• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.851-859
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• 2021

Small PWR can be used for power generation and heating. Considering that small PWR has the characteristics of flexible operating conditions and complex operating environment, the controller designed based on single power level is difficult to achieve the ideal control of small PWR in the whole range of core power range. To solve this problem, a flexible switching controller based on fuzzy controller and LQG/LTR controller is designed. Firstly, a core fuzzy multi-model suitable for full power range is established. Then, T-S fuzzy rules are designed to realize the flexible switching between fuzzy controller and LQG/LTR controller. Finally, based on the core power feedback principle, the core flexible switching control system of small PWR is established and simulated. The results show that the flexible switching controller can effectively control the core power of small PWR and the control effect has the advantages of both fuzzy controller and LQG/LTR controller.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.466-473
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• 2021

Fire protection is one of important issues to ensure safety and reduce risks of nuclear power plants (NPPs). While robust programs to shut down commercial reactors in any fires have been successfully maintained, the concept and associated regulatory requirements are constantly changing or strengthening by lessons learned from operating experiences and information all over the world. As part of this context, it is necessary not only to establish specific fire hazard assessment methods reflecting the characteristics of research reactors and educational reactors but also to make decisions based on advancement encompassing numerical analyses and experiments. The objectives of this study are to address fire simulation in the control room of an educational reactor and to discuss integrity of digital console in charge of main operation as well as analysis results through comparison. Three electrical fire scenarios were postulated and twenty-four thermal analyses were carried out taking into account two turbulence models, two cable materials and two ventilation conditions. Twelve supplementary thermal analyses and six subsequent structural analyses were also conducted for further examination on the temperature and heat flux of cable and von Mises stress of digital console, respectively. As consequences, effects of each parameter were quantified in detail and future applicability was briefly discussed. On the whole, higher profiles were obtained when Deardorff turbulence model was employed or polyvinyl chloride material and larger ventilation condition were considered. All the maximum values considered in this study met the allowable criteria so that safety action seems available by sustained integrity of the cable linked to digital console within operators' reaction time of 300 s.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.549-561
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• 2021

Indonesia has policies and strategies for the management of radioactive waste and spent nuclear fuel that arises from the use of nuclear research and development facilities, including three research reactors, and the use of radioisotopes in medicine and industries. The Indonesian government has provided extensive facilities such as an independent regulatory organization (BAPETEN) and a centralized radioactive waste management organization (CRWT-BATAN). Further, the presence of regulations and several international conventions guarantee the protection of the public from all risks due to handling radioactive waste and spent nuclear fuel. However, the sustainability of radioactive waste management in the future faces various challenges, such as disposal issues related to not only to site selection but also financing of radioactive waste management. Likewise, the problem of transportation persists; as an archipelago country, Indonesia still struggles to manage the infrastructure required for the transport of radioactive materials. The waste from the production of the radioisotopes, especially from the production of 99Mo, requires special attention because BATAN has never handled it. Indonesia should also resolve the management of NORM from various activities. In Indonesia, the definition of radioactive waste does not include NORM. Therefore, the management of this waste needs revision and improvement on the regulations, infrastructure, and technology.

• Nuclear Engineering and Technology
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• v.55 no.8
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• pp.3102-3113
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• 2023

Following the Fukushima nuclear disaster, the simulation of accidents in the spent fuel pool has become more noticeable. Despite the low amount of decay heat power, the consequences of the accidents in a spent fuel pool (SFP) can be severe due to the high content of long-lived radionuclides and lack of protection by the pressure vessel. In this study, the loss-of-cooling accident (LOFA) for the VVER-1000/V446 spent fuel pool is simulated by employing RELAP5 and MELCOR 1.8.6 as the best estimate and severe accident analysis codes, respectively. For two cases with different total power levels, decay heat of spent fuels is calculated by ORIGEN-II code. For modeling SFP of a VVER-1000, a qualified nodalizations are considered in both codes. During LOFA in SFP, the key sequences such as heating up of the pool water, boiling and reducing the water level, uncovering the spent fuels, increasing the temperature of the spent fuels, starting oxidation process (generating Hydrogen and extra power), the onset of fuel melting, and finally releasing radionuclides are studied for both cases. The obtained results show a reasonable consistency between the RELAP5 and MELCOR codes, especially before starting the oxidation process.

• Nuclear Engineering and Technology
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• v.55 no.2
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• pp.690-695
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• 2023

Recommendations are given to improve the efficiency of radiation protection of transport casks for SNF transportation. The attenuation of ${\gamma}$-quanta of long-lived isotopes ¹³⁴Cs, ^137mBa(¹³⁷Cs), ¹⁵⁴Eu and ⁶⁰Co by optimizing the thicknesses and arrangement of layers of Fe and Pb radiation shields of transport casks is studied. The fixed radiation shielding mass (fixed mass thickness) is chosen as the main optimization criterion. The effect of the placement order of Fe and Pb layers in a combined two-layer radiation shield with an equivalent thickness of 30 cm is studied in detail. It is shown that with the same mass thicknesses of the Fe and Pb layers, the placement of Fe in the first layer, and Pb - in the second one provides more than twofold attenuation of ${\gamma}$-quanta compared to the reverse placement: Pb - in the first layer, Fe - in the second. The increase in the efficiency of attenuation of ${\gamma}$-quanta for TC with combined shielding of Fe and Pb is shown to be achieved by designing the first layer of radiation shielding around the canister with SNF from Fe of the maximum possible thickness.

• Nuclear Engineering and Technology
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• v.55 no.1
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• pp.238-247
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• 2023

The present work represents a technical guideline for decommissioning a disused teletherapy machine model Theratron-780 and contains category one ⁶⁰Co radioactive source. The first section predicts the dose rate from the source in case of normal and radiological emergency situations via FLUKA-MC simulation code. Moreover, the dose assessment for the occupational during the whole process is calculated and compared to the measured values. A suggested cordoned area for safety and security in a radiological emergency is simulated. The second section lists the whole process's technical procedures, including (preview, dismantle, securing, transport and storage) of the disused teletherapy machine. Results show that the maximum obtained accumulated dose for occupational were found to be 24.5 ± 4.9 μSv in the dismantle and securing process in addition to 3.5 ± 1.8 μSv during loading on the transport vehicle and unloading at the storage facility. It was found that the measured accumulated dose for workers is in good agreement with the estimated one by uncertainty not exceeding 5% in normal operating conditions.

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.908-918
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• 2023

The Saudi National Atomic Energy project aims to adopt peaceful nuclear technologies and be part of the country's energy mix. As emerging nuclear energy, it is essential to understand public concerns and acceptability of nuclear energy, as well as the factors influencing acceptance to develop nuclear energy policy and implement nuclear energy programs. The purpose of this study is to analyze the public attitudes and acceptance of nuclear energy among Saudi Arabian citizens by utilizing protection motivation theory and theory of planned behavior. A total of 1,404 participants answered a questionnaire which was distribute by convenience sampling approach. A Structural Equation Modeling framework was constructed and analyzed to understand public behavior toward building the country's first Nuclear Power Plant (NPP). Before analyzing the data, the model was validated. The research concluded that the benefits of nuclear power plants were essential in determining people's acceptance of NPPs. Surprisingly, the effect of the perceived benefits was found higher than the effect of the perceived risks to the acceptance. Furthermore, the public's participation in this study revealed that the NPPs location has a significant impact on their acceptance. Based on the finding, several policy implementations were suggested. Finally, the study's model results would benefit scholars, government agencies, and the business sector in Saudi Arabia and worldwide.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.1966-1973
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• 2023

Virtual reality technology has been widely used in the field of nuclear and radiation safety, dose rate computing in virtual environment is essential for optimizing radiation protection and planning the work in radioactive-controlled area. Because the CPU-based gamma dose rate computing takes up a large amount of time and computing power for voxelization of volumetric radioactive source, it is inefficient and limited in its applied scope. This study is to develop an efficient gamma dose rate computing code and apply into fast virtual simulation. To improve the computing efficiency of the point kernel algorithm in the reference (Li et al., 2020), we design a GPU-based computing framework for taking full advantage of computing power of virtual engine, propose a novel voxelization algorithm of volumetric radioactive source. According to the framework, we develop the GPPK(GPU-based point kernel gamma dose rate computing) code using GPU programming, to realize the fast dose rate computing in virtual world. The test results show that the GPPK code is play and plug for different scenarios of virtual simulation, has a better performance than CPU-based gamma dose rate computing code, especially on the voxelization of three-dimensional (3D) model. The accuracy of dose rates from the proposed method is in the acceptable range.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.2246-2255
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• 2023

Evacuation time estimation (ETE) is crucial for the effective implementation of resident protection measures as well as planning, owing to its applicability to nuclear emergencies. However, as confirmed in the Fukushima case, the ETE performed by nuclear operators does not reflect behavioral features, exposing thus, gaps that are likely to appear in real-world situations. Existing research methods including surveys and interviews have limitations in extracting highly feasible behavioral features. To overcome these limitations, we propose a VR-based immersive experiment system. The VR system realistically simulates nuclear emergencies by structuring existing disasters and human decision processes in response to the disasters. Evacuation behavioral features were quantitatively extracted through the proposed experiment system, and this system was systematically verified by statistical analysis and a comparative study of experimental results based on previous research. In addition, as part of future work, an application method that can simulate multi-level evacuation dynamics was proposed. The proposed experiment system is significant in presenting an innovative methodology for quantitatively extracting human behavioral features that have not been comprehensively studied in evacuation. It is expected that more realistic evacuation behavioral features can be collected through additional experiments and studies of various evacuation factors in the future.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3677-3691
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• 2023

Since containment building is the major shielding structure to ensure safety of nuclear power plant, the structural behavior and ultimate pressure capacity of containments must be studied in depth. This paper addresses ambiguous issue of determining free-field position for liner failure by suggesting an expanded free-field region and comparing internal pressure capacities obtained by test data, conservative assumption and suggested free-field region. For this purpose, a practical approach to determine the free-field position for the evaluation of liner tearing is carried out. The maximum principal strain histories versus internal pressure capacities among different free-field positions at various azimuths and elevations are compared with those at the equipment hatch as a conservative assumption. The comparison shows that there are considerable differences in the internal pressure capacity at liner failure within the expanded free-field region compared to the vicinity of the equipment hatch. Additionally, this study proposes an approximate correlation with conservative factors by considering the expanded free-field ranges and material characteristics to determine realistic failure criteria for liner. The applicability of the proposed correlation is demonstrated by comparing the internal pressure capacities of full-scale containment buildings following liner failure criteria according to RG 1.216 and an approximate correlation.