• Nuclear Engineering and Technology
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• 제55권7호
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• pp.2578-2590
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• 2023

In this work, a 24-month two-batch fuel management strategy for the APR1400 using LEU + has been investigated, where enrichments of 5.9 and 5.2 w/o are utilized in lieu of the conventional 4-5 w/o UO₂ fuel. In addition, an Accident Tolerant Fuel (ATF) clad based on the swaging technology is applied to APR1400 fuel assemblies. In this special ATF clad design, both outer and inner SS316 layers protect the conventional zircaloy clad. Erbia (Er₂O₃) is introduced as a burnable absorber with two-fold goals to lower the critical boron concentration in the long-cycle LEU + loaded core as well as to handle the LEU + fuel in the existing front-end fuel facilities without renewing the license. Two types of fuel assemblies with different loading of gadolinia (Gd₂O₃) are considered to control both the reactivity and the core radial power distribution. The erbia burnable absorber is uniformly admixed with UO₂ in all fuel pins except for the gadolinia-bearing ones. In this study, two core designs were devised with different erbia loading, and core performance and safety parameters were evaluated for each case in comparison with a core design without any burnable absorbers. The core analysis was done using the two-step method. First, cross-sections are generated by the SERPENT 2 Monte Carlo code, and the 3-D neutronic analysis is performed with an in-house multi-physics nodal code KANT.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.846-854
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• 2024

FEMAXI-ATF is being developed for fuel performance modeling of SiC cladded UO2 fuel with focuses on modeling pellet-cladding mechanical interactions (PCMI). The code considers probability distributions of mechanical strengths of monolithic SiC (mSiC) and SiC fiber reinforced SiC matrix composite (SiC/SiC), while it models pseudo-ductility of SiC/SiC and propagation of cladding failures across the wall thickness direction in deterministic manner without explicitly modeling cracks based on finite element method in one-dimensional geometry. Some hypothetical BWR power ramp conditions were used to test sensitivities of different model parameters on the analyzed PCMI behavior. The results showed that propagation of the cladding failure could be modeled by appropriately reducing modulus of elasticities of the failed wall element, so that the mechanical load of the failed element could be re-distributed to other intact elements. The probability threshold for determination of the wall element failure did not have large influence on the predicted power at failure when the threshold was varied between 25 % and 75 %. The current study is still limited with respect to mechanistic modeling of SiC failure as it only models the propagation of the cladding wall element failure across the homogeneous continuum wall without considering generations and propagations of cracks.

• Nuclear Engineering and Technology
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• 제53권4호
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• pp.1236-1249
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• 2021

Although there are still controversial opinions and uncertainty on application of SiC_f/SiC composite cladding as next-generation cladding material for its great oxidation resistance in high temperature steam environment and other outstanding advantages, it cannot deny that SiC_f/SiC cladding is a potential accident tolerant fuel (ATF) cladding with high research priority and still in the engineering design stage for now. However, considering its disadvantages, such as low irradiated thermal conductivity, ductility that barely not exist, further evaluations of its in-pile behaviors are still necessary. Based on the self-developed code we recently updated, relevant thermohydraulic and mechanical models in FROBA-ATF were applied to simulate the cladding behaviors under normal and accident conditions in this paper. Even through steady-state performance analysis revealed that this kind of cladding material could greatly reduce the oxidation thickness, the thermal performance of UO₂-SiC was poor due to its low inpile thermal conductivity and creep rate. Besides, the risk of failure exists when reactor power decreased. With geometry optimization and dopant addition in pellets, the steady-state performance of UO₂-SiC was enhanced and the failure risk was reduced. The thermal and mechanical performance of the improved UO₂-SiC was further evaluated under Loss of coolant accident (LOCA) and Reactivity Initiated Accident (RIA) conditions. Transient results showed that the optimized ATF had better thermal performance, lower cladding hoop stress, and could provide more coping time under accident conditions.

• Composites Research
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• 제35권3호
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• pp.161-174
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• 2022

SiC 섬유강화 복합체는 경수형 원자로의 안전성을 획기적으로 향상시킬 수 있는 사고저항성 핵연료 피복관 소재이다. 지르코늄 합금 피복관 및 금속기반 사고저항성 핵연료 피복관에 비해, 중대 사고 환경에서도 우수한 구조적 안정성을 가지고 부식 속도가 매우 낮아, 사고 시 원자로의 온도를 낮추고 사고 진행을 늦출 수 있다. 본 논문에서는 현재 개발되고 있는 사고저항성 SiC 복합체 핵연료 피복관의 개념 및 가동/사고환경에서의 다양한 특성, 상용화를 위해 해결해야 할 다양한 이슈에 대해서 소개하고자 한다.

• Nuclear Engineering and Technology
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• 제52권1호
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• pp.1-13
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• 2020

At present, the Department of Energy (DOE) in Unite State are directing the efforts of developing accident tolerant fuel (ATF) technology. As the first barrier of nuclear fuel system, the material selection of fuel rod cladding for ATFs is a basic but very significant issue for the development of this concept. The advanced cladding is attractive for providing much stronger oxidation resistance and better in-pile behavior under sever accident conditions (such as SBO, LOCA) for giving more coping time and, of course, at least an equivalent performance under normal condition. In recent years, many researches on in-plie or out-pile physical properties of some suggested cladding materials have been conducted to solve this material selection problem. Base on published literatures, this paper introduced relevant research backgrounds, objectives, research institutions and their progresses on several main potential claddings include triplex SiC, FeCrAl and MAX phase material Ti3SiC2. The physical properties of these claddings for their application in ATF area are also reviewed in thermohydraulic and mechanical view for better understanding and simulating the behaviors of these new claddings. While most of important data are available from publications, there are still many relevant properties are lacking for the evaluations.

• Nuclear Engineering and Technology
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• 제52권9호
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• pp.2054-2063
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• 2020

The high-temperature oxidation behavior of ZrSi₂ used as a coating material for nuclear fuel cladding was investigated for developing accident-tolerant fuel cladding of light water reactors. Bulk ZrSi₂ samples were prepared by spark plasma sintering. In situ X-ray diffraction was conducted in air at 900, 1000, and 1100 ℃ for 20 h. The microstructures of the samples before and after oxidation were examined by scanning electron microscopy and transmission electron microscopy. The results showed that the oxide layer of zirconium silicide exhibited a layer-by-layer structure of crystalline ZrO₂ and amorphous SiO₂, and the high-temperature oxidation resistance was superior to that of Zircaloy-4 owing to the SiO₂ layer formed. ZrSi₂ was coated on the Zircaloy-4 tube surface using laser 3D printing, and the coated tube was oxidized for 2000 s at 1200 ℃ under a vapor/argon mixture atmosphere. The outer surface of the coated tube was hardly oxidized (10-30 ㎛), while the inner surface of the uncoated tube was significantly oxidized to approximately 300 ㎛.

• Nuclear Engineering and Technology
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• 제51권1호
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• pp.1-12
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• 2019

One of proposed Accident Tolerant Fuel (ATF) concept is fully ceramic micro-encapsulated fuel (FCMF). FCMF using uranium mononitride (UN) has better safety aspects than $UO_2$ pellet fuel although it might not have a better neutronic performance due to the presence of matrix and high neutron-induced interaction of $^{14}N$. Before implementing UN-FCMF technology in Small-PWR, further research must be taken place to make sure the proposed design of fuel assembly has inherent safety features and maintain the fuel performance. This study focusses on the neutronic analysis of UN-FCMF based fuel assembly using Serpent and SCALE codes. It is shown in the proposed fuel assembly design has inherent safety features with respect to the fuel temperature reactivity coefficient, void reactivity coefficient, and moderator temperature reactivity coefficient. It is noted that the use of FCMF leads to a lower ratio of burnup to $^{235}U$ enrichment ratio compared to the $UO_2/Zr$ fuel.

• 방사성폐기물학회지
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• 제17권1호
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• pp.37-46
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• 2019

사고저항성 핵연료의 일환으로 $UO_2$ 입자가 세라믹 셀 벽으로 둘러싸인 미세구조를 갖는 세라믹 미소셀 $UO_2$ 소결체를 개발 중이다. 이는 핵분열생성물들을 $UO_2$ 펠렛 내에 포집하여 펠렛 외부로의 방출을 저감함으로써 봉내압 상승을 완화하고 응력부식균열 발생률을 낮춘다. 생성량이나 방사능 측면에서 위험한 핵분열생성물 중 하나로 여겨지는 세슘은 세라믹 미소셀소결체 내에서 셀 물질과 화학반응 하여 포집될 수 있다. 따라서, 세슘 포집능은 해당 화학반응의 열역학적, 속도론적 특성에 의해 결정된다. 역으로, 미소셀 소결체의 조성설계 시 해당 반응에 대한 열역학적 예측이 필수적이다. 본 연구는 세라믹 현재 개발 중인 여러 미소셀 조성(Si-Ti-O, Si-Cr-O, Si-Al-O)에 대해 세슘의 포집능을 평가하는 열역학적 계산을 다룬다. 계산에 앞서 먼저 HSC Chemistry를 이용해 세슘과 셀 물질의 물리/화학적 상태를 정의한 후, LWR 정상운전 모사환경에서 계산된 세슘포텐셜(${\Delta}G_{Cs}$)과 산소포텐셜(${\Delta}G_{O_2}$)에 근거하여 세슘포집 반응성을 평가하였다. 계산 결과에 근거하면, 세슘 포집반응은 상기 모든 조성에서 자발적일 것으로 예상되며 이로써 조성설계의 근거를 제시함과 동시에 세슘의 포집능을 평가하는 효과적인 방법을 제공한다.

• Nuclear Engineering and Technology
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• 제54권5호
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• pp.1549-1559
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• 2022

As an accident tolerant fuel (ATF) concept, oxide dispersion strengthened Zircaloy-4 (ODS Zry-4) cladding has been developed to enhance the mechanical properties of cladding using laser processing technology. In this study, a simulation technique was established to investigate the mechanical properties and effects of Y₂O₃ particles for the ODS Zry-4. A 3D representative volume element (RVE) model was developed considering the parameters of the size, shape, distribution and volume fraction (VF) of the Y₂O₃ particles. From the 3D RVE model, the Young's modulus, coefficient of thermal expansion (CTE) and creep strain rate of the ODS Zry-4 were effectively calculated. It was observed that the VF of Y₂O₃ particles had a significant effect on the aforementioned mechanical properties. In addition, the predicted properties of ODS Zry-4 were applied to a simulation model to investigate cladding deformation under a transient condition. The ODS Zry-4 cladding showed better performance, such as a delay in large deformation compared to Zry-4 cladding, which was also found experimentally. Accordingly, it is expected that the simulation approach developed here can be efficiently employed to predict more properties and to provide useful information with which to improve ODS Zry-4.

• Nuclear Engineering and Technology
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• 제52권3호
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• pp.508-519
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• 2020

During transients or accidents, the reactor core is uncovered, and droplets entrained above the quench front collides with the uncovered fuel rod surface. Droplet impact cooling can reduce the peak cladding temperature. Besides zirconium-based cladding, versatile accidental tolerant fuel (ATF) claddings, including FeCrAl, have been proposed to increase the accident coping time. In order to investigate the effect of surface properties on droplet impact cooling of cladding surfaces, the droplet impact phenomena are photographed on the FeCrAl and zircaloy-4 (Zr-4) surfaces under different conditions. On the oxidized FeCrAl surface, the Leidenfrost phenomenon is not observed even when the surface temperature is as high as 550 ℃ with We > 30. Comparison of the impact behaviors observed on different materials shows that nucleate and transition boiling is more intensive on surfaces with larger thermal conductivity. The Leidenfrost point temperature (LPT) decreases with the solid thermal effusivity (${\sqrt{k{\rho}C_p}}$). However, the CHF temperature is relatively insensitive to the surface oxidation and Weber number. Droplet spreading diameter is analyzed quantitatively in the film boiling stage. Based on the energy balance a correlation is proposed for droplet maximum spreading factor. A mechanistic model is also developed for the LPT based on homogeneous nucleation theory.