• Nuclear Engineering and Technology
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• 제42권6호
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• pp.636-655
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• 2010

A thermal-hydraulic code, named CUPID, has been being developed for the realistic analysis of transient two-phase flows in nuclear reactor components. The CUPID code development was motivated from very practical needs, including the analyses of a downcomer boiling, a two-phase flow mixing in a pool, and a two-phase flow in a direct vessel injection system. The CUPID code adopts a two-fluid, three-field model for two-phase flows, and the governing equations are solved over unstructured grids with a semi-implicit two-step method. This paper presents an overview of the CUPID code development and assessment strategy. It also presents the code couplings with a system code, MARS, and, a three-dimensional reactor kinetics code, MASTER.

• 대한기계학회논문집B
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• 제30권9호
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• pp.825-833
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• 2006

To investigate the moderator coolability for CANDU-6 reactors, a test facility (HU-KINS) has been manufactured as a 1/8 scaled-down of a calandria tank. In the design of the test facility, a scaling law was developed in such a way to consider the thermal-hydraulic characteristics of a CANDU-6 moderator. The proposed scaling law takes into consideration of the energy conservation, the dynamic similitude such as dimensionless numbers, Archimedes number (Ar) and Reynolds number (Re), and thermal-hydraulic properties similitude. Using this proposed scaling law, the thermal-hydraulic scaling analyses of similar test facilities such as the SPEL (1/10 scale) and the STERN (1/4 scale), have been identified. As a result, in the case of the SPEL, while the energy conservation is well defined, the similarities of Ar and the heat density are not well considered. As for the similarity of the STERN, while both the energy conservation and the characteristics of Ar are well defined, the heat density is not. In the meanwhile, the HU-KINS test facility with 1/8 length scaled-down is well similitude in compliance with all similarities of the energy conservation, the fluid dynamics and thermal-hydraulic properties. To verify the adequacy of the similarities in terms of thermal-hydraulics, a computational fluid dynamic (CFD) analysis has been conducted using the CFX-5 code. As the results of the CFD analyses, the predicted flow patterns and variation of axial properties inside the calandria tank are well consistant with those of previous studies performed with FLUENT and this implies that the present scaling method is acceptable.

• 한국전산유체공학회지
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• 제21권4호
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• pp.71-77
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• 2016

The CUPID code is a transient, three-dimensional, two-fluid, thermal-hydraulic code designed for a component-scale analysis of nuclear reactor components. The primary objective of this study is to assess the applicability of CUPID to single-phase turbulent flow analyses of $2{\times}2$ rod bundle subchannel. The bulk velocity at the inlet varies from 1.0 m/s up to 2.0 m/s which is equivalent to the fully turbulent flow with the range of Re=12,500 to 25,000. Adiabatic single-phase flow is assumed. The velocity profile at the exit region is quantitatively compared with both experimental measurement and commercial CFD tool. Three different boundary conditions are simulated and quantitatively compared each other. The calculation results of CUPID code shows a good agreement with the experimental data. It is concluded that the CUPID code has capability to reproduce the turbulent flow behavior for the $2{\times}2$ rod bundle geometry.

• Nuclear Engineering and Technology
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• 제56권6호
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• pp.2099-2112
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• 2024

This study examined the effects of CRUD on reflood heat transfer behaviors of nuclear fuel rods during a loss-of-coolant-accident (LOCA) in a pressurized water reactor using a best-estimate thermal-hydraulic analysis code. Changes in thermal properties and boiling heat transfer characteristics of the CRUD layer were extensively reviewed, and a set of correction factors to reflect the changes was implemented into the code. A heat structure layer reflecting the effects of CRUDs on the properties was added to the outer surface of the fuel cladding. Numerical simulations were conducted to examine the effects of CRUDs on reflood cooling of overheated fuel rods for representative separate and integral effect tests, FLECHT-SEASET and LOFT. In LOFT analysis, the average cladding temperature was increased due to the low thermal conductivity of CRUD during steady-state operation; however, in both analyses, the peak cladding temperature decreased, and the quenching time was reduced. Obtained results revealed that when the porous CRUD layer is deposited on the fuel cladding, two opposite effects appear. Low thermal conductivity of the CRUD layer always increases fuel temperature during normal operation; however, its hydrophilic porous structures may contribute to accelerated reflood cooling of fuel rods during a LOCA.

• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2743-2759
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• 2020

A detailed computational fluid dynamics (CFD) simulation analysis model was developed using ANSYS CFX 16.1 and analyzed to simulate the basic design and internal flow characteristics of a 180 MW small modular reactor (SMR) with a natural circulation flow system. To analyze the natural circulation phenomena without a pump for the initial flow generation inside the reactor, the flow characteristics were evaluated for each output assuming various initial powers relative to the critical condition. The eddy phenomenon and the flow imbalance phenomenon at each output were confirmed, and a flow leveling structure under the core was proposed for an optimization of the internal natural circulation flow. In the steady-state analysis, the temperature distribution and heat transfer speed at each position considering an increase in the output power of the core were calculated, and the conceptual design of the SMR had a sufficient thermal margin (31.4 K). A transient model with the output ranging from 0% to 100% was analyzed, and the obtained values were close to the T_hot and T_cold temperature difference value estimated in the conceptual design of the SMR. The K-factor was calculated from the flow analysis data of the CFX model and applied to an analysis model in RELAP5/MOD3.3, the optimal analysis system code for nuclear power plants. The CFX analysis results and RELAP analysis results were evaluated in terms of the internal flow characteristics per core output. The two codes, which model the same nuclear power plant, have different flow analysis schemes but can be used complementarily. In particular, it will be useful to carry out detailed studies of the timing of the steam generator intervention when an SMR is activated. The thermal and hydraulic characteristics of the models that applied porous media to the core & steam generators and the models that embodied the entire detail shape were compared and analyzed. Although there were differences in the ability to analyze detailed flow characteristics at some low powers, it was confirmed that there was no significant difference in the thermal hydraulic characteristics' analysis of the SMR system's conceptual design.

• Nuclear Engineering and Technology
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• 제35권1호
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• pp.45-54
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• 2003

A hyperbolic two-phase, two-fluid equation system developed in the previous work has been implemented in an existing nuclear safety analysis code, MARS. Although the implicit treatment of interfacial pressure force term introduced in momentum equation of the hyperbolic equation system is required to enhance the numerical stability, it is very difficult to implement in the code because it is not possible to maintain the existing numerical solution structure. As an alternative, two-step approach with stabilizer momentum equations has been selected. The results of a linear stability analysis by Von-Neumann method show the equivalent stability improvement with fully-implicit solution method. To illustrate the applicability, the new solution scheme has been implemented into the best-estimate thermal-hydraulic analysis code, MARS. This paper also includes the comparisons of the simulation results for the perturbation propagation and water faucet problems using both two-step method and the original solution scheme.

• Nuclear Engineering and Technology
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• 제41권10호
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• pp.1347-1360
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• 2009

A multi-dimensional component for the thermal-hydraulic system analysis code, MARS, was developed for a more realistic three-dimensional analysis of nuclear systems. A three-dimensional and two-fluid model for a two-phase flow in Cartesian and cylindrical coordinates was employed. The governing equations and physical constitutive relationships were extended from those of a one-dimensional version. The numerical solution method adopted a semi-implicit and finite-difference method based on a staggered-grid mesh and a donor-cell scheme. The relevant length scale was very coarse compared to commercial computational fluid dynamics tools. Thus a simple Prandtl's mixing length turbulence model was applied to interpret the turbulent induced momentum and energy diffusivity. Non drag interfacial forces were not considered as in the general nuclear system codes. Several conceptual cases with analytic solutions were chosen and analyzed to assess the fundamental terms. RPI air-water and UPTF 7 tests were simulated and compared to the experimental data. The simulation results for the RPI air-water two-phase flow experiment showed good agreement with the measured void fraction. The simulation results for the UPTF downcomer test 7 were compared to the experiment data and the results from other multi-dimensional system codes for the ECC delivery flow.

• 한국안전학회지
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• 제27권5호
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• pp.229-234
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• 2012

Operator's action time is evaluated from MAAP4 analysis used in conventional probabilistic safety assessment(PSA) of a nuclear power plant. MAAP4 code which was developed for severe accident analysis is too conservative to perform a realistic PSA. A best-estimate code such as RELAP5/MOD3, MARS has been used to reduce the conservatism of thermal hydraulic analysis. In this study, operator's action time of core cooling recovery operation is evaluated by using the MARS code, which its Fussell-Vessely(F-V) value was evaluated as highly important in a small break loss of coolant(SBLOCA) event and loss of component cooling water(LOCCW) event in previous PSA. The main conclusions were elicited : (1) MARS analysis provides larger time window for operator's action time than MAAP4 analysis and gives the more realistic time window in PSA (2) Sufficient operator's action time can reduce human error probability and core damage frequency in PSA.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.379-384
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• 1996

Thermal-hydraulic performance has been analyzed for an advanced reactor core loaded with hexagonal fuel assemblies. Currently available CHF prediction models and data base for triangular lattice bundles have been thoroughly reviewed, and as a result the KfK-3 CHF correlation with limit CHFR of 1.235 has been determined to be most appropriate. The pressure drop model in COBRA-IV-I code has been modified for the analysis of triangular lattice rod bundles. In view of maximizing the thermal margin, the geometry of a hexagonal fuel assembly, such as rod diameter and rod pitch, has been optimized with a fixed fuel assembly cross sectional area The optimum value of the moderator-to-fuel volume ratio is estimated to lie between 0.65 to 1 with 9.5 mm rod diameter. The thermal margin of these hexagonal fuel assemblies in the AP600 core has been evaluated and compared with that of square lattice fuel assemblies such as VANTAGE-5H and KOFA. The analysis result shows that the performances of hexagonal fuel assemblies are more favorable than the square fuel assemblies in the aspect of steady-state overpower margin.

• Nuclear Engineering and Technology
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• 제55권9호
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• pp.3388-3400
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• 2023

Detailed analysis of the neutron pathway through matter inside the nuclear reactor core is exceedingly needed for safety and economic considerations. Due to the constant development of high-performance computing technologies, neutronics analysis using computer codes became more effective and efficient to perform sophisticated neutronics calculations. In this work, a commercial pressurized water reactor (PWR) presented by Virtual Environment for Reactor Applications (VERA) Core Physics Benchmark are modeled and simulated using a high-fidelity simulation of OpenMC code in terms of criticality and fuel pin power distribution. Various problems have been selected from VERA benchmark ranging from a simple two-dimension (2D) pin cell problem to a complex three dimension (3D) full core problem. The development of the code capabilities for reactor physics methods has been implemented to investigate the accuracy and performance of the OpenMC code against VERA SCALE codes. The results of OpenMC code exhibit excellent agreement with VERA results with maximum Root Mean Square Error (RMSE) values of less than 0.04% and 1.3% for the criticality eigenvalues and pin power distributions, respectively. This demonstrates the successful utilization of the OpenMC code as a simulation tool for a whole core analysis. Further works are undergoing on the accuracy of OpenMC simulations for the impact of different fuel types and burnup levels and the analysis of the transient behavior and coupled thermal hydraulic feedback.