2상류이젝터를 이용하는 $CO_{2}$ 냉동사이클의 성능해석

이윤환,Lee Yoon-Hwan 한국마린엔지니어링학회 2005 한국마린엔지니어링학회지 Vol.29 No.8

The $CO_{2}$ refrigeration cycle is expected to reduce the compressor work and increase the COP by applying two-phase ejector as a device for the recovery of dissipated expansion energy. In this study, the performance of the cycle was simulated and effects of the ejector shapes on the performance of the $CO_{2}$ refrigeration cycle were investigated. The following results were obtained through the cycle simulation. The COP of the $CO_{2}$ refrigeration cycle with two-phase ejector flow which expansion is occured in the isentropic manner is increased by a maximum of 24 $\%$ than the basic cycle with expansion valve If the velocity nonequilibrium in the mixing process is assumed the COP of the cycle is increased with the increase of the length and the decrease of the section area of the mixing tube. The best cycle performance is obtained when the divergent angle of diffuser is 7.

국내 연구용원자로 PSA 수행을 위한 초기사건 선정 및 빈도 분석

이윤환,Lee, Yoon-Hwan 한국안전학회 2021 한국안전학회지 Vol.36 No.2

This paper presents the results of an initiating event analysis as part of a Level 1 probabilistic safety assessment (PSA) for at-power internal events for the Korea Research Reactor (KRR). The PSA methodology is widely used to quantitatively assess the safety of research reactors (RRs) in the domestic nuclear industry. Initiating event frequencies are required to conduct a PSA, and they considerably affect the PSA results. Because there is no domestic database for domestic trip events, the safety of RRs is usually assessed using foreign databases. In this paper, operating experience data from the KRR for trip events were collected and analyzed in order to determine the frequency of specific initiating events. These frequencies were calculated using two approaches according to the event characteristics and data availability: (1) based on KRR operating experience or (2) using generic data.

국내 연구용원자로 전출력 내부사건 1단계 확률론적안전성평가

이윤환,장승철,Lee, Yoon-Hwan,Jang, Seung-Cheol 한국안전학회 2021 한국안전학회지 Vol.36 No.3

This report documents the results of an at-power internal events Level 1 Probabilistic Safety Assessment (PSA) for a Korea research reactor (KRR). The aim of the study is to determine the accident sequences, construct an internal level 1 PSA model, and estimate the core damage frequency (CDF). The accident quantification is performed using the AIMS-PSA software version 1.2c along with a fault tree reliability evaluation expert (FTREX) quantification engine. The KRR PSA model is quantified using a cut-off value of 1.0E-15/yr to eliminate the non-effective minimal cut sets (MCSs). The final result indicates a point estimate of 4.55E-06/yr for the overall CDF attributable to internal initiating events in the core damage state for the KRR. Loss of Electric Power (LOEP) is the predominant contributor to the total CDF via a single initiating event (3.68E-6/yr), providing 80.9% of the CDF. The second largest contributor is the beam tube loss of coolant accident (LOCA), which accounts for 9.9% (4.49E-07/yr) of the CDF.

리스크정보 최적화를 통한 국내 연구용원자로의 안전성 향상

이윤환,장승철,Lee, Yoon-Hwan,Jang, Seung-Cheol 한국안전학회 2022 한국안전학회지 Vol.37 No.2

This paper describes an attempt to improve and optimize the operational safety level of a domestic research reactor by conducting a probabilistic safety assessment (PSA) under full-power operating conditions. The PSA was undertaken to assess the level of safety at an operating research reactor in Korea, to evaluate whether it is probabilistically safe and reliable to operate, and to obtain insights regarding the requisite procedural and design improvements for achieving safer operation. The technical objectives were to use the PSA to identify the accident sequences leading to core damage, and to conduct sensitivity analyses based thereon to derive insights regarding potential design and procedural improvements. Based on the dominant accident sequences identified by the PSA, eight types of sensitivity analysis were performed, and relevant insights for achieving safer operation were derived. When these insights were applied to the reactor design and operating procedure, the risk was found to be reduced by approximately ten times, and the safety was significantly improved. The results demonstrate that the PSA methodology is very effective for improving reactor safety in the full-power operating phase. In particular, it is a highly suitable approach for identifying the deficiencies of a reactor operating at full power, and for improving the reactor safety by overcoming those deficiencies.

Aspects of Preliminary Probabilistic Safety Assessment for a Research Reactor in the Conceptual Design Phase

이윤환,Lee, Yoon-Hwan The Korean Society of Safety 2019 한국안전학회지 Vol.34 No.3

This paper describes the work and results of the preliminary Probabilistic Safety Assessment (PSA) for a research reactor in the design phase. This preliminary PSA was undertaken to assess the level of safety for the design of a research reactor and to evaluate whether it is probabilistically safe to operate and reliable to use. The scope of the PSA described here is a Level 1 PSA which addresses the risks associated with core damage. After reviewing the documents and its conceptual design, eight typical initiating events are selected regarding internal events during the normal operation of the reactor. Simple fault tree models for the PSA are developed instead of the detailed model at this conceptual design stage. A total of 32 core damage accident sequences for an internal event analysis were identified and quantified using the AIMS-PSA. LOCA-I has a dominant contribution to the total CDF by a single initiating event. The CDF from the internal events of a research reactor is estimated to be 7.38E-07/year. The CDF for the representative initiating events is less than 1.0E-6/year even though conservative assumptions are used in reliability data. The conceptual design of the research reactor is designed to be sufficiently safe from the viewpoint of safety.

가연성 물질 저장시설의 안전성 분석

이윤환(Yoon Hwan Lee),진영호(Young Ho Jin) 大韓環境工學會 1996 대한환경공학회지 Vol.18 No.7

소화조의 수리학적 부하가 소화슬러지의 탈수성에 미치는 영향

이윤환(Yoon Hwan Lee),조광명(Kwang Myeung Cho),이면주(Myun Joo Lee) 大韓環境工學會 1996 대한환경공학회지 Vol.18 No.6

VIPEX를 이용한 가상 원자력시설의 핵심구역 파악 분석

이윤환 ( Yoon Hwan Lee ),정우식 ( Woo Sik Jung ),이진홍 ( Jin Hong Lee ) 한국안전학회(구 한국산업안전학회) 2011 한국안전학회지 Vol.26 No.4

The urgent VAI(Vital Area Identification) method development is required since ``The Act of Physical Protection and Radiological Emergency`` that is established in 2003 requires an evaluation of physical threats in nuclear facilities and an establishment of physical protection in Korea. The KAERI(Korea Atomic Energy Research Institute) has developed the VAI methodology and VAI software called as VIPEX(Vital area Identification Package EXpert) for identifying the vital areas. This study is to demonstrate the applicability of KAERI`s VAI methodology to a hypothetical facility, and to identify the importance of information of cable and piping runs when identifying the vital areas. It is necessarily needed to consider cable and piping runs to determine the accurate and realistic TEPS(Top Event Prevention Set). If the information of cable and piping runs of a nuclear power plant is not considered when determining the TEPSs, it is absolutely impossible to acquire the complete TEPSs, and the results could be distorted by missing it. The VIPEX and FTREX(Fault Tree Reliability Evaluation eXpert) properly calculate MCSs and TEPSs using the fault tree model, and provide the most cost-effective method to save the VAI and physical protection costs.

PSA기법을 이용한 원자력시설의 핵심구역 파악

이윤환 ( Yoon Hwan Lee ),정우식 ( Woo Sik Jung ),황미정 ( Mee Jeong Hwang ),양준언 ( Joon Eon Yang ) 한국안전학회 2009 한국안전학회지 Vol.24 No.5

PSA를 이용한 연구용 원자로 안전성 향상 방안 도출

이윤환 ( Yoon-hwan Lee ) 한국안전학회(구 한국산업안전학회) 2018 한국안전학회지 Vol.33 No.5

This paper describes design improvement to a research rector for safety enhancement using Probabilistic Safety Assessment (PSA). This PSA under reactor design was undertaken to assess the level of safety for the design of a research reactor and to evaluate whether it is probabilistically safe to operate and reliable to use. The scope of the PSA reported here is a Level 1 PSA, which addresses the risks associated with the core damage. The technical objectives of this study were to identify accident sequences leading to core damage and to derive design improvement from the dominant accident sequences through the sensitivity analysis. The AIMS-PSA¹⁾ and FTREX²⁾ were used for the this PSA of the research reactor. The criterion for inclusion was all sequences with a point estimate frequency greater than a truncation value of 1.0E-14/yr. The final result indicates a point estimate of 6.79E-05/yr for the overall Core Damage Frequency (CDF) attributable to internal initiating events for the research reactor under design. Based on the dominant accident sequences from the PSA, the seven kinds of sensitivity analysis were performed and some design improvement items were derived. When the five methods to improve the safety were all applied to the reactor design and emergency operating procedure, its risk was reduced to about 1.21E-06/yr from 6.79E-05/yr. The contribution of LOCA and LOEP with high CDF were significantly reduced by the sensitivity analysis. The safety of the research reactor was well improved and the risk was reduced than before adapting the design improvement gotten from the sensitivity analysis. The present study indicated that the research reactor has the well-balanced safety in regard to each initiating event contribution to CDF. The PSA methodology is very effective to improve reactor safety in a conceptual design phase and especially, Risk-informed design(RID) is very nice way to find the deficiencies of research reactor under design and to improve the reactor safety by solving them.