• Proceedings of the Korean Nuclear Society Conference
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• 1997.05b
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• pp.445-450
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• 1997

Self-Powered Neutron Detectors(SPNDs) are currently used to estimate the power generation distribution and fuel burn-up in several nuclear power reactors in Korea. In this paper, Monte Carlo simulation is accomplished to calculate the escape probability of beta particle as a function of their birth position fur the typical geometry of rhodium-based SPNDs. Also, a simple numerical method calculates the initial generation rate of beta particles and the change of generation rate due to rhodium burn-up. Using the simulation and the numerical method, the burn-up profile of rhodium density and the neutron sensitivity are calculated as a function of burn-up time in the reactor. The sensitivity of the SPNDs decreases non-linearly due to the high absorption cross-section and the non-uniform burn-up of rhodium in the emitter rod. In addition, for improvement of some properties of rhodium-based SPNDs which are currently used, this paper presents a new material. The method used here can be applied to the analysis of other types of SPNDs and will be useful in the optimum design of new SPNDs for long term usage.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.430-438
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• 2021

The spatial distribution of neutron flux or reaction rate was calculated by cell or mesh tally in traditional Monte Carlo simulation. However, either cell or mesh tally leads to the increase of memory consumption and simulation time. In this paper, the function expansion tally (FET) method was developed in Reactor Monte Carlo code RMC to solve this problem. The FET method was applied to the tallies of neutron flux distributions of uranium block and PWR fuel rod models. Legendre polynomials were used in the axial direction, while Zernike polynomials were used in the radial direction. The results of flux, calculation time and memory consumption of different expansion orders were investigated, and compared with the mesh tally. Results showed that the continuous distribution of flux can be obtained by FET method. The flux distributions were consistent with that of mesh tally, while the memory consumption and simulation time can be effectively reduced. Finally, the convergence analysis of coefficients of polynomials were performed, and the selection strategy of FET order was proposed based on the statistics uncertainty of the coefficients. The proposed method can help to determine the order of FET, which was meaningful for the efficiency and accuracy of FET method.

• Journal of the Korean Society of Systems Engineering
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• v.16 no.2
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• pp.38-46
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• 2020

The accurate measurement of critical heat flux (CHF) in flow boiling is important for the safety requirement of the nuclear power plant to prevent sharp degradation of the convective heat transfer between the surface of the fuel rod cladding and the reactor coolant. In this paper, a System Engineering approach is used to develop a model that predicts the CHF using machine learning. The model is built using artificial neural network (ANN). The model is then trained, tested and validated using pre-existing database for different flow conditions. The Talos library is used to tune the model by optimizing the hyper parameters and selecting the best network architecture. Once developed, the ANN model can predict the CHF based solely on a set of input parameters (pressure, mass flux, quality and hydraulic diameter) without resorting to any physics-based model. It is intended to use the developed model to predict the DNBR under a large break loss of coolant accident (LBLOCA) in APR1400. The System Engineering approach proved very helpful in facilitating the planning and management of the current work both efficiently and effectively.

• Journal of Energy Engineering
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• v.24 no.1
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• pp.149-163
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• 2015

In this study, we conducted study on the confirmation of thermal-hydraulic safety for Mast assembly with Computational Fluid Dynamics(CFD) analysis. Before performing the natural convection analysis for the Mast assembly by using CFD code, we validated the CFD code against two benchmark natural convection data for the evaluation of turbulence models and confirmation of its applicability to the natural convection flow. From the first benchmark test which was performed by Betts et al. in the simple rectangular channel, we selected standard k-omega turbulence model for natural convection. And then, calculation performance of CFD code was also investigated in the sub-channel of rod bundle by comparing with PNL(Pacific Northwest Laboratory) experimental data and prediction results by MATRA and Fluent 12.0 which were performed by Kwon et al.. Finally, we performed main natural convection analysis for fuel assembly inside the Mast assembly by using validated turbulence model. From the calculation, we observed stable natural circulation flow between the mast assembly and pool side and evaluated the thermal-hydraulic safety by calculating the departure from nucleate boiling ratio.

• Proceedings of the Materials Research Society of Korea Conference
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• 2011.05a
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• pp.15-15
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• 2011

As you know, boron compounds, borax ($Na_2B_4O_5(OH)_4{\cdot}8H_2O$) etc. were known thousands of years ago. As for natural boron, it has two naturally occurring and stable isotopes, boron 11 ($^{11}B$) and boron 10 ($^{10}B$). The neutron absorption $^{10}B$ is included about 19~20% with 80~81% $^{11}B$. Boron is similar to carbon in its capability to form stable covalently bonded molecular networks. The mass difference results in a wide range of ${\beta}$ values between the $^{11}B$ and $^{10}B$. The $^{10}B$ isotope, stable with 5 neutrons is excellent at capturing thermal neutrons. For example, it is possible to decrease a thermal neutron required for the nuclear reaction of uranium 235 ($^{235}U$). If $^{10}B$ absorbs a neutron ($^1n$), it will change to $^7Li+^1{\alpha}$ (${\alpha}$ ray, like $^4He$) with prompt ${\gamma}$ ray from $^{11}B$ $^{11}B$ (equation 1). $$^{10}B+^1n\;{\rightarrow}\;^{11}B\;{\rightarrow}\; prompt \;{\gamma}\;ray (478 keV), \;^7Li+4{\alpha}\;(4He)\;\;\;\;{\cdots}\; (1)$$ If about 1% boron is added to stainless steel, it is known that a neutron shielding effect will be 3 times the boron free steel. Enriched boron or $^{10}B$ is used in both radiation shielding and in boron neutron capture therapy. Then, $^{10}B$ is used for reactivity control and in emergency shutdown systems in nuclear reactors. Furthermore, boron carbide, $B_4C$, is used as the charge of a nuclear fission reaction control rod material and neutron cover material for nuclear reactors. The $B_4C$ powder of natural B composition is used as a charge of a control material of a boiling water reactor (BWR) which occupies commercial power reactors in nuclear power generation. The $B_4C$ sintered body which adjusted $^{10}B$ concentration is used as a charge of a control material of the fast breeder reactor (FBR) currently developed aiming at establishment of a nuclear fuel cycle. In this study for new boron compound, silicon boride ceramics for capturing thermal neutrons, preparation and characterization of both silicon tetraboride ($SiB_4$) and silicon hexaboride ($SiB_6$) and ceramics produced by sintering were investigated in order to determine the suitability of this material for nuclear power generation. The relative density increased with increasing sintering temperature. With a sintering temperature of 1,923 K, a sintered body having a relative density of more than 99% was obtained. The Vickers hardness increased with increasing sintering temperature. The best result was a Vickers hardness of 28 GPa for the $SiB_6$ sintered at 1,923K for 1 h. The high temperature Vickers hardness of the $SiB_6$ sintered body changed from 28 to 12 GPa in the temperature range of room temperature to 1,273 K. The thermal conductivity of the SiB6 sintered body changed from 9.1 to 2.4 W/mK in the range of room temperature to 1,273 K.

• Korean Journal of Materials Research
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• v.18 no.12
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• pp.645-649
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• 2008

Graphite for the nuclear reactor is used to the moderator, reflector and supporter in which fuel rod inside of nuclear reactor. Recently, there are many researches has been performed on the various characteristics of nuclear graphite, however most of them are restricted to the structural and the mechanical properties. Therefore we focused on the thermal property of nuclear graphite. This study investigated the thermal emissivity following the oxidation degree of nuclear graphite with IG-11 used as a sample. IG-11 was oxidized to 6% and 11% in air at 5 l/min at $600^{\circ}C$. The porosity and thermal emissivity of the sample were measured using a mercury porosimeter and by an IR method, respectively. The thermal emissivity of an oxidized sample was measured at $100^{\circ}C$, $200^{\circ}C$, $300^{\circ}C$, $400^{\circ}C$ and $500^{\circ}C$. The porosity of the oxidized samples was found to increase as the oxidation degree increased. The thermal emissivity increased as the oxidation degree increased, and the thermal emissivity decreased as the measured temperature increased. It was confirmed that the thermal emissivity of oxidized IG-11 is correlated with the porosity of the sample.

• Journal of the Korean Society for Nondestructive Testing
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• v.23 no.4
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• pp.342-348
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• 2003

Ultrasonic backscattering profiles of the Zr plates(with a thickness of 1.32mm) with/without Be-Zr alloy layer(with a thickness of $100{\mu}m$) were measured at various incidence positions to evaluate the characteristics of Be diffusion layer. Four principal subprofiles were observed in the backward ultrasound radiated from leaky Lamb waves. The angles and the intensities of the subprofile peaks decreased by the stiffening effect of Be layer. Generation and change of the subprofiles were explained by the acoustical property, collective group velocity and leaky factor difference of the plates under consideration. Backward radiation subprofiles turned out to be an useful method for evaluating thin diffusion layers on plates.

• Journal of Ocean Engineering and Technology
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• v.17 no.5 s.54
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• pp.66-75
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• 2003

Long cylinder, subjected to internal pressure, is important in the analysis and design of nuclear fuel rod structures. In many cases, long cylinder problems have been considered as a plane strain condition. However, strictly speaking, long cylinder problems are not plane strain problems, but rather a general plane strain (GPS) condition, which is a combination of a plane strain state and a uniform strain state. The magnitude of the uniform axial strain is required, in order to make the summation of the axial force zero. Although there has been the GPS element, this paper proposes a general technique to solve long cylinder problems, using several pseudo-general plane strain (PGPS) elements. The conventional GPS elements and PGPS elements employed are as follows: axisymmetric GPS element (GA3), axisymmetric PGPS element (PGA8/6), 2-D GPS element (GIO), 3-D PGPS element (PG20/16), and reduced PGPS elements (RPGA6, RPG20/16). In particular, PGPS elements (PGA8/6, PG20/16) can be applied in periodic structure problems. These finite elements are tested, using several kinds of examples, thereby confirming the validity of the proposed finite element models.

• Nuclear Engineering and Technology
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• v.47 no.4
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• pp.398-406
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• 2015

In this study, R-123 flow boiling experiments were carried out to investigate the effects of nanoparticle deposition on heater surfaces on flow critical heat flux (CHF) and boiling heat transfer. It is known that CHF enhancement by nanoparticles results from porous structures that are very similar to layers of Chalk River unidentified deposit formed on nuclear fuel rod surfaces during the reactor operation period. Although previous studies have investigated the surface effects through surface modifications, most studies are limited to pool boiling conditions, and therefore, the effects of porous surfaces on flow boiling heat transfer are still unclear. In addition, there have been only few reports on suppression of wetting for decoupled approaches of reasoning. In this study, bare and $Al_2O_3$ nanoparticle-coated surfaces were prepared for the study experiments. The CHF of each surface was measured with different mass fluxes of $1,600kg/m^2s$, $1,800kg/m^2s$, $2,100kg/m^2s$, $2,400kg/m^2s$, and $2,600kg/m^2s$. The nanoparticle-coated tube showed CHF enhancement up to 17% at a mass flux of $2,400kg/m^2s$ compared with the bare tube. The factors for CHF enhancement are related to the enhanced rewetting process derived from capillary action through porous structures built-up by nanoparticles while suppressing relative wettability effects between two sample surfaces as a highly wettable R-123 refrigerant was used as a working fluid.

• Journal of the Korean Ceramic Society
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• v.42 no.12 s.283
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• pp.808-815
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• 2005

[ $Bi_xCe_{l-x}O_{2-x/2}$ ](BD C : Bismuth Doped Ceria) powders with x = 0.1, 0.2, and 0.3 were synthesized using the Glycine Nitrate Process (GNP). They were then calcined at $500^{\circ}C$ for 2 hand sintered in a pellet or rod form at 900, 1000 or $1100^{\circ}C$ for 4 h for characterization as the alternative electrolyte material for intermediate temperature solid oxide fuel cells. The BDC powder consisted of a single phase of $CeO_2-Bi_2O_3$ solid solution in the as-synthesized state as well as in the as-calcined state with a mean powder size of 4.5nm in the former state and 6.5 - 10.1nm in the latter. On the contrary, the second phase of $\alpha-Bi_2O_3$ was observed to have been formed in the sinter with its amount increasing roughly with increasing temperature or $Bi_2O_3$ content. The BOC powder was superior in sinterability to other alternative electrolyte materials such as GDC, ScSZ, and LSGM with the minimum sintering temperature for a relative density of $95\%$ or larger as low as $1100^{\circ}C$. The ionic conductivity of BOC increased with $Bi_2O_3$ content and the maximum value of 0.119 S/cm was obtained at $800^{\circ}C$ for $Bi_{0.3}Ce_{0.7}O_{1.85}$.