• Nuclear Engineering and Technology
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• 제38권8호
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• pp.697-732
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• 2006

In this paper, a historical review of the developments in the safety of LWR power plants is presented. The paper reviews the developments prior to the TMI-2 accident, i.e. the concept of the defense in depth, the design basis, the large LOCA technical controversies and the LWR safety research programs. The TMI-2 accident, which became a turning point in the history of the development of nuclear power is described briefly. The Chernobyl accident, which terrified the world and almost completely curtailed the development of nuclear power is also described briefly. The great international effort of research in the LWR design-base and severe accidents, which was, respectively, conducted prior to and following the TMI-2 and Chernobyl accidents is described next. We conclude that with the knowledge gained and the improvements in plant organisation/management and in the training of the staff at the presently-installed nuclear power stations, the LWR plants have achieved very high standards of safety and performance. The Generation 3+LWR power plants, next to be installed, may claim to have reached the goal of assuring the safety of the public to a very large extent. This review is based on the historical developments in LWR safety that occurred primarily in USA, however, they are valid for the rest of the Western World. This review can not do justice to the many fine contributions that have been made over the last fifty years to the cause of LWR safety. We apologize if we have not mentioned them. We also apologize for not providing references to many of the fine investigations, which have contributed towards LWR safety earning the conclusions that we describe just above.

• Journal of Gastric Cancer
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• 제15권4호
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• pp.231-237
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• 2015

Purpose: Various laparoscopic wedge resection (LWR) techniques requiring gastrotomy for gastrointestinal stromal tumors (GISTs) of the stomach have been applied to facilitate tumor resection and preserve the remnant gastric volume. However, there is the possibility of cancer cell dissemination during these procedures. The aim of this study was to assess the oncologic safety of LWR with gastrotomy (LWR-G) compared to LWR without luminal exposure. Materials and Methods: Clinicopathologic and operative results of 193 patients who underwent LWR for gastric GIST were retrospectively analyzed from 2003 to 2013. We stratified the patients into two groups: LWR-G and LWR without gastrotomy (LWR-C). Clinicopathologic features, short-term outcomes, and long-term outcomes were compared. Results: A total of 26 patients underwent LWR-G, and 167 patients underwent LWR-C. The LWR-G group showed significantly more anterior wall-located (n=10, 38.5%), intraluminal (n=20, 76.9%), and ulcerative (n=13, 50.0%) tumors than the LWR-C group (n=33, 19.8%; n=96, 57.5%; n=46, 27.5%, respectively). Postoperative short-term outcomes did not differ between the two groups. When tumor staging was compared, no statistical difference was noted. There was no recurrence in the LWR-G group, while 2 patients in the LWR-C group experienced recurrence. The two recurrences in the LWR-C group were found in the liver and in the remnant stomach at 63 and 12 months after the operation, respectively. No gastric GIST-related death was recorded in any group during the study period. Conclusions: LWR-G for gastric GIST is an oncologically safe procedure even for masses with ulcerations.

• Nuclear Engineering and Technology
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• 제39권4호
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• pp.257-272
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• 2007

Supercritical water-cooled reactors (SCWRs) are recognized as a Generation IV reactor concept. The Super LWR is a pressure-vessel type thermal spectrum SCWR with downward-flow water rods and is currently under study at the University of Tokyo. This paper reviews Super LWR safety. The fundamental requirement for the Super LWR, which has a once-through coolant cycle, is the core coolant flow rate rather than the coolant inventory. Key safety characteristics of the Super LWR inhere in the design features and have been identified through a series of safety analyses. Although loss-of-flow is the most important abnormality, fuel rod heat-up is mitigated by the "heat sink" and "water source" effects of the water rods. Response of the reactor power against pressurization events is mild due to a small change in the average coolant density and flow stagnation of the once-through coolant cycle. These mild responses against transients and also reactivity feedbacks provide good inherent safety against anticipated-transient-without-scram (ATWS) events without alternative actions. Initiation of an automatic depressurization system provides effective heat removal from the fuel rods. An "in-vessel accumulator" effect of the reactor vessel top dome enhances the fuel rod cooling. This effect enlarges the safety margin for large LOCA.

• 디지털융복합연구
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• 제13권6호
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• pp.185-195
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• 2015

본 연구는 일 대학병원의 환자안전문화 정착을 위해 리더십 워크라운드(Leadership WalkrRounds, LWR) 프로그램을 개발 적용한 후 의료직의 환자안전문화 인식정도를 비교한 것이다. 리더십 워크라운드는 미국 의료질향상 연구소(IHI)의 도구와 미시건 대학의 환자안전 라운드(Patient Safety Rounds, PSRs)를 기반으로 준비-일정계획-운영-보고-문제해결 5단계로 구성하였다. 대상자의 평균 연령은 30.9세, 방사선사가 55.2%, 간호사가 26.0%, 약무직이 18.8%였다. 중재 후 환자안전문화 인식 총점은 2.63점에서 3.36점으로 유의하게 증가하였다(p<.001). 약무직, 여자, 30대 및 근무경력 1년 이하 그룹에서 인식정도가 가장 많이 상승하였다. 환자안전문화의 모든 영역에서 인식정도가 유의하게 상승했는데(p<.001), 특히 안전사고보고영역(p<.001), 의사소통영역(p<.001)이 가장 많이 상승하였다. 새로운 형태의 리더십과 환자안전관리 개념의 융합으로 의료 질관리에 새로운 관리방안을 시도한 리더십 워크라운드는 의료인의 환자안전문화 인식 향상에 유용한 프로그램으로 활용할 수 있을 것이다.

• 대한기계학회논문집A
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• 제35권3호
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• pp.281-287
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• 2011

얇은 관 탄성좌굴 공식의 불확실성을 고려하기 위해, 공식을 구성하는 파라미터인 튜브재료의 탄성계수, 푸아송 비, 튜브 두께 및 지름의 불확실성을 분석하였다. 본 연구는 원자로에서 연소되는 핵연료봉과 같이 사용 중 함몰을 엄격히 방지하고 있는 얇은 관의 설계신뢰도를 향상시키는 데에 중요하다. 분석 방법은 각각의 파라미터가 변화할 수 있는 범위를 충분히 포함할 수 있는 최소의 탄성좌굴 안전율을 구하고 이를 선형적으로 합하여 최종의 최소안전율을 구하였다. 최소 안전율에 가장 큰 영향을 미치는 파라미터는 관의 두께로 나타났다. 두께가 얇을수록 더 큰 최소안전율이 필요하며 예로 적용한 지르코늄 합금관의 경우, 두께가 0.254 와 0.87 mm 일 때 최소안전율은 각각 1.547 과 3.487 로 나타났다.

• Nuclear Engineering and Technology
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• 제41권6호
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• pp.753-764
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• 2009

JAEA started the LSTF experiments in 1985 for the fourth stage of the ROSA Program (ROSA-IV) for the LWR thermal-hydraulic safety research to identify and investigate the thermal-hydraulic phenomena and to confirm the effectiveness of ECCS during small-break LOCAs and operational transients. The LSTF experiments are underway for the ROSA-V Program and the OECD/NEA ROSA Project that intends to resolve issues in thermal-hydraulic analyses relevant to LWR safety. Six types of the LSTF experiments have been done for both the system integral and separate-effect experiments among international members from 14 countries. Results of four experiments for the ROSA Project are briefly presented with analysis by a best-estimate (BE) code and a computational fluid dynamics (CFD) code to illustrate the capability of the LSTF and codes to simulate the thermal-hydraulic phenomena that may appear during SBLOCAs and transients. The thermal-hydraulic phenomena dealt with are coolant mixing and temperature stratification, water hammer up to high system pressure, natural circulation under high core power condition, and non-condensable gas effect during asymmetric SG depressurization as an AM action.

• Nuclear Engineering and Technology
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• 제54권12호
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• pp.4731-4742
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• 2022

RBMK fuel assemblies differ from other LWR FA due to a specific arrangement of the fuel rods, the low enrichment, and the used burnable absorber - erbium. Therefore, there is a challenge to adapt modeling tools, developed for other LWR types, to solve RBMK problems. A set of 10 different depletion simulation schemes were tested to estimate the impact on reactivity and spent fuel composition of possible SCALE code options for the neutron transport modelling and the use of different nuclear data libraries. The simulations were performed using cross-section libraries based on both, VII.0 and VII.1, versions of ENDF/B nuclear data, and assuming continuous energy and multigroup simulation modes, standard and user-defined Dancoff factor values, and employing deterministic and Monte Carlo methods. The criticality analysis with burn-up credit was performed for the SFP loaded with RBMK-1500 FA. Spent fuel compositions were taken from each of 10 performed depletion simulations. The criticality of SFP is found to be overestimated by up to 0.08% in simulation cases using user-defined Dancoff factors comparing the results obtained using the continuous energy library (VII.1 version of ENDF/B nuclear data). It was shown that such discrepancy is determined by the higher U-235 and Pu-239 isotopes concentrations calculated.

• Nuclear Engineering and Technology
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• 제53권8호
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• pp.2600-2609
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• 2021

Low cycle fatigue behaviors of SA508-3 low-alloy steel were investigated in room-temperature air, high-temperature air and in light water reactor (LWR) water environments. The fatigue mean curve and design curve for the low-alloy steel are developed based on the fatigue data in room-temperature and high-temperature air. The environmental fatigue model for low-alloy steel is developed by the environmental fatigue correction factor (F_en) methodology based on the fatigue data in LWR water environments with the consideration of effects of strain rate, temperature, and dissolved oxygen concentration on the fatigue life.

• Nuclear Engineering and Technology
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• 제45권6호
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• pp.811-820
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• 2013

SiC has been under investigation as a potential cladding for LWR fuel, due to its high melting point and drastically reduced chemical reactivity with liquid water, and steam at high temperatures. As SiC is a brittle material its behavior during the reflood phase of a Loss of Coolant Accident (LOCA) is another important aspect of SiC that must be examined as part of the feasibility assessment for its application to LWR fuel rods. In this study, an experimental assessment of thermal shock performance of a monolithic alpha phase SiC tube was conducted by quenching the material from high temperature (up to $1200^{\circ}C$) into room temperature water. Post-quenching assessment was carried out by a Scanning Electron Microscopy (SEM) image analysis to characterize fractures in the material. This paper assesses the effects of pre-existing pores on SiC cladding brittle fracture and crack development/propagation during the reflood phase. Proper extension of these guidelines to an SiC/SiC ceramic matrix composite (CMC) cladding design is discussed.

• Nuclear Engineering and Technology
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• 제48권1호
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• pp.1-15
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• 2016

For a long time, a top priority in the nuclear industry was the safe, reliable, and economic operation of light water reactors. However, the development of accident-tolerant fuel (ATF) became a hot topic in the nuclear research field after the March 2011 events at Fukushima, Japan. In Korea, innovative concepts of ATF have been developing to increase fuel safety and reliability during normal operations, operational transients, and also accident events. The microcell $UO_2$ and high-density composite pellet concepts are being developed as ATF pellets. A microcell $UO_2$ pellet is envisaged to have the enhanced retention capabilities of highly radioactive and corrosive fission products. High-density pellets are expected to be used in combination with the particular ATF cladding concepts. Two concepts-surface-modified Zr-based alloy and SiC composite material-are being developed as ATF cladding, as these innovative concepts can effectively suppress hydrogen explosions and the release of radionuclides into the environment.