• Nuclear Engineering and Technology
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• 제55권10호
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• pp.3898-3906
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• 2023

The paper notes that the most common way of handling spent nuclear fuel (SNF) of power reactors is its temporary long-term dry storage. At the same time, the operation of the dry spent fuel storage facilities almost never use the modern capabilities of information systems in safety control and collecting information for the next studies under implementation of aging management programs. The author proposes a structure of an information system that can be implemented in a dry spent fuel storage facility with ventilated storage containers. To control the thermal component of spent fuel storage safety, a database structure has been developed, which contains 5 tables. An algorithm for monitoring the thermal state of spent fuel was created for the proposed information system, which is based on the comparison of measured and forecast values of the safety criterion, in which the level of heating the ventilation air temperature was chosen. Predictive values of the safety criterion are obtained on the basis of previously published studies. The proposed algorithm is an implementation of the information function of the system. The proposed information system can be used for effective thermal monitoring and collecting information for the next studies under the implementation of aging management programs for spent fuel storage equipment, permanent control of spent fuel storage safety, staff training, etc.

• Nuclear Engineering and Technology
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• 제56권8호
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• pp.2893-2905
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• 2024

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. 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. In the present research, the thermal-hydraulic behavior of the PKL facility is investigated in the presence of the SACO passive safety system during an extended SBO with Loss of AC Power accident scenario. This SBO can be categorized into three distinct phases depending on the activation of the SACO system and the refilling process of the SACO pool. The first phase is depressurizing using primary and secondary relief valves, the second phase is cooling down using SACO system, and the third phase is the refilling of 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. Furthermore, this study provides sensitivity analysis for the assessments of system codes on the selection of maximum time step.

• Journal of Platform Technology
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• 제11권6호
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• pp.13-20
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• 2023

Regularly inspecting vehicle tires' condition is imperative for driving safety and comfort. Poorly maintained tires can pose fatal risks, leading to accidents. Unfortunately, manual tire visual inspections are often considered no less laborious than employing an automatic tire inspection system. Nevertheless, an automated tire inspection method can significantly enhance driver compliance and awareness, encouraging routine checks. Therefore, there is an urgency for automated tire inspection solutions. Here, we focus on developing a deep learning (DL) model to predict cracked tires. The main idea of this study is to demonstrate the comparative analysis of DenseNet121, VGG-19 and EfficientNet Convolution Neural Network-based (CNN) Transfer Learning (TL) and suggest which model is more recommended for cracked tire classification tasks. To measure the model's effectiveness, we experimented using a publicly accessible dataset of 1028 images categorized into two classes. Our experimental results obtain good performance in terms of accuracy, with 0.9515. This shows that the model is reliable even though it works on a dataset of tire images which are characterized by homogeneous color intensity.

• Nuclear Engineering and Technology
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• 제52권9호
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• pp.1983-1989
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• 2020

The human reliability analysis is a method by which, in general terms, the human impact to the safety and risk of a nuclear power plant operation can be modelled, quantified and analysed. It is an indispensable element of the PSA process within the nuclear industry nowadays. The paper herein presents a sensitivity study of the human reliability analysis performed on a real nuclear power plant-specific probabilistic safety assessment model. The analysis is performed on a pre-selected set of post-initiator operator actions. The purpose of the study is to investigate the impact of these operator actions on the plant risk by altering their corresponding human error probabilities in a wide spectrum. The results direct the fact that the future effort should be focused on maintaining the current human reliability level, i.e. not letting it worsen, rather than improving it.

• Nuclear Engineering and Technology
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• 제41권8호
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• pp.1065-1072
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• 2009

As the use of software increases at nuclear power plants (NPPs), the necessity for including software reliability and/or safety into the NPP Probabilistic Safety Assessment (PSA) rises. This work proposes an application procedure of software reliability growth models (RGMs), which are most widely used to quantify software reliability, to NPP PSA. Through the proposed procedure, it can be determined if a software reliability growth model can be applied to the NPP PSA before its real application. The procedure proposed in this work is expected to be very helpful for incorporating software into NPP PSA.

• Nuclear Engineering and Technology
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• 제46권3호
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• pp.447-460
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• 2014

The results of this study, suggest public communication to promote the use of radiation as follows: first, suitable information for the recipient's perception patterns should be provided, as there is a difference in risk perception and acceptance between the experts and the public. Thus, information on the necessity of nuclear power should be provided to the public, while information based on technical risks is provided by the experts. Second, since the levels of perception, knowledge, and attitudes increased highly for sectors which use radiation after the class, classes should be provided continuously to increase students' perception, knowledge, and attitude, which are all preemptive variables which induce positive behavioral changes. Third, since the seven sectors which use radiation are highly correlated, arguments for the necessity of other sectors should be based on the necessity of the medical sector.

• Nuclear Engineering and Technology
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• 제41권10호
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• pp.1223-1234
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• 2009

The outlines of seismic PSA implementation standards and seismic hazard evaluation procedure were shown. An overview of the cause investigation of seismic motion amplification on the Niigata-ken Chuetsu-oki (NCO) earthquake was also shown. Then, the contents for improving the seismic hazard evaluation methodology based on the lessons learned from the NCO earthquake were described. (1) It is very important to recognize the effectiveness of a fault model on the detail seismic hazard evaluation for the near seismic source through the cause investigation of the NCO earthquake. (2) In order to perform and proceed with a seismic hazard evaluation, the Japan Nuclear Energy Safety Organization has proposed the framework of the open deliberation rule regarding the treatment of uncertainty which was made so as to be able to utilize a logic tree. (3) The b-value evaluation on the "Stress concentrating zone," which is a high seismic activity around the NCO hypocenter area, should be modified based on the Gutenberg-Richter equation.

• Nuclear Engineering and Technology
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• 제41권7호
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• pp.921-928
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• 2009

As part of an evaluation for an in-vessel severe accident management strategy, a coolant injection into the reactor vessel under depressurization of the reactor coolant system (RCS) has been evaluated in detail using the SCDAP/RELAP5 computer code. A high-pressure sequence of a small break loss of coolant accident (SBLOCA) has been analyzed in the Optimized Power Reactor (OPR) 1000. The SCDAP/RELAP5 results have shown that safety injection timing and capacity with RCS depressurization timing and capacity are very effective on the reactor vessel failure during a severe accident. Only one train operation of the high pressure safety injection (HPSI) for 30,000 seconds with RCS depressurization prevents failure of the reactor vessel. In this case, the operation of only the low pressure safety injection (LPSI) without a HPSI does not prevent failure of the reactor vessel.

• Nuclear Engineering and Technology
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• 제42권3호
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• pp.259-270
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• 2010

New generation nuclear reactors are designed using advanced safety analysis methods. A thorough understanding of different interacting physical phenomena is necessary to avoid underestimation and overestimation of consequences of off-normal transients in the reactor safety analysis results. This feature requires a multiphysics reactor simulation model. In this context, a coupled dynamics model based on a multiphysics formulation is developed indigenously for the transient analysis of large pressurized VVER reactors. Major simplifications are employed in the model by making several assumptions based on the physics of individual phenomenon. Space and time grids are optimized to minimize the computational bulk. The capability of the model is demonstrated by solving a series of international (AER) benchmark problems for VVER reactors. The developed model was used to analyze a number of reactivity transients that are likely to occur in VVER reactors.

• Nuclear Engineering and Technology
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• 제46권4호
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• pp.467-474
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• 2014

The aim of this paper is to describe a relatively recent international agreement on the widely debated concepts of: (i) attributing effects to low dose radiation exposure situations that have occurred in the past and, (ii) inferring radiation risk to situations that are planned to occur in the future. An important global consensus has been recently achieved on these fundamental issues at the level of the highest international intergovernmental body: the General Assembly of the United Nations. The General Assembly has welcomed with appreciation a scientific report on attributing health effects to radiation exposure and inferring risks that had been prepared the United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR) following a formal request by the General Assembly.