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[en] This study was undertaken to investigate the relationships film and processing solution at different processing temperature. Three kinds of periapical film were used for this study. They included EP-21 film, DF-58, and A film EAch film was processed by automatic film processor with RD-III, X-dol 90, and A processing solutions at 68 degree, 74 degree, 80 degree, 86 degree, and 92 degree F. Film density was measured with the densitometer, and base plus fog density, film relative speed, film contrast, and subject contrast were evaluated. The following results were obtained: 1. As the processing temperature was increased, base plus density was increased. Inadequate base plus fog densities were obtained with three films in combination with three processing solutions at 92 degree F. 2. Lowest base plus fog densities were obtained with A film, followed in ascending order by EP-21, and DF-58 film i n combination with A or RD-III processing solutions. The sequence of base plus fog densities was in ascending order by EP-21, A, and DF-58 film in combination with X-dol 90 processing solution. 3. The sequence of film relative speed values was in ascending order of EP-21, A, and DF-58 film in combination with A and RD-III processing solutions, respectively.
[en] A project for the development of a new research reactor (NRR) was launched by KAERI in 2012. It has two purposes: 1) providing a facility for radioisotope production, neutron transmutation doping, and semiconductor wafer doping, and 2) obtaining a standard model for exporting a research reactor (RR). The instrumentation and control (I and C) design should reveal an appropriate architecture for the NRR export. The adoption of a graded approach (GA) was taken into account to design the I and C and architecture. Although the GA for RRs is currently under development by the IAEA, it has been recommended and applied in many areas of nuclear facilities. The Canadian Nuclear Safety Commission allows for the use of a GA for RRs to meet the safety requirements. Germany applied the GA to a decommissioning project. It categorized the level of complexity of the decommissioning project using the GA. In the case of 10 C.F.R. Part 830 830.7, a contractor must use a GA to implement the requirements of the part, document the basis of the GA used, and submit that document to U.S. DOE. It mentions that a challenge is the inconsistent application of GA on DOE programs. RG 1.176 states that graded quality assurance brings benefits of resource allocation based on the safety significance of the items. The U.S. NRC also applied the GA to decommissioning small facilities. The NASA published a handbook for risk informed decision making that is conducted using a GA. ISATR67.04.09 2005 supplements ANSI/ISA.S67.04.01. 2000 and ISA RP67.04.02 2000 in determining the setpoint using a GA. The GA is defined as a risk informed approach that, without compromising safety, allows safety requirements to be implemented in such a way that the level of design, analysis, and documentation are commensurate with the potential risks of the reactor. The IAEA is developing a GA through DS351 and has recommended applying it to a reactor design according to power and hazarding level. Owing to the wide range of RR utilization, the safety requirements for RRs may not be required to be applied to every RR in the same way. DS351 also states that the way in which the requirements are demonstrated to be met for a multipurpose and high power RR might be very different from the way in which the requirements are demonstrated to be met for a RR with very low power and very low associated radiological hazards to the facility staff, the public, and the environment. The GA should not compromise safety or waive the safety requirements. The GA is not a quantitative method but rather a qualitative method to determine the scope and level of application of the safety requirements to the design of a RR. It adopts a systematic approach and engineering judgment for the determination. The GA is applicable in all stages of the RR lifetime. Any grading during the lifetime should ensure that safety functions are maintained and that there are no radiological hazards to the operators and public. The grading activities should be based on a safety analysis, regulatory requirements, and engineering judgment. In DS351, the GA activities consist of two steps: 1) categorizing a facility into a range of the highest to the lowest risk, which is an initial grading of the facility, and 2) grading the system, structure, and components important to safety, which is a more detailed grading of the facility. As an example of the GA, fewer inspections and hold points for a 100 kW RR than those for a 5 MW RR can be determined. For the application of the GA to the I and C design of an RR, Rah man proposed the GA to develop the digital MMIS (Man Machine Interface System) for RRs regarding cyber security, software V and V, and human factors engineering. However, it did not show the specific design decisions. Suh presented the overall I and C architecture for the NRR, but it has a lack of rationale for the design decision making. This paper presents a strategy to make a design decision for NRR I and C systems. According to the characteristics and safety analysis of the NRR, the proper design level should be determined to avoid an over design
[en] As Man Machine Interface System (MMIS) of commercial power plant is fully digitalized, that of research reactor is also changed from analog to digital system. Thus, research reactor is furnished with Information Processing System (IPS) conducting the role of information layer between control layer and operation layer of MMIS. Basically, the IPS acquires and processes reactor data from other I and C subsystems. The goal of IPS is to provide reactor status information to operator personnel in control room. In this paper, we define functional roles of IPS in research reactor and describe an applicable system models to the IPS. Finally, we propose adequate system architecture for the IPS by taking account of economic feasibility and effectiveness factors
[en] NRR) by 2016 was launched by KAERI in 2012. The purposes of the project are to meet domestic industrial needs of a research reactor and to secure an internationally competitive NRR. For instrumentation and control (I and C) systems of the NRR, it shall also be designed to secure the competition of the NRR. The I and C should account for the competition in terms of three aspects: safety, performance and cost. A I and C conceptual design activity should be carried out at the early stage of the project to figure out the I and C on the high level. Constructing I and C architecture is to accomplish the high level design. During the I and C architecture construction, the three aspects should be taken into account. This paper conceptually constructs the I and C architecture for the NRR by comparing to the cases of a Jordan training and research reactor (JRTR) project and a RA-10 multipurpose research reactor. The JRTR is an on-going project launched by KAERI and DAEWOO consortium in 2011. The I and C of the JRTR was digitalized based on the I and C functions of the HANARO research reactor, as shown in Fig. 1. The RA-10 was launched by INVAP of Argentina in 2011. The digital I and C developed by INVAP was built in an OPAL of Australia, as shown in Fig. 2
[en] It can provide precise status information of plant to operator, make preventive maintenance, give better operational aid and eliminate human errors. In the near future, this technology will be introduced to nuclear power plant through overcoming some weakness. In this study, current status of art of 4th industrial revolution technology and the strength and weakness of its technology when introducing them to the nuclear power plant. Although the current 4th industrial revolution technology has some weakness, they are widely used and being developed in various industrial fields and society to improve safety and economy.
[en] Application of digital I and C has increased in nuclear industry since last two decades but lack of experience, innovative and naive nature of technology and insufficient failure information raised questions on its use. The issues has been highlighted due to the use of digital I and C which were not relevant to analog. These are the potential weakness of digital systems for Common Cause Failure, threat to system security and reliability due to inter channel communication, need for highly integrated control room and difficulty to assess the digital I and C reliability. In the existing scenario, HANARO and JRTR have hybrid I and C systems (digital plus analog) whereas OPAL is fully digitalized. In order to authenticate the choice of fully digital I and C architecture for research reactor, it is required to perform assessment from risk point of view, cyber security as well other issues. The architecture assessment method and restrictions are discussed in the next part of article
[en] Though research reactors are small in size yet they are important in terms of industrial applications and R and D, educational purposes. Keeping the eye on its importance, Korean government has intention to upgrade and extend this industry. Presently, Korea is operating only HANARO at Korea Atomic Energy Research Institute (KAERI) and AGN-201K at Kyung Hee University (KHU), which are not sufficient to meet the current requirements of research and education. In addition, we need self-sufficiency in design and selfreliance in design and operation, as we are installing research reactors in domestic as well as foreign territories for instance Jordan. Based on these demands, KAERI and universities initiated a 5 year research project since December 2011 collaboratly, for the deep study of reactor core, thermal hydraulics, materials and instrumentation and control (I and C). This particular study is being carried out to develop highly reliable advanced digital I and C systems using a grading approach. It is worth mentioning that next generation research reactor should be equipped with advance state of the art digital I and C for safe and reliable operation and impermeable cyber security system that is needed to be devised. Moreover, human error is one of important area which should be linked with I and C in terms of Man Machine Interface System (MMIS) and development of I and C should cover human factor engineering. Presently, the digital I and C and MMIS are well developed for commercial power stations whereas such level of development does not exist for research reactors in Korea. Since the functional and safety requirements of research reactors are not so strict as commercial power plants, the design of digital I and C systems for research reactors seems to be graded based on the stringency of regulatory requirements. This paper was motivated for the introduction of those missions, so it is going to describe the general overview of digital I and C systems, the graded approaches, and future plans of the project
[en] The reactor protection system (RPS) in a research reactor is a well-known conventional setpoint-based protection system. The RPS performs protective actions with the generation of alarms when the measurement values exceed the setpoints. The RPS has disadvantages in that alarms are not generated before the measurement values exceed the setpoints; they are generated at the time of protection actions are performed. In addition, each alarm has a direct relation with signals, not accidents, so it is difficult to identify the accident type quickly. Thus, new methods are required to diagnose and classify accidents. We propose a deep-learning-based alarm system. The proposed alarm system is modeled with convolutional and fully connected neural networks. The proposed scheme is designed from safety analysis in the safety analysis report. We prepare various datasets and scenarios for training and test. The results show that the proposed alarm system provides fast diagnosis alarms with probability values.