilitate a better understanding of mitigations
systems performance and their impact on progression of an accident.
In support of such activities, THAI research program aims at investigating open questions relevant for
safety assessment of water-cooled reactors, with specific focus on containment safety research under
severe accident conditions. Experiments are conducted in the frame of national project sponsored by
the German Federal Ministry for the Environment, Nature Conservation, Nuclear Safety and
Consumer Protection (BMUV), and international projects which run under the auspices of OECD
Nuclear Energy Agency (NEA). The ongoing THAI national project involves experimental research
related to water-cooled small modular reactor relevant topics, and OECD/NEA THEMIS project has a
specific focus on investigating effect of late phase conditions on hydrogen risk and source term related
issues, among others.
The present paper will discuss a selection of THAI experiments related to active and passive safety
systems performance, impact of late phase conditions (e.g., CO, O2 lean atmosphere), and fission
products remobilization from surfaces and/or water pools. The experimental database facilitating
assessment and validation of safety analyses tools towards mitigation, and management of severe
accidents will be highlighted with selected examples of international code benchmarks conducted
based on THAI data. Finally, planned re-orientation of THAI experimental research in the frame of
future national and OECD/NEA THEMIS follow-up projects will be briefly discussed with emphasis
on passive safety systems related investigations including those relevant for water cooled SMRs.

The Chernobyl and Fukushima nuclear accidents have significantly impacted the public's perception of nuclear energy and its potential benefits. To fully harness nuclear technology's potential for contributing to the Sustainable Development Goals (SDGs), as highlighted by the IAEA bulletin of September 2016, we must restore public confidence in this energy source. Emergency preparedness and response is a cardinal principle of radiation protection, and the literature review indicates that most research and energy reactors worldwide have modelled hypothetical accident release scenarios. However, most of these models only considered respirable gaseous radionuclides released from the reactor core and failed to consider other possible exposure media, routes, and scenarios. In this paper, we propose a framework for using the RESRAD family of codes- ONSITE, OFFSITE, RDD, and BIOTA- to improve emergency preparedness and response planning, especially in Africa, where site-specific data is lacking. By critically analysing related literature through a desk review, we demonstrate the application of these codes in biodiversity conservation, protecting people and the environment, and safeguarding future generations. Our proposed frameworks, if implemented, will build public confidence in nuclear energy projects in Africa, restoring public confidence and consequently help in solving the lingering energy crisis on the continent.

Nuclear reactors and associated structural facilities are designed to withstand various types of loadings including thermal, vibrational and fatigue etc. Despite meticulous design incorporations, these structures can be damaged during operation due to unforeseen scenarios and can cause imminent danger in retrospect of nuclear radiation leakages, loss of critical infrastructure and life. To avoid catastrophic outcomes, continuous structural health monitoring (SHM) of such structures ensure safety and efficient maintenance protocols. Finite Element Analysis (FEA) provides a valuable insight into structural performance during the design phase and can be validated through experimental twin. Optimal sensor placement to provide critical insight with reduced sensors can be cost effective. A successive full field estimation using a reduced FE model based on the OSP locations can be useful to visualize comprehensive structural behavior and assist in addressing the issues in sensor-less regions. A D-optimal (determinant optimization) method is shown to be used in tandem with System Equivalent Reduction Expansion Process (SEREP) for sensor placement and full field estimation of required measurand, respectively. Three test cases for simple structural elements such as beam and plate with various boundary conditions are shown to produce full field strain fields for the structures, when applied with an impact load to be later determined for its magnitude and location. The results produce fairly accurate reconstruction of the structural dynamic response when observed at different time instances/windows along with identification of impact magnitude and position. This concept is proposed to be extended to structures such a reactors, boilers, and piping in nuclear facilities to provide a detailed and comprehensive monitoring of these structures and to avoid damages and failure due to unknown cause(s) while in service and to identify their location and magnitude.

Due to their high specific heat, low viscosity, and good diffusivity, supercritical fluids have the potential to be ideal coolants. In addition, the supercritical heat transfer properties are shown to result in up to 45% efficiency in nuclear power generation, which operates nearly at 550°C. However, understanding the heat transfer for fluids under supercritical conditions has been a challenge. In this regard, a wide range of experiments with different parameters has been performed to understand the peculiar heat transfer characteristics. The generated experimental database could serve as a reference to assess the prediction capabilities of the Reynolds-Averaged Navier-Stokes (RANS) based computational fluid dynamics (CFD) approach under supercritical conditions. RANS is the most widely used modeling approach and has been an industrial workhorse for decades. This work aims to study the heat transfer characteristics of supercritical Carbon Dioxide (sCO2) in horizontal tubes using different RANS models. Various flow conditions are modeled to study the impact on the heat transfer coefficient. A large-temperature variation is also expected in some conditions due to stratification. In such cases, the impact of buoyancy on the heat transfer coefficient is also explored. The prediction capabilities of selected RANS models will be assessed against the experimental reference data and will be presented in a full-length article.

As demonstrated in the Fukushima catastrophe, a malfunction of a nuclear power plant may result in the release of elevated quantities of noble gases, I-131, and Cs-137. Consequently, the key components of decision-support mechanisms for emergency readiness and response to hazardous nuclear incidents entail the evaluation of the potential danger to the populace through the modeling of the dissemination of radionuclides into the surroundings. Therefore, a meticulous appraisal of the environmental repercussions of the discharge is imperative and must be determined with confidence, particularly in the vicinity of the source of the release and in the vicinity of the plant edifices, where direct consequences may impact the personnel involved in accessing the facility during exigent circumstances.
Several computational software tools for atmospheric dispersion, based on either Gaussian plume or regional Lagrangian models, have been developed to evaluate accidental scenarios involving the release of radionuclides. However, these models' predictive accuracy can be limited, especially in the near-field region. This is because the associated physics of pollutants' dispersion may not be adequately captured, as these models either partially account for the nuclear plant's buildings (e.g., Gaussian plume with building effects accounted for) or ignore them altogether (e.g., standard Gaussian plume model and regional models). Additionally, it is crucial to consider key UAE environment characteristics, including the arid ground surface topology and the atmospheric boundary layer stability regime. To address these shortcomings, the present study aims to use computational fluid dynamics (CFD) modeling methodology to accurately account for such conditions and assess their influence on the discharge and dispersion of radionuclides from the nuclear power plant to the environment. Hence, this study focuses on (i) atmospheric micro-scale (200 m to 5 km) meteorological phenomena, (ii) ground and built-in environment topologies, and (iii) undertaking parametric assessments to validate the CFD predictions.

The detection of corrosion and cracks in nuclear power plants is a critical task that requires accurate and efficient monitoring systems. Traditional inspection methods can be time-consuming and may not be able to detect defects in hard-to-reach areas. In recent years, machine learning and deep learning techniques have emerged as promising alternatives for the detection of corrosion and cracks in nuclear power plants.
This paper will compare the latest research on machine learning and deep learning techniques for corrosion and crack detection in nuclear power plants. It includes an overview of the different machine learning and deep learning algorithms that have been applied in this field. This article also investigates the effect of different input features and transfer learning techniques on the accuracy of corrosion and crack detection models. Additionally, a systematic review of publicly available datasets for corrosion and crack detection in nuclear power plants will be presented.

Radiation detection systems are essential for upholding nuclear security and safety. The main objective of this study is to design and develop a new radiation detection system that can locate the position of a radioactive source using Geiger-Muller (GM) detectors. Three GM detectors are utilized in this system, and their relative angles are known. The detectors collect data on radiation exposure levels in the immediate area surrounding the detectors and send it to a Raspberry Pi 4 processor. After that, a MATLAB algorithm analyzes the data using Curve Similarity techniques to determine the source's position. To provide a user-friendly experience when using the system, a graphical user interface (GUI) was designed. The benefit of the Curve Similarity technique is that it offers great accuracy, rapid response, and low cost. It can be applied in fields like environmental monitoring, homeland security, and nuclear power plant safety, including locating lost radioactive materials and other radiation security and safety concerns.

Gate simulation modeling was used to simulate the system to test the Curve Similarity method in locating the source. A Cs-137 point source was used to study the system response. The results show that the system can detect a point source accurately within a few centimeters. Following the simulation process, an experimental setup is prepared to test the radiation detection system in a laboratory environment using the methods that were developed.

Characterizing and modeling multi-phase flow is a complicated scientific and technical phenomenon represented by a variety of interrelated elements. Yet, the introduction of dimensionless numbers used to grasp gas-liquid flow is a significant step in controlling and improving the multi-phase flow area. Such as the SL (Slippage number), a dimensionless number defined as the ratio of the difference in gravitational forces between slip and no-slip conditions to the inertial force of the gas. The fact that plotting SL versus Frm provides a single acceptable curve for all of the data provided proves that SL may be used to realize the behavior of gas-liquid flow. This paper creates a numerical link between SL and Froud mixing number using vertical gas-liquid flow, and then utilizes that relationship to validate its reliability in practice. An improved correlation in drift flux model generated from the experimental data, and its rationality has been verified. The method in this paper is to approach for predicting the void fraction in bubbly flow, in through SL/Frm relation and the limitations of this method, as well as areas for development, are stated.

Severe accidents (SA) mitigation strategy in Nordic Boiling Water Reactors (BWRs) may lead to ex-vessel steam explosion (SE) scenario. The molten corium falls from the reactor vessel lower head that failed due to thermal and mechanical loads into a pool of water to form a coolable debris bed after fragmenting and quenching the corium melt. Ultimately, preventing containment failure and the release of radioactivity into the environment.

During corium fragmentation in water, a vapor film is formed around the melt which prevents the direct melt water contact and limits the heat transfer between the two liquids. In case of vapor film collapse, in a phase called (triggering) an explosive transfer of energy from the melt to the volatile coolant may occur as a shock wave that traverses along the water body causing steam explosion that affects the containment integrity.

The numerical instability of old SE codes causes a large spread in predictions of SE loads. In this work, we develop a code that utilizes improved numerical methods to assess steam explosions and their uncertainty.

In this paper, we address the triggering and propagation of a shock wave generated in a SE scenario. We build a numerically stable code using WENO solver with AUSM+-up and Godunov flux schemes to model pressure propagation in a multiphase domain.

We compare the results of shock wave propagation obtained with the experimental results from KROTOS facility and with the results from TEXAS-V code. We discuss the results and how they contribute to the enhancement of triggering and propagation modeling in a SE code.

The integrity of the reactor pressure vessel (RPV) is of the utmost importance in nuclear reactor safety. One of the cases that need to be studied more thoroughly is the formation of a corium pool which could happen during severe accident scenarios. The melted core is a mixture of multiple melted materials, giving the fluid a unique flow structure. Additionally, the presence of high-temperature variations within the corium pool causes an extremely high Rayleigh number for a natural convection flow regime. This unique problem requires significant research efforts. This paper studies a semi-oval vessel with internal heat generation representing high-temperature melted core and corresponds to the famous BALI experiments. In this context, a wide range of RANS (Reynolds-Averaged Navier-Stokes) based CFD (Computational Fluid Dynamics) simulations are performed to better understand the complex thermal-hydraulics phenomena in a corium pool. Additionally, a comparative study is performed to assess the prediction capabilities of different RANS models.

Turbulent heat transfer is an extremely complex phenomenon and is critical in scientific and industrial applications. It becomes much more challenging in a buoyancy-influenced flow regime, particularly for non-unity Prandtl number (Pr) fluids. In this article, an effort has been put forward to assess the prediction capabilities of different Reynolds-Averaged Navier-Stokes (RANS) based turbulence models for a mixed convection flow regime. In this regard, a fixed Richardson number (Ri = 0.5) case is considered at three different Prandtl number fluids (Pr= 1, 0.1, and 0.01). The considered flow configuration is a parallel plate arrangement with differentially heated top and bottom walls. Two different classes of turbulent heat flux models, i.e., based on Simple Gradient Diffusion Hypothesis and Algebraic formulations, are compared with the available reference DNS (Direct Numerical Simulation) database. The prediction capabilities for these modeling approaches are assessed and will be extensively discussed in this full-length paper.

It is an unfortunate fact that no single turbulence model is universally accepted as being superior for all classes of problems. The choice of turbulence model depends on considerations such as the physics of the flow, the established practice for a specific class of problem, the level of accuracy required, the available computational resources, and the amount of time available for the simulation. To make the most appropriate choice of model for certain application, one needs to understand the capabilities and limitations of the various options. Therefore, the validation study presented in this paper aimed to assess the capabilities of different turbulence models for the prediction of turbulent flow and heat transfer in a tightly spaced bare rod bundle.
In fact, a comprehensive CFD approach toward the accurate prediction of the turbulent flow and heat transfer in a tightly spaced rod bundles was developed. Since the experimental database was not available, the numerical experiment was performed in order to generate the high fidelity reference database by means of Direct Numerical Simulations (DNS). In the first step numerical experiment was designed, later DNS was performed. Finally, the validation of lower-order turbulent models was performed. For the validation purposes six commonly used turbulent models implemented in ANSYS Fluent software were chosen. In the validation study the turbulent flow and heat transfer profiles were compared qualitatively and quantitatively against the obtained DNS results.

The heat produced in the nuclear fuel rod is dissipated by the coolant running through the channels in the fuel assembly. The flow between fuel assembly rods shows oscillating behavior, having a noticeable effect on the cooling process. Additionally, the flow effects extend to the fuel assembly causing vibration in its structural system. The design and reliable operation of nuclear systems depend heavily on a comprehensive understanding of flow and temperature in a fuel assembly. In aiming to enhance the nuclear reactor's efficiency, safety and stability, a thorough understanding of fuel assembly coolant is crucial. Therefore, this study analyzes the flow between bare rod bundle fuel assembly configuration utilizing advanced computational fluid dynamics (CFD) approaches. In this regard, a hybrid (LES/RANS) turbulence modelling approach has been adopted to study a square lattice bare rod bundle configuration. By minimizing the overall computational cost, the best aspects of Large Eddy Simulation (LES) and Reynolds-Averaged Navier-Stokes (RANS) are employed. The obtained results are thoroughly compared with the available reference Direct Numerical Simulation (DNS) database of a closely-spaced bar rod bundle based on the well-known Hooper experiment. The hybrid methodology is evaluated through a qualitative comparison of the velocity field with the DNS database. Additionally, the prediction of the flow pulsation is analyzed numerically. The findings in this work justify the usage of hybrid (RANS/LES) for these types of complex flow configurations and show its reliability.

We report on the results for two buoyancy-driven benchmark cases, heat-driven square cavity at Ra = 1011, and Rayleigh-Bénard convection at Ra = 109, by using the elliptic-blending eddy viscosity k-ε-ζ-f model (or just ζ-f), and three different formulations of the turbulent heat flux, namely the Simple Gradient Diffusion Hypothesis (SGDH), the General Gradient Diffusion Hypothesis (GGDH) and the Algebraic Flux Model (AFM). The ζ-f model is well-posed for computing turbulent heat transfer since it contains an approximation of the normal Reynolds stress in the wall-normal direction that is needed in GGDH and AFM formulations. Furthermore, the modeling of the wall-blocking effect by using the elliptic-relaxation approach is physically more sound than the commonly used damping functions. This work is motivated by the recently held 17th ERCOFTAC SIG15/MONACO2025 workshop on turbulent natural convection flows in differentially heated cavities, Manceau (2023), which demonstrated superior performance of the ζ-f model in predicting the main flow features for the selected cases.

The Reynolds-Averaged Navier-Stokes (RANS) equations, crucial in predicting
turbulent flows through computational fluid dynamics (CFD), involve the decomposition
of flow variables into time-averaged and fluctuating components using Reynolds
decomposition applied to the Navier-Stokes equations. Predicting turbulent stresses
accurately necessitates turbulence models due to flow complexity, which are mathematical
or empirically based. Commonly used models include k − and k − , focused on
turbulent kinetic energy and dissipation rate, and turbulent kinetic energy and specific
dissipation rate, respectively. These models offer strengths and weaknesses, and their
selection is dependent on simulation specifics and result accuracy. The first review section
delves into model variations, discussing closure term functions and constants. Machine
learning (ML) enhances turbulent models by enabling data-driven closures and rapid
precise predictions. The second part explores ML's role in enhancing turbulent models—
predicting quantities, optimizing models, and creating efficient reduced-order models. This
ML integration in modelling holds the potential for improved accuracy, efficiency, and cost
reduction. Challenges and prospects in this field are also addressed in this review.

The discrete-ordinates-method (DOM) is one of the primary numerical techniques for the solution of the Boltzmann Transport Equation (BTE) as well as other transport equations that have been derived from the BTE. These transport equations are used for the study of various phenomenon such as the radiation heat transfer in a participating media, heat conduction via phonon transport at the micro/nanometre scales, neutron transport theory, etc. DOM has traditionally been used in rectangular coordinates, however, relatively recently, the DOM technique has been fully extended to be applicable in structured, general, orthogonal as well as non-orthogonal coordinate systems. Orthogonal and non-orthogonal coordinate systems can be successfully used to cover complex, irregular geometrical domains with a structured mesh and hence an efficient solution procedure based on the DOM can be developed.

This is a preliminary investigation of safety characteristics of the Stable Salt Reactor (SSR), which is a fast-spectrum molten salt reactor proposed by Moltex Energy. The reference Moltex fuel composition and other publicly released information have been utilized in this work. Regarding the liquid salt fuel, two more TRU (transuranics) compositions are also considered for improved core performances. Since the SSR concept adopts an on-power refueling scheme, a pseudo-equilibrium state was envisioned based on the linear reactivity model. It was found that reactivity change induced by withdrawal or insertion of fuel assembly could exceed 2 $, which insinuates possible safety or operational issues pertaining to the on-power refueling scheme that has not yet been thoroughly investigated. This work finds that the coolant temperature coefficient (CTC) is clearly positive ~ 2 pcm/K, the Doppler effect of the fuel salt is ~ -0.5 pcm/K and its temperature-dependent expansion coefficient is about -10 pcm/K, leading to fuel temperature coefficient (FTC) of the SSR to be around -10.5 pcm/K. Through point kinetic analyses coupled with a simplified lumped heat balance model, the transient behavior of the SSR system was estimated based on the evaluated temperature coefficients of the reactivity. Although the change in the reactivity is expected to be significant, the feasibility of on-power refueling strategy for the SSR has been found. All of the neutronic calculations were performed using the SERPENT 2 Monte Carlo code with nuclear data library ENDF/B-VII.1.

The fundamental energy transformation that occurs in a nuclear power plant involves converting the kinetic energy of fission process fragments into thermal energy. The reactor coolant is maintained at temperatures typically between 300-400°C, which limits the efficiency of energy conversion from thermal to mechanical and then to electrical to around 30-35%. The average energy of neutrons produced during fission is approximately 2.0 MeV or 1.55×10^10 K. However, these high-energy neutrons eventually slow down, resulting in a loss of their energy quality, until they reach a much lower energy level of 50 meV or 600K.

It is possible to convert the energy of neutrons during the moderation process into electrical energy by incorporating carbon nuclei into a p-i-n diamond Schottky diode. This technology works similarly to betavoltaic devices that produce electrical energy by collecting electron-hole pairs generated from the energy of beta particles released during radioactive decay. However, the high radiation levels present in a nuclear reactor can overcome the inefficiencies associated with the thin layers’ geometry of betavoltaic devices. Moreover, diamond material exhibits remarkable radiation, thermal, and chemical tolerance, making it an ideal candidate for use within intense radiation fields of the nuclear reactor.

A Schottky diode is proposed to be formed using diamond as the interaction material and graphite as the conductor. Beryllium and oxygen, which have low neutron absorption cross sections, will be used as doping elements. The device will serve both as a direct energy conversion device and a neutron moderator. The concept is to be explored in terms of known nuclear radiation flux characteristics and performance of diamond as a radiation detector material in different radiation sources contexts.

The SMART (System-integrated Modular Advanced ReacTor) reactor is a small, pressurized water reactor that utilizes integral pressurized water coolant, which offers many advantages over traditional designs. In this paper, a neutronic analysis of the SMART modular reactor fuel using the DeCART2D computer code is performed, considering annular and mixed oxide (MOX) fuel types. The study modeled and analyzed the behavior of the SMART reactor fuel under different operating conditions, including the use of annular fuel design and MOX fuel. DeCART2D code is a two-dimensional neutron transport code that uses the method of characteristics to solve the neutron transport equation, to simulate the neutronic behavior of these fuel types. The study explores the use of annular fuel design for the SMART reactor in several areas, which are thermal efficiency, peak fuel temperature, fuel burnup, neutron flux distribution, reactivity, and power distribution.
Overall, the neutronic analysis of the SMART modular reactor fuel using the DeCART2D computer code provides valuable insights into the behavior of the reactor fuel and can inform the design and operation of the SMART reactor. The findings can also contribute to the development of advanced fuel designs for small modular reactors, with potential applications in both existing and future nuclear power plants.

Source term evaluation constitutes an important feature and element for Severe Accident Management (SAM) strategy. For assessment of SAM efficiency it is crucial to identify phenomena and parameters that present major contributions to the uncertainty in the magnitude and timing of the releases, assess their sensitivity, and quantify the uncertainty. In this work source term evaluation and uncertainty quantification were performed using MELCOR for two accident scenarios, large break LOCA and station blackout, that leads to containment failure due to ex-vessel phenomena at RPV melt-through. Preliminary screening was performed using best-estimate and bounding assessment, where parameters were varied one-at-a-time. Dakota was used to perform Morris sensitivity analysis, followed by uncertainty quantification of the cesium and iodine release fractions using point-estimate values of phenomenological uncertain parameters that were identified to affect the accident progression, release paths and magnitude of release. It was observed from the sensitivity indices that during LOCA, melt candling, fission product diffusion in the fuel and bubble characteristic models are phenomenologically important. Whereas, aerosol dynamics, vapour diffusivity, hygroscopic aerosol and bubble characteristic models were phenomenologically important during SBO accident. The melt debris release characteristics was shown to affect fission product release in both accident scenarios. For the uncertainty quantification, parameters were sampled using Monte-Carlo and Latin-Hypercube sampling methods. 95th percentiles for cesium and iodine releases were computed with empirical CDFs and Wilks’ methods. The results of the study provide valuable insights into the impact of MELCOR models, modelling parameters, and sensitivity coefficients on code predictions.

Severe Accidents (SA) are the most complex, extreme, and unlikely accident scenarios that might occur in a Nuclear Power Plant (NPP). However, no matter how unlikely they are, the five accidents that already occurred in commercial NPPs (TMI-2 [1], Chernobyl-4 [2] and Fukushima Daiichi (Units 1 through 3) [3]), highlight that “improbable” does not mean “impossible” and, given their potential consequences, their investigation is necessary to prevent and/or mitigate them.
Early this century, the Severe Accident Research NETwork of Excellence (SARNET) was born as an EC project and a decade later it became the Technical Area 2 (TA2) of NUGENIA, the SNETP (Sustainable Nuclear Energy Technology Platform) pillar devoted to research on Gen. II and Gen. III Light Water Reactors (LWRs). During these years, NUGENIA TA2 has produced meaningful advances in methodologies used for managing severe accidents, from enabling severe accident analytical tools to characterize the mitigation devices performance, passing through extension of some databases and exploration of new methodologies. For these achievements, the support of the different EURATOM Framework programs has been instrumental.
The proposed paper describes the most recent achievements from SNETP/NUGENIA research (2019-2023) related to phenomena occurring during a severe accident. . Besides, the commitment to knowledge dissemination through courses and conferences is highlighted. Finally, there is a consensus that any investigation on severe accidents to be launched in the coming years should have a direct impact on either reducing the uncertainties associated to their modelling or on optimizing their management, or on both. A specific project is currently running to identify and prioritize the issues deserving further research in the coming 10 years.

Determination of a source term is an essential part in the chain of nuclear safety analyses and serves as a cornerstone to following analyses of radiological consequences. Aside the design basis accidents like large break LOCA, events where primary coolant bypasses the containment may be of high importance. Such accidents are steam generator tube ruptures, where a flow of activities from primary to the secondary circuit and consequently through SDA directly into the environment surrounding the power plant may cause non negligible radiological consequences. Within EU R2CA project, UJV developed a methodology and computational tool, which uses existing or new results of relevant transients from thermal hydraulic codes such as RELAP5 or ATHLET and with application of balance equations calculates the source term. The methodology incorporates several typical physical phenomena occurring during the transport of activities between the primary and secondary circuit such as partitioning and flashing. In this paper, basics of the methodology will be presented, including sample application on a steam generator tube rupture of a VVER-1000/V-320 unit calculated with RELAP5.

This work aimed to validate a MELCOR model of the Reactor Cavity Cooling System (RCCS) by conducting Computational Fluid Dynamics (CFD) simulations. The development of a passive heat removal system design falls under the category of safety systems, which require guaranteed functionality and verification during the licensing process for the construction and operation of nuclear reactors, both under normal operating conditions and during accident scenarios. In the High Temperature Gas-cooled Reactor (HTGR), the containment structure differs from typical Light Water Reactors (LWRs) and is typically designed to be non-leaktight confinement. Therefore, the RCCS plays a crucial role as a safety function, aiming to control reactivity by maintaining fuel temperature below a limit (set at 1600°C) during an accident scenario with increased fission product release probability.

Based on the validated model of RCCS, a series of variant simulations were performed. The simulations were aimed at studying and analyzing various scenarios related to the RCCS, considering its effectiveness in heat removal and reactivity control during accident conditions. The results obtained from these simulations contribute to the understanding of the system's behaviour, performance, and safety features. The findings of this study have implications for the design, construction, and operation of future nuclear reactors, particularly in terms of safety system performance and accident mitigation strategies.

Since the beginning of nuclear industry, corrosion issues have been a major concern. Less than two years after start-up, stress corrosion cracking occurred on the stainless steel tubing of the steam generator of the prototype for the Nautilus in USA (1953); more recently, last year in 2022, several French Pressurised Water Reactors were shut down due to stress corrosion cracking of stainless steels pipes in a safety injection circuit. We propose to underline the complex corrosion mechanisms linked to the aggressive environments (high temperature and high pression water environments) and to present briefly the three main corrosion phenomena occurring in Light Water Reactors (LWRs), after a short overview of the basic designs and materials of the boiling water reactors (BWRs) and pressurized water reactors (PWRs):
- general corrosion of zirconium cladding which limits the life-time of fuel elements to generally 3 cycles;
- flow-accelerated corrosion (FAC) of carbon steel components, which is the only corrosion phenomenon that has led to several deaths in PWRs;
- stress corrosion cracking (SCC) of nickel base alloys (“the Coriou effect”) and of stainless steels including irradiation-assisted stress corrosion cracking (IASCC); SCC phenomena has led to the replacement of major components like steam generators or pressurised vessel heads.
Finally, the corrosion future will be discussed as BWRs and PWRs are extending their period of operation up to 60 and 80 years, and even more.

Through a series of projects (HTR-PL, Gemini+, GOSPOSTRATEG-HTR) the National Centre for Nuclear Research (NCBJ), Świerk, Poland, got involved into the small-scale High Temperature Gas-cooled Reactor (HTGR) technology. The objective is to replace the existing fossil fueled plants working for chemical and petrochemical industry by the nuclear reactors in order to reduce CO2 emission in Poland.
The task of this talk will be first short introduction into the HTGR technology – its inherent safety properties, the basic construction components, and an ability of high temperature process heat production. Then, some presentation of the current stage of the work on the concept of the 30 MWth research reactor to be build at the NCBJ site which would also serve a demonstrator of HTGR-SMR technology for Polish industry will be given. The reactor’s main technical specifications, its mission, its research, experimental, and utility objectives will be presented. Important steps of the reactor’s basic design and preliminary safety report preparation performed in collaboration with Japan Atomic Energy Agency (an HTTR operator) will also be briefly presented. A breakthrough of such a project would be a new boost to the nuclear in Poland after the only Polish working research reactor “Maria” constructed nearly 50 years ago.

The Saudi Low Power Research Reactor (LPRR) is one of the unique Saudi Vision 2030 projects. It is a pool-type multipurpose low power nuclear reactor designed and developed by King Abdulaziz City for Science and Technology (KACST) with the participation of international expertise in accordance with the best and highest international safety standards. The reactor is under construction and expected to be fully operated by mid of 2024. In line with the Vision 2030, the Saudi LPRR will thoroughly contribute to the innovative design and development of the nuclear reactor industry in the Kingdom, developing and qualifying competencies and building human capacity to operate nuclear power reactors, and transferring nuclear energy technologies. The LPRR is equipped with several attractive facilities with various capabilities and designed as a multipurpose low power reactor to support demands on research, education, training for building human capital, and developing the competencies to support the country’s national atomic energy project.
In this paper, the main characteristic of the Saudi LPRR will be presented in terms of the reactor core design, main utilization, and training capabilities.

This paper reviews recent experiences conducted worldwide tackling underutilization challenges faced by research reactors. A nuclear power program with ambitious goals, requires a well-established nuclear infrastructure and robust national framework. And to achieve that in a sustainable manner, countries aim to develop well-utilized research reactors. This paper explores case studies from different counties and sheds special light on the backwards flow approach to determine functional specifications, technical specifications (i.e. reactor core size, geometry, neutron flux, irradiation positions, fuel type, required irradiation duration) based on the captured national needs and aspirations through a set of analyses. Additionally, utilization requirements are presented using the mentioned approach considering three main applications which are radioisotopes production, neutron transmutation doping, and material testing.

In a nuclear reactor, one of the most important parts is the cladding, which is used to cover the fuel rod. The cladding will face many factors that impact the mechanical structure of the materials while in operation due to its position between the reactor coolant and the nuclear fuel. This cladding should be made of highly corrosion-resistant material with low thermal neutron absorbance.
The zirconium alloy is the preferred material for use as cladding in nuclear reactors. As previously stated, the zirconium alloy is highly corrosion-resistant, and a thermal neutron absorption cross-section is estimated to be 0.18 x 10-24 cm2. In the reactors, we are concerned with three types of zirconium alloys: Zr-Sn-Fe-Ni-Cr alloy (Zircaloy 2), Zr-Sn-Fe-Cr alloy (Zircaloy 4), and Zr-Sn-NbFe-Cr alloy (Zircaloy 2.5).
This material is exposed to pressure from coolant, temperature, and the irradiation process; as well as changes in the crystallization in the microstructure of the alloy. Thus, the zirconium alloy was found to achieve the purpose.
The article reviews the radiation effect on zirconium alloys in nuclear power plants and emphasizes the impact of radiation on the material's mechanical structure. It also explains some phenomena that occur within the cladding during reactor operation and impact the material's quality and life span.

The development of Generation IV Sodium-cooled Fast Reactors (SFRs) presents a crucial challenge in the form of Core Disruptive Accidents (CDAs). Boron migration is a significant concern in CDA evaluation since the eutectic reaction between boron carbide (B4C) and Stainless Steel (SS) can lead to forming a molten pool that could potentially relocate widely in the core, increasing the neutron absorption of the disrupted core. This study employs a quantitative and high-resolution method using Radiative heating to visualise the eutectic behaviour of boron migration and subsequent melt structure. The experiments were conducted using a B4C pellet with SS tubes in the temperature range of 1473 K−1645 K to observe the long-duration melting and its relocation behaviour for the first time. Eutectic melting was observed when the interface temperature reached around 1240 °C. Two melting mechanisms were observed: the SS peeling off the B4C pellet and forming a melting drop containing boride structures. The B4C pellet broke into multiple pieces due to thermal stress, and the entire SS melted during long-duration heating, while a small amount of B4C pellet was absorbed during the eutectic reaction, mixing with the SS melt. A detailed composition analysis using scanning electron microscopy-energy dispersive X-ray spectroscopy (SEM-EDX) revealed the microstructures of the molten drop after cooling at a cooling rate of 50 ºC/s. The high atomic boron zone was rich in chromium and iron. It did not contain nickel, suggesting chromium is more likely to coexist with boron in the solidified microstructure. Different crystal phases were confirmed using EDX analysis, which could help determine suitable decommissioning procedures. The study provides a deeper understanding of the behaviour of boron migration and its relocation in SFRs, shedding light on the mechanisms underlying the eutectic reaction between B4C and SS in the event of CDAs.

The in box-LOCA (Loss of Coolant Accident) is a key safety concern to be addressed in the design of the WCLL-BB (Water-Cooled Lead-Lithium Breeding Blanket). In this accident, a rupture of the tubes in the coolant circuit would cause a severe exothermic chemical reaction occurring between the primary coolant (water) and the Lead-Lithium alloy. The experimental facility LIFUS5/Mod3 was realized at ENEA Brasimone RC to investigate the phenomena associated with the interaction between lithium-lead and water. One of the main objectives of the experimental campaign is to provide data to develop and validate the SIMMER codes to be used on fusion scenarios related to the WCLL-BB safety design.
In the present work, post-test numerical simulations carried out by the University of Pisa with the codes SIMMER-III and SIMMER-IV are analyzed and compared against new experimental data obtained in the LIFUS5/Mod3 experimental campaign. It is shown that the code is capable of correctly capturing the main phenomena involved in the experiments, even though the quantitative predictions are still not satisfactory; however, the quantitative results presented and analyzed here are a crucial asset for the ongoing development and validation of the chemical model and are used to address the further steps needed to fully validate the codes.

In the European fusion project, the ENEA division of Fusion and technology for Nuclear Safety and
Security Department coordinates the Italian program supported by third parties (universities, Italian
research institutes and industries).
In particular, the ENEA Experimental Engineering Division with headquarter in the Brasimone RC,
design, develops and characterizes technologies and materials related to the Water Cooled Lithium
Lead (WCLL) Breeding Blanket (BB) and the connected ancillary system.
The WCLL is a BB concept featured by pressurized water as the coolant and Lithium-Lead Eutectic
(LLE) enriched at 90% in 6 Li as the breeder, neutron multiplier and tritium carrier.
ENEA research team is involved in the design of a multipurpose experimental infrastructure
conceived to investigate and qualify the water and LLE technologies for the DEMO BB and balance
of plant systems.
The proper design of the WCLL-BB is related to the capability of extraction of tritium in the BB
and to be rerouted as fuel in the plasma chamber to achieve the self-sustainability goal. To
successfully meet this task, experimental activities are carried out to characterize different T-
extraction technologies (e.g. Gas-Liquid Contactor-GLC an Permeator Against Vacuum-PAV
systems). In this tests hydrogen is used in the experiments as “cold surrogate” of T for safety issue.
Special attention is dedicated to development, optimization and characterization of structural and
functional materials having low activation, chemical compatibility, enough ductility, and creep
resistance at higher temperatures and dpa’s. Moreover, the protective coatings represent an
important R&D area to enhance materials functionality.
Other relevant research tasks are dedicated to the development of reliable instruments to control key
operative parameters, such as flow rate (development of a novel thermal flow meter), H
concentration in LLE (development of Hydrogen Permeation Sensor) and to the characterization of
physical properties such as the tritium solubility and diffusivity in LLE.

Two simulation programs SDTrimSP and RDS-BASIC were used to study the variation of sputtering energy threshold values (Eth) of beryllium metal irradiated by Helium, Tritium, and Deuterium ions when bombarded at various angles of incidence. The study aims to mimic the actual condition that the beryllium first wall of the International Thermonuclear Experimental Reactor (ITER) is subjected to in regular operation conditions. In all of the studied irradiation systems, we found that increasing the angle of incidence causes Eth values to decrease gradually until they reach their minimum values at an angle range between 40o and 70o. The Eth minimum values were found to be (10% to 35%) lower than their normal incidence value Eth(0o). These results were discussed theoretically and compared with one suggested theoretical model.

Due to their high specific heat, low viscosity, and good diffusivity, supercritical fluids have the potential to be ideal coolants. However, understanding the heat transfer for fluids under supercritical conditions has been a challenge. To understand the peculiar heat transfer characteristics, a wide range of experiments with different range of parameters and geometrical configurations has been conducted. The generated experimental data can be used as a reference to expand and assess the prediction accuracy of computational fluid dynamics models under supercritical conditions. Out of these models, RANS is the most widely used and consumes less computational power relative to other models. In this paper, natural circulation heat transfer of supercritical carbon dioxide will be investigated using RANS approach. To validate the prediction accuracy of RANS model, an extensive comparative study with experimental data will be presented in a full length article.

The fretting wear at the grid assemblies in fuel assemblies as a result of flow-induced vibration (FIV) is one of the main causes of fuel failures in the Light Water Reactor (LWR). Therefore, accurately predicting FIV is crucial to mitigating this issue, and a computationally efficient simulation method is necessary. In this regard, the Unsteady Reynolds-Averaged Navier-Stokes (URANS) approach is applied as a promising and efficient simulation method for FIV prediction. While previous studies have primarily relied on Large Eddy Simulation (LES) for the fluid domain, URANS provides an attractive alternative due to its lower computational demands, especially for a strong 2-way fluid-structure interaction (FSI) coupling. This paper aims to explore efficient approaches for benchmarking axial FIV for nuclear applications, focusing on a high-stiffness rod subjected to axial turbulence flow. The experiments conducted at Vattenfall AB, which involved a steel beam constrained on both ends, were used to predict the damping of vibration in an axial turbulent flow. The results show that URANS and even the laminar fluid model accurately predict the added mass effects of the vibration. The paper then examines self-exciting axial FIV and compares the experiments of axial FIV over a cantilevered rod at the University of Manchester using different URANS models and convective momentum equation divergence schemes. In both variations for the closure of the URANS model, which are the eddy viscosity model, the k-𝛚 SST model and the Reynolds Stresses Model (RSM), the Launder, Reece and Rodi (LRR) model, accurately predicted the frequency of vibration, while the former predicted with a lower amplitude of vibration. Modification of the k-𝛚 SST model with a higher-order cubic interpolation for the convective of the momentum equations could achieve accurate amplitude of vibration.

This study was part of a project that aims to investigate direct and scattered neutron flux, specifically, thermal and epithermal energy ranges around an unshielded PetTrace 880 cyclotron's target in several locations at a medical facility in the western region of Saudi Arabia. Efforts were made to characterize the neutron flux in a diagonal formation using a 3D-printed holder array within the allowable safe distance from the cyclotron target. The foil activation method was employed to estimate the thermal and epithermal neutron fluxes. The experimental design aimed to: 1) irradiate three batches of several sets of gold foils with and without cadmium filters, 2) collect the activated foils and finally 3) measure the photons emitted from the activated foils using a calibrated Broad Energy Germanium (BEGe) Detector. Initially, bare gold foils were irradiated to measure the direct and scattered thermal and epithermal neutron fluxes. Then, a second batch of one-side cadmium-covered gold foils was irradiated to account for the direct epithermal and scattered neutron fluxes. The third foil batch setup was made slightly different than the first two, where gold foils are inserted between two cadmium filters in a sandwich setup before the irradiation. The objective of the third batch irradiation was to isolate the thermal neutrons from the direct and scattered epithermal neutron fluxes. The mean thermal neutron flux was found to be in the order of 1.1E+06 ± 1.1E+05 neutrons cm-2s-1 at almost all investigated locations. The mean epithermal neutron flux was found to be in the order of 6.7E+05 ± 1.7E+04 neutrons cm-2s-1 at all investigated locations. The contribution from the scattered neutrons was found to be reduced by 20-50% after applying the sandwich setup. Further experiments will investigate neutrons with higher energies in the fast range to fully characterize the area around the cyclotron's target.

The scintillation detectors are one of the best for gamma spectroscopy. Thus, this experiment aims to determine the resolution and Efficiency of the detection system composed of a SiPM board in addition to three types of scintillation crystals. Two of the crystals are inorganic, CaF2 and LYSO. The third one is organic plastic scintillator crystal, NE102A. 4 different sources are studied in this experiment: Am-241 and Ba-133 for their low energies, Cs-137 for its mid-range Co-60 for its high energy. Each crystal was tested with each source to determine the resolution and Efficiency. For the CaF2, the resolution for the Am-241 and Cs-137 is 113.14% and 36.78%, respectively. As for the LYSO the resolution for the Am-241, Ba-133 (30.8 and 81 keV), and Cs-137 (32 and 662 keV) is 35.58%, 73.74%, 21.97%, 91.52%, and 11.10%, respectively. Moreover, the NE102A plastic scintillator crystal's resolution for the Am-241 is 37.5%. In addition, the Efficiency was calculated for each source using different crystals; for the CaF2, the absolute Efficiency for the Am-241 and Cs-137 is 2.76E-03% and 2.15E-04%, respectively. As for the LYSO, the absolute Efficiency for the Am-241 and Cs-137 (662 keV) is 0.104% and 0.013%, respectively. Moreover, in the NE102A plastic scintillator crystal, the absolute Efficiency for the Am-241 is 1.62E-03%

This study investigates the impact of simulating different fuel shapes for a material testing reactor (MTR) using the OpenMC code. Two 2-dimensional bare fuel element models were constructed: one with a curved fuel element shape to represent the actual dimensions of the MTR, and another with a simplified flat fuel element shape with the same amount of fuel as the curved model. The neutron distribution and k-eigenvalue were calculated and compared between the two models. The neutron distribution and k-eigenvalue showed only slight differences due to shape changes. The results indicate that simulating the MTR fuel as flat elements provides a satisfactory approximation of the real shape. However, it may introduce discrepancies for in-depth simulation studies.

The accelerating growth of the population and the constant development of technology created a high energy demand. However, these involved factors increased energy consumption, therefore increasing the greenhouse effect and triggering the atmospheric pollutants., Thus creating demand for a more sustainable and cleaner source of energy, which emerged the role of nuclear energy. The nuclear industry itself faced challenges revolving waste management where it needed advanced technology, and detailed plans to safely dispose of the radioactive waste. This paper investigates the geology of the Kingdom of Saudi Arabia to site the first Deep Geological Repository (DGR) using the Geographical Information System (GIS). The geology of the kingdom looks promising for such big-scale projects, The geology of the southern side of the kingdom looks promising for such big-scale project which centers within the empty quarter and the northern side in An Nafud desert which meets the requirements of DGR such as depth ranging up 1000m, rare possibilities of earthquakes, and geological formations. This paper contributes to the Saudi nuclear program for the waste management plans, generated from the commercial and research reactors following the 2030 vision to use nuclear and renewable energy resources.

Nuclear power became one of the major contributors to the worldwide energy mix, with a total share of 10% from thirty-three countries operating nuclear power reactors. Owing to the development of safe and advanced nuclear technologies, many countries are planning to embark on nuclear power. Moreover, it provides reliable and low-carbon power, which is well suited to address certain environmental challenges, such as those stated by the International Treaty on Climate Change signed in Paris in 2015. The agreement resembles international cooperation toward carbon neutrality. The Kingdom of Saudi Arabia (KSA) is one of the signatories of the Paris Treaty and is committed to reaching carbon neutrality by 2060. Furthermore, the kingdom has launched its peaceful nuclear power program to diversify the country’s energy mix, reduce greenhouse gas emissions, and meet the increasing demand for electricity. However, considering the region’s environmental conditions and other security concerns, several challenges are imposed on the kingdom’s first-ever Nuclear Power Plant (NPP). Therefore, this article attempts to address potential challenges that might face NPPs in the kingdom and provide lessons learned from past experience in NPP operation.

This senior project seeks to monitor Environmental Radioactivity (ERA) within Heqal village in Saudi Arabia, which has a high incidence of cancer. The research project will measure the radiation levels in the region's air, water, sand, stones, and plants. The study will utilize a Geiger counter and other radiation detection apparatus in the field work. The collected data will be analyzed to ascertain the village's radiation exposure-level. The significance of this research lies in the fact that it will aid in identifying potential sources of radiation exposure that may be contributing to the village's elevated cancer rates. It will also provide policymakers and public health officials with essential information for developing strategies to reduce radiation exposure and prevent future cases of cancer. Samples are collected from various locations within the village, including residential areas, agricultural fields, and water sources, as part of the research methodology. Using chemical separation technics in the laboratory, the samples will be analyzed for their radioactive content.
The results of this study are expected to substantially contribute to our understanding of in the villages with a high incidence of cancer. Additionally, the findings can inform future research on radiation monitoring and management strategies for comparable communities worldwide. Overall, by identifying potential sources of nuclear radiation exposure and devising effective prevention strategies, this senior project is a crucial step towards enhancing public health outcomes.

Research reactors are nuclear reactors used for radiopharmaceuticals production, basic and fundamental research, development of nuclear power technologies, education and training. They have usually low-temperature and low-pressure design with light water as the cooling medium. Since they serve as a neutron-producing machine for various medical, scientific and industrial applications, they have high neutron fluxes, usually at least an order of magnitude higher than the power nuclear reactors. Their cores can be filled with experimental devices that mimic the conditions in both fusion and advanced fission reactors. In the paper, the examples of the aforementioned applications are presented. MARIA's flexible design is described as an example of medium-sized reactor for countries with ambitions in fields of nuclear medicine and advanced power reactors technologies.

The TRIGA Mark II research reactor at the Jožef Stefan Institute in Slovenia, achieved first criticality in 1966. Since then the reactor has been playing an important role in developing nuclear technology. The reactor has been mainly used for research, education of university students, training of operators of the Krško nuclear power plant (start of operation in 1983) and other nuclear specialists, isotope production, and beam applications. Despite the age of the reactor, there is a wide range of research activities going on in the three main areas. 1) Reactor Physics activities are related to verification and validation of computer codes and nuclear data, testing and development of experimental equipment used for core physics tests at the Krško NPP, neutron radiography, neutron activation studies, development of bio-dosimeters, radiation hardness studies, safeguards activities. 2) Some of the environmental studies are using reactor for neutron activation analyses and for production of radioactive tracers. 3) Reactor is being used by particle physics department for radiation hardness studies of ATLAS detector in CERN. The future of nuclear technology in Slovenia is focused on new NPPs, while the research community is looking forward to a possible new nuclear reactor. The basic initiatives are at a very preliminary stage: the primary choice is dual-core pool-type reactor, with a zero-power core and a separate MW-size core, cooled and moderated with light water. Such a reactor will be capable of supporting the European fleet of existing and future nuclear power plants, including small modular reactors based on pressurized water reactor technology. Another option would be hosting one or more micro reactors with electrical and/or heating power producing capability. In this way the knowledge and infrastructure available for research and development could offer stronger support towards demonstration of prototype small modular reactors in prototype future electrical grids.

Radioisotopes are one of the essential cornerstones of modern medicine. They may be used for both diagnostic and therapeutic purposes. Here we present a description of a compact high field superconducting magnet used for a 30 MeV cyclotron with a magnetic field 2 times higher than conventional H- cyclotrons developed recently. This magnet will be a modern, state of the art design, which, because of the higher magnetic field, smaller, lower maintenance, lighter weight and lower power consumption than any other magnet available. The purpose of this design to help cyclotron to provide a sustainable supply of the critical Imaging Isotope F-18 and N13, to eliminate the need for supply from other production facilities for small centers. This paper mainly focuses on the simulation results of the Magnet proposed for our project as the Magnet design is 60% of the overall TAAC30 cyclotron Design.

The site selection for a nuclear reactor is a crucial safety and security activity with escalating demands. The prospect of using seismic data to estimate the operational behavior of the nuclear reactor is intriguing. Visualization and analysis of continuous data recorded at a seismic station away from the reactor site can be helpful in this regard. In this work, we investigate the potential for inferring the impact of an operational reactor from seismic data acquisition. The obtained data exhibit an apparent relationship between seismic features and reactor primary operational frequency. The short-time frequency transform is utilized to analyze the frequency components. The outcomes are helpful in choosing the potential nuclear reactor site in an oil and gas-rich region.

—This study presents the computing stress intensity factors (SIF) due to mechanical
stress generated under tensile loading, regarding semi-elliptic surface crack initiated inside a
finite plate. The analysis is further extended to study the effect of mechanical stresses on SIF for
a reactor pressure vessel (RPV) having an elliptic corner surface crack at the location of the
cylinder-nozzle intersection which is considered the point of highest stress concentration. The
specimen considered for the finite plate having a semi-elliptic surface crack is stainless steel
under tensile loading of 200 MPa and for RPV having an elliptic corner surface crack at the
location of cylinder-nozzle intersection under design pressure of 17.16 MPa, the material
parameters correspond to SA-533 Grade B Class-1. The propagation of the crack depends upon
the material’s fracture toughness if SIF under mechanical loading exceeds the material’s fracture
toughness, then the crack propagates leading to failure. The results obtained regarding SIF for a
finite plate having a semi-elliptic surface crack considering worst case scenario is 56 MPa√𝑚
and for RPV with elliptic surface crack is 141.7 MPa√𝑚, which is below the fracture toughness
of the material showing safe design. This study is done using the extended finite element method
(XFEM) in open-source software (SALOME MECA) to exemplify its application and accuracy.
The results are validated for both cases with a difference of less than 4% for the finite plate and
6% for RPV. The difference in results is due to limitations in computational power and mesh
refinement.

The author has recently developed a model to predict an upper-bound fatigue crack growth (FCG) rate for a metastable stainless steel immersed in various hydrogenated environments at ambient temperature and proposed that the high-frequency regime behavior is governed by hydrogen diffusion and can be adequately described by da/dN ∝ 1/(f) where  varied depending on the alloy microstructure, applied stress intensity factor (K), loading ratio, and the environment. The model predicts a frequency-independent upper-bound FCG rate for solution-annealed 304 type stainless steel as a function of the maximum stress intensity factor (Kmax). The model prediction shows good agreement with FCG data for 304 stainless steel in different hydrogenated environments.
The main objectives of this paper are two-fold. First, to compile fatigue crack growth (FCG) data generated at variable frequency from the open literature for different austenitic stainless steels exposed to simulated light water reactors (LWR) environments and to characterize the time-dependency of the possible governing crack-tip processes. Second, an attempt will be carried out to check the applicability of the prediction model to FCG data of stainless steels in LWR conditions taking into consideration the high-temperature effects on the mechanical properties and hydrogen diffusivity. The modified model prediction will be compared with available data in the literature. An upper-bound FCG prediction model for stainless steel exposed to hydrogenated water environments can be used for service-life assessment methods.

Nuclear reactor pressure vessel is a critical component of any nuclear power plant. The vessel is a thick wall container that withstands internal pressure caused by the reactor activity. This also plays important role to provide necessary barrier to keep radioactive materials out of the environment. Considering the functions, high strength low steel is usually proper candidate for making the vessel. However, reactor operation generates subatomic particle neutron under operation causing embrittlement and degradation of the steel vessel. Another source of degradation of the pressure vessel steel is by temper embrittlement due to its exposure at high temperature. Hydrogen embrittlement further deteriorates the situation. In this research work, the effect of hydrogen embrittlement on the fatigue crack growth behaviour of 2.25Cr-1.0Mo pressure vessel steel before and after temper embrittlement has been studied. Experimental results revealed that both hydrogen embrittlement and temper embrittlement contribute in enhancing crack growth and also in changing the fracture morphology.

The molten salt nuclear reactors, as an innovative type of Gen IV reactors, have a great
potential due to their safety and efficiency. However, the synergetic effects of irradiation and
corrosion of structural materials is not a fully explored area. Using a novel simultaneous
corrosion-irradiation facility, a FLiNaK molten salt in a pure nickel corrosion cell is exposed to a
proton beam through a 25m foil sample of the tested commercial alloys. Cross-section of each
foil were specifically prepared using dry ion polishing to preserve salt-filled features and
characterized. In order to evaluate the effects of irradiation on corrosion, assessments of the
corrosion depth and corrosion acceleration factor have been undertaken. The net corrosion effect
is determined by competing, simultaneous processes of acceleration via salt chemistry and
deceleration from radiation enhanced diffusion, found to be greatly dependent on alloying
elements and associated self-healing. Furthermore, we have discovered a subset of conditions,
practically useful for nuclear structural materials, where radiation damage decelerates or has no
effect on corrosion.

The advent of nuclear microreactors represents a significant leap in atomic technology, offering a unique combination of safety, versatility, and efficiency. As Saudi Arabia embarks on diversifying its energy sources and reducing carbon emissions, nuclear microreactors emerge as a promising solution, due to their compactness. This talk will delve into nuclear microreactors’ technical intricacies and operational mechanisms, underscoring their compact size, transportability, and rapid deployment capabilities.

We will explore how microreactors, with their inherent safety features and reduced need for on-site infrastructure, can revolutionize energy accessibility in remote and off-grid locations across Saudi Arabia, making unlivable places livable. The presentation will provide insights into the advanced cooling systems, passive safety measures, and modular design that characterize these reactors, ensuring a reduced environmental footprint and heightened security.

Moreover, the talk will address the specific energy needs of Saudi Arabia, highlighting how nuclear microreactors can support industrial processes, desalination plants, and urban energy grids and provide reliable 24/7 clean energy for megaprojects, such as the Line, Sindalah, Oxagon, etc. We will discuss the regulatory framework and safety protocols necessary for deploying these reactors, aligning with the Kingdom's Vision 2030 for sustainable energy.

Finally, the presentation will glimpse the future, ongoing research and development in nuclear microreactors and discuss solutions from companies like Aalo Atomics. We will explore potential advancements in fuel efficiency, waste management, and integration with renewable energy sources, setting the stage for a new era of nuclear energy in Saudi Arabia.

Liquid metal fast reactors represent one of the most promising proposals for the upcoming GEN IV of nuclear power plants. They indeed allow for both breeding processes and increased plant efficiencies: nevertheless, several challenges still need to be overcome. During the last years the European Union launched several projects in support of the development of such a technology: the University of Pisa joined the common effort providing numerical analyses addressing the thermal-hydraulics aspects of LFMBRs.
In particular, system thermal-hydraulics codes and CFD approaches were considered for the analysis of both normal operating conditions and accidental scenarios. Buoyancy-induced phenomena were particularly addressed aiming at understanding the capabilities of passive cooling systems. Both forced and natural circulation conditions were investigated: the results of the calculations were validated and compared against available experimental results showing in general good predicting capabilities.
The present paper reports the recent numerical activities performed at the University of Pisa in support of Gen IV LMFBRs. The addressed experimental facilities and experimental data are presented discussing the limits and capabilities of the adopted modelling techniques being STH, CFD and coupled STH/CFD applications. The obtained results are considered as a basis for the suggestion of best practice guidelines for the simulation of some of the NPP primary system components paying particular attention to the required computational resources and expected/required refinement of the adopted model. In addition, the capabilities of pre-test analyses in support of the final design of experimental facilities are highlighted. The future perspectives and foreseen developments are eventually resumed.

In the framework of the GEN IV innovative nuclear system, the ENEA Brasimone RC is supporting
the technological development of innovative nuclear system cooled by heavy liquid metals (HLM)
and, in particular, by liquid lead.
ENEA Brasimone RC is currently involved in several HORIZON2020 project such as PATRICIA,
PASCAL, ANSELMUS and INNUMAT.
Since 2013 ENEA has been member of the Fostering Alfred CONstruction (FALCON) international
consortium in partnership with ANSALDO NUCLEARE and RATEN-ICN supporting the
advanced lead-cooled fast reactor European demonstrator (ALFRED) to fully demonstrates the LFR
technology viability.
Moreover, in 2022 ENEA and newcleo signed a framework agreement with the intention of
exchanges information and knowledge for the construction of a lead-cooled nuclear prototype
outside Italy. The main goal of this cooperation is the construction at the ENEA Brasimone site of
an electrical prototype of the Lead-cooled Fast Reactor (LFR) system, to allow studying the thermo-
dynamic, mechanical and functional performances.
That said, ENEA Brasimone RC host one of the largest European fleets of experimental facilities
aiming at investigating HLM thermal-hydraulics, coolant chemistry control, corrosion behavior for
structural materials, and material properties in HLM environment, as well as at developing
corrosion-protective coatings, components, instrumentation, and innovative systems, supported by
experiments and numerical tools.
Particular attention is paid to research activities on sever accident precursors in GEN IV lead-cooled
fast reactors such as steam generator tube rupture and core flow blockages.
Moreover, experimental data on the simulation of postulated accidents are used to support the
development and the validation of numerical tools for specific application to liquid metals.

Heat is the world’s largest energy end use, accounting for almost half of global energy consumption in 2021. Half of this heat is used in industrial processes, and half is used in buildings for space and water heating. For most major industrial heat applications, nuclear energy is the only credible non-carbon option. Desalination is one attractive use for nuclear generated heat because the process requires lower-temperature heat than other industrial processes. Heat from light water reactors is suitable for desalination. The NDP-400 (Nuclear Desalination Plant) is a small advanced integral reactor that produces 400 MW of thermal power at a system pressure of 15 bar. The reactor’s coolant temperature is below 200 °C, which is relatively low compared with those of conventional PWRs for heat production. The NDP-400 offers economic benefits through system simplification. It has a compact printed circuit heat exchanger, reduced component size, and a lightweight design made possible by the low system pressure. The many of main components and reactor core in the NDP-400 have been proved in the development of Korea’s SMART. The reactor’s fuel cycle is 36 months long. The core power peaks and power distributions are comparable to conventional PWRs. The reactor’s safety features include a multi loop system, gravity-driven safety injection system, and low system pressure. These features maximize the inherent safety of the NDP-400. K.A.CARE and KAERI have collaborated on the development of a preliminary NDP-400 design since 2021. This paper presents the results of this design collaboration.

Radiation Hazards in the Oil & Gas Industry in KSA: A Comprehensive Analysis and Best Practices for Control and Management

Khaled Al-Qahtani [1]
Abdulkarim Mukhrish [2], Seraj Albahrani [1], Sufyan Alrubayyi [1], Afaque Shams [1]

[1] Mechanical Engineering Department, King Fahd University of Petroleum & Minerals, Dhahran 31261, Saudi Arabia.
[2] Chemical Engineering Department, King Fahd University of Petroleum & Minerals, Dhahran 31261, Saudi Arabia.

Abstract
Radiation hazards in the oil and gas industry present a significant concern in the Kingdom of Saudi Arabia due to the potential risks they pose to human health and the environment. By understanding the specific context of radiation hazards, control, and management practices in the Saudi Arabian oil and gas industry, the paper can investigate the various sources of radiation, including naturally occurring radioactive materials (NORM) and technologically enhanced naturally occurring radioactive materials (TENORM), and their prevalence in the industry. With an in-depth understanding of radioactive materials, the oil and gas industry in the Kingdom of Saudi Arabia, with the support of regulatory bodies, will be able to demonstrate an appropriate standard for radiation safety and radioactive material management to meet the interests of the oil and gas industry and to ensure public safety by emphasizes the need for comprehensive risk assessment and mitigation measures tailored to the local context. Overall, this Paper provides insights into the unique aspects of radiation hazards, control, and management practices in the Saudi Arabian oil and gas industry, with a focus on promoting safety and sustainability.

This study aims to investigate the issues related to stress corrosion cracking (SCC) in nuclear power plant systems. The study will explore the relationship between materials and environmental variables which cause the onset and progression of SCC in different materials commonly used in piping systems and related infrastructure.
Stress corrosion cracking is a significant issue in the nuclear power industry, leading to costly downtime and repair, and can have serious safety implications. The findings of this study can provide insights into the mechanisms underlying SCC and inform the development of strategies for preventing and mitigating its effects on piping systems.
By understanding the factors that contribute to SCC and how they can be mitigated, industries can develop strategies to prevent corrosion-related failures and ensure safe and reliable operation. The results of this study will contribute to the ongoing efforts to improve the reliability and safety of nuclear power plants and other industries that face similar corrosion-related challenges.

It is recommended that probabilistic safety assessment (PSA) should be involved in licensing Generation-IV reactors. Human reliability analysis (HRA) is an important component of PSA process. In the existing PSA models, consideration is being taken for actions of operator recovery and accident management for the plant recovery from a degraded state or core damage situation. As happened in Fukushima event, these actions can be strictly prohibited by releases at other facilities. The HRA models for single units do not consider such a scenario. There is necessity for the HRA of a multi-unit site to consider situation where the site is contaminated with radioactive materials and accident management action required to be conducted in this environment.
In general, HRA is an area of high uncertainty, as compared to PSA inputs that are derived from plant-specific data. Quantification of operator reliability in severe conditions further increases the uncertainties due to the conditions and stresses the operators might be likely to work under. The impact on PSA model refers to that techniques for estimating human reliability in extreme conditions are still developing and the quantitative estimates have large uncertainties. In some situations, conservative biases are used. In other cases, “best estimates” are employed.
Assumed conservative human errors can present a conservative bias in the computed risk results and should be assessed for masking effects. In cases where “best estimate” values are employed, sensitivity studies on the mean unreliability values are required to examine the effect of a change in failure probability on key risk metrics.
This paper highlights the unique issues that have to be considered in applying HRA for a multi-unit SMR. It presents a holistic framework for conducting HRA in the context of PSA in order to examine the hazards that threaten a multi-unit nuclear site.

In this study we develop a model that can predict the condensation heat transfer coefficient (HTC) during free-fall condensation on vertical tube surfaces in the presence of non-condensable gases (NCG). The aim is to compile a comprehensive database that includes a wide range of geometric values and operating conditions. The study is specifically motivated by the need to establish a generalized model/correlations that can predict the condensation heat transfer performance of the passive containment cooling system used in nuclear power plants. This passive cooling system eliminates heat from the containment vessel in case of an accident by condensing water vapor with gravity-driven force. To develop the model, we used MATLAB’s neural network toolbox to build an artificial neural network model, specifically a multi-layer perceptron network. The model predicts the HTC during the condensation process of two types of NCG (air and nitrogen) as well as pure steam. The inclusion of pure steam data aims to improve the accuracy of predictions under conditions where light gases are present. The dataset used for the model was constructed from 1,613 data points obtained from various experimental sources. The input layer receives various parameters, including P_tot,∆T_sub,W_nc,L,D_h. The output data is the condensation HTC. The input data was normalized by scaling each feature between the range of 0 to 1, whereas the output data was subjected to a transformation using the natural logarithm. The resulting machine learning model exhibited outstanding performance when predicting the condensation HTC. The findings of this study will represent a significant advancement in the analysis of large amounts of data from experiments and simulations, enabling the identification of complex patterns and relationships. Findings from the present study would serve as tools for the nuclear industry for designing and modelling Design Basis Accident (DBA) and Design Extension Conditions (DEC) scenarios.

Carbon emissions are a significant concern in today’s society due to their contribution to air pollution and global warming. As the entire world looks for greener energy sources, hydrogen fuel cells show great potential for a more sustainable future. Even though Saudi Arabia is the one of the world’s biggest exporters of crude oil, the Kingdom has developed its plan to reach carbon-neutrality following The Paris Agreement. Saudi Arabia has also developed its Saudi National Atomic Energy Project (SNAEP) and is looking towards adding nuclear power to their energy mix. Hydrogen can be produced by a variety of ways, and by understanding how it is produced and how it can be used as a clean energy source, a study on Saudi Arabia’s ability to produce Hydrogen as part of its SNAEP can be performed to help the country archive several goals with one project, bringing the Kingdom closer to its 2030 vision. This paper begins by exploring the energy landscape in Saudi Arabia by studying the country`s energy background, consumption, current sources, and challenges. Furthermore, it elaborates on different methods to produce hydrogen and how nuclear energy can be implemented to produce hydrogen. The feasibility study provides an economic assessment that compares the operation cost of producing hydrogen using different technologies and energy sources. Moreover, it sheds light on potential challenges in implementing these technologies in Saudi Arabia. It was found that implementing a high temperature electrolysis process that uses electricity from a nuclear power plant along with heat can lead to higher economic and technical efficiencies compared to other methods. Additionally, hydrogen production using nuclear power plants can supply extra revenue to the kingdom. Overall, this paper investigates the technical and economic feasibility of using nuclear energy to produce hydrogen in Saudi Arabia.

The Reactor Pressure Vessel (RPV) is a crucial component in nuclear systems. One of the mechanisms that can jeopardize the integrity of RPV is a transient condition called Pressurized Thermal Shock (PTS). During accidents such as Loss of Coolant Accidents (LOCA), emergency systems are initiated to mitigate the accident. The cold water injected by the Emergency Core Cooling System (ECCS) causes a quick cooling of the downcomer and internal surface of the RPV. As a result of this rapid cooling, large temperature gradients and thermal stresses occur, representing the PTS phenomenon. A proper understanding of this scenario is essential for the long-term safe operation of a nuclear power plant. In this regard, we have obvious limitations of experimental means. Similarly, one-dimensional thermal-hydraulic system codes don't add value in understanding this complex three-dimensional phenomenon. In that respect, advanced numerical methods such as Computational Fluid Dynamics (CFD) could play an important role in better understanding the thermal hydraulics of a PTS-type scenario. In the framework of CFD, direct numerical simulation (DNS) is considered as the most accurate method, which comes at the cost of large computational resources. This, in turn, limits the application of DNS to relatively low-Reynolds number flow cases. Whereas the large eddy simulation (LES), which is also considered as the high-fidelity method in CFD, once validated, could be beneficial in simulating high-Reynolds number flows. Therefore, in this present study, LES has been used to study the simplified PTS scenario. The obtained results are compared with the available DNS database. The obtained results indicate that, if performed correctly, LES can be used as a reference to validate low-order turbulence modelling approaches, such as RANS-based models

I have discussed this with Mr. Afaque Shams and I will submit my abstract in a couple of weeks.

I have discussed this with Mr. Afaque Shams and I will submit my abstract in a couple of weeks.

Accurate prediction of thermodynamic properties is an essential tool in any engineering setting. It is ever more valuable in settings where experimental data for the desired system is scarce – a repeating theme in mixtures found in novel nuclear reactors, notably in Molten Salt Reactors (MSRs) and CANDU reactors. SAFT equations of state promise to offer accurate estimates of a ultitude of physical properties of interest including isobaric heat capacity, viscosity, and solubility under varying operating conditions. The objective of this research is to expand the database of modelled pure substances and mixtures by SAFT-𝛾-Mie to systems applicable to prominent nuclear reactors. Initial modelling studies on mixtures of Deuterium and heavy water in CANDU reactors will be presented, together with details of plans for expansion of the database to incorporate various Molten Salt Reactor mixtures. Statistical Associating Fluid Theory (SAFT) is a family of equations of state (EoS) that pride themselves in their theoretical approach, building EoS using perturbation theory and Wertheim’s Thermodynamic Perturbation Theory (TPT1). The SAFT approach improves upon the prediction of crucial thermophysical properties of fluids that cubic EoS tended to struggle with, with non-spherical and ssociating fluids showing the greatest improvement. SAFT-𝛾-Mie is an EoS that uses a Group Contribution (GC) approach, allowing for the modelling of heterogenous chains unlike other variations of SAFT, which model all groups assuming an identical size. GC’s main premise is the prediction of physical properties of a molecule based on the individual contribution from each functional group present. Hence, the aforementioned models present an effective tool to be used alongside available experimental data in the extreme conditions found in the nuclear industry.

Despite the amount of water on our planet, water scarcity is a crucial problem facing the world. Thus, many countries resort to seawater desalination to solve this problem. Some of the most popular desalination technologies in the world are Reverse Osmosis desalination (RO), Multi Effect Desalination (MED) and Multi-Stage Flash desalination (MSF). These technologies are operated by heat and electricity, which are conventionally provided through fossil fuels. The world’s largest desalinated water producer by a significant margin is Saudi Arabia. Consequently, with the prospect of reducing the dependence on fossil fuels and shifting to more environmentally friendly sources, nuclear desalination will be a contributing factor. This study discusses the process of coupling Nuclear Power Plants (NPPs) with RO desalination plants in Saudi Arabia. As well as provide valuable recommendations to utilize the country’s new nuclear ventures in its already existing desalination infrastructure.

The high complexity inherent to reactor severe accidents has highly conditioned the capability to reliably model these scenarios from the beginning of the nuclear age. Nonetheless, noticeable accomplishments have been made along decades: from the initial simple correlations developed for the nuclear plant siting to the current integrated engineering tools, which are capable of simulating challenging multiphysic scenarios for days.The last generation of these analytical tools is numerically robust, fairly extensive in the amount of phenomena presently considered and largely validated to the extent possible. This being said, there is still room for meaningful improvement and uncertainty and sensitivity analysis (UaSA) might play a role in it, if properly adapted to the severe accident domain. This paper briefly walks the reader along all these topics and discusses that time has come for BEPU (Best Estimate Plus Uncertainty) application into the severe accident domain,although some additional work still needs to be done to achieve a systematic methodology in which engineering judgement will be more necessary than ever.

Pebble Bed Reactors (PBR) are a Generation-IV reactor design which are a subject of extensive research – due to their increased mixing and turbulence characteristics. Packed beds have sophisticated yet randomized geometry with vacant spaces, increasing the flow complexity in PBR cores. This experimental research on a facility of randomly packed pebble bed spheres investigates the complex flow phenomena to simulate fluid dynamics within a PBR core. By utilizing Particle Image Velocimetry (PIV) on the near-wall boundary and among the spheres, high-fidelity velocity measurements were carried out. In this facility, Matched Index of Refraction (MIR) provides a clear view of the packed spheres to analyze the flow at microscopic scales with precise resolution. The flow was investigated for various Reynolds numbers (𝑅𝑒) by utilizing PIV measurements. In order to provide a comprehensive profile of the flow and geometry that was studied, a three-dimensional reconstruction of the flow was carried out. This serves the primary purpose of providing the geometry for validation of simulation as well as the secondary purpose of illustrating the geometry of the packed bed. The experiments investigated isothermal conditions to examine the differences in flow dynamic patterns within packed spheres. The results characterize first- and second-order flow statistics including instantaneous velocity magnitude, mean velocity, velocity fluctuations, and Reynolds stresses. Moreover, the effect of different Reynolds number and heat flux boundaries and spheres were investigated. Increasing 𝑅𝑒 and heat flux were observed to affect the fluid dynamics within the pebble bed geometry, highlighting an increase in turbulence and flow mixing between spheres and within the gaps. This experimental campaign provides unique high-fidelity data sets for computational fluid dynamics model development and validation.

Plenum-to-plenum thermal hydraulics in a Prismatic Modular Reactor (PMR) during loss of flow accidents (LOFA) has been of high interest for several researchers. Accordingly, this study aims to study the effect of uniform and non-uniform heating configurations on the buoyancy driven air in a vertical dual channel plenum-to-plenum facility (P2PF) representing the PMR core geometry. Advanced measurement techniques such as hot wire anemometry, T-type thermocouples, and micro-foil sensors are integrated to enable thermal and flow fields measurements at different axial and radial locations along the electrically heated channel of the P2PF. Results show that surface temperature as well as air temperature and velocity increase from the leading edge followed by a reduction within the axial range of (0.6 < Z/L < 0.8) depending on the heating profile. This reduction could be attributed to the conduction heat losses through the flange connecting the hot channel to the upper plenum in addition to expected flow reversal and back mixing at the exit of the hot channel leading to flow destabilization. Indeed, it is concluded that the distribution of air temperature and velocity along the heated channel is significantly affected by the heating intensity and configuration applied to the channel. Present findings will overcome the lack of insufficient experimental data for studying the thermal hydraulics in PMR and for validating computational fluid dynamics codes.

The purpose of this investigation was to study bubble dynamics in an adiabatic air-water two-phase flow, mimicking the Loss of Coolant Accident (LOCA) scenario in a Pressurized Water Small Modular Reactor (PWSMR). A 5x5 rod bundle, with each rod 9.5 mm in diameter and a pitch-to-Diameter ratio (P/D) of 1.33, was used to represent the fuel rods. A 4-point fiber optical probe was used to obtain detailed data on bubble dynamics, including local void fraction, bubble velocity, bubble chord length, interfacial area concentration, and bubble passage frequency at a wide range of water and air superficial velocities. The experiment was conducted at various axial and radial locations before and right after the spacer grids to assess the impact of spacer grid mixing vanes. Higher void fraction, bubble passage, and specific interfacial area were observed in the subchannels compared to those obtained in the gap between the rods at all conditions. However, no significant differences were observed for the bubble chord length and bubble velocity. Moreover, Computational Fluid Dynamics (CFD) simulations were utilized to validate the data collected, which revealed a significant agreement between the experimental and numerical results. This study offers valuable insights into the behavior of air-water two-phase flow in rod bundles, which is critical to improving the safety and effectiveness of nuclear reactor design and operation.

The UAE's nuclear energy program, which began in 2008 with the publication of the "UAE Policy on the Peaceful Use of Nuclear Energy" document, has prioritized the safe, reliable, and economic operation of its APR-1400 nuclear power plants. Following the Fukushima events, the development of Accident Tolerant Fuels (ATFs) has become a focus in the nuclear fuel research and development community. ATFs are designed to withstand a significant loss of active cooling in the reactor core for a longer time period, resulting in increased safety compared to the existing fuel system while maintaining or improving normal operation performance. The purpose of this study is to investigate the feasibility of using ATFs as nuclear fuel in the APR1400 and assess their impact on the plant's safety and operation parameters. To accomplish this goal, a variety of analyses and assessments will be necessary, including neutronics, thermal-hydraulic, thermo-mechanical/chemical, and fuel performance analyses for multiple candidate ATF concepts. The investigation focuses on the APR1400 reactor, which is the reactor of choice in the UAE. The research aims to provide a review of ongoing research and potential ATF concepts for use in the near future in APR1400 nuclear power plants.

Hybrid microwave (HMW) processing was used to investigate the sintering behavior of 8 wt.% yttria-stabilized zirconia (YSZ) ceramic material. HMW sintering offers several advantages over conventional sintering methods such as but not limited to lower sintering temperatures, shorter times, selective and uniform heating as well as improved materials properties. 8 wt.% YSZ is a ceramic material and has the fluorite structure as uranium dioxide (UO2). Commercial 8% YSZ powder was used in the current study where it was pressed and molded into green body disks using uniaxial press before it was sintered using HMW heating using 2.45 GHz microwave frequency combined with conventional heating at the same cavity. Similar YSZ green disks were sintered at 3 different temperatures: 1350oC, 1400oC and 1450oC using the same holding time of 2 hours before cooling down to room temperature. The HMW sintered disks were then characterized using X-ray diffraction (XRD), scanning electron microscopy (SEM) and density measurement using Archimedes method. The crystal size of HMW sintered samples was calculated to be around 49-51 nm using Scherrer method at (111) plane at the sintered temperatures. Furthermore, SEM micrographs of the HMW sintered samples showed the typical sintering pattern of YSZ with its well-defined formed grains and grain boundaries. The samples with higher HMW sintering temperature exhibited the highest achieved densities values. HMW processing was used successfully to sinter 8 wt.% YSZ samples in a much lower temperature and shorter time compared to conventional sintering as inferred from the indexed XRD patterns and SEM micrographs of the sintered samples. Hybrid microwave processing is considered as a promising technique to sinter YSZ materials.

The irradiation with highly energetic (MeV-GeV) heavy ions can lead to the modifications of the properties of different materials [1]. Among the observed effects, the creation of surface nanostructures in different materials were demonstrated [2]. Recently, slow (eV-keV) highly charged ions (HCIs) were successfully utilized for the creation of similar surface nanostructures in various solids [3,4]. However, HCI exhibit unique feature by altering only the top surface layers without modifying the deeper layers, which can not be avoided for MeV-GeV heavy ions. Based on both the type of the material and ion beam parameters (charge state, kinetic energy, potential energy, etc..), nanostructures of different shapes (pits, caldera-like, hillocks) and sizes were obtained. This paper reviews the research progress of HCI-induced nanostructuring and the used theoretical approaches for understanding the creation mechanisms of the fabricated surface structures.

[1] A.S. El-Said, S. Rao, S. Akhmadaliev, Stefan Facsko, Physical Review Applied 13, 044073 (2020).
[2] A.S. El-Said, R.A. Wilhelm, R. Heller, Sh. Akhmadaliev, E. Schumann, M. Sorokin, S. Fascko, C. Trautmann, Nuclear Instruments and Methods B 382, 86 (2016).
[3] A.S. El-Said, R.A. Wilhelm, R. Heller, S. Facsko, Nuclear Instruments and Methods B 460, 137 (2019).
[4] A.S. El-Said, R. A. Wilhelm, R. Heller, M. Sorokin, S. Facsko, F. Aumayr, Physical Review Letters 117, 126101 (2016).

The Kingdom of Saudi Arabia (KSA) has prepared many strategies and plans that will be implemented to conserve natural resources according to the Kingdom's 2030 vision which promotes 17 sustainable development goals; the 6th goal of this strategy focuses on ensuring water quality for all. Several regions within the KSA were reported to have higher radionuclide concentrations in groundwater. The concentration of the radioactive radionuclides in groundwater depends mainly on the prevailing geochemical, geological, and hydrogeological conditions. This study aims to investigate a high background area of Natural radioactivity in Kingdom of Saudi Arabia (KSA) in Hail and Qassim by using Geographic Information System (GIS) and Remote Sensing (RS). An integrated approach that combines remote sensing data and tools, hydrogeological investigations, field observations, and geochemical analyses will be developed and applied to answer these four research questions over Hail and Qassim regions in central parts of KSA: (1) How do the geological conditions (e.g., lithology, structures) affect the radioactivity of groundwater? (2) how can we using Land use and Land Cover (LULC) tools to calculate the natural resources area, (3) how can we using Geographic Information System (GIS) to monitor radioactivity, (4) assess the risk of exposure due to natural radioactivity to estimate ( radiological effects like the radium equivalent (Raeq), the absorbed dose rate (Dr), external hazard index (Hex), internal hazard index (Hin), representative gamma hazard index (Iγ) and the total annual effective dose equivalent (AEDE)) according to the following expression (UNSCEAR, 2000). Results from this research will be reported in a GIS format (e.g., raster and shapefiles) and will be available, upon request, for locals, researchers from universities and institutions, and decision makers in different KSA’s governmental agencies.

Despite the acknowledged advantages that computer simulations can offer to Radiographic Testing, simulation solutions are uncommon in the industry. This paper presents an effective strategy for the simulation of Radiographic Testing techniques using Geant4 Application for Tomographic Emission (GATE), which is a medical imaging software toolkit.
Models for X-ray emission, test specimen geometry, radiation interaction with the test specimen, and image receptors were created in GATE. Double Wall Single Image technique was set up to evaluate the simulation models.
The results showed that GATE can afford a reliable and cost-effective solution in the simulation of Radiographic Testing procedures.

In the industrial sector, the focus in recent years has been on enhancing production and minimizing human error. Therefore, engineers have used non-destructive testing to evaluate materials by combining AI technologies with non-destructive tests. One of the field applications of NDT is the detection of welding defects.
Therefore, the use of neural networks enhanced the accuracy of detecting defects in industrial welding radiation testing. CNN with a triple classification for welding defects was suggested. Cut the original images to 150 × 150 pixels in the first step. The images were then divided into three categories: training, testing, and validation.
In the triple classification experiment (Crack, other types of Defects, No Defects), the CNN model had 6 layers and 9,667 parameters. Model accuracy approached 0.92% after 800 epochs. The F1 factors of Crack, other types of Defects, and No Defects were 100%, 91%, and 90%, respectively.
The article provides methods used by CNN in detecting welding defects and highlights the potential to improve defect detection accuracy.

Flow reconstruction of single-phase buoyant jet from sparse temperature measurements

Xicheng Wang, YiMeng Chan, Dmitry Grishchenko, Pavel Kudinov
Division of Nuclear Engineering, Royal Institute of Technology, KTH
Corresponding Address
xicheng@kth.se, ymchan@kth.se, dmitrygr@kth.se, pkudinov@kth.se
ABSTRACT
Measurement of the velocity field in thermal-hydraulic experiments is of great importance for phenomena interpretation and code validation. Direct measurement by means of Particle Image Velocimetry (PIV) is challenging in some multiphase's tests where the measurement system would be strongly affected by the phase interaction. A typical example can refer to the test with steam injection into a water pool where the rapid collapse of bubbles and significant temperature gradient makes it impossible to obtain main flow information in a relatively large steam flux.

The goal of this work is to investigate the capability of the use of data-driven methods for the flow reconstruction of the jet induced by steam condensation from sparse temperature measurement with ThermoCouples (TCs). The reconstruction is divided into 2 subtasks, that are (i) from sparse temperature measurements to complete temperature field, and (ii) from complete temperature field to complete velocity field. For the first subtask, several data-driven methods (e.g., linear regression, neural network) will be studied to build the mapping from measurement space to latent space with reduced dimensionality. Then, the relation between latent space and complete space will be encoded by Proper Orthogonal Decomposition (POD). For the second subtask, Physics Informed Neural Networks (PINNs) and linear regression are considered. The paper starts with a single-phase buoyant jet to narrow down the uncertainties of the proposed framework. Validated CFD scheme is used to generate the training data.

KEYWORDS: Data-driven, flow reconstruction, buoyant flow, physics informed neural networks.

Large gas bubbles separated by the liquid slugs are the main characteristic of the slug flow regime. We have analyzed the stagnant Taylor bubble in the vertical isothermal counter-current flow with high speed videos at 100 to 800 frames per second. A single Taylor bubble was captured in each experiment through dynamical balance of the bubble drag in the downward liquid flow. Bubbles of around two to six diameters length were observed in two pipes of 12.4 and 26 mm diameter. Liquid Reynolds numbers in front of the bubble were around 1000 (laminar flow) and 6000 (turbulent flow) in small and large diameter pipe, respectively. Video frequencies at around 400 Hz were found sufficient to capture all temporal fluctuations of the bubble interface. Algorithms for two-phase interface recognition have been developed and applied on the images of the cap and the body of the Taylor bubbles. We have shown stable axisymmetric Taylor bubble in laminar flow and asymmetric bubbles in turbulent liquid flow. Even the long time averaging of up to 10 minutes did not produced axisymmetric time-averaged shape of the bubble in turbulent liquid flow. In turbulent regime we have observed bubble of bullet-train shape with the thinnest liquid film on the belly of the bullet-train shape bubble. Azimuthal position of the bubble's belly is randomly determined during the injection of the bubble into the test section. In addition, dynamics of the tiny disturbance waves with tenth of mm amplitudes has been tracked along the interface of the Taylor bubbles in laminar and turbulent cases. Stable standing waves were observed in the laminar case and traveling waves in the turbulent liquid experiments. Cross-correlations of time-dependent interface fluctuations were measured at different spatial positions in turbulent flow to determine propagation speeds of the traveling interface waves.

Multiphase flow systems, usually a flow system involving multiple phases, are and have been for decades a phenomenon of particular interest to researchers and scientists globally. This is due to the extensive application of such systems in the physical, chemical, biomedical and petrochemical industries. In the nuclear power industry, accurate prediction of transient phenomenon occurring in the reactor core, primary and secondary loops largely involving two-phase flows is paramount for safety analysis and control. The safety system analysis codes used employ the two-fluid model which treats the thermal hydraulic properties of each fluid separately and couples both systems by the introduction of interfacial transfer terms. Understanding of interfacial structures and deducing the interfacial area concentration leads to correct modeling of the interfacial terms. Since bubbles play a significant role in a two-phase flow thermal-hydraulics properties, the effects of bubble formation and growth on flow properties or system performance cannot be overemphasized. This work therefore introduces an algorithm for detecting bubbles in a two-phase flow using a multiple thresholding approach to detect bubbles in layers based on pixel intensity stemming from their relative distance away from the capturing source. Parameters from the detected bubbles serves as the bedrock for the deduction of other secondary parameters needed to accurately deduce the interfacial area concentration and hence the interfacial transfer terms.

Nuclear energy has long been recognized as a low-carbon emission technology. However, the widespread adoption of nuclear power plants (NPPs) is hindered by the complexity of their man-machine-network integration systems, the occurrence of various faults, insufficient automation, and the challenges faced by human operators. In recent years, the development and utilization of artificial intelligence (AI) technology has presented both prospects and complexities in enhancing the functionality and security of nuclear reactors. This exponential growth of AI offers novel avenues to optimize the operation and ensure the safety of nuclear reactors. Artificial intelligence (AI) technologies have the potential to address the limitations and enhance the functionality of NPPs. In this review different AI techniques are investigated which can contribute to NPPs Accident Scenario Identification. By exploring the applications of AI in this context, we shed light on the potential benefits and advancements that can be achieved by integrating AI technologies into nuclear energy systems.

Passive safety systems are integrated into the latest generation of Light Water Reactors (LWRs), including small modular reactors. This paper employs the US-NRC TRACE thermal hydraulic code to examine the performance of a passive safety condenser known as SACO, designed to serve as the ultimate heat sink for dissipating decay heat during accident scenarios. The TRACE model is constructed with reference to the PKL/SACO test facility, which is an integral testing facility replicating a four-loop Western-type KWU pressurized water reactor (PWR). The PKL facility maintains a 1:1 height scaling and a 1:145 power and volume scaling. The safety condenser (SACO) is interconnected with the PKL facility via the secondary side of steam generator 1, effectively serving as a third natural circulation cooling loop during accident scenarios. The modeling of the PKL/SACO facility involves the use of both 1D and 3D TRACE components. Specifically, the SACO water pool is represented as a 3D TRACE VESSEL component, while all other facility components are represented as 1D TRACE components, including PIPE, VALVE, FILL, BREAK, and single junction. Previously, a series of parametric investigations had been conducted aimed at validating the PKL/SACO TRACE model. In the present research, the thermal-hydraulic behavior of the PKL facility is investigated in the presence of the SACO passive safety system during a Station Black Out (SBO) with Extended Loss of AC Power (LEAP) accident scenario. The SBO scenario entails an extended and prolonged transient process, which can be categorized into three distinct phases depending on the activation of the SACO system and the refilling process of the SACO pool. The findings indicate that the SACO system effectively manages to dissipate all decay heat, even though there is temporary evaporation of the SACO water pool.

One of the main requirements to the reactor safety is the assurance that the appropriate margins of fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences. CPR (critical power ratio) defined as the ratio of the critical power to the bundle power at the reactor condition of interest is widely used as a figure of merit for expressing BWR thermal margin. Traditionally, CPR is predicted by applying empirical CPR-correlations developed by fuel vendors from the actual fuel design experimental data. The recent trend in new fuel bundle development is the enhanced application of subchannel codes to fuel designing. The paper discusses the applicability of CTF (COBRA-TF) subchannel code for CPR prediction in comparison to the classical CPR correlations. The CPR is estimated by subchannel analysis from the critical condition that is defined when the subchannel code CTF predicts critical heat flux at any location of the FAs during iterative simulations with a power ramp. The obtained CTF results are then compared to the CPR estimations obtained from 3-D core simulations using the fuel design specific CPR correlation. Finally, the uncertainties in CPR estimation by CTF are analyzed by applying classical Wilks formula and available uncertainties distributions for the most important parameters that affects CPR, i.e. power, coolant flow rate and system pressure.

Saudi Arabia plans to diversify its energy sector by focusing on low-carbon generation and nuclear energy is being explored in this context. Siting Nuclear Power Plants necessitates careful consideration of many factors such as water access, hazard exposure, proximity to anticipated load, and possible coproduct generation (e.g., desalination). Existing studies have only explored large light water reactors and identified limited coastal sites as possible due to consequent water requirements. However, small and micro modular reactors could mitigate this issue and open up a larger swath of the country. Here, we consider the added potential for nuclear when considering different sizes of reactors. We perform country-scale geospatial analyses at high resolution, using multiple reactor scales and corresponding criteria following internationally recognized guidelines. We identify viable sites across Saudi Arabia, and we show that decreasing the size of the reactors significantly increase the potential locations of nuclear power plants. We use existing infrastructure density to indicate the most economically promising sites. The power and siting flexibility of smaller reactors makes them good candidates to replace existing power generation infrastructure. This path can help Saudi Arabia develop a successful civilian nuclear program.

Medical cyclotrons have become a crucial component in the healthcare industry, particularly in the field of nuclear medicine. The Kingdom of Saudi Arabia (KSA) has seen significant advancements in the development of the cyclotron facility in the medical sector in recent years. This review paper provides an overview of the life cycle of medical cyclotrons in the KSA, highlighting the various stages from conception to operation.

The KSA has made significant investments in the development of medical cyclotron facilities, with several hospitals and research centers having established their own facilities. These facilities have enabled the local production of radiopharmaceuticals, reducing the reliance on imported products, and enhancing the quality of patient care.

This paper discusses the life cycle of a medical cyclotron, which starts with the planning phase, where a feasibility study is conducted to evaluate the potential demand for the facility. The second phase involves the design and construction of the cyclotron facility, including the installation of the cyclotron, target assembly, and supporting equipment. The third phase is the commissioning of the facility, which involves the testing and validation of the equipment and processes.

The fourth phase is the operational phase, where the facility is used for producing radioactive isotopes for medical applications. The final phase is the decommissioning of the facility, which involves the safe disposal of the equipment and the restoration of the site to its original state.

This abstract will provide valuable insights into the life cycle of medical cyclotrons in the KSA and highlight the importance of these facilities in enhancing the quality of patient care. The increasing demand for radiopharmaceuticals makes the development of medical cyclotron facilities a critical priority for the healthcare industry in the KSA.

The purpose of this reseach is to provide a dosimeter films for industrial and medical applications using different types of dyes with polyvinyl alcohol as support (using a concentration of 1 ml and 3 thicknesses of (10, 15, 20 µm), and PVA films were exposure to gamma rays. The films were exposure to different dosages range (from 5-60 kGy). UV/VIS spectrophotometer was used to measure the colour change of film dosimeter before and after exposure to gamma rays (at 250 to 800nm), Also the film stability was studied at different time periods 1, 7 and 14 days is done for all films. However, in this study, a natural dye (RO-H, RO-C) was also used (400 to 800 nm). The RO-H film gives better result than Ro-C, which need further study with different parameters.

Ionizing radiation shielding is required to prevent or mitigate the radiological risks resulting therefrom. Low Z materials such as polyethylene are preferable for neutron shielding, while high Z materials such as lead are preferable for photons (gamma and x-rays). Concrete is a conventional shielding material that is used to shield against either photons or neutrons. Although concrete is cheap and can be easily formed, it is responsible for 8% of carbon dioxide emissions. If volcanic silica rocks take the role of concrete in radiation shielding, this will help reduce the level of carbon dioxide emission. Monte Carlo code Fluka was used to simulate the experiment setup and calculate the exposure rate on the other side of the shielding samples. The obtained results showed that the linear, mass attenuation, and absorption coefficients of volcanic silica are almost like those of concrete. These results reveal that volcanic silica rocks could be used similarly to concrete for the shield against X-rays diagnostic range up to 250 keV.

Reactor physics is the study of the behavior of neutrons in nuclear reactors. Monte Carlo codes are a popular method for simulating the behavior of neutrons in complex geometries. One of the Monte Carlo codes is OpenMC code which is an open source code developed by the Massachusetts Institute of Technology. It designed to simulate neutron transport in various geometries allowing flexible geometry and material specifications as well as a range of neutron cross section libraries. VERA Core Physics benchmark aims to validate the performance of computational tools used in the design and analysis of nuclear reactors. The benchmark includes a set of core physics problems such as the prediction of criticality, power distribution and control rod worth. In this paper, Analysis of 2D Hot Zero Power (HZP) Beginning of Cycle (BOC) Quarter Core problem of VERA benchmark will be simulated using OpenMC code and study the capability to predict the results of criticality, pin power distribution as well as control rod worth. The simulations of OpenMC code show a good agreement with the VERA benchmark. These results provide confidence in the capability of OpenMC simulations in high fidelity calculations for generating data to analyze reactor core physics.

Vitreous materials in form of both relatively homogeneous glasses and glass crystalline materials (GCM) incorporating crystalline disperse phases are currently the most reliable wasteforms effectively used on industrial scale for nuclear waste immobilisation. Glasses are solid state materials with a topologically disordered atomic structure which can be considered as true solid solutions i.e., solutions being frozen via vitrification to a solid state without phase separation. Nuclear waste vitrification is attractive because of technological and compositional flexibility enabling hazardous elements to be safely immobilised providing a glassy material characterised by high corrosion resistance, mechanical and radiation durability, as well as effectively reducing the volume of the resulting wasteform. Borosilicate and to a lesser extent phosphate glasses are the overwhelming world-wide choice for the immobilization of high-level radioactive wastes (HLW) resulting from used nuclear fuel reprocessing and low- and intermediate level radioactive wastes (LILW) such as those from operation of nuclear power plants and legacy waste. Vitrification is a mature technology which has been used on an industrial scale for more than fifty years. Continued advances in devising durable vitreous wasteforms and improving nuclear waste vitrification technologies provides key solutions in enabling widespread deployment of nuclear energy.

Condensation passive containment cooling systems in combination with small modular reactors are pivotal for removing excessive heat from reactors during steam release accident scenarios. However, the presence of non-condensable gases (NCG) significantly reduces the heat transfer, causing safety concerns. This is mainly caused by the accumulation of NCG during condensation. A better understanding of the physics of the accumulation of NCG owing to condensation leads to the design of efficient heat removal and better coolant recovery mechanisms. Additionally, the accumulation of NCG (mainly hydrogen and oxygen) in the main evaporator of a nuclear reactor containment may cause explosion safety concerns. In a recent study, jet impingement condensation (JIC) was reported as an excellent method for achieving high condensation rates owing to the improved thinning of the diffusion boundary layer associated with the accumulation of NCG near the liquid-vapor interface. The high convective heat and mass transfer coefficients associated with impinging jets have been successfully utilized in several industrial applications, such as drying processes, electronic cooling, and cooling of turbine blades, to name a few. Designing an optimal JIC system requires understanding its physics at varying parameters, such as jet size, jet-to-surface distance, jet speed, jet-to-surface temperature/concentration difference, and environmental temperature/concentration. In this study, we present a numerical model to investigate the influence of these parameters on the initial stages of the condensation process. We look mainly at the existence of Breath Figure spots, which are defined as the effective areas at which condensation occurs. The numerical model was validated against experimental and semi-analytical work on a single round jet issued from a tube at a prescribed temperature, NCG concentration, and flow rate. Our model provides an effective numerical tool for better understanding the physics of efficient condensation in the presence of NCG.

Steam generator tube rupture (SGTR) is one of the safety issues for pool type lead-cooled fast reactors (LFR). After high-pressure water is injected into the lead pool, the subsequently generated steam bubble would transport to the core and affect the heat transfer performance. This paper addresses tracking the bubble motion using an Eulerian-Lagrangian method in CFD based on the 1/8 centrosymmetric scale geometric model of ELSY primary system. The steady and transient bubble distributions under different leakage heights are obtained. Furthermore, the simulation results are predicted by machine learning, where Gaussian Process Regression (GPR) is employed for steady conditions and backpropagation neural network (BPNN) for transient conditions. For steady conditions, the prediction results by the kernel function of ardrationalquadratic show the best accuracy in predicting the percentages of bubbles reaching the core, top of the steam generator, and staying in the system. Four typical transient cases of bubble accumulation in the core are selected for BPNN prediction. All the cases are well predicted with R2>0.99. In conclusion, machine learning algorithms have great potential to predict bubble transport in the primary system after SGTR.

The research findings presented in the literature confirmed that the rotated triangle array is inherently fluidelastically unstable in two-phase flow, especially in the transverse direction. On the other hand, recent work confirmed that the rotated square array is fluidelastically stable for all tested void fractions in two-phase flow, except for 97%. The quasi-steady analysis showed a significant reduction in damping for 97% void fraction compared to lower void fractions. The quasi-steady model, however, could not resolve the issue of the increase in tube bundle vibrations in the transverse direction for 97% void fraction. Hence, further analysis is required to deeply look into the array using the unsteady theory. In this work, the unsteady fluid forces were measure for a rotated square array with P/D=1.64. The advantage the unsteady theory has is taking into consideration the variation of the fluid force phase with reduced flow velocity. This is not encountered in the quasi-steady theory where the fluid force phase is always assumed to be constant. Unlike the quasi-static force measurements, the unsteady fluid dynamic force component and the vibration modes of the tubes are taken into account in the unsteady theory. The results of this work add a deeper understanding to the rotated square array dynamic behaviour. An array that showed a stable behaviour in two-phase flow. This study aimed in part to analyse the APR1400 steam generator tube bundle in single and two-phase cross-flow.

This work presents an overview of Nagra’s AMAC (Advanced Methodology for Activation Characterization), a Monte Carlo (MCNP) based methodology used to simulate the neutron propagation as well as the activation processes during the operation of a nuclear power plant (NPP). By a high-resolution tally-system it is possible to characterize the activation distribution for each major component of the NPP as well as for the surrounding building structures. The resulting nuclide inventories are stored in an NPP specific activation atlas and can be particularly used for NPP decommission planning purposes but also for handling and shielding issues during the plant operation. Attached user-friendly modules to AMAC allow for visualization of the activation distribution and the nuclide activity balancing and bookkeeping for all relevant NPP components. Furthermore, they enable the development of segmentation strategies followed by optimized automated packaging concepts for the decommission waste into an adaptable list of available container types. Thus, AMAC is expected to lead to significant cost savings by reducing the number of expensive waste containers as well as by significantly limiting the number of radiological measurements during NPP decommissioning. The methodology has been validated by activation foil campaigns where samples have been placed at selected locations within the NPP building for a full operation cycle irradiation. Alternatively, destructive sampling followed by radiological analyses have been also conducted for validation and scaling purposes. AMAC has been applied to all Swiss NPPs and research reactors as well as for corresponding nuclear facilities in Germany and South-Korea. Furthermore, its results are the basis for the periodic Swiss cost study for the definition of the size of the NPP decommission and disposal funds. In summary, the application of AMAC leads to the detailed knowledge of the radiological situation caused by activation processes of an NPP during operation and for decommission planning.

In this search for new sources of clean energy to replace the consumption of fossil fuels, nuclear energy comes as a solution, as efficient, clean, and highly beneficial as an energy source. All these benefits come with a cost of byproducts that could harm the environment such as radioactive Iodine. In this study, we propose the use of functionalized Polyethylene as an absorbent for the capture of radioactive iodine. The study will focus on the functionalization of PE by gamma rays that will produce peroxides on the surface of PE, followed by the functionalization with various heteroaromatic moieties that will create efficient binding sites on the surface of PE to capture Iodine. The synthesized functionalized PE will be characterized by various spectroscopic techniques such as NMR, and FT-IR. Thermal stability will be evaluated by TGA and DSC. The adsorption capacities will be investigated by UV adsorption. The kinetics and mechanistic details of the adsorption process will be investigated against different variables such as time, concentration, and heat. The study will provide a new direction that will help in the use of nuclear energy in the kingdom.

Nuclear energy is not sustainable if its waste is not effectively managed, considering its global significance, accounting for 10% of electricity generation and climate change mitigation efforts. Saudi Arabia's entry into nuclear power underscores this shift towards a cleaner energy future.
This poster discuss the type of the nuclear waste coming from the nuclear power plant, and presents a case study focused on nuclear waste generated by two Pressurized Water Reactors in Saudi Arabia. It outlines three essential waste management phases: Fuel Pool, Dry Storage, and Deep Geological Disposal. Calculations were made to size the fuel pool and discuss the type of the dry storage cask.
Recommendations from the study include keeping spent fuel in a pool for at least five years, proximity of dry storage facilities to the power plant, horizontal casks for safe transportation, and deep geological disposal as a long-term solution.
To make DGD feasible, stable site selection, stringent regulation, long-term containment, adherence to global guidelines, secure waste transport, and adaptability are necessary. Ensuring nuclear waste is efficiently managed is crucial for the continued use of nuclear energy as a sustainable, clean power source.

A valuable method for figuring out the concentrations of various elements in natural materials, either online or in situ, without changing their chemical forms in the matter, is the prompt gamma-ray neutron activation analysis (PGNAA). The current study illustrates three different techniques that have been adopted at KFUPM that sought the yield of PGNAA systems that records gamma rays from neutron inelastic scatterings and neutron thermal capture reactions. These techniques are portable neutron generators (DD-reaction), Am-Be source, and neutron accelerator (DT-reaction). This work will clarify the advantages and disadvantages of each system and the improvement of each one to measure the minimum detectable concentrations (MDC) of heavy contaminants in environmental samples. Results and comparisons will be discussed.

The Kingdom of Saudi Arabia is exploring the use of advanced nuclear power plants for hydrogen production as a sustainable and clean alternative to conventional hydrogen production methods. In this paper, a techno-economic model for hydrogen production using advanced nuclear power plant technologies in Saudi Arabia is presented. In this model, the capital and operating costs of advanced nuclear reactor technology, as well as the costs of hydrogen production plant were considered. Two distinct hydrogen production approaches are evaluated; a standalone electrolysis process and an integrated system that combines high-temperature steam electrolysis (HTSE) with nuclear cogeneration. The cogeneration process involves the production of both hydrogen and electricity from the same advanced nuclear power plant, utilizing electricity and high-temperature steam produced by the reactor. In this study, the Hydrogen Economic Evaluation Program (HEEP) was employed to evaluate various scenarios, including variations in the capital cost of the advanced nuclear power plants, the hydrogen production cost, and other factors such as the cost of nuclear fuel, operation and maintenance costs, and safety considerations. To ensure the safe and efficient integration of advanced nuclear power plant-based hydrogen production into Saudi Arabia's energy system, the technical implications of factors such as the design and operation of the nuclear reactor, the compatibility of the HTSE unit with the reactor, and the balance of plant requirements for hydrogen production were analyzed. The study provides valuable data-driven insights into the feasibility and economic viability of utilizing advanced nuclear power plants for hydrogen production in Saudi Arabia and can support in the decision-making process regarding investments in this emerging field.

Maintaining a constant power output of a nuclear power reactor, NPP, is preferred due to operational factors such as xenon poisoning and optimizing the capital investment. However, constant output does not match variable demand, requiring making it a baseload source or storing excess energy giving larger flexibility. Compressed air energy storage (CAES) is an ideal solution due to its high capacity, power rating, long lifespan (~40Y), and technical maturity. This makes it a suitable choice for nuclear power plants to store and utilize excess energy.

To incorporate CAES technology, the nuclear power plant utilizes two tanks of phase change material (PCM): one for hot and one for cold. The nuclear heat is utilized to power an absorption chiller process, which produces cold ice storage. The heat that has degraded by 20-40°C is then used to charge the hot PCM storage tank with a suitable temperature that is lower than the reactor output.

To optimize the integration of CAES in NPP for Saudi Arabia, the desalination process has been incorporated into the CAES charging process. This involves utilizing saline water to absorb heat during the compression process through direct contact heat exchange. By using a cold PCM tank, the compression process can be made to be nearly isothermal just above the freezing point of water, resulting in the conversion of compression heat into fresh water while keeping the compression work low and near the isothermal limit.

During the expansion process, the compressed air regains its original state by expanding nearly isothermally, with energy coming from the hot PCM tank powered by nuclear heat. This setup ensures that the CAES process does not interfere with the nuclear power operation, thereby making the CAES system more effective and efficient in Saudi Arabia.