핵연료 설계 기술 자립 및 고성능 개량 핵연료 개발

전경락 한국원자력산업회의 2005 원자력산업 Vol.25 No.7

웨스팅하우스형 원전용 개량핵연료(ACE$^{TM}$) 개발

전경락 한국원자력산업회의 2005 원자력산업 Vol.25 No.2

실감형 Networked Virtual Environment의 사실성 증진를 위한 Non Player Character의 지능적 제어 프레임워크

전경구(Kyungkoo Jun),성미영(Meeyoung Sung),이상락(Sangrak Lee) 한국HCI학회 2006 한국HCI학회 학술대회 Vol.2006 No.2

PROGRESS IN NUCLEAR FUEL TECHNOLOGY IN KOREA

송건우,전경락,장영기,박주환,구양현 한국원자력학회 2009 Nuclear Engineering and Technology Vol.41 No.4

During the last four decades, 16 Pressurized Water Reactors (PWR) and 4 Pressurized Heavy Water Reactors (PHWR) have been constructed and operated in Korea, and nuclear fuel technology has been developed to a self-reliant state. At first, the PWR fuel design and manufacturing technology was acquired through international cooperation with a foreign partner. Then, the PWR fuel R&D by Korea Atomic Energy Research Institute (KAERI) has improved fuel technology to a self-reliant state in terms of fuel elements, which includes a new cladding material, a large-grained UO₂ pellet, a high performance spacer grid, a fuel rod performance code, and fuel assembly test facility. The MOX fuel performance analysis code was developed and validated using the in-reactor test data. MOX fuel test rods were fabricated and their irradiation test was completed by an international program. At the same time, the PWR fuel development by Korea Nuclear Fuel (KNF) has produced new fuel assemblies such as PLUS7 and ACE7. During this process, the design and test technology of fuel assemblies was developed to a self-reliant state. The PHWR fuel manufacturing technology was developed and manufacturing facility was set up by KAERI, independently from the foreign technology. Then, the advanced PHWR fuel, CANFLEX(CANDU Flexible Fuelling), was developed, and an irradiation test was completed in a PHWR. The development of the CANFLEX fuel included a new design of fuel rods and bundles.. The nuclear fuel technology in Korea has been steadily developed in many national R&D programs, and this advanced fuel technology is expected to contribute to a worldwide nuclear renaissance that can create solutions to global warming.

CONTACT FORCE MODEL FOR A BEAM WITH DISCRETELY SPACED GAP SUPPORTS AND ITS APPROXIMATED SOLUTION

박남규,서정민,전경락 한국원자력학회 2011 Nuclear Engineering and Technology Vol.43 No.5

This paper proposes an approximated contact force model to identify the nonlinear behavior of a fuel rod with gap supports;also, the numerical prediction of interfacial forces in the mechanical contact of fuel rods with gap supports is studied. The Newmark integration method requires the current status of the contact force, but the contact force is not given a priori. Taylor’s expansion can be used to predict the unknown contact force; therefore, it should be guaranteed that the first derivative of the contact force is continuous. This work proposes a continuous and differentiable contact force model with the ability to estimate the current state of the contact force. An approximated convex and differentiable potential function for the contact force is described, and a variational formulation is also provided. A numerical example that considers the particularly stiff supports has been studied, and a fuel rod with hardening supports was also examined for a realistic simulation. An approximated proper solution can be obtained using the results, and abrupt changes from the contacting state to non-contacting state, or vice versa, can be relieved. It can also be seen that not only the external force but also the developed contact force affects the response.

STRUCTURAL INTEGRITY EVALUATION OF NUCLEAR FUEL WITH REDUCED WELDING CONDITIONS

박남규,박준규,서정민,김규태,전경락 한국원자력학회 2009 Nuclear Engineering and Technology Vol.41 No.3

Welding is required for a connection between two different components in the nuclear fuel of a pressurized water reactor. This work relies on a mechanical experiment and analytic results to investigate the structural integrity of nuclear fuel in a situation where some components are not welded to each other. A series of lateral vibration tests are performed in a test facility, and the test structures are examined in terms of dynamic behavior. In the tests, the displacement signal at every grid structure that sustains fuel rods is measured and processed to identify the dynamic properties. The fluid-elastic stability of the structure is also analyzed to evaluate susceptibility to a cross flow with an assumed conservative cross flow distribution. The test and analysis results confirm that the structural integrity can be maintained even in the absence of some welding connections.

대학수학능력시험 공통과학 중 화학 영역의 문항 및 응시자 응답 분석

홍미영(Mi Young Hong),전경문(Kyung Moon Jeon),이양락(Yang Rak Lee),이범홍(Bum Hong Yi) 한국과학교육학회 2002 한국과학교육학회지 Vol.22 No.2

In this study, the students` responses on the chemistry items of in the general science of College Scholastics Ability Test (CSAT) implemented for the past 3 years since 1999 were investigated. The number of items by content and inquiry process, the average percent correct by content and inquiry process, the distribution of items by the level of percent correct, and the items with high and/or low percent correct were analysed. There were the fewest items in `environment` area, especially in `ozon layer`, no test item had been made. The most difficult content area was `acid rain` in environment. By inquiry process, the most number of items belonged to `analyzing & interpreting data` , and `identifying problems & formulating hypothesis` was the most difficult process. No test item came under the level of `very difficult`, and many items under the `easy` or `very easy` level. Students were generally poor at solving test items demanding several concepts, and very good at simply requiring basic concept treated in lower grade. Educational implications are discussed.

ANSYS를 이용한 개량핵연료 집합체 상세모델 개발에 관한 연구

전상윤(Jeon Sang-Youn),전경락(Jeon Kyeong-Lak),이영신(Lee Young-Shin) 대한기계학회 2004 대한기계학회 춘추학술대회 Vol.2004 No.10

The fuel assembly seismic and LOCA analyses are performed to verify that the stresses of fuel assembly components will be within its allowable stress and/or maintain coolable geometry so as to assure a safe shutdown of the reactor. The fuel assembly core model for seismic and LOCA analysis is accomplished by simplified spring-mass model, which is established using fuel assembly detailed model. The fuel assembly detailed model is usually established by using a fuel vendor's special computer program. In this study, the fuel assembly detailed model using a commercial finite element analysis program, ANSYS, was established to replace WECAN finite element program of Westinghouse Electric Co. for fuel assembly seismic and LOCA analysis. To verify ANSYS model, fuel assembly modal analyses were performed and the results were compared with the WECAN analysis results and the fuel assembly mechanical test results. The differences of analysis results between ANSYS and WECAN model were less than 1% and the analysis results are well coincident with the test results.

웨스팅하우스형 원전용 개량핵연료 ACE7™의 노내 연소성능

장영기(Young Kim Jang),전경락(Kyeong Lak Jeon) 대한기계학회 2010 대한기계학회 춘추학술대회 Vol.2010 No.5

Advanced nuclear fuels of ACE7™ which were developed for 16×16 and 17×17 Westinghouse type nuclear power plants completed their in-reactor verification tests in 2008 and 2009. The targets were to develop advanced fuels from the economy and safety points of view in comparison to the nuclear fuels being supplied currently. Three times of assembly-wise inspections after each cycle and rod-wise inspections after discharge using the ACE7™ Lead Test Assemblies (LTAs) were performed in Kori units 2&3. The results of the evaluation confirmed to meet their design criteria. The 16ACE7™ started to supply to Kori unit 2 from 2008 and the 17ACE7™ started to supply 17×17 Westinghouse type plants starting Kori unit 4 from 2009 commercially, respectively. The hot cell examinations on ACE7T LTA rods and skeleton are being prepared for further verification in detail. These fuels are also being examined continuously through the surveillance programs during the region implementation.

MECHANICAL AND IRRADIATION PROPERTIES OFZIRCONIUM ALLOYS IRRADIATED IN HANARO

권오현,엄경보,김재익,서정민,전경락 한국원자력학회 2011 Nuclear Engineering and Technology Vol.43 No.1

Irradiation Growth, Irradiation Test, Mechanical Property, HANARO, Performance These experimental studies are carried out to build a database for analyzing fuel performance in nuclear power plants. Inparticular, this study focuses on the mechanical and irradiation properties of three kinds of zirconium alloy (Alloy A, Alloy Band Alloy C) irradiated in the HANARO (High-flux Advanced Neutron Application Reactor), one of the leading multipurposeresearch reactors in the world. Yield strength and ultimate tensile strength were measured to determine the mechanicalproperties before and after irradiation, while irradiation growth was measured for the irradiation properties. The samples forirradiation testing are classified by texture. For the irradiation condition, all samples were wrapped into the capsule (07M-13N)and irradiated in the HANARO for about 100 days (E > 1.0 MeV, 1.11021n/cm2). These tests and results indicate that the mechanical properties of zirconium alloys are similar whether unirradiated orirradiated. Alloy B has shown the highest yield strength and tensile strength properties compared to other alloys in irradiatedcondition. Even though each of the zirconium alloys has a different alloying content, this content does not seem to affect themechanical properties under an unirradiated condition and low fluence. And all the alloys have shown the tendency toincrease in yield strength and ultimate tensile strength. Transverse specimens of each of the zirconium alloys have a slightlylower irradiation growth tendency than longitudinal specimens. However, for clear analysis of texture effects, further testingunder higher irradiation conditions is needed.