• 비파괴검사학회지
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• 제22권1호
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• pp.59-64
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• 2002

CANDU형 핵연료채널 압력관(Zr-2.5%Nb)과 calandria관(Zircaloy-2)에 대한 원격장 와전류탐상의 자계분포특성을 파악하기 위하여 유한요소 해석을 수행하였다. 압력관과 칼란드리아관의 전자기장 분포와 위상각 해석을 통하여 최적 검사 주파수와 감지코일의 위치를 평가하였다. 또한 축대칭 구조물(Al-ring과 Al-block)이 공존시 파라미터해석을 통하여 원격장 와전류의 특성을 평가하였다.

• 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.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3379-3397
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• 2021

In this study, a multi-physics modeling method was developed to analyze a nuclear fuel rod's thermo-mechanical behavior especially for high temperature anisotropic creep deformation during ballooning and burst occurring in Loss of Coolant Accident (LOCA). Based on transient heat transfer and nonlinear mechanical analysis, the present work newly incorporated the nuclear fuel rod's special characteristics which include gap heat transfer, temperature and burnup dependent material properties, and especially for high temperature creep with material anisotropy. The proposed method was tested through various benchmark analyses and showed good agreements with analytical solutions. From the validation study with a cladding burst experiment which postulates the LOCA scenario, it was shown that the present development could predict the ballooning and burst behaviors accurately and showed the capability to predict anisotropic creep behavior during the LOCA. Moreover, in order to verify the anisotropic creep methodology proposed in this study, the comparison between modeling and experiment was made with isotropic material assumption. It was found that the present methodology with anisotropic creep could predict ballooning and burst more accurately and showed more realistic behavior of the cladding.

• Nuclear Engineering and Technology
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• 제51권6호
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• pp.1525-1531
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• 2019

The indigenous nuclear power program of India is based mainly on a series of Pressurised Heavy Water Reactors (PHWRs). A burst correlation for Indian PHWR fuel claddings has been developed and empirical burst parameters are determined. The burst correlation is developed from data available in literature for single-rod transient burst tests performed on Indian PHWR claddings in inert environment. The heating rate and internal overpressure were in the range of 7 K/s-73 K/s and 3 bar-80 bar, respectively, during the burst tests. A burst criterion for inert environment, which assumes that deformation is controlled by steady state creep, has been developed using the empirical burst parameters. The burst criterion has been validated with experimental data reported in literature and the prediction of burst parameters is in a fairly good agreement with the experimental data. The burst criterion model reveals that increasing the heating rate increases the burst temperature. However, at higher heating rates, burst strain is decreased considerably and an early rupture of the claddings without undergoing considerable ballooning is observed. It is also found that the degree of anisotropy has significant influence on the burst temperature and burst strain. With increasing degree of anisotropy, the burst temperature for claddings increases but there is a decrease in the burst strain. The effect of anisotropy in the ${\alpha}$-phase is carried over to ${\alpha}+{\beta}$-phase and its effect on the burst strain in the ${\alpha}+{\beta}$-phase too can be observed.

• Nuclear Engineering and Technology
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• 제53권3호
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• pp.741-751
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• 2021

There has been a deep interest in trying to find better-performing fuel clad motivated by the desire to decrease the likelihood of the reactor barrier failure like what happened in Fukushima in recent years. In this study, the effect of move towards accident tolerant fuel (ATF) cladding as the most attracting concept for improving reactor safety is investigated for SMART modular reactor. These reactors have less production cost, short construction time, better safety and higher power density. The SiC and FeCrAl materials are considered as the most potential candidate for ATF cladding, and the results are compared with Zircaloy cladding material from reactor physics point of view. In this paper, the calculations are performed by generating PMAX library by DRAGON lattice physics code to be used for further reactor core analysis by PARCS code. The differential and integral worth of control and safety rods, reactivity coefficient, power and temperature distributions, and boric acid concentration during the cycle are analyzed and compared from the conventional fuel cladding. The rod ejection accident (REA) is also performed to study how the power changed in response to presence of the ATF cladding in the reactor core. The key quantitative finding can be summarized as: 20 ℃ (3%) decrease in average fuel temperature, 33 pcm (3%) increase in integral rod worth and cycle length, 1.26 pcm/℃ (50%) and 1.05 pcm/℃ (16%) increase in reactivity coefficient of fuel and moderator, respectively.

• 방사성폐기물학회지
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• 제20권2호
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• pp.161-170
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• 2022

The effect of oxidation time on the characteristics and mechanical properties of spent nuclear fuel cladding was investigated using Raman spectroscopy, tube rupture test, and tensile test. As oxidation time increased, the Raman peak associated with the tetragonal zirconium oxide phase diminished and merged with the Raman peak associated with the monoclinic zirconium oxide phase near 333 cm^-1. Additionally, the other tetragonal zirconium oxide phase peak at 380 cm^-1 decreased after 100 d of oxidation, whereas the zirconium monoclinic oxide peak became the dominant peak. The oxidation time had no effect on the tube rupture pressure of the oxidized zirconium alloy tube. However, the yield and tensile stresses of the oxidized nuclear fuel cladding tube decreased after 100 d of oxidation. The results of the scanning electron microscopy and transmission electron microscopy were represented with the in-situ Raman analysis result for the oxide characteristics generated on the cladding of spent nuclear fuel.

• 방사성폐기물학회지
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• 제20권3호
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• pp.357-364
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• 2022

The safety of a KTC-360 transport cask, a large-capacity pressurized heavy-water reactor transport cask that transports CANDU spent nuclear fuel discharged from the reactor after burning in a pressurized heavy-water reactor, must be demonstrated under the normal transport and accident conditions specified under transport cask regulations. To confirm the thermal integrity of this cask under normal transport and accident conditions, high-temperature and fire tests were performed using a one-third slice model of an actual KTC-360 cask. The results revealed that the surface temperature of the cask was 62℃, indicating that such casks must be transported separately. The highest temperature of the CANDU spent nuclear fuel was predicted to be lower than the melting temperature of Zircaloy-4, which was the sheath material used. Therefore, if normal operating conditions are applied, the thermal integrity of a KTC-360 cask can be maintained under normal transport conditions. The fire test revealed that the maximum temperatures of the structural materials, stainless steel, and carbon steel were 446℃ lower than the permitted maximum temperatures, proving the thermal integrity of the cask under fire accident conditions.

• Nuclear Engineering and Technology
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• 제55권7호
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• pp.2454-2465
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• 2023

Film boiling may lead to burnout of the heating element. Even though burnout does not occur, the heating element is subject to deformation because it is not sufficiently strong to withstand external loads. In particular, the ballooning and rupture of a tube under film boiling are important phenomena in the field of nuclear reactor safety. If the tube-type cladding of nuclear fuel ruptures owing to high internal pressure and thermal load, radioactive materials inside the cladding are released to the coolant. Therefore, predicting the ballooning and rupture is important. This study presents numerical simulations to predict the ballooning behavior and rupture time of a horizontal tube at high internal pressure under saturated film boiling. To do so, a multi-step coupled simulation of conjugated film boiling heat transfer and ballooning using creep model is adopted. The numerical methods and models are validated against experimental values. Two different nonuniform heat flux distributions and four different internal pressures are considered. The three-step simulation is enough to obtain a convergent result. However, the single-step simulation also successfully predicts the rupture time. This is because the film boiling heat transfer characteristics are slightly affected by the tube geometry related to creep ballooning.

• 한국진공학회지
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• 제6권S1호
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• pp.96-109
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• 1997

In order to simulate the corrosion behavior of Zr-4 alloy in pressurized water reactors it was implanted (or bombarded) with 190ke V $Zr^+\; and \;Ar^+$ ions at liquid nitrogen temperature and room temperature respectively up to a dose of $5times10^{15} \sim 8\times10^{16} \textrm{ions/cm}^2$ The oxidation behavior and electrochemical vehavior were studied on implanted and unimplanted samples. The oxidation kinetics of the experimental samples were measured in pure oxygen at 923K and 133.3Pa. The corrosion parameters were measured by anodic polarization methods using a princeton Applied Research Model 350 corrosion measurement system. Auger Electron Spectroscopy (AES) and X-ray Photoelectric Spectroscopy (XPS) were employed to investigate the distribution and the ion valence of oxygen and zirconium ions inside the oxide films before and after implantation. it was found tat: 1) the $Zr^+$ ion implantation (or bombardment) enhanced the oxidation of Zircaloy-4 and resulted in that the oxidation weight gain of the samples at a dose of $8times10^{16}\textrm{ions/cm}^2$ was 4 times greater than that of the unimplantation ones;2) the valence of zirconium ion in the oxide films was classified as $Zr^0,Zr^+,Zr^{2+},Zr^{3+}\; and \;Zr^{4+}$ and the higher vlence of zirconium ion increased after the bombardment ; 3) the anodic passivation current density is about 2 ~ 3 times that of the unimplanted samples; 4) the implantation damage function of the effect of ion implantation on corrosion resistance of Zr-4 alloy was established.

• Nuclear Engineering and Technology
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• 제52권4호
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• pp.742-754
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• 2020

Zircaloy cladding oxidation is an important phenomenon for both design basis accident and severe accidents, because it results in cladding embrittlement and rapid fuel temperature escalation. For this reason during the last decade, many experts have been conducting experiments to identify the oxidation phenomena that occur under design basis accidents and to develop mathematical analysis models. However, since the study of design extension conditions (DEC) is relatively insufficient, it is essential to develop and validate a physical and mathematical model simulating the oxidation of the cladding material at high temperatures. In this study, the QUENCH-05 and -06 experiments were utilized to develop the best-fitted oxidation model and to validate the SPACE code modified with it under the design extension condition. It is found out that the cladding temperature and oxidation thickness predicted by the Cathcart-Pawel oxidation model at low temperature (T < 1853 K) and Urbanic-Heidrick at high temperature (T > 1853 K) were in excellent agreement with the data of the QUENCH experiments. For 'LOCA without SI' (Safety Injection) accidents, which should be considered in design extension conditions, it has been performed the evaluation of the operator action time to prevent core melting for the APR1400 plant using the modified SPACE. For the 'LBLOCA without SI' and 'SBLOCA without SI' accidents, it has been performed that sensitivity analysis for the operator action time in terms of the number of SIT (Safety Injection Tank), the recovery number of the SIP (Safety Injection Pump), and the break sizes for the SBLOCA. Also, with the extended acceptance criteria, it has been evaluated the available operator action time margin and the power margin. It is confirmed that the power can be enabled to uprate about 12% through best-estimate calculations.