• Nuclear Engineering and Technology
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• 제53권9호
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• pp.2847-2858
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• 2021

When a nuclear reactor with rectangular fuel assemblies runs for a long time, impurities and debris may be taken into coolant channels, which may cause flow blockage, and the blocked fuel assemblies might be destroyed. Therefore, the purpose of this study is to perform a thermal-hydraulic analysis of a rectangular fuel assembly by STAR-CCM+, under the condition of one subchannel with 80% blockage ratio. A rectangular fuel assembly of the International Atomic Energy Agency (IAEA) 10 MW material test reactor (MTR) is chosen. In view of the gasket material taken into the coolant channel is close to the single side of the coolant channel, in the flow blockage accident of the Oak Ridge Research Reactor (ORRR), a new blockage category called single side blockage is attempted. The blockage positions include inlet, middle and outlet, and the blockage is set as a cuboid. It is found by simulations that the blockage redistributes the mass flow rate, and large vortices appear locally. The peak temperature of the cladding is maximum, when the blockage is located at the single side of the coolant channel inlet, and no boiling occurs in all blockage cases. Moreover, as the height of the blockage increases, the damage caused by the blockage increases slightly.

• Nuclear Engineering and Technology
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• 제53권2호
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• pp.455-465
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• 2021

The Molten Salt Reactor (MSR), one of the six advanced reactor types of the 4th generation nuclear energy systems, has many impressive features including economic advantages, inherent safety and nuclear non-proliferation. This paper introduces a system analysis code named TREND, which is developed and used for the steady and transient simulation of MSRs. The TREND code calculates the distributions of pressure, velocity and temperature of single-phase flows by solving the conservation equations of mass, momentum and energy, along with a fluid state equation. Heat structures coupled with the fluid dynamics model is sufficient to meet the demands of modeling MSR system-level thermal-hydraulics. The core power is based on the point reactor neutron kinetics model calculated by the typical Runge-Kutta method. An incremental PID controller is inserted to adjust the operation behaviors. The verification and validation of the TREND code have been carried out in two aspects: detailed code-to-code comparison with established thermal-hydraulic system codes such as RELAP5, and validation with the experimental data from MSRE and the CIET facility (the University of California, Berkeley's Compact Integral Effects Test facility).The results indicate that TREND can be used in analyzing the transient behaviors of MSRs and will be improved by validating with more experimental results with the support of SINAP.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2000년도 춘계학술대회논문집
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• pp.679-686
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• 2000

Reactor Noise is defined as the fluctuations of measured instrumentation signals during full-power operation of reactor which have informations on reactor system dynamics such as neutron kinetics, thermal-hydraulics, and structural dynamics. Reactor noise analyses of ex-core neutron detector signals have been performed to monitor the vibration modes of reactor internals such as fuel assembly and Core Support Barrel in Yonggwang 3&4 Nuclear Power Plant. A real time mode separation technique have been developed and applied for the analyses. It has been found that the first vibration mode frequency of the fuel assembly was around 2.5 Hz, the beam and shell mode frequencies of CSB(Core Support Barrel) 8 Hz and 14.5 Hz, respectively. Also the analyses data base have been constructed for the continuous monitoring and diagnose of the reactor internals.

• Nuclear Engineering and Technology
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• 제44권3호
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• pp.249-260
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• 2012

Main focus of the combined nuclear research activities at Aachen University (RWTH) and the Research Center J$\ddot{u}$lich (J$\ddot{U}$LICH) is the experimental and analytical investigation of containment phenomena and processes. We are deeply convinced that reliable simulations for operation, design basis and beyond-design basis accidents of nuclear power plants need the application of so-called lumped-parameter (LP) based codes as well as computational fluid dynamics (CFD) codes in an indispensable manner. The LP code being used at our institutions is the GRS code COCOSYS and the CFD tool is ANSYS CFX mostly used in German nuclear research. Both codes are applied for safety analyses especially of beyond design accidents. Focal point of the work is containment thermal-hydraulics, but source term relevant investigations for aerosol and iodine behavior are performed as well. To increase the capability of COCOSYS and CFX detailed models for specific features, e.g. recombiner behavior including chimney effect, building condenser, and wall condensation are developed and validated against facilities at different scales. The close connection between analytical and experimental activities is notable and identifying feature of the RWTH/J$\ddot{U}$LICH activities.

• Nuclear Engineering and Technology
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• 제55권6호
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• pp.2315-2324
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• 2023

Space nuclear reactors are becoming popular in deep space exploration owing to their advantages of high-power density and stability. Following the fourth-generation nuclear reactor technology, a conceptual design of the dual drum-controlled space molten salt reactor (D²-SMSR) is proposed. The reactor concept uses molten salt as fuel and heat pipes for cooling. A new reactivity control strategy that combines control drums and safety drums was adopted. Critical physical characteristics such as neutron energy spectrum, neutron flux distribution, power distribution and burnup depth were calculated. Flow and heat transfer characteristics such as natural convection, velocity and temperature distribution of the D²-SMSR under low gravity conditions were analyzed. The reactivity control effect of the dual-drums strategy was evaluated. Results showed that the D²-SMSR with a fast spectrum could operate for 10 years at the full power of 40 kWth. The D²-SMSR has a high heat transfer coefficient between molten salt and heat pipe, which means that the core has a good heat-exchange performance. The new reactivity control strategy can achieve shutdown with one safety drum or three control drums, ensuring high-security standards. The present study can provide a theoretical reference for the design of space nuclear reactors.

• Nuclear Engineering and Technology
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• 제38권1호
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• pp.33-44
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• 2006

This paper presents a review of small-to-medium-sized, pressurized-water-cooled nuclear power reactors whose major primary coolant systems are integrated into a reactor pressure vessel, the concepts categorized as Integrated Primary System Nuclear Reactors (IPSRs). Typical examples of these proposals of interest in this review are CAREM, SMART, IRIS and IMR, all of which are being aimed at the near term deployment. Emphasis is placed on thermal hydraulic aspects. A brief characterization of the IPSR concepts is made and comparisons of plant key parameters are shown. Discussions will follow for the core cooling under rated power conditions and natural circulation heat removal on the basis of the design data available in the public domain.

• Nuclear Engineering and Technology
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• 제55권12호
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• pp.4685-4694
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• 2023

The ability to calculate the material density sensitivity coefficients of power with respect to the material density has broad application prospects for accelerating Monte Carlo-Thermal Hydraulics iterations. The second-order material density sensitivity coefficients for the general Monte Carlo score have been derived based on the differential operator sampling method in this paper, and the calculation of the sensitivity coefficients of cell power scores with respect to the material density has been realized in continuous-energy Monte Carlo code RMC. Based on the power-density sensitivity coefficients, the sensitivity coefficients of power scores to some other physical quantities, such as power-boron concentration coefficients and power-temperature coefficients considering only the thermal expansion, were subsequently calculated. The effectiveness of the proposed method is demonstrated in the power-density coefficients problems of the pressurized water reactor (PWR) moderator and the heat pipe reactor (HPR) reflectors. The calculations were carried out using RMC and the ENDF/B-VII.1 neutron nuclear data. It is shown that the calculated sensitivity coefficients can be used to predict the power scores accurately over a wide range of boron concentration of the PWR moderator and a wide range of temperature of HPR reflectors.

• Nuclear Engineering and Technology
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• 제52권4호
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• pp.721-733
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• 2020

In nuclear engineering, the occurrence of critical heat flux (CHF) is complicated for rod bundle, and it is much more difficult to predict the CHF when it is in natural circulation under motion condition. In this paper, the dryout-type CHF is investigated for the rod bundle in a natural circulation loop under rolling motion condition based on the coupled analysis of subchannel method, a one-dimensional system analysis method and a CHF mechanism model, namely the three-fluid model for annular flow. In order to consider the rolling effect of the natural circulation loop, the subchannel model is connected to the one-dimensional system code at the inlet and outlet of the rod bundle. The subchannel analysis provides the local thermal hydraulic parameters as input for the CHF mechanism model to calculate the occurrence of CHF. The rolling motion is modeled by additional motion forces in the momentum equation. First, the calculation methods of the natural circulation and CHF are validated by a published natural circulation experiment data and a CHF empirical correlation, respectively. Then, the CHF of the rod bundle in a natural circulation loop under both the stationary and rolling motion condition is predicted and analyzed. According to the calculation results, CHF under stationary condition is smaller than that under rolling motion condition. Besides, the CHF decreases with the increase of the rolling period and angular acceleration amplitude within the range of inlet subcooling and mass flux adopted in the current research. This paper can provide useful information for the prediction of CHF in natural circulation under motion condition, which is important for the nuclear reactor design improvement and safety analysis.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.374-379
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• 1997

The thermal margin of CANDU-6 reactor is estimated by the CCP, which is dependent on fuel channel hydraulics and the CHF of fuel bundle. This paper intents to describe the characteristics of CCP behavior for the CANDU-6 channel in which CANFLEX-NU fuel bundles are assumed to be loaded. Also, it includes the thermal margin evaluation of the CANDU-6 channel loaded with a mixed CANFLEX-NU and 37-element fuel bundles as a simulation of the partial loading of CANFLEX-NU fuel bundle in the CANDU-6 reactor. For the mixed fuel channels, the effects of axial flux distribution(AFD) on CCP were investigated by using the AFD tilted in the downstream. The CCP of CANFLEX-NU fuel bundle was found to be improved by the CHF enhancement, despite of the slight flow decrease, in case of both full and partial loading, compared with those of a standard 37-element fuel bundle.

• Nuclear Engineering and Technology
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• 제37권5호
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• pp.469-478
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• 2005

The subchannel analysis capability of MARS, a multi-dimensional thermal-hydraulic system code, has been enhanced. In particular, the turbulent mixing and void drift models for the flow-mixing phenomena in rod bundles were improved. Then, the subchannel analysis feature was combined with the existing coupled system thermal-hydraulics (T/H) and 3D reactor kinetics calculation capability of MARS. These features allow for more realistic simulations of both the hot channel behavior and the global system T/H behavior. Using the coupled features of MARS, a coupled analysis of a main steam line break (MSLB) is carried out for demonstration purposes. The results of the calculations are very reasonable and promising.