Evaluation of Neutron Capture Gamma-ray Spectra for ⁸⁹Y, ⁹³Nb, ¹²⁷I, ¹³³Cs, ¹⁴¹Pr, ¹⁹⁷Au, ^natTl, and ²⁰⁹Bi

KIM, Hyeong Il,YI, Mi Ja,LEE, Young-Ouk Atomic Energy Society of Japan 2007 Journal of nuclear science and technology Vol.44 No.8

<P>The neutron capture gamma-ray spectra for 8 nuclides, <SUP>89</SUP>Y, <SUP>93</SUP>Nb, <SUP>127</SUP>I, <SUP>133</SUP>Cs, <SUP>141</SUP>Pr, <SUP>197</SUP>Au, <SUP>nat</SUP>Tl, and <SUP>209</SUP>Bi, were calculated by using the Hauser-Feshbach statistical model, and their results were compared with the available experimental data. Two dominant ingredients to perform the statistical calculation were the level densities described by the Gilbert-Cameron approach with an improved systematics, and the gamma-ray transmission coefficients described by gamma-ray strength functions. Although various gamma-ray strength functions with a Lorentzian formula have been developed by using the photonuclear data or a microscopic analysis, they have failed to reproduce the occasional anomalous bumps observed near or below a neutron binding. In this work, we could reproduce the bumps well by adding a Lorentzian with an energy-temperature dependent width into a giant electric dipole resonance with an enhanced generalized Lorentzian. In addition, we introduced a correction function so as to compensate for the shortcomings due to missing levels or level-cuts.</P>

Neutron-Induced Data Evaluation for Selected Fission Products

LEE, Yong Deok,PARK, Joo-Hwan,LEE, Jung Won Atomic Energy Society of Japan 2008 Journal of nuclear science and technology Vol.45 No.6

<P>The neutron-induced cross section data for 19 high-priority fission products were evaluated in the fast energy region using the following models: spherical and deformed optical model, multi-step compound and direct (MSC and MSD), pre-equilibrium exiton and Hauser-Feshbach models with a width fluctuation. The calculations were compared to recently measured data and to currently evaluated files. The results in the fast energy region were converted into the ENDF-6 format and merged with the evaluated resonance part. For consistency, the background was included with the merging energy region. The nuclear data set involves the (n, tot), (n, n), (n, n′), (n, 2n), (n, 3n), (n, nα), (n, np), (n, γ), (n, p), and (n, α) cross sections from the thermal to 20 MeV energy range. The results showed good agreement with the measured data. A considerable improvement was achieved for most reactions and nuclei. Format and physics checking codes were applied to each full data set and the NJOY code was used to perform a processing check on the individual cross sections.</P>

Analysis of X-ray spectra in 14.5-GHz ECR ion source for optimizing operation conditions

Lee, Cheol Ho,Chang, Dae-Sik,Oh, Byung-Hoon,Kim, Yong-Kyun Atomic Energy Society of Japan 2016 Journal of nuclear science and technology Vol.53 No.12

<P>An electron cyclotron resonance (ECR) ion source operating at 14.5 GHz was developed for the generation of charged ions at the Korea Atomic Energy Research Institute (KAERI). Experiments were carried out to study the plasma inside the ECR ion source by analyzing the X-ray spectra generated by it. The X-ray energy distribution and electron energy inside the plasma chamber are influenced by the status of the heated plasma. That status depends on various operation parameters such as microwave power, injected gas-pressure, and solenoid and trim coil currents. X-ray spectra were recorded to find the correlation between the plasma and the X-rays for variations in the operation parameters. A standard NaI(Tl) detector was used for that purpose. The X-ray energy distribution was studied in the range of 100-500 W for radiofrequency power. The influence of the injected gas pressure and the mirror ratio in the emission of X-rays were analyzed.</P>

Benchmark Calculations of a Radiation Heat Transfer for a CANDU Fuel Channel Analysis using the CFD Code

KIM, Hyoung Tae,RHEE, Bo Wook,PARK, Joo Hwan Atomic Energy Society of Japan 2006 Journal of nuclear science and technology Vol.43 No.11

<P>To justify the use of a commercial Computational Fluid Dynamics (CFD) code for a CANDU fuel channel analysis, especially for the radiation heat transfer dominant conditions, the CFX-10 code is tested against three benchmark problems which were used for the validation of a radiation heat transfer in the CANDU analysis code, a CATHENA. These three benchmark problems are representative of the CANDU fuel channel configurations from a simple geometry to a whole fuel channel geometry. For the solutions of the benchmark problems, the temperature or the net radiation heat flux boundary conditions are prescribed for each radiating surface to determine the radiation heat transfer rate or the surface temperature, respectively by using the network method.The Discrete Transfer Model (DTM) is used for the CFX-10 radiation model and its calculation results are compared with the solutions of the benchmark problems. The CFX-10 results for the three benchmark problems are in close agreement with these solutions, so it is concluded that the CFX-10 with a DTM radiation model can be applied to the CANDU fuel channel analysis where a surface radiation heat transfer is a dominant mode of the heat transfer.</P>

Out-of-pile and In-pile Perfomance of Advanded Zirconium Alloys (HANA) for High Burn-up Fuel

JEONG, Yong Hwan,PARK, Sang-Yun,LEE, Myung-Ho,CHOI, Byung-Kwon,BAEK, Jong-Hyuk,PARK, Jeong-Yong,KIM, Jun-Hwan,KIM, Hyun-Gil Atomic Energy Society of Japan 2006 Journal of nuclear science and technology Vol.43 No.9

<P>The performance of the advanced Zr alloys (HANA) for a high burn-up fuel has been evaluated in the out-of-pile and in-pile conditions. The corrosion resistance of the HANA claddings was superior to Zicaloy-4 in a PWR-simulating loop condition. The improved corrosion resistance of the HANA claddings was attributed to the fine distribution of the precipitate. HANA claddings showed a higher creep resistance as compared to Zircaloy-4 from the thermal creep test. The deformation behavior of HANA in a LOCA condition was similar to Zircaloy-4. Threshold ECR value of HANA was higher than the conventional value of 17% in Zircaloy-4, which is mainly due to the fact that the Nb decreases the oxidation rate as well as the hydrogen pickup. Fretting wear test revealed that HANA claddings have a similar wear resistance to Zircaloy-4. From the irradiation test up to burn-up of about 12 GWd/MtU, HANA claddings showed a better corrosion resistance as well as a better creep resistance than Zircaloy-4. The in-pile corrosion resistance of the HANA claddings was improved by 40–50% as compared to Zircaloy-4 on the basis of the oxide thickness measurements.</P>

Turbulence-induced Heat Transfer in PBMR Core Using LES and RANS

LEE, Jung-Jae,YOON, Su-Jong,PARK, Goon-Cherl,LEE, Won-Jae Atomic Energy Society of Japan 2007 Journal of nuclear science and technology Vol.44 No.7

<P>This paper introduces the results of numerical simulations on flow fields and relevant heat transfer in the pebble bed reactor (PBR) core. In the core, since the coolant passes a highly complicated random flow path with a high Reynolds number, an appropriate treatment of the turbulence is required. A set of simple experiments for the flow over a circular cylinder with heat transfer was conducted to finally select the large eddy simulation (LES) and <I>k</I>-ω model among the considering Reynolds-averaged Navier-Stokes (RANS) models for PBR application. Using these models, the PBR cores, whose geometries were simplified to the body-centered cubical (BCC) and face-centered cubical (FCC) structures, were simulated. A larger pressure drop, a more random flow field, a higher vorticity magnitude and a higher temperature at the local hot spots on the pebble surface were found in the results of the LES than in those of RANS for both geometries. In cases of the LES, the flow structures were resolved up to the grid scales. Irregular distributions of the flow and local heat transfer were found in the BCC core, while relatively regular distributions for the FCC core. The turbulent nature of the coolant flow in the pebble core evidently affected the fuel surface temperature distribution.</P>

Structural Integrity Confirmation of a Once-through Steam Generator from the Viewpoint of Flow Instability

KANG, Han-Ok,SEO, Jae-Kwang,KIM, Yong Wan,YOON, Juhyeon,KIM, Keungkoo Atomic Energy Society of Japan 2007 Journal of nuclear science and technology Vol.44 No.1

<P>Helically-coiled once-through steam generators have been utilized for an integral type reactor showing several benefits such as high quality steam generation, geometric compactness, and compensation for a thermal expansion. Steam generator operations with unstable two-phase flow conditions on the tube-side may cause degradation of the tube materials and curtail the lifetime of the component. Based on existing experimental results for a once-through steam generator, its structural integrity was confirmed from the viewpoint of flow instability. The work was composed of three items, the prevention of static instability between the module steam/feedwater pipes, tube inlet orifice sizing against a dynamic instability between the heated coils, and a thermal-cyclic stress analysis for an overall component lifetime evaluation. The static thermo-hydraulic calculation for the steam generator cassette showed that while the prevention of the static instability was satisfied for the power operational mode, special care should be taken during the startup/cooling operational modes. The tube inlet orifice size was determined based on the orifice coefficient concept and existing experimental data for once-through steam generators. The thermal-cyclic stress evaluation for the heated tube revealed that the maximum alternating stress intensity was lower than the allowable fatigue limit value of the tube material.</P>

Separation of Pure LiCl-KCl Eutectic Salt from a Mixture of LiCl-KCl Eutectic Salt and Rare-Earth Precipitates by Vacuum Distillation

EUN, Hee-Chul,YANG, Hee-Chul,CHO, Yong-Jun,PARK, Hwan-Seo,KIM, Eung-Ho,KIM, In-Tae Atomic Energy Society of Japan 2007 Journal of nuclear science and technology Vol.44 No.10

<P>In this study, the vacuum distillation of LiCl-KCl eutectic salt in a mixture of LiCl-KCl eutectic salt and rare-earth precipitates was carried out to evaluate the vaporization characteristics of LiCl-KCl eutectic salt. It was confirmed that the required time for salt vaporization was reduced by a reduction in the pressure. It appeared that the vaporization of LiCl-KCl eutectic salt containing rare-earth precipitates was decreased in comparison with that of pure salt because the salt adhered to the fine particles of the rare-earth precipitates. However, the distillation of the salt was almost achieved by elevating the surface area and further reducing the pressure. The distilled salt from the mixture consisted of 43.7 wt% LiCl and 56.3 wt% KCl. It is thought that the recovered salt can be reused because its composition is similar to the mixed ratio (44.2 wt% LiCl: 55.8 wt% KCl) of the salt used in an electrorefining process.</P>

Analysis of Leaching Behavior of Simulated LILW Glasses by Using the MCC-1 Test Method

KIM, Cheon-Woo,PARK, Jong-Kil,HWANG, Tae-Won Atomic Energy Society of Japan 2011 Journal of nuclear science and technology Vol.48 No.7

<P>Glasses developed for the treatment of low- and intermediate-level radioactive waste (LILW) from nuclear power plants were evaluated by using the Material Characterization Center-1 (MCC-1) leaching method. Tests were conducted at temperatures of 40, 70, and 90°C for three weeks in pH buffer solutions spanning the range from pH 4 to pH 11. Normalized mass losses and forward dissolution rates of major glass elements (B, Na, Al, Si, Co, Cs) were analyzed under each leaching condition. From these data, the forward rate equations depending on pH and temperature were defined using a nonlinear regression method. This equation provided an overall diagram of the leach rate with these parameters (<I>i.e.</I>, pH and temperature). The forward dissolution rates of the glasses were found to have a V-shaped pH dependence. The glasses in the pH ranges were found to have a forward dissolution rate below 10 g/m<SUP>2</SUP>·d, when the temperatures were between 40 and 90°C and the leachant condition was pH 4–11. Except for the DG2 glass, the minimum forward dissolution rate (0.01–1 g/m<SUP>2</SUP>·d) was obtained at approximately pH 7–8. Compared with previously reported results, the developed glasses showed relatively high forward dissolution rates at the neutral region, while showing similar or lower rates compared with other glasses and ceramic waste forms at both extremes of pH.</P>

Electrodeposition Characteristics of Uranium in Molten LiCl–KCl Eutectic and its Salt Distillation Behavior

LEE, Jong-Hyeon,KANG, Young-Ho,HWANG, Sung-Chan,SHIM, Joon-Bo,AHN, Byung-Gil,KIM, Eung-Ho,PARK, Seong-Won Atomic Energy Society of Japan 2006 Journal of nuclear science and technology Vol.43 No.3

<P>Electrorefining experiments with a crucial anode containing U, elemental rare earths; Gd, Nd and Ce or Nd<SUB>2</SUB>O<SUB>3</SUB> were carried out in KCl–LiCl eutectic melt at 500°C. Partitioning behavior of the components according to the applied voltage or current was investigated at various initial U concentrations in a molten salt. Elemental REs concentrations in the cathode deposits increased as the applied voltage decreased at a low initial UCl<SUB>3</SUB> concentration, while they were maintained at a very low level at a higher concentration than ∼6 wt%. Nd<SUB>2</SUB>O<SUB>3</SUB> also shows a similar electrochemical behavior with the elemental REs. This means that rare earth oxides which are inherently incorporated in a U ingot can be partitioned during an electrorefining. The dependency of the applied current on the microstructure of the deposit was also discussed. From a variation of the cell resistance according to the rotation speed of the anode, a concentration polarization was detected below 3 wt% of UCl<SUB>3</SUB> in the molten salt, but not above 6 wt%. Pure uranium ingot was obtained after a distillation of the salt from the deposit.</P>