• Nuclear Engineering and Technology
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• 제51권6호
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• pp.1487-1503
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• 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
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• 제55권6호
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• pp.2230-2245
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• 2023

KANT (KAIST Advanced Nuclear Tachygraphy) is a PWR core simulator recently developed at Korea Advance Institute of Science and Technology, which solves three-dimensional steady-state and transient multigroup neutron diffusion equations under Cartesian geometries alongside the incorporation of thermal-hydraulics feedback effect for multi-physics calculation. It utilizes the standard Nodal Expansion Method (NEM) accelerated with various Coarse Mesh Finite Difference (CMFD) methods for neutronics calculation. For thermal-hydraulics (TH) calculation, a single-phase flow model and a one-dimensional cylindrical fuel rod heat conduction model are employed. The time-dependent neutronics and TH calculations are numerically solved through an implicit Euler scheme, where a detailed coupling strategy is presented in this paper alongside a description of nodal equivalence, macroscopic depletion, and pin power reconstruction. For validation of the steady, transient, and depletion calculation with pin power reconstruction capacity of KANT, solutions for various benchmark problems are presented. The IAEA 3-D PWR and 4-group KOEBERG problems were considered for the steady-state reactor benchmark problem. For transient calculations, LMW (Lagenbuch, Maurer and Werner) LWR and NEACRP 3-D PWR benchmarks were solved, where the latter problem includes thermal-hydraulics feedback. For macroscopic depletion with pin power reconstruction, a small PWR problem modified with KAIST benchmark model was solved. For validation of the multi-physics analysis capability of KANT concerning large-sized PWRs, the BEAVRS Cycle1 benchmark has been considered. It was found that KANT solutions are accurate and consistent compared to other published works.

• Nuclear Engineering and Technology
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• 제55권11호
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• pp.4274-4281
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• 2023

Blanket is of vital importance for engineering application of the fusion reactor. Nuclear heat deposition in materials is the main heat source in blanket structure. In this paper, the three-dimensional method for thermal-hydraulics/neutronics coupling analysis is developed and applied for the full-scale module of the helium-cooled ceramic breeder tritium breeding blanket (HCCB TBB) designed for China Fusion Engineering Test Reactor (CFETR). The explicit coupling scheme is used to support data transfer for coupling analysis based on cell-to-cell mapping method. The coupling algorithm is realized by the user-defined function compiled in Fluent. The three-dimensional model is established, and then the coupling analysis is performed using the paralleled Coupling Analysis of Thermal-hydraulics and Neutronics Interface Code (CATNIC). The results reveal the relatively small influence of the coupling analysis compared to the traditional method using the radial fitting function of internal heat source. However, the coupling analysis method is quite important considering the nonuniform distribution of the neutron wall loading (NWL) along the poloidal direction. Finally, the structure optimization of the blanket is carried out using the coupling method to satisfy the thermal requirement of all materials. The nonlinear effect between thermal-hydraulics and neutronics is found during the blanket structure optimization, and the tritium production performance is slightly reduced after optimization. Such an adverse effect should be thoroughly evaluated in the future work.

• Nuclear Engineering and Technology
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• 제46권4호
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• pp.481-488
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• 2014

An analysis of the uncertainty quantification related to LBLOCA using the Monte-Carlo calculation has been performed and compared with the tolerance level determined by the Wilks' formula. The uncertainty range and distribution of each input parameter associated with the LOCA phenomena were determined based on previous PIRT results and documentation during the BEMUSE project. Calulations were conducted on 3,500 cases within a 2-week CPU time on a 14-PC cluster system. The Monte-Carlo exercise shows that the 95% upper limit PCT value can be obtained well, with a 95% confidence level using the Wilks' formula, although we have to endure a 5% risk of PCT under-prediction. The results also show that the statistical fluctuation of the limit value using Wilks' first-order is as large as the uncertainty value itself. It is therefore desirable to increase the order of the Wilks' formula to be higher than the second-order to estimate the reliable safety margin of the design features. It is also shown that, with its ever increasing computational capability, the Monte-Carlo method is accessible for a nuclear power plant safety analysis within a realistic time frame.

• Nuclear Engineering and Technology
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• 제29권1호
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• pp.35-44
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• 1997

Critical too-Phase flow rates of subcooled water through Short Pipes (L 140039n) with small diameters (D$\leq$7.15 min) have been experimentally investigated for wide ranges of subcooling (0~199$^{\circ}C$) and pressure (0.5~2.0 MPa). To examine the effects of various parameters (i.e., the location of flashing inception, the degree of subcooling, the stagnation temperature and pressure, and the pipe size) on the critical two-phase flow rates of subcooled water through short pipes with small diameters, a total of 135 runs were made for various combinations of test parameters using four different L/D test sections. Experimental results that show effect of various parameters on subcooled critical two phase flow rates are presented in the form of graphs such as the dimensionless mass flux ( $G^{*}$) versus the dimensionless subcooling ( $T_{sub}$$^{*}$) curve. An empirical correlation expressed in terms of a dimensionless subcooling is also obtained for subcooled two-phase flow rates through present test sections. Comparisons between the mass fluxes calculated by present correlation and a total of 755 selected experimental data points of 9 different investigators show that the agreement is fairly good except for very low subcooling data obtained from small L/D (less than 10) orifices.s.s.s.

• Nuclear Engineering and Technology
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• 제29권1호
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• pp.45-57
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• 1997

The direct contact condensation phenomenon, which occurs when steam is injected into the subcooled water, has been experimentally investigated. Two plume shapes in the stable condensation regime are found to be conical and ellipsoidal shapes depending on the steam mass flux and the liquid subcooling. Divergent plumes, however, are found when the subcooling is relatively small. The measured expansion ratio of the maximum plume diameter to the injector inner diameter ranges from 1.0 to 2.3. By means of fitting a large amount of measured data, an empirical correlation is obtained to predict the steam plume length as a function of a dimensionless steam mass flux and a driving potential for the condensation process. The average heat transfer coefficient of direct contact condensation has been found to be in the range 1.0~3.5 ㎿/$m^2$.$^{\circ}C$. Present results show that the magnitude of the average condensation heat transfer coefficient depends mainly on the steam mass fin By using dynamic pressure measurements and visual observations, six regimes of direct contact condensation have been identified on a condensation regime map, which are chugging, transition region from chugging to condensation oscillation, condensation oscillation, bubbling condensation oscillation, stable condensation, and interfacial oscillation condensation. The regime boundaries are quite clearly distinguishable except the boundaries of bubbling condensation oscillation and interfacial oscillation condensation.

• Nuclear Engineering and Technology
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• 제54권3호
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• pp.888-895
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• 2022

Lately, ocean nuclear power plants have attracted attention as one of diverse uses of nuclear power plants. Because ocean nuclear power plants are movable or transportable, it is necessary to analyze the thermal hydraulics in a moving frame of reference, and computer codes have been developed to predict thermal hydraulics in large moving systems. The purpose of this study is to incorporate a three dimensional dynamic motion model into the SPACE code (Safety and Performance Analysis CodE) so that the code is able to analyze thermal hydraulics in an ocean nuclear power plant. A rotation system that describes three-dimensional rotations about an arbitrary axis was implemented, and modifications were made to the one-dimensional momentum equations to reflect the rectilinear and rotational acceleration effects. To demonstrate the code's ability to solve a problem utilizing a rotational frame of reference, code calculations were conducted on various conceptual problems in the two-dimensional and three-dimensional pipeline loops. In particular, the code results for the three-dimensional pipeline loop with a tilted rotation axis agreed well with the multi-dimensional CFD results.

• Nuclear Engineering and Technology
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• 제44권8호
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• pp.831-846
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• 2012

KAERI has developed a two-phase CFD code, CUPID, for a refined calculation of transient two-phase flows related to nuclear reactor thermal hydraulics, and its numerical models have been verified in previous studies. In this paper, the CUPID code is validated against experiments on the downcomer boiling and moderator flow in a Calandria vessel. Physical models relevant to the validation are discussed. Thereafter, multi-scale thermal hydraulic analyses using the CUPID code are introduced. At first, a component-scale calculation for the passive condensate cooling tank (PCCT) of the PASCAL experiment is linked to the CFD-scale calculation for local boiling heat transfer outside the heat exchanger tube. Next, the Rossendorf coolant mixing (ROCOM) test is analyzed by using the CUPID code, which is implicitly coupled with a system-scale code, MARS.

• 대한기계학회논문집B
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• 제33권9호
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• pp.643-657
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• 2009

This paper introduces the current nuclear experimental research activities in KAERI, the unique nuclear research institute in Korea, and the universities in Korea to solve and assess the issues which have been faced in the design of new reactors such as APR1400, SMART, GEN-IV reactors as well as fusion reactor. Also the experimental evaluations of safety for operating nuclear plants have been presented. The nuclear thermalhydraulic experiments performed in such organizations are classified the fundamental test, the separated effect test, and the integral effect test with ATLAS and SNUF. Introduction is deployed according to institutes. Finally, the future works and the direction of research voyage in the nuclear thermal-hydraulic field were suggested.

• 한국시뮬레이션학회:학술대회논문집
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• 한국시뮬레이션학회 2004년도 춘계학술대회 논문집
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• pp.101-108
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• 2004

KEPRI has developed an Nuclear Steam Supply System(NSSS) thermal-hydraulics simulation program (called ARTS-KORI), based on the best-estimate system code, RETRAN, as a part of the development project for the KORI unit 1 Nuclear Power Plant Simulator. A number of code modifications, such as simplifications and removing of discontinuities of the physical correlations, were made in order to change the RETRAN code as an nuclear Steam Supply System thermal-hydraulics engine in the simulator. Some simplified models and a backup system were also developed. This paper briefly presents the results of non-integrated standalone test of ARTS-KORI.