Establishment of the design stress intensity value for the plate-type fuel assembly using a tensile test

Kim, Hyun-Jung,Tahk, Young-Wook,Jun, Hyunwoo,Kong, Eui-Hyun,Oh, Jae-Yong,Yim, Jeong-Sik Korean Nuclear Society 2021 Nuclear Engineering and Technology Vol.53 No.3

In this paper, the design stress intensity values for the plate-type fuel assembly for research reactor are presented. Through a tensile test, the material properties of the cladding (aluminum alloy 6061) and structural material (aluminum alloy 6061-T6), in this case the yield and ultimate tensile strengths, Young's modulus and the elongation, are measured with the temperatures. The empirical equations of the material properties with respect to the temperature are presented. The cladding undergoes several heat treatments and hardening processes during the fabrication process. Cladding strengths are reduced compared to those of the raw material during annealing. Up to a temperature of 150 ℃, the strengths of the cladding do not significantly decrease due to the dislocations generated from the cold work. However, over 150 ℃, the mechanical strengths begin to decrease, mainly due to recrystallization, dislocation recovery and precipitate growth. Taking into account the uncertainty of the 95% probability and 95% confidence level, the design stress intensities of the cladding and structural materials are established. The presented design stress intensity values become the basis of the stress design criteria for a safety analysis of plate-type fuels.

Uncertainty and sensitivity analyses for fuel temperature evaluations of U-Mo/Al plate-type dispersion fuel

Sweidan, Faris B.,Tahk, Young-wook,Yim, Jeong-Sik,Ryu, Ho Jin Elsevier 2018 Annals of nuclear energy Vol.120 No.-

<P><B>Abstract</B></P> <P>U-Mo/Al plate-type dispersion fuel is a promising candidate for the conversion of research reactor fuels from highly enriched to low-enriched uranium due to its high uranium density. The fuel temperature is a very important parameter, as it affects the performance of the fuel through various aspects, such as the formation of an interaction layer (IL) between the fuel particles and the matrix, swelling, and the release of fission gas. For these reasons, the fuel temperature as a function of the fission density was calculated for two representative heat flux profiles using best-estimate values and Monte Carlo simulations. Uncertainty and sensitivity analyses which utilized the uncertainties of the critical parameters were then conducted to determine the upper (maximum) and lower (minimum) bounds of the fuel temperature for the selected heat flux profiles. The uncertainty analysis used common uncertainty propagation approaches and a probabilistic sensitivity analysis (Monte Carlo simulation), randomly sampling numbers following a Gaussian distribution. Lastly, the Pearson correlation coefficient was used to identify the input uncertainties which influence the fuel temperature most in the sensitivity analysis. These analyses contribute to safety analyses and to the licensing process, as they are used in best-estimate approaches that apply realistic assumptions complemented with uncertainty analyses, such as the Best Estimate Plus Uncertainty (BEPU) approach.</P> <P><B>Highlights</B></P> <P> <UL> <LI> Uncertainty and sensitivity analyses on U-Mo/Al dispersion fuel were conducted. </LI> <LI> Uncertainties of key parameters on the fuel temperature were evaluated. </LI> <LI> The fuel temperature as a function of the fission density was calculated. </LI> <LI> Pearson correlation coefficient was used to identify the influence of the input uncertainties. </LI> </UL> </P>

FUEL PERFORMANCE CODE COSMOS FOR ANALYSIS OF LWR UO2 AND MOX FUEL

BYUNG-HO LEE,구양현,JAE-YONG OH,Jin-SikCheon,YOUNG-WOOK TAHK,손동성 한국원자력학회 2011 Nuclear Engineering and Technology Vol.43 No.6

The paper briefs a fuel performance code, COSMOS, which can be utilized for an analysis of the thermal behavior and fission gas release of fuel, up to a high burnup. Of particular concern are the models for the fuel thermal conductivity, the fission gas release, and the cladding corrosion and creep in UO2 fuel. In addition, the code was developed so as to consider the inhomogeneity of MOX fuel, which requires restructuring the thermal conductivity and fission gas release models. These improvements enhanced COSMOS’s precision for predicting the in-pile behavior of MOX fuel. The COSMOS code also extends its applicability to the instrumented fuel test in a research reactor. The various in-pile test results were analyzed and compared with the code’s prediction. The database consists of the UO2 irradiation test up to an ultra-high burnup, power ramp test of MOX fuel, and instrumented MOX fuel test in a research reactor after base irradiation in a commercial reactor. The comparisons demonstrated that the COSMOS code predicted the in-pile behaviors well, such as the fuel temperature, rod internal pressure, fission gas release, and cladding properties of MOX and UO2 fuel. This sufficient accuracy reveals that the COSMOS can be utilized by both fuel vendors for fuel design, and license organizations for an understanding of fuel in-pile behaviors.

FUEL PERFORMANCE CODE COSMOS FOR ANALYSIS OF LWR UO₂ AND MOX FUEL

Lee, Byung-Ho,Koo, Yang-Hyun,Oh, Jae-Yong,Cheon, Jin-Sik,Tahk, Young-Wook,Sohn, Dong-Seong Korean Nuclear Society 2011 Nuclear Engineering and Technology Vol.43 No.6

The paper briefs a fuel performance code, COSMOS, which can be utilized for an analysis of the thermal behavior and fission gas release of fuel, up to a high burnup. Of particular concern are the models for the fuel thermal conductivity, the fission gas release, and the cladding corrosion and creep in $UO_2$ fuel. In addition, the code was developed so as to consider the inhomogeneity of MOX fuel, which requires restructuring the thermal conductivity and fission gas release models. These improvements enhanced COSMOS's precision for predicting the in-pile behavior of MOX fuel. The COSMOS code also extends its applicability to the instrumented fuel test in a research reactor. The various in-pile test results were analyzed and compared with the code's prediction. The database consists of the $UO_2$ irradiation test up to an ultra-high burnup, power ramp test of MOX fuel, and instrumented MOX fuel test in a research reactor after base irradiation in a commercial reactor. The comparisons demonstrated that the COSMOS code predicted the in-pile behaviors well, such as the fuel temperature, rod internal pressure, fission gas release, and cladding properties of MOX and $UO_2$ fuel. This sufficient accuracy reveals that the COSMOS can be utilized by both fuel vendors for fuel design, and license organizations for an understanding of fuel in-pile behaviors.

DROP IMPACT ANALYSIS OF PLATE-TYPE FUEL ASSEMBLY IN RESEARCH REACTOR

Kim, Hyun-Jung,Yim, Jeong-Sik,Lee, Byung-Ho,Oh, Jae-Yong,Tahk, Young-Wook Korean Nuclear Society 2014 Nuclear Engineering and Technology Vol.46 No.4

In this research, a drop impact analysis of a fuel assembly in a research reactor is carried out to determine whether the fuel plate integrity is maintained in a drop accident. A fuel assembly drop accident is classified based on where the accident occurs, i.e., inside or outside the reactor, since each occasion results in a different impact load on the fuel assembly. An analysis procedure suitable for each drop situation is systematically established. For an accident occurring outside the reactor, the direct impact of a fuel assembly on the pool bottom is analyzed using implicit and explicit approaches. The effects of the key parameters, such as the impact velocity and structural damping ratios, are also studied. For an accident occurring inside the reactor, the falling fuel assembly may first hit the fixing bar at the upper part of the standing fuel assembly. To confirm the fuel plate integrity, a fracture of the fixing bar should be investigated, since the fixing bar plays a role in protecting the fuel plate from the external impact force. Through such an analysis, the suitability of an impact analysis procedure associated with the drop situation in the research reactor is shown.

연구로 핵연료집합체 낙하사고에서의 낙하 충격 해석

김현중(Hyun-Jung Kim),임정식(Jeong-Sik Yim),이병호(Byung-Ho Lee),오제용(Jae-Young Oh),탁영욱(Young-Wook Tahk) 대한기계학회 2013 대한기계학회 춘추학술대회 Vol.2013 No.5