• 기술사
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• 제30권5호
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• pp.82-95
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• 1997

A Study of Improvement in Reactor Thermal Power Measurement Method using Kalman Filter. The objectives of the safety analysis of nuclear power plants are to maintain the surface temperature of fuel and fuel cladding within limit value in case of Loss of Coolant accident (LOCA) so that it ensures the safety and reliability of nuclear power plants. The new technique evaluating the reactor power and improvement of existing plant system increase the safety margin of nuclear power plant operation, and accordingly, economic effect will be anticipated. Hereby, 1 would like to introduce reactor power measurement method using Kalman filter that enables to calculate the reactor power more precisely combining the parameters, for example, turbine output as the 1 st stage pressure of high pressure turbine, and reactor power using energy equilibrium relation. It is expected that the new technique will enhance the accuracy of measurement of reactor power and maintain the reliability of nuclear power operation by increasing operational safety margin, and gain the economic benefit

• Nuclear Engineering and Technology
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• 제37권4호
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• pp.309-316
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• 2005

The contents of a new book currently in preparation are described. The dearth of books in the field of nuclear materials has left both students in nuclear materials classes and professionals in the same field without a resource for the broad fundamentals of this important sub-discipline of nuclear engineering. The new book is devoted entirely to materials problems in the core of light-water reactors, from the pressure vessel into the fuel. Key topics deal with the $UO_2$ fuel, Zircaloy cladding, stainless steel, and of course, water. The restriction to LWR materials does not mean a short monograph; the enormous quantity of experimental and theoretical work over the past 50 years on these materials presents a challenge of culling the most important features and explaining them in the simplest quantitative fashion. Moreover, LWRs will probably be the sole instrument of the return of nuclear energy in electric power production for the next decade or so. By that time, a new book will be needed.

• Nuclear Engineering and Technology
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• 제56권2호
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• pp.728-744
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• 2024

The presence of hydrogen absorbed by zirconium-based cladding materials during reactor operation can trigger degradation mechanisms and endanger the rod integrity. Ensuring the durability of the rods in extended time-frames like dry storage requires anticipating hydrogen behavior using numerical modeling. In this context, the present paper describes a hydrogen post-processing tool for Falcon - HYPE, a PSI's in-house tool able to calculate hydrogen uptake, transport, thermochemistry, reorientation of hydrides and hydrogen-related failure criteria. The tool extracts all necessary data from a Falcon output file; therefore, it can be considered loosely coupled to Falcon. HYPE has been successfully validated against experimental data and applied to reactor operation and interim storage scenarios to present its capabilities.

• Nuclear Engineering and Technology
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• 제42권5호
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• pp.576-581
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• 2010

KAERI (Korea Atomic Energy Research Institute) fabricated MOX (Mixed Oxide) fuel pellets as a cooperation project with PSI (Paul Scherrer Institut) for an irradiation test in the Halden reactor. The MOX pellets were fitted into fuel rods that included instrumentation for measurement in IFE (Institutt for Energiteknikk). The fuel rods were assembled into the test rig and irradiated in the Halden reactor up to 50 MWd/kgHM. The irradiated fuel rods were transported to the IFE, where ceramography was carried out. The fuel rods were cut transversely at the relatively higher burn-up locations and then the radial cross sections were observed. Micrographs were analyzed using an image analysis program and grain sizes along the radial direction were measured by the linear intercept method. Radial cracks in the irradiated MOX were observed that were generally circumferentially closed at the pellet periphery and open in the hot central region. A circumferential crack was formed along the boundary between the dark central and the outer regions. The inner surface of the cladding was covered with an oxide layer. Pu-rich spots were observed in the outer region of the fuel pellets. The spots were surrounded by many small pores and contained some big pores inside. Metallic fission product precipitates were observed mainly in the central region and in the inside of the Pu spots. The average areal fractions of the metallic precipitates at the radial cross section were 0.41% for rod 6 and 0.32% for rod 3. In the periphery, pore density smaller than 2 ${\mu}m$ was higher than that of the other regions. The grain growth occurred from 10 ${\mu}m$ to 12 ${\mu}m$ in the central region of rod 6 during irradiation.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.973-979
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• 2024

In the framework of the Generation IV research and development project, in which the French Commission of Alternative and Atomic Energies (CEA) is involved, a main objective for the design of Sodium-cooled Fast Reactor (SFR) is to meet the safety goals for severe accidents. Among the severe ones, the Unprotected Transient OverPower (UTOP) accidents can lead very quickly to a global melting of the core. UTOP accidents can be considered either as slow during a Control Rod Withdrawal (CRW) or as fast. The paper focuses on fast UTOP accidents, which occur in a few milliseconds, and three different scenarios are considered: rupture of the core support plate, uncontrolled passage of a gas bubble inside the core and core mechanical distortion such as a core flowering/compaction during an earthquake. Several levels and rates of reactivity insertions are also considered and the thermal-mechanical behavior of an ASTRID fuel pin from the ASTRID CFV core is simulated with the GERMINAL code. Two types of fuel pins are simulated, inner and outer core pins, and three different burn-up are considered. Moreover, the feedback from the CABRI programs on these type of transients is used in order to evaluate the failure mechanism in terms of kinetics of energy injection and fuel melting. The CABRI experiments complete the analysis made with GERMINAL calculations and have shown that three dominant mechanisms can be considered as responsible for pin failure or onset of pin degradation during ULOF/UTOP accident: molten cavity pressure loading, fuel-cladding mechanical interaction (FCMI) and fuel break-up. The study is one of the first step in fast UTOP accidents modelling with GERMINAL and it has shown that the code can already succeed in modelling these type of scenarios up to the sodium boiling point. The modeling of the radial propagation of the melting front, validated by comparison with CABRI tests, is already very efficient.

• Nuclear Engineering and Technology
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• 제20권3호
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• pp.170-178
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• 1988

중수로형 핵연료의 피복재로 사용되는 Zircaloy-4관의 결함검사를 위한 자동초음파 탐상 장치가 개발되었다. 이 장치에는 중심진동수가 14 MHz이고 대역폭이 11MHz인 집속 초음파 펄스를 사용한 수침 펄스-에코우 탐상기술과 특별히 고안된 시험수조 이송식 초음파주사 기술이 적용되었다 같은 크기와 방향을 갖는 관내외면 결함들을 같은 높이의 초음파 신호로 검출하기 위한 초음파 빔의 최적입사각은 26도이었다. Zircaloy-4피복관의 최대 허용 결함인, 깊이가 관두께의 10%인 0.04 mm이고, 길이가 0.76 mm인 축방향 및 길이가 0.38 mm인 원주방향 V형 인공결함들이 관내외면에 개재된 표준시험관을 사용하여 이 장치의 성능시험을 수행하였다. 그 결과 인공 표준시험관내의 모든 결함들을 매우 우수한 재현성을 갖고 분당 약 1m의 속도로 검출할 수 있었으며 이때의 신호 대 잡음 비는 축방향 결함에 대해서는 20 dB, 원주방향 결함에 대해서는 12 dB 이상이었다.

• Nuclear Engineering and Technology
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• 제54권12호
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• pp.4798-4808
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• 2022

The porous corrosion deposits (known as CRUD) adhered to the cladding have an important effect on the heat transfer from fuel rods to coolant in PWRs. The vapor film is the main constituent in the two-phase film boiling model. This paper presents a vapor film thickness correlation, associated with CRUD porosity, CRUD chimney density, CRUD particle size, CRUD thickness and heat flux. The dependences of the vapor film thickness on the various influential factors can be intuitively reflected from this vapor film thickness correlation. The temperature, pressure, and boric acid concentration distributions in CRUD can be well predicted using the two-phase film boiling model coupled with the vapor film thickness correlation. It suggests that the vapor thickness correlation can estimate the vapor film thickness more conveniently than the previously reported vapor thickness calculation methods.

• 한국세라믹학회지
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• 제50권6호
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• pp.359-363
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• 2013

Silicon carbide and its composites are being considered as a nuclear fuel cladding material for LWR nuclear reactors because they have a low neutron absorption cross section, low hydrogen production under accident conditions, and high strength at high temperatures. The SiC composite cladding tube considered in this study consists of three layers, monolith CVD SiC - $SiC_f$/SiC composite -monolith CVD SiC. The volume fraction of SiC fiber and surface roughness of the composite layer affect mechanical and corrosion properties of the cladding tube. In this study, various types of SiC fiber preforms with tubular shapes were fabricated by a filament winding method using two types of Tyranno SA3 grade SiC fibers with 800 filaments/yarn and 1600 filaments/yarn. After chemical vapor infiltration of the SiC matrix, the surface roughness and fiber volume fraction were measured. As filament counts were changed from 800 to 1600, the surface roughness increased but the fiber volume fraction decreased. The $SiC_f$/SiC composite with a bamboo-like winding pattern has a smaller surface roughness and a higher fiber volume fraction than that with a zigzag winding pattern.

• Nuclear Engineering and Technology
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• 제55권11호
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• pp.4066-4076
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• 2023

Pressure resistance welding is usually used to seal the connection between the cladding tube and the end plug made of zirconium alloy. The seal welded joint has a direct effect on the service performance of the fuel rod cladding structure. In this paper, the pressure resistance welded joints of zirconium alloy tube-plug structure were obtained by thermal-mechanical simulation experiments. The microstructure and microhardness of the joints were both analyzed. The effect of processing parameters on the microstructure was studied in detail. The results showed that there was no β-Zr phase observed in the joint, and no obvious element segregation. There were different types of Widmanstätten structure in the thermo-mechanically affected zone (TMAZ) and heat affected zone (HAZ) of the cladding tube and the end plug joint because of the low cooling rate. Some part of the grains in the joint grew up due to overheating. Its size was about 2.8 times that of the base metal grains. Due to the high dislocation density and texture evolution, the microhardnesses of TMAZ and HAZ were both significantly higher than that of the base metal, and the microhardness of the TMAZ was the highest. With the increasing of welding temperature, the proportion of recrystallization in TMAZ decreased, which was caused by the increasing of strain rate and dislocation annihilation.

• Nuclear Engineering and Technology
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• 제18권4호
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• pp.267-272
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• 1986

핵연료 피복관 제조 및 사용 시에 필요한 자료를 얻기 위하여 지르칼로이-4재료에서 가공과 열처리의 영향을 조사하였다. 지르칼로이-4 재료는 저가공도에서는 경도가 급격히 증가하지만 10% 이상 가공도 에서는 점진적으로 증가하였다. 냉간가공된재료의 재결정은 가공도가 30%, 60%, 80%로 증가함에 따라서 64$0^{\circ}C$, 59$0^{\circ}C$, 555$^{\circ}C$에서 각각 완료되었다. $\beta$구역에서 열처리한후에 노냉, 공냉, 수냉을하였을 때 냉각속도가 증가함에 따라서 경도는 증가하고, 조직은 coarse widmanstatten($\alpha$) $\longrightarrow$ fine parallel plate($\alpha$) $\longrightarrow$ martensite($\alpha$$^{'}$)순으로 변화한다. 변화한다.