핵연료 피복관용 다중층 SiC 복합체 튜브의 Hoop Stress 전산모사 연구

이현근,김대종,박지연,김원주,Lee, Hyeon-Geun,Kim, Daejong,Park, Ji Yeon,Kim, Weon-Ju 한국세라믹학회 2014 한국세라믹학회지 Vol.51 No.5

Silicon carbide-based ceramics and their composites have been studied for application to fusion and advanced fission energy systems. For fission reactors, $SiC_f$/SiC composites can be applied to core structural materials. Multilayered SiC composite fuel cladding, owing to its superior high temperature strength and low hydrogen generation under severe accident conditions, is a candidate for the replacement of zirconium alloy cladding. The SiC composite cladding has to retain its mechanical properties and original structure under the inner pressure caused by fission products; as such it can be applied as a cladding in fission reactor. A hoop strength test using an expandable polyurethane plug was designed in order to evaluate the mechanical properties of the fuel cladding. In this paper, a hoop strength test of the multilayered SiC composite tube for nuclear fuel cladding was simulated using FEA. The stress caused by the plug was distributed nonuniformly because of the friction coefficient difference between the inner surface of the tube and the plug. Hoop stress and shear stress at the tube was evaluated and the relationship between the concentrated stress at the inner layer of the tube and the fracture behavior of the tube was investigated.

Experimental and numerical investigation on thermo-mechanical behavior of fuel rod under simulated LOCA conditions

Yadav, Ashwini Kumar,Shin, Chang Hwan,Lee, Sung Uk,Kim, Hyo Chan Elsevier 2018 Nuclear engineering and design Vol.337 No.-

<P><B>Abstract</B></P> <P>The data from out-of-pile single rod experiments can provide vital information about transient temperature variation and deformation rate of the cladding which can be extensively used to assess the flow blockage under Loss of Coolant Accidents (LOCA). With this motivation, in the present investigation experiments were conducted at low heating rates (2–8 K/s) to simulate a LOCA scenario. The un-irradiated Zircaloy-4 clad tube was internally heated using a tungsten heater with alumina pellets as a fuel rod simulator in an inert atmosphere. The ballooning initiation led to a significant decrease in temperature rise rate owing to increase in gap width between pellet and cladding. A maximum hoop strain of 93% at 1080 K in the α-phase and minimum hoop strain of 28% at 1340 K in the β-phase was observed. The Azimuthal Temperature Difference (ATD) over the clad tube was below 18 K for all tests and therefore ballooning was essentially symmetrical. Based on the experimental results, a burst criterion has been developed to predict the burst by a code named ‘TRAFR’ (Transient Response Analysis of Fuel Rod). The predictions based on plasticity and creep deformation models were compared with experimental results. At high temperature, the PLASTIC model predicted too low burst strain due to hardening effect. The burst strain predictions by the CREEP model were in good agreement with the experimental results in all temperature zones. The time-dependent behavior of creep phenomena led to a gradual rise in hoop stress until burst. The predictions with CREEP model can be improved by optimizing stress exponent value at high temperature in β-phase where superplastic behavior of Zircaloy-4 is governed by dislocation climb induced creep. However creep behavior is more complex and as such no simple law is established.</P> <P><B>Highlights</B></P> <P> <UL> <LI> A burst criterion for the Zircaloy-4 clad tube at low heating rates (2–8 K/s) developed using fuel simulator. </LI> <LI> Cladding temperature, fill pressure and radial deformation were recorded for code validation. </LI> <LI> A code was developed using finite difference method and plastic/creep equations. </LI> <LI> The predictions by CREEP model were better due to gradual rise in hoop stress until burst. </LI> </UL> </P>

Numerical modeling of fuel rod transient response under out of pile test conditions

Yadav, Ashwini Kumar,Shin, Chang-Hwan,Lee, Chan,Lee, Sung-Uk,Kim, Hyo Chan Elsevier 2019 Progress in nuclear energy Vol.113 No.-

<P><B>Abstract</B></P> <P>As per revised Emergency core cooling system (ECCS) acceptance criteria, a precise prediction of fuel rod behavior is essential for realistic safety analysis of nuclear reactor. In this context, a one-dimensional code name ‘TRAFR’ (Transient Response Analysis of Fuel Rod) is developed to simulate the thermo-mechanical behavior of Zircaloy-4 cladding under transient conditions. The transient simulations for inert and oxidizing atmosphere were performed under out of pile test conditions and the predicted burst strains were in good agreement with the experiments conducted in past. Under inert atmosphere, the cladding rupture was delayed and burst strain was higher in all the phases due to the absence of oxidation kinetics. In the oxidizing atmosphere, the burst strain was considerably small at high temperature in mix phase (α+β) and β-phase due to increment in cladding strength and reduction in ductility. For the same internal pressure and clad surface boundary conditions, the temperature of failure was higher for oxidizing atmosphere due to heat generation by exothermic reactions at the surface of the cladding. The clad surface temperature rise rate decreased before the burst with increment in gap width between pellet and cladding owing to decrement in gap conductance. The code under-predicted the burst strain in the mix phase (α+β) and β-phase by ‘Baker-Just’ model. The reason for such deviation was abrupt parabolic oxide growth prediction with higher exothermic heat generation and subsequent faster reduction in cladding thickness which made ‘Baker-Just’ model approach more conservative than ‘Cathcart-Pawel’ model.</P> <P><B>Highlights</B></P> <P> <UL> <LI> A fully-coupled 1-D code named ‘TRAFR’ is developed to access thermo-mechanical behavior of cladding. </LI> <LI> The effect of oxidation was significant at high temperature resulting into lower burst strains. </LI> <LI> The conservative approach of ‘Baker-Just’ model led to abrupt parabolic oxide growth prediction. </LI> <LI> In the absence of oxidation the cladding was able to sustain against internal stresses for longer duration. </LI> </UL> </P>

PVD-Be와 비정질 Zr-Be 합금을 용가재로 사용한 Zircaloy-4의 브레이징 접합부의 비교 연구

황용화(Hwang, Yong-Hwa),김재용(Kim, Jae-Yong),이형권(Lee, Hyung-Kwon),고진현(Koh, Jin-Hyun),오세용(Oh, Se-Yong) 한국산학기술학회 2006 한국산학기술학회논문지 Vol.7 No.2

중수로형 핵연료 제조공정 중 연료봉 피복관에 간격체와 지지체 등의 부착물이 브레이징으로 접합된다. 본 연구에서는 베릴륨을 물리 증착법(PVD)으로 접합될 부착물의 표변에 증착한 것과 비정질 용가재[Zr₁-xBex(0.3≤x≤0.5)]를 사용하여 브레이징된 접합부의 미세조직과 경도 등의 특성을 비교하고 브레이장 온도가 접합부에 미치는 영향 조사하였다. 비정질 용가재에 의한 정합층의 두께는 PVD-Be의 경우와 비교하여 더 얇았고, Be 함량이 감소할수록 접합층의 두께는 감소하였으며 모재의 침식은 거의 없었다. PVD-Be의 경우 공정 반응, 액상 출현, 모세관 현상과 확산으로 브레이징 되나 비정질 합금은 용가재 만이 용융되어 액상 접합되는 것으로 사료된다. PVD-Be 접합부의 미세 조직은 계변에서 수지상이 형성되어 내부로 성장하나, 비정질 합금에 의한 접합부는 석출된 제2상들이 구상으로 구성 되며 브레이징 온도가 증가할수록 구상은 더욱 커졌다. 비정질 합금 접합부의 경도는 Be 함량이 감소할수록 경도는 증가하였다. 본 연구에 사용된 비정질 합금 중 합금은 접합부에서 Be의 모재로의 확산이 적어 부드러운 계면과 모재의 침식이 없었고 높은 경도 때문에 핵연료 피복재 접합에 가장 적합한 용가재로 사료된다. Brazing is an important manufacturing process in the fabrication of Heavy Water Reactor fuel rods, in which bearing and spacer pads are joined to Zircaloy-4 cladding tubes. The physical vapor deposition(PVD) technique is currently used to deposit metallic Be on the surfaces of pads as a filler metal. Amorphous Zr-Be binary alloys which are manufactured by rapid solidification process are under developing to substitute the conventional PVD-Be coating. In the present study, brazed joint with PVD and amorphous alloys of Zr₁-xBe, (0.3≤;x≤0.5) as filler metals are compared by mechanism, microstructure and hardness. The thickness of brazed joint with amorphous alloys became much smaller than that of PVD-Be. The erosion of base metal did not occur in the brazed joint with amorphous alloys. The brazing mechanism for PVD-Be seems to be Be diffusion into Zr-4 with capillary action resulting from eutectic reaction while that for amorphous alloys are associated with the liquid phase formation in the brazed joint. The brazed joint microstructure with PVD-Be consists of dendrite while that with amorphous alloys is globular. The Zr0.7Beo.3 alloy shows the smooth interface with little erosion in the base metal and is recommended a most suitable brazing filler metal for Zircaloy-4.

피복관 재료와 지지격자 형상에 따른 핵연료봉 피복관의 마멸거동에 관한 연구

김대정(Dae-Jung Kim),용석주(Suk-Ju Yong),정성훈(Sung-Hoon Jeong),이영제(Young-Ze Lee),김일규(Il-Kyu Kim) 한국트라이볼로지학회 2003 한국트라이볼로지학회 학술대회 Vol.36 No.-

Fretting is the small-amplitude oscillatory motion, usually tangential, between two solid surfaces in contact. fretting wear is wear arising as a result of fretting. In nuclear power plant, high flow rates can induce vibration of the nuclear fuel rod cladding tubes resulting in fretting wear damage due to contacts between the tubes and their supports. In this paper the sliding wear' tests and fretting wear tests were performed in water at room temperature. Sliding tests with the pin-on-disk type tribometer were done using nuclear fuel rod cladding tube materials of Zirconium and Zircaloy-4 alloy series against STS 304 under various applied loads. Fretting tests were done using some different tube materials and supporting grid shapes under various applied loads. From the results of sliding and fretting wear tests, the wear of tube materials can be predictable by obtaining wear coefficient using the work rate model. Depending on various normal load, tube materials, and supporting grid shapes, distinctively different wear scars of fretting and stick-slip mechanism can occur.

급속응고된 비정질 Zr-Be 합금 용가재를 이용한 Zircaloy-4의 브레이징 특성

김상호,고진현,박춘호,김성규,Kim, Sang-Ho,Go, Jin-Hyeon,Park, Chun-Ho,Kim, Seong-Gyu 한국재료학회 2002 한국재료학회지 Vol.12 No.2

This study was conducted to investigate the brazing characteristics between Zircaloy-4 nuclear fuel cladding tubes and bearing pads with filler metals of amorphous $Zr_{1-x}Be_x$(0.3$\leq$x$\leq$0.5) binary alloy, in which they were produced in the ribbon form by the melt-spinning metod. The crystallization behavior, stability, hardness and micro-structure of brazed zone were examined by X-ray diffraction, differential scanning calorimetry, micro-Vickers hardness test, optical microscopy, and transmission electron microscopy. $Zr_{1-x}Be_x$(0.3$\leq$x$\leq$0.4) amorphous alloys were crystallized to $\alpha$-Zr with increasing the temperature, and the rest were transformed to ZrBe$_2$at higher temperatures. On the other hand, $Zr_{1-x}Be_x$(0.4$\leq$x$\leq$0.5) amorphous alloys were crystallized to $\alpha$-Zr and ZrBe$_2$, simultaneously. The thickness of the layer brazed with amorphous alloy was increased with increasing the beryllium content due to the higher diffusion of Be. The morphology of brazed layer with PVD Be filler metal showed dendrite while that brazed with amorphous alloys appeared globular. Micro-Vickers hardness of brazed zone increased as the beryllium content of filler metal was decreased.

부식막이 핵연료 피복관의 프레팅 마멸에 미치는 영향

김진선(Jinseon Kim),이영제(YoungZe Lee) 한국트라이볼로지학회 2009 한국트라이볼로지학회 학술대회 Vol.2009 No.6

The nuclear fuel cladding tubes which are a key component in nuclear power plant are damaged by the fretting wear caused by FIVs. Those are also interfered with severe environments such as high temperature, high pressure, radioactivity etc. Therefore the fretting wear tests were conducted using the nuclear cladding tube treated by the corrosion under radioactivity environment. As the load increased, the corrosive layer was fractured quickly. In case that corroded layers were existed, the maximum friction forces and wear depths were relatively lower.

지르칼로이-4 브레이징용 비정질 Ti-Be 용가재의 결정화 거동 및 접합부 미세조직

김상호,고진현,박춘호,Kim, Sang-Ho,Go, Jin-Hyeon,Park, Chun-Ho 한국재료학회 2002 한국재료학회지 Vol.12 No.4

Three different ribbons of amorphous $Til_{1-x}Be_x$ alloys such as $Ti_{0.59}Be_{0.41},\;Ti_{0.61}Be_{0.39}\;and\;Ti_{0.63}Be_{0.37}$ were made by melt-spinning method to be used as brazing filler metals for joining Zircaloy-4 nuclear fuel cladding tubes, and their crystallization behavior as well as microstructure of the brazed zone were examined. The crystallization behavior was investigated in teams of thermal stability, crystallization temperature and activation energy. The crystallization of the $Ti_{1-x}Be_x$ alloys proceeded in two steps by the formation of ${\alpha}$</TED>-Ti at a lower temperature and of TiBe at a higher temperature. The crystallization temperature and activation energy of $Ti_{1-x}Be_x$ alloys were higher and larger than those of $Zr_{1-x}Be_x$ alloys and PVD Be. Those resulted thinner joining layer with $Ti_{1-x}Be_x$ alloys, which kept sound thickness of Zircaloy-4 nuclear fuel cladding tubes after brazing. But in the brazed zones made by $Ti_{1-x}Be_x$ filler metals, a little solid-solution layers composed of Zr and Ti were formed toward the Zr cladding tube and Zr was detected in the brazed zones. Microstructure of brazed zone was changed from globular to dentrite with decreasing Be content in the $Ti_{1-x}Be_x$ filler metal.

핵연료 피복재 튜브의 원격장와전류 탐상을 위한 차폐된 관통형 탐촉자의 수치해석적 설계

신영길,신상호 한국비파괴검사학회 2001 한국비파괴검사학회지 Vol.21 No.6

본 논문에서는 핵연료 피복재 튜브를 검사하기 위한 차폐된 관통형 원격장와전류 탐촉자의 설계과정을 설명하고, 이 탐촉자에 의한 결함신호의 특성을 조사하였다. 먼저, 자기 에너지가 튜브 내부로 관통될 수 있도록 여자코일 외부를 전기적으로 절연된 얇은 철 박판을 적층시켜 차폐시켰다. 그리고 유한요소 해석을 통하여 차폐의 효과와 탐상주파수를 연구하였으며, 센서코일의 위치를 결정하였다. 그러나 이렇게 설계된 탐촉자를 사용하여 예측된 결함신호는 센서코일이 결함을 지날 때의 결함지시가 명확하지 않았으며, 여자코일이 결함을 지날 때의 결함지시도 차폐체로부터의 영향이 나타나는 등 여자코일로부터 자속이 직접적으로 센서코일에 영향을 미친다는 사실을 알게 되었다. 따라서 센서코일도 여자코일과 같은 형태로 차폐시켰는데 이 차폐의 효과는 놀라울 정도로 결함신호의 특성을 향상시켰다. 최종적으로 설계된 탐촉자를 사용하여 수치 모델링을 수행한 결과는 관내삽입 원격장와전류 탐촉자를 사용하였을 때의 신호와 매우 흡사한 신호특성을 보였다. 즉, 위상신호는 내부결함과 외부결함에 대하여 거의 동일한 민감도를 보였으며, 위상신호의 세기와 결함의 깊이 사이에 선형적인 관계가 있음이 관찰되었다. This paper explains the process of designing a shielded encircling remote field eddy current (RFEC) probe to inspect nuclear fuel cladding tubes and investigates resulting signal characteristics. To force electromagnetic energy from exciter coil to penetrate into the tube, exciter coil is shielded outside by laminations of iron insulated electrically from each other. Effects of shielding and the proper operating frequency are studies by the finite element analysis and the location for sensor coil is decided. However, numerically simulated signals using the designed probe do not clearly show the defect indication when the sensor passes a defect and the other indication appeared as the exciter passes the defect is affected by the shape of shielding structure, which demonstrates that the sensor is directly affected by exciter fields. For this reason, the sensor is also shielded outside and this shielding dramatically improves signal characteristics. Numerical modeling with the finally designed probe shows very similar signal characteristics to those of inner diameter RFEC probe. That is, phase signals show almost equal sensitivity to inner diameter and outer diameter defects and the linear relationship between phase signal strength and defect depth is observed.

핵연료 피복관의 충격-프레팅 마모 기구에 미치는 주파수 영향

이영호(Young-Ho Lee),김형규(Hyung-Kyu Kim),김현길(Hyun-Gil Kim) 대한기계학회 2021 대한기계학회 춘추학술대회 Vol.2021 No.4

가동 중 원전 구조물에서 빠른 냉각수의 흐름으로 발생하는 관 구조물의 진동(즉, 유체유발진동)은 상대적으로 얇고 긴 구조물을 가진 시켜 이를 지지하는 구조물과 상호작용이 발생, 최종적으로 접촉부의 마모 손상을 일으킨다. 또한 구조물내 손상 및 유지보수과정에서 잔류할 수 있는 이물질 또한 마모를 발생시키는 중요한 원인으로 알려져 있다. 현재까지 핵연료 피복관, 증기발생기 전열관 및 제어봉 등 각각의 관 구조물의 접촉부에서 발생하는 마모 손상 평가를 위한 실험적 연구가 진행중이다. 그러나 두 물체의 접촉조건은 일정한 하중이 계속 작용하는 경우와 간격에 의한 불규칙적인 하중으로 구분할 수 있으나 후자의 경우 실험적 연구가 많지 않다. 본 연구에서는 핵연료 피복관이 축방향으로 가진 될 때 3N 이하의 낮은 충격하중을 측면에 적용시켜 충격 주파수에 따른 마모량의 변화를 관찰하였다. 이를 위해 여러가지 진폭 및 주파수를 발생시킬 수 있는 충격-프레팅 마모시험 장치를 제작하였고 간격 및 주파수에 따른 마모량 변화를 분석하였다. 시험결과로부터 충격 주파수가 증가함에 따라 피복관의 마모량은 급격히 증가하였으며, 충격 하중에 따른 전단 하중의 크기는 지지격자 스프링의 강성과 주파수에 매우 민감한 경향을 보였다. 충격 하중조건에서 핵연료 피복관 손상 영역은 기존 미끄럼 혹은 프레팅 마모와는 달리 국부적인 표면 소성 변형층의 파괴로 인한 연삭 혹은 흡착 마모기구가 급격히 진행됨을 확인하였다. 특히 충격에 의한 접촉 발생 시 피복관 및 지지격자의 인가 주파수 차이에 따라 마모량이 현저히 증가함을 관찰하였다. 이상의 결과를 바탕으로 세부적인 마모기구를 분석하였으며, 본 연구에 적용된 시험방법을 이물질 마모를 포함한 증기발생기 전열관 및 제어봉으로 확장하여 연구를 수행 중이다.