• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.259.1-259
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• 2001

• Nuclear Engineering and Technology
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• v.55 no.1
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• pp.310-323
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• 2023

The k-eigenvalue neutronics/thermal-hydraulics coupling calculation is a key issue for reactor design and analysis. Jacobian-free Newton-Krylov (JFNK) method, featured with super-linear convergence rate and high efficiency, has been attracting more and more attention to solve the multi-physics coupling problem. However, it may converge to the high-order eigenmode because of the multiple solutions nature of the k-eigenvalue form of multi-physics coupling issue. Based on our previous work, a modified JFNK with a line search method is proposed in this work, which can find the fundamental eigenmode together with thermal-hydraulics feedback in a wide range of initial values. In detail, the existing modified JFNK method is combined with the line search strategy, so that the intermediate iterative solution can avoid a sudden divergence and be adjusted into a convergence basin smoothly. Two simplified 2-D homogeneous reactor models, a PWR model, and an HTR model, are utilized to evaluate the performance of the newly proposed JFNK method. The results show that the performance of this proposed JFNK is more robust than the existing JFNK-based methods.

• Journal of Radiation Protection and Research
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• v.42 no.2
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• pp.114-129
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• 2017

Background: The purpose of this paper is to confirm the event timings and the magnitude of fission product aerosol release from the Fukushima accident. Over a few hundreds of technical papers have been published on the environmental impact of Fukushima Daiichi accident since the accident occurred on March 11, 2011. However, most of the research used reverse or inverse method based on the monitoring of activities in the remote places and only few papers attempted to estimate the release of fission products from individual reactor core or from individual spent fuel pool. Severe accident analysis code can be used to estimate the radioactive release from which reactor core and from which radionuclide the peaks in monitoring points can be generated. Materials and Methods: The basic material used for this study are the initial core inventory obtained from the report JAEA-Data/Code 2012-018 and the given accident scenarios provided by Japanese Government or Tokyo Electric Power Company (TEPCO) in official reports. In this research a forward method using severe accident progression code is used as it might be useful for justifying the results of reverse or inverse method or vice versa. Results and Discussion: The release timing and amounts to the environment are estimated for volatile radioactive fission products such as noble gases, cesium, iodine, and tellurium up to 184 hours (about 7.7 days) after earthquake occurs. The in-plant fission product behaviors and release characteristics to environment are estimated using the severe accident progression analysis code, MELCOR, for Fukushima Daiichi accident. These results are compared with other research results which are summarized in UNSCEAR 2013 Report and other technical papers. Also it may provide the physically based arguments for justifying or suspecting the rationale for the scenarios provided in open literature. Conclusion: The estimated results by MELCOR code simulation of this study indicate that the release amount of volatile fission products to environment from Units 1, 2, and 3 cores is well within the range estimated by the reverse or inverse method, which are summarized in UNSCEAR 2013 report. But this does not necessarily mean that these two approaches are consistent.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.193-198
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• 1996

In this study the RETRAN-03 code was modified to analyze the thermal-hydraulic transients under three-dimensional ship motions for the application to the future marine reactors. First Japanese nuclear ship MUTSU reactor have been analyzed under various ship motions to verify this code. As results, typical thermal-hydraulic characteristics of marine reactors such as flow rate oscillations and S/G water level oscillations are successfully simulated at various conditions.

• Journal of computational fluids engineering
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• v.21 no.4
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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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• v.52 no.8
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• pp.1603-1610
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• 2020

Due to ever-growing advancements in computers and relatively easy access to them, many efforts have been made to develop high-fidelity, high-performance, multi-physics tools, which play a crucial role in the design and operation of nuclear reactors. For this purpose in this study, the neutronic Monte Carlo and thermal-hydraulic sub-channel codes entitled MCNP and COBRA-EN, respectively, were applied for external coupling with each other. The coupled code was validated by code-to-code comparison with the internal couplings between MCNP5 and SUBCHANFLOW as well as MCNP6 and CTF. The simulation results of all code systems were in good agreement with each other. Then, as the second problem, the core of the VVER-1000 v446 reactor was simulated by the MCNP4C/COBRA-EN coupled code to measure the capability of the developed code to calculate the neutronic and thermohydraulic parameters of real and industrial cases. The simulation results of VVER-1000 core were compared with FSAR and another numerical solution of this benchmark. The obtained results showed that the ability of the MCNP4C/COBRA-EN code for estimating the neutronic and thermohydraulic parameters was very satisfactory.

• Nuclear Engineering and Technology
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• v.54 no.2
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• pp.532-545
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• 2022

In-depth convergence analyses for neutronics/thermal-hydraulics (T/H) coupled calculations are performed to investigate the performance of nonlinear methods based on the Fixed-Point Iteration (FPI). A simplified neutronics-T/H coupled system consisting of a single fuel pin is derived to provide a testbed. The xenon equilibrium model is considered to investigate its impact during the nonlinear iteration. A problem set is organized to have a thousand different fuel temperature coefficients (FTC) and moderator temperature coefficients (MTC). The problem set is solved by the Jacobi and Gauss-Seidel (G-S) type FPI. The relaxation scheme and the Anderson acceleration are applied to improve the convergence rate of FPI. The performances of solution schemes are evaluated by comparing the number of iterations and the error reduction behavior. From those numerical investigations, it is demonstrated that the number of FPIs is increased as the feedback is stronger regardless of its sign. In addition, the Jacobi type FPIs generally shows a slower convergence rate than the G-S type FPI. It also turns out that the xenon equilibrium model can cause numerical instability for certain conditions. Lastly, it is figured out that the Anderson acceleration can effectively improve the convergence behaviors of FPI, compared to the conventional relaxation scheme.

• Advances in Energy Research
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• v.5 no.2
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• pp.91-105
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• 2017

Thermal hydraulic (TH) analysis of nuclear power reactors is utmost important. In this way, the numerical codes that preparing TH data in reactor core are essential. In this paper, a subchannel analysis of a Russian pressurized water reactor (WWER1000) core with enhanced numerical code is carried out. For this, in fluid domain, the mass, axial and lateral momentum and energy conservation equations for desired control volume are solved, numerically. In the solid domain, the cylindrical heat transfer equation for calculation of radial temperature profile in fuel, gap and clad with finite difference and finite element solvers are considered. The dependence of material properties to fuel burnup with Calza-Bini fuel-gap model is implemented. This model is coupled with Isotope Generation and Depletion Code (ORIGEN2.1). The possibility of central hole consideration in fuel pellet is another advantage of this work. In addition, subchannel to subchannel and subchannel to rod connection data in hexagonal fuel assembly geometry could be prepared, automatically. For a demonstration of code capability, the steady state TH analysis of a the WWER1000 core is compromised with Thermal-hydraulic analysis code (COBRA-EN). By thermal hydraulic parameters averaging Fuel Assembly-to-Fuel Assembly method, the one sixth (symmetry) of the Boushehr Nuclear Power Plant (BNPP) core with regular subchannels are modeled. Comparison between the results of the work and COBRA-EN demonstrates some advantages of the presented code. Using the code the thermal modeling of the fuel rods with considering the fission gas generation would be possible. In addition, this code is compatible with neutronic codes for coupling. This method is faster and more accurate for symmetrical simulation of the core with acceptable results.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.257.2-257
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.209.1-209
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• 2002