• Nuclear Engineering and Technology
• /
• v.51 no.6
• /
• pp.1487-1503
• /
• 2019

The methods and performance of a pin-level nuclear reactor core thermal-hydraulics (T/H) code ESCOT employing the drift-flux model are presented. This code aims at providing an accurate yet fast core thermal-hydraulics solution capability to high-fidelity multiphysics core analysis systems targeting massively parallel computing platforms. The four equation drift-flux model is adopted for two-phase calculations, and numerical solutions are obtained by applying the Finite Volume Method (FVM) and the Semi-Implicit Method for Pressure-Linked Equation (SIMPLE)-like algorithm in a staggered grid system. Constitutive models involving turbulent mixing, pressure drop, and vapor generation are employed to simulate key phenomena in subchannel-scale analyses. ESCOT is parallelized by a domain decomposition scheme that involves both radial and axial decomposition to enable highly parallelized execution. The ESCOT solutions are validated through the applications to various experiments which include CNEN $4{\times}4$, Weiss et al. two assemblies, PNNL $2{\times}6$, RPI $2{\times}2$ air-water, and PSBT covering single/two-phase and unheated/heated conditions. The parameters of interest for validation include various flow characteristics such as turbulent mixing, spacer grid pressure drop, cross-flow, reverse flow, buoyancy effect, void drift, and bubble generation. For all the validation tests, ESCOT shows good agreements with measured data in the extent comparable to those of other subchannel-scale codes: COBRA-TF, MATRA and/or CUPID. The execution performance is examined with a mini-sized whole core consisting of 89 fuel assemblies and for an OPR1000 core. It turns out that it is about 1.5 times faster than a subchannel code based on the two-fluid three field model and the axial domain decomposition scheme works as well as the radial one yielding a steady-state solution for the OPR1000 core within 30 s with 104 processors.

• Nuclear Engineering and Technology
• /
• v.55 no.8
• /
• pp.3017-3029
• /
• 2023

In the late in-vessel phase of a nuclear reactor severe accident, the internal heat transfer and crust evolution during the debris bed melting process have important effects on the thermal load distribution along the vessel wall, and further affect the reactor pressure vessel (RPV) failure mode and the state of melt during leakage. This study coupled the phase change model and large eddy simulation to investigate the variations of the temperature, melt liquid fraction, crust and heat flux distributions during the debris bed melting process in the hypothetical severe accident of HPR1000. The results indicated that the heat flow towards the vessel wall and upper surface were similar at the beginning stage of debris melting, but the upward heat flow increased significantly as the development of the molten pool. The maximum heat flux towards the vessel wall reached 0.4 MW/m². The thickness of lower crust decreased as the debris melting. It was much thicker at the bottom region with the azimuthal angle below 20° and decreased rapidly at the azimuthal angle around 20-50°. The maximum and minimum thicknesses were 2 and 90 mm, respectively. By contrast, the distribution of upper crust was uniform and reached stable state much earlier than the lower crust, with the thickness of about 10 mm. Moreover, the sensitivity analysis of initial condition indicated that as the decrease of time interval from reactor scram to debris bed dried-out, the maximum debris temperature and melt fraction became larger, the lower crust thickness became thinner, but the upper crust had no significant change. The sensitivity analysis of in-vessel retention (IVR) strategies indicated that the passive and active external reactor vessel cooling (ERVC) had little effect on the internal heat transfer and crust evolution. In the case not considering the internal reactor vessel cooling (IRVC), the upper crust was not obvious.

• Nuclear Engineering and Technology
• /
• v.56 no.6
• /
• pp.2029-2038
• /
• 2024

A deep understanding of the characteristics and mechanism of geyser boiling and capillary pumping is necessary to optimize a high-temperature sodium heat pipe. In this work, the Volume of Fluid (VOF) two-phase model and the capillary force model in the mesh wick were used to model the complex phase change and fluid flow in the heat pipe. Computational Fluid Dynamics (CFD) simulations successfully predicted the process of bubble nucleation, growth, aggregation, and detachment from the wall in the liquid pool of the evaporation section of the heat pipe in horizontal and tilted states, as well as the reflux phenomenon of capillary suction within the wick. The accuracy and stability of the capillary force model within the wick were verified. In addition, the causes of geyser boiling in heat pipes were analyzed by extracting the oscillation distribution of heat pipe wall temperature. The results show that adding the capillary force model within the wick structure can reasonably simulate the liquid backflow phenomenon at the condensation; Under the horizontal and inclined operating conditions of the heat pipe, the phenomenon of local dry-out will occur, resulting in a sharp increase in local temperature. The speed of bubble detachment and the timely reflux of liquid sodium (condensate) replenishment in the wick play a vital role in the geyser temperature oscillation of the tube wall. The numerical simulation method and the results of this study are anticipated to provide a good reference for the investigation of geyser boiling in high-temperature heat pipes.

• The Korean Journal of Nuclear Medicine
• /
• v.3 no.1
• /
• pp.19-32
• /
• 1969

Radioisotope renography was carried out in 564 cases consisting of 150 normal controls, 140 hypertensives, 102 hypertensive nephropathys, 62 chronic renal diseases, non-functioning kidneys. It was aimed to study which parameter of the renogram is most applicable to any definite disease of the kidney. The analytical methods adopted were; Tobe, Spencer, Krueger, Matchida and Takeuchi. In the non-functioning kidney groups, the hemograms and serum nitrogen series were also studied to evaluate the relationships between the renograms and renal anemia. The parameters were; time of maximum amplitude (Tmax), half-time of maximum amplitude ($T\frac{1}{2}$), Kac value calculated from these two parameters in Tobe's method, slopes of Band C phase, B/A and B/C values in Spencer's method, total concentration (T.C.), minute concentration (M.C.) and minute excretion (M.E.) in Krueger's method, Matchida's K value and Takeuchi's renal function Index (R.F.I.). Following were the results: 1. In general, marked differences in the patterns of the renogram were observed between the normal controls and nephropathys. In Tobe's method, each parameter showed statistically significant delay or decrease in patients with hypertensive nephropathys and chronic renal diseases. In Spencer's method, slopes of B and C phase and B/C, also showed the statistically significant decrease in patients with hypertension, hypertensive nephropathys and chronic renal diseases. In Krueger's method, M.C. and ME showed the statistically significant differences between the control and patients with hypertension, hypertensive nephropathys and chronic renal diseases, In Matchida's method, K value showed the statistically significant differences between the control and patients with hypertensive nephropathys and chronic renal diseases. 2. It appeared, therefore, that Tobe's $T\frac{1}{2}$, Kac value, Spencer's slopes of Band C phase, B/A, B/C values, Krueger's T.C., M.C., and M.E. values, Matchida's K value are useful for the differentiation of various renal diseases, however, qualitative analysis of the renogram with one or two parameters is not accurate. 3. In bilateral non-functioning kidney groups, a positive correlation between anemia and nitrogen retention was observed, although the quantitative assessment of the degree of non-functioning was impossible.

• Nuclear Engineering and Technology
• /
• v.54 no.11
• /
• pp.4072-4083
• /
• 2022

Ultrasonic impact treatment (UIT) is carried out on the Ni-based alloy stainless steel pipe gas tungsten arc welding (GTAW) girth weld, the differences of microstructure, microhardness and shear strength distribution of the joint before and after ultrasonic shock are studied by microhardness test and shear punch test. The results show that after UIT, the plastic deformation layer is formed on the outside surface of the Ni-based alloy overlayer, single-phase austenite and γ type precipitates are formed in the overlayer, and a large number of columnar crystals are formed on the bottom side of the overlayer. The average microhardness of the overlayer increased from 221 H V to 254 H V by 14.9%, the shear strength increased from 696 MPa to 882 MPa with an increase of 26.7% and the transverse average residual stress decreased from 102.71 MPa (tensile stress) to -18.33 MPa (compressive stress), the longitudinal average residual stress decreased from 114.87 MPa (tensile stress) to -84.64 MPa (compressive stress). The fracture surface has been appeared obvious shear lip marks and a few dimples. The element migrates at the fusion boundary between the Ni-based alloy overlayer and the austenitic stainless steel joint, which is leaded to form a local martensite zone and appear hot cracks. The welded joint is cooled by FA solidification mode, which is forming a large number of late and skeleton ferrite phase with an average microhardness of 190 H V and no obvious change in shear strength. The base metal is all austenitic phase with an average microhardness of 206 H V and shear strength of 696 MPa.

• Nuclear Engineering and Technology
• /
• v.55 no.5
• /
• pp.1519-1526
• /
• 2023

For the first time Aluminium-BariumeZinc oxide nanocomposite (ZABONC) was synthesized by solution combustion method where calcination was carried out at low temperatures (600℃) to study the electromagnetic (EM) (X/γ) radiation shielding properties. Further for characterization purpose standard techniques like PXRD, SEM, UV-VISIBLE, FTIR were used to find phase purity, functional groups, surface morphology, and to do structural analysis and energy band gap determination. The PXRD pattern shows (hkl) planes corresponding to spinel cubic phase of ZnAl₂O4, cubic Ba(NO₃)₂, α and γ phase of Al₂O₃ which clearly confirms the formation of complex nano composite. From SEM histogram mean size of nano particles was calculated and is in the order of 17 nm. Wood and Tauc's relation direct energy band gap calculation gives energy gap of 2.9 eV. In addition, EM (X/γ) shielding properties were measured and compared with the theoretical ones using standard procedures (NaI (Tl) detector and multi channel analyzer MCA). For energy above 356 keV the measured shielding parameters agree well with the theory, while below this value slight deviation is observed, due to the influence of atomic/crystallite size of the ZABONC. Hence synthesized ZABONC can be used as a shielding material in EM (X/γ) radiation shielding.

• Nuclear Engineering and Technology
• /
• v.51 no.3
• /
• pp.884-893
• /
• 2019

Spent fuel racks are steel structures used in the storage of the spent fuel removed from the nuclear power reactor. Rack units are submerged in the depths of the spent fuel pool to keep the fuel cool. Their free-standing design isolates their bases from the pool floor reducing structural stresses in case of seismic event. However, these singular features complicate their seismic analysis which involves a transient dynamic response with geometrical nonlinearities and fluid-structure interactions. An accurate estimation of the response is essential to achieve a safe pool layout and a reliable structural design. An analysis methodology based on the hydrodynamic mass concept and implicit integration algorithms was developed ad-hoc, but some dispersion of results still remains. In order to validate the analysis methodology, vibration tests are carried out on a reduced scale mock-up of a 2-rack system. The two rack mockups are submerged in free-standing conditions inside a rigid pool tank loaded with fake fuel assemblies and subjected to accelerations on a unidirectional shaking table. This article compares the experimental data with the numerical outputs of a finite element model built in ANSYS Mechanical. The in-phase motion of both units is highlighted and the water coupling effect is detailed. Results show a good agreement validating the methodology.

• Journal of the Korean Society of Systems Engineering
• /
• v.14 no.2
• /
• pp.58-66
• /
• 2018

This paper represents a systematic approach aimed at improving the performance of the proportional integral (PI) controller for the Advanced Power Reactor (APR) 1400 Feedwater Control System (FWCS). When the performance of the PI controller offers superior control and enhanced robustness, the steam generator (SG) level is properly controlled. This leads to the safe operation and increased the availability of the nuclear power plant. In this paper, a systems engineering approach is used in order to design a novel PI controller for the FWCS. In the reverse engineering stage, the existing FWCS configuration, especially the characteristics of the feedwater controller as well as the feedwater flow path to each SG from the FWCS, were reviewed and analysed. The overall block diagram of the FWCS and the SG was also developed in the reverse engineering process. In the re-engineering stage, the actual design of the feedwater PI controller was carried out using a genetic algorithm (GA). Lastly, in the validation and verification phase, the existing PI controller and the PI controller designed using GA method were simulated in Simulink/Matlab. From the simulation results, the GA-PI controller was found to exhibit greater stability than the current controller of the FWCS.

• Nuclear Engineering and Technology
• /
• v.51 no.8
• /
• pp.1905-1915
• /
• 2019

This paper investigated the influence of simple support girds on flow, irrespective of having mixing vanes, in a 1 × 3 array rod bundle by using CFD methodology and the most accurate turbulence model which could reflect the actual physics of the flow was determined. In this context, a CFD model was created simulating the experimental studies on a single-phase flow [1] and the results were compared with the experimental data. In the first part of the study, influence of mesh was examined. Tetra, hybrid and poly type meshes were analyzed and convergence study was carried out on each in order to determine the most appropriate type and density. k - ε Standard and RSM LPS turbulence models were used in this section. In the second part of the study, the most appropriate turbulence model that could reflect the physics of the actual flow was investigated. RANS based turbulence models were examined using the mesh that was determined in the first part. Velocity and turbulence intensity results obtained on the upstream and downstream of the spacer grid at -3d_h, +3d_h and +40d_h locations were compared with the experimental data. In the last section of the study, the behavior of flow through the spacer grid was examined and its prominent aspects were highlighted on the most appropriate turbulence model determined in the second part. Results of the study revealed the importance of mesh type. Hybrid mesh having the largest number of structured elements performed remarkably better than the other two on results. While comparisons of numerical and experimental results showed an overall agreement within all turbulence models, RSM LPS presented better results than the others. Lastly, physical appearance of the flow through spacer grids revealed that springs has more influence on flow than dimples and induces transient flow behaviors. As a result, flow through a simple support grid was examined and the most appropriate turbulence model reflecting the actual physics of the flow was determined.

• Korean Journal of Metals and Materials
• /
• v.50 no.9
• /
• pp.659-667
• /
• 2012

Alloy 617 is a Ni-base superalloy and a candidate material for the intermediate heat exchanger (IHX) of a very high temperature gas reactor (VHTR) which is one of the next generation nuclear reactors under development. The high operating temperature of VHTR enables various applications such as mass production of hydrogen with high energy efficiency. Alloy 617 has good creep resistance and phase stability at high temperatures in an air environment. However, it was reported that the mechanical properties decreased at a high temperature in an impure helium environment. In this study, high-temperature corrosion tests were carried out at $850^{\circ}C-950^{\circ}C$ in a helium environment containing the impurity gases $H_2$, CO, and $CH_4$, in order to examine the corrosion behavior of Alloy 617. Until 250 h, Alloy 617 specimens showed a parabolic oxidation behavior at all temperatures. The activation energy for oxidation in helium environment was 154 kJ/mol. The SEM and EDS results elucidated a Cr-rich surface oxide layer, Al-rich internal oxides and depletion of grain boundary carbides. The thickness and depths of degraded layers also showed a parabolic relationship with time. A normal grain growth was observed in the Cr-rich surface oxide layer. When corrosion tests were conducted in a pure helium environment, the oxidation was suppressed drastically. It was elucidated that minor impurity gases in the helium would have detrimental effects on the high-temperature corrosion behavior of Alloy 617 for the VHTR application.