• Applied Chemistry for Engineering
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• v.28 no.3
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• pp.345-350
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• 2017

The sintered density of uranium oxide pellets for pressurized water reactors is generally analyzed with pellet's samples completed with the sintering process. In this paper, the sintered density was analyzed by the newly developed method measuring the chromatography of ammonium diuranate, a precursor of uranium oxide, by a spectrophotometer (CM-5, Konica Minolta) before completing the sintering process. As a result of the sintered density analysis based on the brightness, color coordinate values (L, a, b) obtained from five ammonium diuranate samples by a spectrophotometer and the trend line of sintered density analyzed by a previous method, the sintered density with respect to the L value was observed with 0.9967 of the decision factor $R^2$. In case of a value, $R^2$ value was 0.9534 indicating lower reliability than that of the L value. However, b value with $R^2$ value of 0.4349 showed a very low correlation.

• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.178-187
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• 2021

Dual-cooled annular fuel allows a significant increase in power density while maintaining or improving safety margins. However, the dual-cooled design brings much higher Zircaloy charge in reactor core, which could cause a great threaten of hydrogen explosion during severe accidents. Hence, an innovative fuel combined dual-cooled annular geometry and SiC cladding was proposed for the first time in this study. Capabilities of fuel design and behavior simulation were developed for this new fuel by the upgrade of FROBA-ANNULAR code. Considering characteristics of both SiC cladding and dual-cooled annular geometry, the basic fuel design was proposed and preliminary proved to be feasible. After that, a design optimization study was conducted, and the optimal values of as-fabricated plenum pressure and gas gap sizes were obtained. Finally, the performance simulation of the new fuel was carried out with the full consideration of realistic operation conditions. Results indicate that in addition to possessing advantages of both dual-cooled annular fuel and accident tolerant cladding at the same time, this innovative fuel could overcome the brittle failure issue of SiC induced by pellet-cladding interaction.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 2000.11a
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• pp.105-109
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• 2000

종래의 핵연료소결체가 혼합산화물 혹은 이산화우라늄중 한가지 핵연료만으로 구성한 것과 달리 내부를 저농축 이산화우라늄 핵연료로 채우고 그 외부를 링형태의 혼합산화물 핵연료로 둘러 싼 이중구조를 특징으로 한다. 이러한 형태의 핵연료소결체는 중심영역의 핵분열반응률 줄임으로써 핵분열 기체생성, 핵연료봉 중심온도와 평균온도를 낮추어 준다. 이는 핵분열 기체방출을 낮추어 혼합산화물 핵연료봉 성능을 향상시키고 방출 연소도를 증가시키는 효과가 있다.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.385-386
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• 2005

• Proceedings of the Korean Society for Technology of Plasticity Conference
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• 2004.05a
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• pp.109-113
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• 2004

Slitting device is equipment to separate spent fuel of 250 mm rod cut pellets and hull in order to supply required $UO_2$ pellets through the dry pulverizing/mixing device. For development of its device, We have analyzed slitting programs so that the existing device is modified an appropriate scale in the advanced spent fuel conditioning process. The results of the analysis, we added the automatic separation function of pellets and hull, After slitting. Also, we have concentrated on reducing the operation time so that the support and the body of a slitting blade could have been established in the single structure to be easily maintained. It is based on a design and manufacture of a testing device and we have performed an efficiency evaluation. We have analyzed the results of efficiency tests of the slitting device and get the specification of the slitting device. we complete the basic design of the slitting device by using of these data. Therefore, We apply to a basic data when manufacturing a slitting device.

• Nuclear Engineering and Technology
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• v.35 no.1
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• pp.55-63
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• 2003

Threshold porosity above which fission gas release channels would be formed in the rim egion of high burnup UO$_2$ fuel was estimated by the Monte Carlo method and Hoshen-Kopelman algorithm. With the assumption that both rim pore and rim grain can be represented by cube, pore distribution in the rim was simulated 3-dimensionally by the Monte Carlo method according to porosity and pore size distribution. Then, using the Hoshen-Kopelman algorithm, the fraction of open rim pores interlinked to the outer surface of a fuel pellet was derived as a function of rim porosity. The simulation showed that porosity of 24-25% is the threshold above which the number of rim pores forming release channels increases very rapidly. On the other hand, channels would not be formed if the porosity is less than about 23.5%. This is consistent with the observation that, for porosity less than 23.5%, almost no fission gas is released in the rim. However, once the rim porosity reaches beyond 25%, extensive open paths would be developed and considerable fission gas release would start in the rim.

• Nuclear Engineering and Technology
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• v.49 no.6
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• pp.1287-1300
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• 2017

Based on the observation that ignoring the angle dependency of multigroup resonance cross sections within a fuel pellet would result in nontrivial underestimation of the spatial self-shielding of flux, a parametrized spectral superhomogenization (SPH) factor library (PSSL) method is developed as a practical means of resolving the problem. Region-wise spectral SPH factors are calculated by the normal and transport corrected SPH iterations after ultrafine group slowing down calculations over various light water reactor pin-cell configurations. The parametrization is done with fuel temperature, U-238 number density, fuel radius, moderator source represented by ${\Sigma}_{mod}V_{mod}$, and the number density ratio of resonance nuclides to that of U-238 in a form of resonance interference correction factors. The parametrization is successful in that the root mean square errors of the interpolated SPH factors over the fuel regions of various pin-cells are within 0.1%. The improvement in reactivity error of the PSSL method is shown to be superior to that by the original SPH method in that the reactivity bias of -200 pcm to -300 pcm vanishes almost completely. It is demonstrated that the environment effect takes only about 4% in the reactivity improvement so that the pin-cell based PSSL method is effective in the assembly problems.

• Advances in Energy Research
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• v.5 no.2
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• pp.91-105
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• 2017

Thermal hydraulic (TH) analysis of nuclear power reactors is utmost important. In this way, the numerical codes that preparing TH data in reactor core are essential. In this paper, a subchannel analysis of a Russian pressurized water reactor (WWER1000) core with enhanced numerical code is carried out. For this, in fluid domain, the mass, axial and lateral momentum and energy conservation equations for desired control volume are solved, numerically. In the solid domain, the cylindrical heat transfer equation for calculation of radial temperature profile in fuel, gap and clad with finite difference and finite element solvers are considered. The dependence of material properties to fuel burnup with Calza-Bini fuel-gap model is implemented. This model is coupled with Isotope Generation and Depletion Code (ORIGEN2.1). The possibility of central hole consideration in fuel pellet is another advantage of this work. In addition, subchannel to subchannel and subchannel to rod connection data in hexagonal fuel assembly geometry could be prepared, automatically. For a demonstration of code capability, the steady state TH analysis of a the WWER1000 core is compromised with Thermal-hydraulic analysis code (COBRA-EN). By thermal hydraulic parameters averaging Fuel Assembly-to-Fuel Assembly method, the one sixth (symmetry) of the Boushehr Nuclear Power Plant (BNPP) core with regular subchannels are modeled. Comparison between the results of the work and COBRA-EN demonstrates some advantages of the presented code. Using the code the thermal modeling of the fuel rods with considering the fission gas generation would be possible. In addition, this code is compatible with neutronic codes for coupling. This method is faster and more accurate for symmetrical simulation of the core with acceptable results.

• Nuclear Engineering and Technology
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• v.42 no.5
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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.

• Journal of the Korean Society for Precision Engineering
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• v.30 no.1
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• pp.113-119
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• 2013

To develop a new nuclear fuel, it needs to make a test fuel rod and carry out burn-up test in the test loop of a research reactor to check the irradiation characteristics of the nuclear fuel. At that time, several sensors such as thermocouples, LVDTs and SPNDs are needed to be attached in and out of the fuel rod and connect them with instrumentation cables. Then, the instrumentation cables deliver the signals measured by the sensors to the measuring device located outside of the reactor pool. In particular, to install a thermocouple in a fuel rod, it needs to drill off holes on the alumina blocks and sintered $UO_2$ pellets. However, because the hardness of a sintered $UO_2$ pellet is 700 Hv (or HRC 61) and that of an alumina block is 1480 Hv, a special drilling machine which adapts a diamond coated drill bit had developed. In this study, several case experiments have been carried out to find an optimal drilling condition of the drilling machine. And, using the optimal drilling condition, minimum numbers of the holes that a drill bit can drill off are verified.