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Thermal-hydraulic analysis of a new conceptual heat pipe cooled small nuclear reactor system

  • Wang, Chenglong;Sun, Hao;Tang, Simiao;Tian, Wenxi;Qiu, Suizheng;Su, Guanghui
    • Nuclear Engineering and Technology
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    • v.52 no.1
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    • pp.19-26
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    • 2020
  • Small nuclear reactor features higher power capacity, longer operation life than conventional power sources. It could be an ideal alternative of existing power source applied for special equipment for terrestrial or underwater missions. In this paper, a 25kWe heat pipe cooled reactor power source applied for multiple use is preliminary designed. Based on the design, a thermal-hydraulic analysis code for heat pipe cooled reactor is developed to analyze steady and transient performance of the designed nuclear reactor. For reactor design, UN fuel with 65% enrichment and potassium heat pipes are adopted in the reactor core. Tungsten and LiH are adopted as radiation shield on both sides of the reactor core. The reactor is controlled by 6 control drums with B4C neutron absorbers. Thermoelectric generator (TEG) converts fission heat into electricity. Cooling water removes waste heat out of the reactor. The thermal-hydraulic characteristics of heat pipes are simulated using thermal resistance network method. Thermal parameters of steady and transient conditions, such as the temperature distribution of every key components are obtained. Then the postulated reactor accidents for heat pipe cooled reactor, including power variation, single heat pipe failure and cooling channel blockage, are analyzed and evaluated. Results show that all the designed parameters satisfy the safety requirements. This work could provide reference to the design and application of the heat pipe cooled nuclear power source.

Analysis of Steady Flow by Main Pipe Arrangement in the Water Distributing Pipe Network (배수관망(配水管網)의 간선배치(幹線配置)에 따른 정류(定流)흐름 해석(解析))

  • Lee, Jeung Seok;Park, Ro Sam;Kim, Jee Hak;Choi, Yun Young;Ahn, Seung Seop
    • Journal of Korean Society of Water and Wastewater
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    • v.13 no.3
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    • pp.73-82
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    • 1999
  • In this study, the optimal analysis for pipe network is performed for the combined ideal pipe network system(CASE 1, CASE 2 and CASE 3) which is composed of 25 nodes, 41 elements, and 1 fixed nodal head with evaluating pressure variation distribution of main and branch in grid composed drainage pipe network. The linear analysis technique used as the analysis method in this study, the KYPIPE being used extensively as the linear technique to design and analysis of pipe network is applied. Firstly, in the analysis of pipe network, the CASE 2 and CASE 3 supply same thing(value) in the result of considering the total flow provided each pipeline, but in the general intension in the case of CASE 2, relative width of supply is more large than CASE 1 and CASE 3. Secondly, in the analysis technique of pipe network, CASE 3 is analysed largest as a result of comparing with same heads, and in the order of their size CASE 2 and CASE 1 were determined but the difference doesn't appear to be obvious. Thirdly, as the result of determining main factor, pressure in the design and analysis of net work. CASE 3 is from Node 3 to 25 than CASE 1 and CASE 2 and it is determined in the order of their size, CASE 2 and CASE 1. Finally, in this study, discharge flow distribution is evaluated in the same condition with 3-type CASE in the case of branch position for designing optimal composed drainage pipe network. As the result of that, branch pipe perform. Therefore, it is thought that the efficient and reasonable management of water supply and sewerage design will be possible if it give all our energies to study at the pipe system design in and out of country in the future.

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DEVELOPMENT OF A WALL-TO-FLUID HEAT TRANSFER PACKAGE FOR THE SPACE CODE

  • Choi, Ki-Yong;Yun, Byong-Jo;Park, Hyun-Sik;Kim, Hee-Dong;Kim, Yeon-Sik;Lee, Kwon-Yeong;Kim, Kyung-Doo
    • Nuclear Engineering and Technology
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    • v.41 no.9
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    • pp.1143-1156
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    • 2009
  • The SPACE code that is based on a multi-dimensional two-fluid, three-field model is under development for licensing purposes of pressurized water reactors in Korea. Among the participating research and industrial organizations, KAERI is in charge of developing the physical models and correlation packages for the constitutive equations. This paper introduces a developed wall-to-fluid heat transfer package for the SPACE code. The wall-to-fluid heat transfer package consists of twelve heat transfer subregions. For each sub-region, the models in the existing safety analysis codes and the leading models in literature have been peer reviewed in order to determine the best models which can easily be applicable to the SPACE code. Hence a wall-to-fluid heat transfer region selection map has been developed according to the non-condensable gas quality, void fraction, degree of subcooling, and wall temperature. Furthermore, a partitioning methodology which can take into account the split heat flux to the continuous liquid, entrained droplet, and vapor fields is proposed to comply fully with the three-field formulation of the SPACE code. The developed wall-to-fluid heat transfer package has been pre-tested by varying the independent parameters within the application range of the selected correlations. The smoothness between two adjacent heat transfer regimes has also been investigated. More detailed verification work on the developed wall-to-fluid heat transfer package will be carried out when the coupling of a hydraulic solver with the constitutive equations is brought to completion.

ASSESSMENT OF CFD CODES USED IN NUCLEAR REACTOR SAFETY SIMULATIONS

  • Smith, Brian L.
    • Nuclear Engineering and Technology
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    • v.42 no.4
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    • pp.339-364
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    • 2010
  • Following a joint OECD/NEA-IAEA-sponsored meeting to define the current role and future perspectives of the application of Computational Fluid Dynamics (CFD) to nuclear reactor safety problems, three Writing Groups were created, under the auspices of the NEA working group WGAMA, to produce state-of-the-art reports on different aspects of the subject. The work of the second group, WG2, was to document the existing assessment databases for CFD simulation in the context of Nuclear Reactor Safety (NRS) analysis, to gain a measure of the degree of quality and trust in CFD as a numerical analysis tool, and to take initiatives to extend the existing databases. The group worked over the period of 2003-2007 and produced a final state-of-the-art report. The present paper summarises the material gathered during the study, illustrating the points with a few highlights. A total of 22 safety issues were identified for which the application of CFD was considered to potentially bring real benefits in terms of better understanding and increased safety. A list of the existing databases was drawn up and synthesised, both from the nuclear area and from other parallel, non-nuclear, industrial activities. The gaps in the technology base were also identified and discussed. In order to initiate new ways of bringing experimentalists and numerical analysts together, an international workshop -- CFD4NRS (the first in a series) -- was organised, a new blind benchmark activity was set up based on turbulent mixing in T-junctions, and a Wiki-type web portal was created to offer online access to the material put together by the group giving the reader the opportunity to update and extend the contents to keep the information source topical and dynamic.

MULTI-SCALE MODELING AND ANALYSIS OF CONVECTIVE BOILING: TOWARDS THE PREDICTION OF CHF IN ROD BUNDLES

  • Niceno, B.;Sato, Y.;Badillo, A.;Andreani, M.
    • Nuclear Engineering and Technology
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    • v.42 no.6
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    • pp.620-635
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    • 2010
  • In this paper we describe current activities on the project Multi-Scale Modeling and Analysis of convective boiling (MSMA), conducted jointly by the Paul Scherrer Institute (PSI) and the Swiss Nuclear Utilities (Swissnuclear). The long-term aim of the MSMA project is to formulate improved closure laws for Computational Fluid Dynamics (CFD) simulations for prediction of convective boiling and eventually of the Critical Heat Flux (CHF). As boiling is controlled by the competition of numerous phenomena at various length and time scales, a multi-scale approach is employed to tackle the problem at different scales. In the MSMA project, the scales on which we focus range from the CFD scale (macro-scale), bubble size scale (meso-scale), liquid micro-layer and triple interline scale (micro-scale), and molecular scale (nano-scale). The current focus of the project is on micro- and meso-scales modeling. The numerical framework comprises a highly efficient, parallel DNS solver, the PSI-BOIL code. The code has incorporated an Immersed Boundary Method (IBM) to tackle complex geometries. For simulation of meso-scales (bubbles), we use the Constrained Interpolation Profile method: Conservative Semi-Lagrangian $2^{nd}$ order (CIP-CSL2). The phase change is described either by applying conventional jump conditions at the interface, or by using the Phase Field (PF) approach. In this work, we present selected results for flows in complex geometry using the IBM, selected bubbly flow simulations using the CIP-CSL2 method and results for phase change using the PF approach. In the subsequent stage of the project, the importance of effects of nano-scale processes on the global boiling heat transfer will be evaluated. To validate the models, more experimental information will be needed in the future, so it is expected that the MSMA project will become the seed for a long-term, combined theoretical and experimental program.

POWER UPRATES IN NUCLEAR POWER PLANTS: INTERNATIONAL EXPERIENCES AND APPROACHES FOR IMPLEMENTATION

  • Kang, Ki-Sig
    • Nuclear Engineering and Technology
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    • v.40 no.4
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    • pp.255-268
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    • 2008
  • The greater demand for electricity and the available capacity within safety margins in some operating NPPs are prompting nuclear utilities to request license modification to enable operation at a higher power level, beyond their original license provisions. Such plant modifications require an in-depth safety analysis to evaluate the possible safety impact. The analysis must consider the thermo hydraulic, radiological and structural aspects, and the plant behavior, while taking into account the capability of the structures, systems and components, and the reactor protection and safeguard systems set points. The purpose of this paper is to introduce international experiences and approaches for implementation of power uprates related to the reactor thermal power of nuclear power plants. The paper is intended to give the reader a general overview of the major processes, work products, issues, challenges, events, and experiences in the power uprates program. The process of increasing the licensed power level of a nuclear power plants is called a power uprate. One way of increasing the thermal output from a reactor is to increase the amount of fissile material in use. It is also possible to increase the core power by increasing the performance of the high power bundles. Safety margins can be maintained by either using fuels with a higher performance, or through the use of improved methods of analysis to demonstrate that the required margins are retained even at the higher power levels. The paper will review all types of power uprates, from small to large, and across various reactor types, including light and heavy water, pressurized, and boiling water reactors. Generally, however, the content of the report focuses on power uprates of the stretch and extended type. The International Atomic Energy Agency (IAEA) is developing a technical guideline on power uprates and side effects of power uprates in nuclear power plants.

ELECTRICAL RESISTANCE IMAGING OF TWO-PHASE FLOW WITH A MESH GROUPING TECHNIQUE BASED ON PARTICLE SWARM OPTIMIZATION

  • Lee, Bo An;Kim, Bong Seok;Ko, Min Seok;Kim, Kyung Youn;Kim, Sin
    • Nuclear Engineering and Technology
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    • v.46 no.1
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    • pp.109-116
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    • 2014
  • An electrical resistance tomography (ERT) technique combining the particle swarm optimization (PSO) algorithm with the Gauss-Newton method is applied to the visualization of two-phase flows. In the ERT, the electrical conductivity distribution, namely the conductivity values of pixels (numerical meshes) comprising the domain in the context of a numerical image reconstruction algorithm, is estimated with the known injected currents through the electrodes attached on the domain boundary and the measured potentials on those electrodes. In spite of many favorable characteristics of ERT such as no radiation, low cost, and high temporal resolution compared to other tomography techniques, one of the major drawbacks of ERT is low spatial resolution due to the inherent ill-posedness of conventional image reconstruction algorithms. In fact, the number of known data is much less than that of the unknowns (meshes). Recalling that binary mixtures like two-phase flows consist of only two substances with distinct electrical conductivities, this work adopts the PSO algorithm for mesh grouping to reduce the number of unknowns. In order to verify the enhanced performance of the proposed method, several numerical tests are performed. The comparison between the proposed algorithm and conventional Gauss-Newton method shows significant improvements in the quality of reconstructed images.

PIV measurement and numerical investigation on flow characteristics of simulated fast reactor fuel subassembly

  • Zhang, Cheng;Ju, Haoran;Zhang, Dalin;Wu, Shuijin;Xu, Yijun;Wu, Yingwei;Qiu, Suizheng;Su, G.H.
    • Nuclear Engineering and Technology
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    • v.52 no.5
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    • pp.897-907
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    • 2020
  • The flow characteristics of reactor fuel assembly always intrigue the designers and the experimentalists among the myriad phenomena that occur simultaneously in a nuclear core. In this work, the visual experimental method has been developed on the basis of refraction index matching (RIM) and particle image velocimetry (PIV) techniques to investigate the detailed flow characteristics in China fast reactor fuel subassembly. A 7-rod bundle of simulated fuel subassembly was fabricated for fine examination of flow characteristics in different subchannels. The experiments were performed at condition of Re=6500 (axial bulk velocity 1.6 m/s) and the fluid medium was maintained at 30℃ and 1.0 bar during operation. As for results, axial and lateral flow features were observed. It is shown that the spiral wire has an inhibitory effect on axial flow and significant intensity of lateral flow mixing effect is induced by the wire. The root mean square (RMS) of lateral velocity fluctuation was acquired after data processing, which indicates the strong turbulence characteristics in different flow subchannels.

Performance analysis of automatic depressurization system in advanced PWR during a typical SBLOCA transient using MIDAC

  • Sun, Hongping;Zhang, Yapei;Tian, Wenxi;Qiu, Suizheng;Su, Guanghui
    • Nuclear Engineering and Technology
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    • v.52 no.5
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    • pp.937-946
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    • 2020
  • The aim in the present work is to simulate accident scenarios of AP1000 during the small-break loss-of-coolant accident (SBLOCA) and investigate the performance and behavior of automatic depressurization system (ADS) during accidents by using MIDAC (The Module In-vessel Degradation severe accident Analysis Code). Four types of accidents with different hypothetical conditions were analyzed in this study. The impact on the thermal-hydraulic of the reactor coolant system (RCS), the passive core cooling system and core degradation was researched by comparing these types. The results show that the RCS depressurization becomes faster, the core makeup tanks (CMT) and accumulators (ACC) are activated earlier and the effect of gravity water injection is more obvious along with more ADS valves open. The open of the only ADS1-3 can't stop the core degradation on the basis of the first type of the accident. The open of ADS1-3 has a great impact on the injection time of ACC and CMT. The core can remain intact for a long time and the core degradation can be prevent by the open of ADS-4. The all results are significant and meaningful to understand the performance and behavior of the ADS during the typical SBLOCA.

Physics study for high-performance and very-low-boron APR1400 core with 24-month cycle length

  • Do, Manseok;Nguyen, Xuan Ha;Jang, Seongdong;Kim, Yonghee
    • Nuclear Engineering and Technology
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    • v.52 no.5
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    • pp.869-877
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    • 2020
  • A 24-month Advanced Power Reactor 1400 (APR1400) core with a very-low-boron (VLB) concentration has been investigated for an inherently safe and high-performance PWR in this work. To develop a high-performance APR1400 which is able to do the passive frequency control operation, VLB feature is essential. In this paper, the centrally-shielded burnable absorber (CSBA) is utilized for an efficient VLB operation in the 24-month cycle APR1400 core. This innovative design of the VLB APR1400 core includes the optimization of burnable absorber and loading pattern as well as axial cutback for a 24-month cycle operation. In addition to CSBA, an Er-doped guide thimble is also introduced for partial management of the excess reactivity and local peaking factor. To improve the neutron economy of the core, two alternative radial reflectors are adopted in this study, which are SS-304 and ZrO2. The core reactivity and power distributions for a 2-batch equilibrium cycle are analyzed and compared for each reflector design. Numerical results show that a VLB core can be successfully designed with 24-month cycle and the cycle length is improved significantly with the alternative reflectors. The neutronic analyses are performed using the Monte Carlo Serpent code and 3-D diffusion code COREDAX-2 with the ENDF/B-VII.1.