• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1698-1711
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• 2022

A combined safety approach, which uses a best-estimate computer code and adopts conservative assumptions for safety systems availability, is developed and applied to the safety margin evaluation for the Loss of Condenser Vacuum (LOCV) of the 1000 MWe Korean Nuclear Power Plant. The Multi-dimensional Analysis of Reactor Safety-KINS standard (MARS-KS) code is selected as a best-estimate code and the PAPIRUS program is used to obtain different initial operational conditions through random sampling of control variables. During an LOCV event, fuel integrity is not threatened by the increase in Departure from Nuclear Boiling Ratio (DNBR). However, the high pressure in the primary coolant system and the secondary system might affect the system integrity. Thus, the peak pressure becomes a major safety concern. Transient analyses are performed for 124 cases of different initial conditions and the most conservative case, which results in the highest system pressure is selected. It is found the suggested methodology gives similar peak pressures when compared to those predicted from existing methodologies. The proposed approach is expected to minimize the time and efforts required to identify the conservative plant conditions in the existing conservative safety methodologies.

• Journal of the Korean Society of Marine Environment & Safety
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• v.16 no.1
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• pp.57-63
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• 2010

The marine safety investigation inquires and concludes the facts and causes of the marine casualties and incidents with the objective of preventing similar roses in the future. Thus, IMO and most states adopt and revise marine safety related conventions and national laws based on the results of marine safety investigation. In particular, IMO recently adopts mandatory IMO Casualty Investigation Code to identify the precise cause of rose with states' cooperation, to report the result of investigation, and to establish new international safety standard helping prevention of similar cases based on the report. The Judgement of the Korean Maritime Safety Tribunal system based on the 'Act on the investigation of and inquiry into marine accidents' is used for the purpose of marine safety investigation in Korea to prove cause of marine accident and to improve marine safety. Therefore, this study examines the Code and compares the Code with the Act to reflect the contents of provisions in the Code into the Act. The study would also be the basic references in relation to revising of marine safety investigation system in Korea Specially, the contents in relation to the independence of investigation authority and mandatory counselling system, and guarantee of seafarers human rights to ensure fairness of investigation would be included.

• Proceedings of the Korean Institute of Navigation and Port Research Conference
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• 2013.06a
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• pp.163-165
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• 2013

IMO(International Maritime Organization) is existed the movement for revising ISM Code so that the maintenance history and the trouble information given trading in a ship can be transferred. An empirical analysis was made on the influence that will have upon shipping industry through surveying on the recognition on ISM Code revision in employees of the relevant field and on the expected problems given being amended ISM Code as the above. In conclusion, the positive effect is judged to be more in the aspect of ship safety, which is the aim of ISM Code, rather than the negative effect, which may take place given being revised ISM Code. In other words, the clean market can be formed through this because fairness is maintained on both sides given trading in a ship by which opening the maintenance record and the trouble history is applied equally to a buyer and a seller. Ships can be reduced a loss of time and cost in preventing similar problems and seeking solution that may appear in important equipments, through this maintenance record. Also, based on these materials, it comes to be available for analyzing a risk of ship and preventing and managing a risk, thereby being increased ability of maintenance and repair in a ship, resulting in being judged to likely contributing to ship safety and environmental-pollution prevention.

• 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.28 no.2
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• pp.185-196
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• 1996

The purpose of this paper is to modify the ASME Code Z-Factor, which is used in the evaluation of circumferential surface crack in nuclear ferritic pipings. The ASME Code Z-Factor is a load multiplier to compensate plastic load with elasto-plastic load. The current ASME Code Z-Factor underestimates pipe maximum load. In this study, the original SC. TNP method is modified first because the original SC. TNP method has a problem that the maximum allowable load predicted from the original SC. TNP method is slightly higher than that measured from the experiment. Then the new Z-Factor is developed using the modified SC. TNP method. The desirability of both the modified SC. TNP method and the new Z-Factor is examined using the experimental results for the circumferential surface crack in pipings. The results show that (1) the modified SC. TNP method is good for predicting the circumferential surface crack behavior in pipings, and (2) the Z-Factor obtained from the modified SC. TNP method well predicts the behavior of circumferential surface crack in ferritic pipings.

• Journal of the Korean Society of Safety
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• v.28 no.5
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• pp.41-48
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• 2013

This paper is intended to provide the background information and justification for Korean highway bridge design code(limit state design)(2012). Limit state design method calculates reliability index and probability of failure through the analysis of the reliability of the experimental database. It has become possible to perform the economical and consistent design by evaluating the safety of a structure quantitatively. In this paper, we used the design specifications of RC slab bridge of superstructure form of Road Design Manual in Part 5 bridge built in highway bridge. This study conducted structural analysis using the method of frame structure theory, design and analysis of bridge by limit state design method, the design code including various standards and Load model applied Korean highway bridge design code limit state design(KHBDC;2012). As a result, it analyzed the effect of safety through comparison. Showing effect of improvement the safety factor and comparing the value of the result, it is determined to be capable of economical design and safety. Furthermore, limit state design method was able to determine many redundant force of cross-section compared with existing design method. It is determined that it can reduce the overall amount because of the reduction of the cross-section and girder depth.

• Nuclear Engineering and Technology
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• v.43 no.2
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• pp.187-194
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• 2011

Korea Nuclear Fuel is developing the X-GEN fuel which shows high performance and robust reliability for the worldwide supply. However, the simplified code systems such as CESEC-III which were developed in 1970s are still used in the current Non-LOCA safety analysis of OPR1000 and APR1400 plants. Therefore, it is essential to secure an advanced safety analysis methodology to make the best use of the merits of X-GEN fuel. To accomplish this purpose, the $\b{i}$ntegrated $\b{s}$afety $\b{a}$nalysis $\b{m}$ethodology (iSAM), is developed by selecting the best-estimate thermal-hydraulic code RETRAN. iSAM possesses remarkable advantages, such as generality, integrity, and designer-friendly features. That is, iSAM can be applied to both OPR1000 and APR1400 plants and uses only one computer code, RETRAN, in the whole scope of the non-LOCA safety analyses. Also the iSAM adopts the unique and automatic initialization and run tool, $\b{a}$utomatic $\b{s}$teady-$\b{s}$tate $\b{i}$nitialization and $\b{s}$afety analysis too l (ASSIST), to enable unhandy designers to use the new design code RETRAN without difficulty. In this paper, a brief overview of the iSAM is given, and the results of applying the iSAM to typical non-LOCA transients being checked during the reload design are reported. The typical non-LOCA transients selected are the single control element assembly withdrawal (SCEAW) accident, the asymmetric steam generator transients (ASGT), the locked rotor (LR) accident, and bank CEA withdrawal (BCEAW) event. Comparison to current licensing results shows a close resemblance; thus, it reveals that the iSAM can be applied to the non-LOCA safety analysis of OPR1000 and APR1400 plants.

• Journal of the Korean Institute of Gas
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• v.26 no.6
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• pp.52-58
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• 2022

According to the government's roadmap for revitalizing the hydrogen economy, various industries carry out projects using liquefied hydrogen as an energy source. However, safety standards necessary for operational demonstration projects are not prepared in Korea, thus, it is necessary to prepare safety standards as soon as possible. Therefore, in order to secure the safety of liquefied hydrogen instrumentation and handling facilities, it is necessary to prepare safety standards that comprehensively consider the risk of liquefied hydrogen. This study aims to prioritize safety standard items using ETA, FMEA, and AHP, which are risk assessment techniques, to present the feasibility of selecting safety standard items.

• Journal of Korean Society of Disaster and Security
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• v.6 no.3
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• pp.29-34
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• 2013

This study is to provide the implementation plan for standard code to efficiently manage curriculum information in disaster and safety education programs across domestic and foreign institutions, and to encourage active information sharing. Projects regarding disaster safety education have been progressed and developed since the opening of NDMI on March 2006. Efficient management and systematic operation for the existing disaster safety education contents are also required. It is essential for both domestic and foreign disaster management organizations to share and utilize their educational contents each other prior to the effect of the Framework Act on the Management of Disaster and Safety starting from February 7th, 2014. As disaster gets more complex and diverse in its types and sizes, the share of information on advanced disaster and safety education system between each countries is becoming more necessary than they ever did before. Therefore, the standard code of disaster safety education curriculum is resulted based on the flexibility that accommodates the change of education environment and extended applications in both domestic and foreign education institutions. The effective application of standard code will be a possible way to improve the disaster safety education system and help to set its correct definition.

• Nuclear Engineering and Technology
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• v.41 no.3
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• pp.307-318
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• 2009

KAERI (Korea Atomic Energy Research Institute) has developed the GAMMA+ code for a thermo-fluid and safety analysis of a VHTR (Very High Temperature Gas-Cooled Reactor). A key safety issue of the VHTR design is to demonstrate its inherent safety features for an automatic reactor power trip and power stabilization during an anticipated transient without scram (ATWS) accident such as a loss of forced cooling by a trip of the helium circulator (LOFC) or a reactivity insertion by a control rod withdrawal (CRW). This paper intends to show the ATWS assessment capability of the GAMMA+ code which can simulate the reactor power response by solving the point-kinetic equations with six-group delayed neutrons, by considering the reactivity changes due to the effects of a core temperature variation, xenon transients, and reactivity insertions. The present benchmark calculations are performed by using the safety demonstration experiments of the 10 MW high temperature gas cooled-test module (HTR-10) in China. The calculation results of the power response transients and the solid core temperature behavior are compared with the experimental data of a LOFC ATWS test and two CRW ATWS tests by using a 1mk-control rod and a 5mk-control rod, respectively. The GAMMA+ code predicts the power response transients very well for the LOFC and CRW ATWS tests in HTR-10.