• Journal of the Korean Vacuum Society
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• v.4 no.S2
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• pp.100-105
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• 1995

Nitrogen implantation process has been applied for improvement of wear resistance of Zircaloy-4 fuel cladding materials. Nitrogen was implanted at 120keV to a total dose range of $1\times 10^{17}$ions/$\textrm{cm}^2$ to $1\times 10^{18}$ions/$\textrm{cm}^2$ at various temperatures between $270^{\circ}C$ and $671^{\circ}C$. The microstructure changes by nitrogen implantation were analyzed by XRD and AES and wear behavior was evaluated by performing ball-on-disc type wear testing at various loads and sliding velocities under unlubricated condition. Nitrogen implantation produced ZrNx nitride above $3\times 10^{17}$ions/$\textrm{cm}^2$ as well as heavy dislocations, which resluted in an increase in microhardness of the implanted surface of up to 1400 $H_k$ from 200 $H_k$ of unimplanted substrate. Hardness was also found to be increased with increasing implantation temperature up to 1760 $H_k$ at $620^{\circ}C$. The wear resistance was greatly improved as total ion dose and implantation temperature increased. The effective enhancement of wear resistance at high dose and temperature is believed to be due to the significant hardening associated with high degree of precipitation of Zr nitrides and generation of prismatic dislocation loops.

• Nuclear Engineering and Technology
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• v.40 no.1
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• pp.21-36
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• 2008

High bum-up fuel technology has been developed through a national R&D program, which covers key technology areas such as claddings, $UO_2$ pellets, spacer grids, performance code, and fuel assembly tests. New cladding alloys were developed through alloy designs, tube fabrication, out-of-pile test and in-reactor test. The new Zr-Nb tubes are found to be much better in their corrosion resistance and creep strength than the Zircaloy-4 tube, owing to an optimized composition and heat treatment of the new Zr-Nb alloys. A new fabrication technology for large grain $UO_2$ pellets was developed using various uranium oxide seeds and a micro-doping of Al. The uranium oxide seeds, which were added to $UO_2$ powder, were prepared by oxidizing and heat-treating scrap $UO_2$ pellets. A $UO_2$ pellet containing tungsten channels was fabricated for a thermal conductivity enhancement. For the fuel performance analysis, new high burnup models were developed and implemented in a code. This code was verified by an international database and our own database. The developed spacer grid has two features of contoured contact spring and hybrid mixing vanes. Mechanical and hydraulic tests showed that the spacer grid is superior in its rodsupporting, wear resistance and CHF performance. Finally, fuel assembly test technology was also developed. Facilities for mechanical and thermal hydraulic tests were constructed and are now in operation. Several achievements are to be utilized soon by the Korea Nuclear Fuel and thereby contribute to the economy and safety of PWR fuel in Korea

• Proceedings of the Materials Research Society of Korea Conference
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• 2003.11a
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• pp.213-213
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• 2003

고온 및 고압의 가혹한 방사선 분위기에서 사용되는 핵연료 피복관은 중성자 조사 및 수소화합물의 생성 등으로 인하여 기계적 성질이 저하된다. 따라서 조사된 핵연료 피복관의 손상기준 확립과 안전성 해석을 위해서는 연성 및 강도 등 기계적 특성을 정확히 이해하여야 할 필요가 있다. 핵연료 피복관의 종 및 횡 방향 인장특성 평가를 위하여 개발된 기존의 다양한 시험법들을 비교하고, 핫셀시험에 적합한 인장시험법을 개발하였다. 피복관의 종방향 인장시편은 튜브시편 또는 게이지부 내에서 균일한 변형률 분포를 얻도록 설계된 도그본 튜브시편(그림 1)을 사용한다. 피복관의 횡방향 인장시험에 사용되는 링시편(그림 2)은 게이지부 내에서 균일한 단축 원환변형율 분포 또는 평면변형율 조건을 나타내도록 설계한다. 연소 또는 조사된 피복관으로부터 시편을 제작하기 위해서는 핫셀 내에서 작업 이 가능한 방전가공기(그림 3)를 사용한다. 피복관의 종방향 인장시험용그립(grip)은 핀-부하형이며, 횡방향 인장시험의 경우는 시험 동안 시편의 곡률이 일정하게 유지 되도록 그립의 형상 및 치수를 결정한다(그림 4). 피복관의 종 및 횡방향 강도와 변형 등 기계적 특성을 평가하기 위한 응력-변형율 곡선은 시험기의 복합 강성(K)을 고려하여 결정한다. 이상과 같이 검토된 인장시험법은 피복관의 안전성 해석(safety analysis)과 관련 규정(regulatory)에서 사용되는 피복관 손상기준(fuel damage criteria)의 개선에 필수적인 자료를 제공한다.

• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 2003.11a
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• pp.64-71
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• 2003

In order to evaluate the effect of alloying elements (Nb and Sn) on the wear resistance of advanced Zr fuel claddings, sliding wear tests have been performed in room temperature air and water and these results were compared with those of commercial alloys such as Zircaloy-4, A and B alloys. As a result, the advanced Zr fuel claddings have a similar wear resistance compared with the commercial alloys. The wear resistance of the advanced Zr fuel claddings is closely releted to the content of Nb and Sn even though the effects of transition elements are involved in deforming wear properties. In the tested specimens with similar Sn content, wear volume became down to a minimum at $0.4\;wt\;\%$ Nb, then rapidly increased at 1.0 wt Nb. This behavior results in the variation of grain size with alloying contents. But Sn did not have a significant effect on the wear volume of advanced Zr fuel claddings below $1.1\;wt\%$. The relationship between alloying elements and wear behaviour was evaluated and discussed using material compatibility factor.

• Journal of Welding and Joining
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• v.7 no.2
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• pp.60-69
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• 1989

The present study was aimed at investigating the effect of welding parameters such as welding current, electrode force(or squeeze force) and parts cleaning on the sound weld, and establishing the most reliable weld conditions for HWP(Heavy Water Reactor) fuel end capping with the resistance upset butt welding. Major results obtained are as follows. 1. The amount of sound weld was increased with increasing weld current(5.0-11KA) because the activated diffusion with increasing heat generation played an important role in eliminating the porosity and weld line in the weld interface. 2. It was found that weld current was not significantly influenced by the electrode force although the increase of it caused a slight increase of weld current and upset deformation. 3. Acetone rinsing before drying for the Zircaloy-4 end cap cleaning produced the reliable sound weld because it would remove the remaining solvent and surface films, and provided the uniform contact between the end cap and the tube. 4. The optimum welding conditions for fuel end capping by a resistance upset hytt welding are obtained as follows. weld current: 10-11KA, electrode force: 62-90KPa parts cleaning: vapor degreasing.rarw.water, acetone rinsing.rarw.drying.

• Nuclear Engineering and Technology
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• v.51 no.6
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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.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.39 no.3
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• pp.249-257
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• 2015

A fuel assembly consists of fuel rods composed of pellets (UO2) and a cladding tube (Zircaloy). The role of the fuel rods in the reactor is to generate heat by nuclear fission, as well as to retain fission products during operation. A simulation method using a computer program was used to evaluate the safety of the nuclear fuel rods. This computer program has been called the fuel performance code. In the analysis of a light water reactor fuel rod, the gap conductance, which depended on the distance between the pellets and cladding tube, mainly influenced the thermomechanical behavior of the fuel rod. In this work, a 3D gap element was proposed to simulate the thermo-mechanical behavior of the nuclear fuel rod, considering the gap conductance. To implement the proposed 3D gap element, a 3D thermo-mechanical module was also developed using FORTRAN90. The asymmetric characteristics of the nuclear fuel rod, such as the MPS (missing pellet surface) and eccentricity, were simulated to evaluate the proposed 3D gap element.

• Journal of the Korean Ceramic Society
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• v.48 no.1
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• pp.69-73
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• 2011

A $ZrO_2$ thin film as a corrosion protective layer was deposited on Zircaloy-4 (Z-4) clad material using $N_2O$ as a reactive gas by RF reactive magnetron sputtering at room temperature. The Z-4 substrate was located in plasma or out of plasma during the $ZrO_2$ deposition process to investigate mechanical and corrosive properties for the plasma immersion. Tetragonal and monoclinic phases were existed in $ZrO_2$ thin film immersed in plasma. We observed that a grain size of the $ZrO_2$ thin film immersed in plasma state is larger than that of the $ZrO_2$ thin film out of plasma state. In addition, the corrosive property of the $ZrO_2$ thin films in the plasma was characterized using the weight gains of Z-4 after the corrosion test. Compared with the $ZrO_2$ thin film immersed out of plasma, the weight gains of $ZrO_2$ thin film immersed in plasma were larger. These results indicate that the $ZrO_2$ film with the tetragonal phase in the $ZrO_2$ can protect the Z-4 from corrosive phenomena.

• Nuclear Engineering and Technology
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• v.55 no.7
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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.