• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.06a
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• pp.205-210
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• 2005

The proposed regulation for low and intermediate level radioactive waste disposal facility, scheduled to be revised, recommends that the waste generator should verify the radioactive waste conforms to the disposal requirements before disposing of it. According to the regulation, the radionuclide concentration of the radioactive waste, and its physical and chemical characteristics and safety must be confirmed prior to the disposal of low and intermediate level radioactive wastes, and the waste generator is required to deliver this information to the disposal facility operator. In addition, the disposal facility operator must assess the safety of the disposal site to establish the SWAC (Site Specific Waste Acceptance Criteria) in consideration of the characteristics of the site, whereas the waste generator must comply with the criteria in managing, disposing of and delivering low and intermediate level radioactive wastes. To abide by the afore-mentioned regulation and criteria, the waste generator must verify that the radioactive wastes to be disposed of are suitable for disposal before they are transported to the disposal facility, and to this end a radioactive waste certification program must be developed. This study conducted an in-depth analysis of the radioactive waste certification programs enforced in countries advanced in atomic energy to develop a draft of a certification program applicable to local power plants, and the program is currently applied as pilot to Uljin Power Plants No. 1 & 2 to prove its applicability. This study is going to analyze the results of the pilot application with a view to developing a radioactive waste certification program suitable to local conditions.

• Fire Science and Engineering
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• v.30 no.5
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• pp.74-81
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• 2016

This study evaluated the habitability of operators for main control room fires in nuclear power plants. Fire modeling (FDS v.6.0) was utilized for a fire safety assessment so that it could determine the performance of the smoke ventilation and operator habitability with the main control room. For this study, it categorized fire scenarios into three cases depending on the conditions in the HVAC system. As a result of fire modelling, it showed that Case 1 (with HVAC) would give rise to the worst situation associated with the absolute temperature, radiative heat flux, optical density, and smoke layer height as deliberating the habitability and smoke ventilation. On the other hand, it showed that Cases 2 (w/o HVAC) and 3 can maintain much safer situations than Case 1. In the case of temperature at 820 s, Cases 2 and 3 were up to approximately 63% greater than Case 1 in the wake of ignition. In addition, the influence of radiative heat flux of Case 1 was even larger than Cases 2 and 3. That is, the radiative heat fluxes of Cases 2 and 3 were approximately 68% higher than Case 1. Furthermore, when it comes to considering the optical density, Case 1 was approximately 93% greater than Cases 2 and 3. Accordingly, it expected that the HVAC system can influence a the performance on the smoke ventilation that can be sustainable for operator habitability. On the other hand, it revealed an inconsecutive pattern for the smoke layer height of Cases 2 and 3 because supply vents and exhaust vents were installed within the same surface.

• The Korean Journal of Nuclear Medicine Technology
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• v.14 no.2
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• pp.177-180
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• 2010

Purpose: As PET test came to be covered by the pay system of medical insurance (July 1, 2006) and the needs for it becoming increased for laboratory purpose, it became necessary to purchase expensive medical equipments to solve those problems. However, as most of equipments that are operated by cyclotron are very expensive as to amount from tens of millions up to hundreds of millions of won, it is difficult to purchase those equipments from the point of medical organizations. It may be possible to self manufacture those equipments with least costs if their parts functions that meets the operators demands. The Nuclear Medicine department of National Cancer Center (NCC) is trying to manufacture and use equipments that can be made with least costs, including introducing 2 medical equipments that can improves the operator's works. Materials and Methods: Example 1: Self production of radioisotope($^{18}F$) divider was fabricated. The NCC's Nuclear Medicine department acquired one acrylic panel, seven 3-way valve, tubing etc. that can be found in the market to make the main body of divider in cooperation with biomedical engineering, and placed them inside hot cell, and installed switching box outside of hot cell to make it possible to control them from outside. This main body of divider were placed in radioisotope transfer line that are manufactured in the cyclotron. Example 2: Self production of $^{18}F$-FDG automated divider was fabricated. The NCC's Nuclear Medicine department used cavro pump syringe that consists the main body of divider in cooperation with biomedical engineering, biomedical engineering developed programs that divides a certain amount. $^{18}F$-FDG automated divider is placed inside hot cell, and cable chords were used in the equipment, and then it was connected to PC outside hot cell to make it possible to control the $^{18}F$-FDG automated divider. Results: From the NCC's Nuclear Medicine department tests that were carried out from March, 2007 until now, we found out that radioisotope can be sent to radiopharmaceuticals composite module we want, and from the tests that are carried out at NCC's Nuclear Medicine department using $^{18}F$-FDG automated divider since August, 2009 it was possible to distribute radiopharmaceuticals into vial intended. Conclusion: Through the two examples above, we found out that costs can be reduced by self manufacturing expensive equipments from NCC's cyclotron room with least costs. Also, it decreased radiation exposure dose on workers, and set up problem solving processes in cooperation with lots of parties related.

• Journal of Radiation Protection and Research
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• v.40 no.3
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• pp.162-167
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• 2015

Although the perfomance indicators of the nuclear power plants in Korea show optimal, it requires detailed analysis and discussion centered on the radiation dose. As analysis methods, analysis on the radiation dose of nuclear power plants over the past five years was assessed by comparing the relevant radiation dose of radiation workers and per capita average annual radiation dose of the world's major nuclear power stations was also analyzed. The radiation workers over the annual radiation dose limit of 50 mSv were not. The contrast ratio of the radiation exposure according to the reactor type was the normal operation of PHWR was 6.2% higher than those of the PWR. This shows the radiation work of PHWR during normal driving operation is much more than those of PWR. According to the Performance Indicators of the World Association of Nuclear Operator, the annual radiation dose per unit in 2013 showed 527 man-mSv of Korea is the best country among the major nuclear power generating states, the world average was 725 man-mSv. The annual per capita radiation dose is about 80% less than 1 mSv of the public dose limit and also the average per capita dose showed a very low level as 0.82 mSv. Workers in related organizations showed 1.07 mSv, the non-destructive inspection agency workers showed 3.87 mSv. The remarkable results were due to radiation reduced program such as development of radiation shielding and radiation protection. In conclusion, the radiation exposured dose of nuclear power plants workers in Korea showed a trend which is ideally reduced. But more are expected to be difficul and the psychological insecurity against the operation of the nuclear power plants is existed to the residents near the nuclear power plants. So the radiation dose reduction policy and radiation dose follow up study of nuclear power plants will be continously excuted.

• Journal of the Korean Society of Systems Engineering
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• v.9 no.1
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• pp.33-45
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• 2013

Systems Engineering Approach was applied to the development of operator-support core management system based on the on-site operation experience and document of core management procedures, which is for enhancing operability and safety in PHWR (Pressurized Heavy Water Reactor) operation. The dissertation and definition of the system were given on th basis of investigating and analyzing the core management procedures. Fuel management, detector calibration, safety management, core power distribution monitoring, and integrated data management were defined as main user's requirements. From the requirements, 11 upper functional requirements were extracted by considering the on-site operation experience and investigating documents of core management procedures. Detailed requirements of the system which were produced by analyzing the upper functional requirements were identified by interviewing members who have responsibility of the core management procedures, which were written in SRS (Software Requirement Specification) document by using IEEE 830 template. The system was designed on the basis of the SRS and analysis in terms of nuclear engineering, and then tested by simulation using on-site data as a example. A model of core power monitoring related to the core management was suggested and a standard process for the core management was also suggested. And extraction, analysis, and documentation of the requirements were suggested as a case in terms of systems engineering.

• Journal of the Ergonomics Society of Korea
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• v.32 no.1
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• pp.9-16
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• 2013

Objective: The aim of this study is to investigate a trend of human error types observed in a series of verification and validation experiments for an Advanced Control Room(ACR) equipped with Lager Display Panel(LDP), Work Station Flat Panel Display(WS FPD), list type Alarm System(AS), Soft Control(SC) and Computerized Procedure System(CPS). Background: Operator behaviors in a fully computerized control room are quite different from those in a traditional hard-wired control room. Operators in an ACR all together monitor plant status and variables through their own interface system such as LDP and WS FPD, are notified of abnormal plant status through their own list type AS, control the plant through their own SC, and follow the structured procedure through their own CPS whereas operators in a traditional control room only separately do their duty directed by their supervisor. Especially the secondary task such as manipulating the user interface of ACR can be an extra burden to all the operators including the supervisor. Method: The Reason's human error classification method was applied to operators' behavioral data collected from a series of verification and validation experiments where operators showed their plant operational behaviors under a couple of harsh scenarios using the ACR simulator. Results: As operators accustomed to the new ACR system, knowledge or rule based mistakes appearing frequently in the early series of experiments decreased drastically in the latest stage of the series. Slip and lapse types of errors were observed throughout the series of experiments. Conclusion: Education and training can be one of the most important factors for the operators accustomed to the traditional control room to be adapted to the new system and to run the ACR successfully. Application: The results of this study implied that knowledge or rule based mistakes can be reduced by training and education but that lapse type errors might be reduced only through innovative improvement in human-system interface design or teamwork culture design including a new leadership style suitable for ACR.

• Nuclear Engineering and Technology
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• v.23 no.3
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• pp.316-320
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• 1991

The Core Operating Limit Supervisory System (COLSS) is digital computer based on-line monitoring system that is designed to assist the operator in monitoring of the Limiting Conditions for Operation. A current COLSS calculates axial power distribution based on in-core detector signals using 5th order Fourier series method. It was found that the 5th elder Fourier series method was not accurate for certain axial power shapes, especially saddle power shapes, resulting in thermal margin decrease. A cubic spline synthesis was applied to the COLSS in order to improve the axial power distribution monitoring for the various axial power shapes. The results showed that the cubic spline synthesis simulated more accurately the axial power shapes, up to 5% in RMS errors, compared to those of the Fourier series.

• Nuclear Engineering and Technology
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• v.20 no.1
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• pp.47-53
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• 1988

The objective of this study is to evaluate the reliability of the High Head Safety Injection System (HHIS) of KNU 5, 6, 7 and 8 following the removal of safety injection signal (s-signal) from the mini-flow bypass line valves of charging/safety injection pumps. The unavailability of HHSIS and the rupture probability of a charging/safety injection pump have been computed for two different cases; with s-signal on and removed. The results show that when the s-signal is removed from the mini-flow bypass line valves, the unavailability of HHSIS slightly increases while the rupture probability of a charging/safety injection pump is significantly reduced. Hence, based upon the results of this study we conclude that it is more reasonable to remove the s-signal from the mini-flow bypass line valves of KNU 5, 6, 7 and 8 in the normal plant operation. And to improve the availability of HHSIS, the modification of operational procedures and the emphasis on operator training are recommended.

• Nuclear Engineering and Technology
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• v.27 no.3
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• pp.292-300
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• 1995

This paper present a new dynamic HRA (Human Reliability Analysis) method and its application for Quantifying the human error probabilities in implementing an accident management action. For comparisons of current HRA methods with the new method, the characteristics of THERP, HCR, and SLIM-MAUD, which are most frequently used methods in PSAs, are discussed. The action associated with the implementation of the cavity flooding during a station blackout sequence is considered for its application. This method is based on the concepts of the quantified correlation between the performance requirement and performance achievement. The MAAP 3.0B code and Latin Hypercube sampling technique are used to determine the uncertainty of the performance achievement parameter. Meanwhile, the value of the performance requirement parameter is obtained from interviews. Based on these stochastic distributions obtained, human error probabilities are calculated with respect to the various means and variances of the timings. It is shown that this method is very flexible in that it can be applied to any kind of the operator actions, including the actions associated with the implementation of accident management strategies.

• Journal of the Ergonomics Society of Korea
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• v.27 no.2
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• pp.73-82
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• 2008

Currently an office in information society is advanced to a digitalized VDT office. The VDT office has a natural luminous system prioritizing visual perception and so importance of lighting design by artificial lighting is emphasized in the VDT office. The purpose of the study is to analyze Illumination and Luminance for lighting design by the standard of ergonomics in the VDT office. The study analyzed Illumination and Luminance in the main control room of an nuclear power plant needing a design of ergonomics ACR(Advanced Control Room) type Ling Ao phase II NPP MCR in Guangdong, China. The study examined the relativity of ACR characteristics to its operator's duties and set up an outline of lighting design, the standard of ergonomics and input data for the analysis. Thus, the study examined appropriateness with the standard of ergonomics setting up the analyzed Illumination and Luminance and drew a conclusion.