• Journal of Radiation Protection and Research
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• v.26 no.3
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• pp.225-229
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• 2001

In Korea, an advanced reactor system of 330MWt power called SMART (System integrated Modular Advanced ReacTor) is being developed by KAERI to supply energy for seawater desalination as well as electricity generation. A shielding design of the SMART core with 49 CEDM is established by a two-dimensional discrete ordinates radiation transport analyses. The DORT two-dimensional discrete ordinates transport code is used to evaluate the SMART shielding designs. Three axial regions represent the SMART reactor assembly, each of which is modeled in the R-Z geometry. The BUGLE-96 library is used in the analyses, which consists of 47 neutron and 20 gamma energy groups. The results indicate that the maximum neutron fluence at the bottom of reactor vessel is $5.89 {\times} 10^{17}\;n/cm^2$ and that on the radial surface of reactor vessel is $4.49 {\times} 10^[16}\;n/cm^2$. These results meet the requirement, $1.0 {\times} 10^{20}\;n/cm^2$, in 10 CFR 50.61 and the integrity of SMART reactor vessel during the lifetime of the reactor is confirmed.

• International Journal of Safety
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• v.3 no.1
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• pp.53-58
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• 2004

SMART-P is a promising advanced small and medium category nuclear power reactor. It is an integral type reactor with a sensible mixture of new innovative design features and proven technologies aimed at achieving a highly enhanced safety and improved economics. The enhancement of the safety and reliability is realized by incorporating inherent safety improving features and reliable passive safety systems. The improvement in the economics is achieved through a system simplification, and component modularization. Preliminary safety analyses on selected limiting accidents confirm that the inherent safety improving design characteristics and the safety system of SMART-P ensure the reactor's safety. SMART-P is an advanced integral pressurized water reactor. The purpose of this study is for the safety analysis of the steam generator module pipe break for the SMART-P. The integrity of the fuel rod is the major criteria of this analysis. As a result of this analysis, the safety of the RCS and the secondary system is guaranteed against the module pipe break of a steam generator of the SMART-P.

• Nuclear Engineering and Technology
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• v.31 no.1
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• pp.9-16
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• 1999

A 330 MWt-sized multi-purpose integral-type reactor, SMART is under development in Korea for the use of nuclear energy other than electricity generation. In this study, various radioactive source terms are estimated for SMART. SMART is different from conventional reactor concepts in operation and design. Therefore Specific Calculation method namely recurrence model is used. This model is based on the change rate in the RC radioactivity materials and operational characteristics of SMART Calculation results show tremendously increase of the levels of RC activity because no cleanup of RC and long term operation.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.30 no.9 s.252
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• pp.858-865
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• 2006

SMART Pilot is a multipurpose small capacity integral type reactor. Main coolant pump (MCP) of SMART Pilot is a canned-motor-type axial pump to circulate the primary coolant between nuclear fuel and steam generator in the primary system. The reactor is designed to operate under condition of $310^{\circ}C$ and 14.7MPa. Thus MCP has to be tested under same operating condition as reactor design condition to verify its performance and safety. In present wort a test apparatus to simulate real operating situations of the reactor has been designed and constructed to test MCP. And then functional tests, performance tests, and endurance tests have been carried out upon a prototype MCP. Canned motor characteristics, homologous head/torque curves, coast-down curves, NPSH curves and lift-time performance variations were obtained from the qualification test as well as hydraulic performance characteristics of MCP.

• The KSFM Journal of Fluid Machinery
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• v.15 no.4
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• pp.5-11
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• 2012

To identify the flow characteristics of the SMART reactor, a flow distribution model test and a numerical simulation are performed in KAERI. Among several part of the SMART reactor, the fuel assemblies are simulated using simulators because of the complexity. The geometries of the core in the SMART reactor and simulator are different, but some similarities are maintained such as the ratio of pressure drop in the vertical and cross directions. There are cross flow holes in each core simulator to reproduce the cross flow of SMART fuel assemblies. To know the flow characteristics of the cross flow, numerical analysis is performed. As the cross flow area is decreased, the pressure drop between inlet and outlet is decreased. Also, when the flow imbalance between two core simulators is constant, the cross flow area does not significantly affect the cross flow.

• Proceedings of the KSME Conference
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• 2001.06a
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• pp.441-446
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• 2001

In the integral type reactor, SMART, all the major components such as steam generators, pressurizer and pumps are located inside the single reactor pressure vessel. The objective of this study is to evaluate the structural integrity for RPV of SMART under the postulated pressurized thermal shock by applying the finite element analysis. Input data for the finite element analysis were generated using the commercial code I-DEAS, and the fracture mechanics analysis was performed using the ABAQUS. The crack configurations, the crack aspect ratio and the clad thickness were considered in the parametric study. The effects of these parameters on the reference nil-ductility transition temperature were also investigated.

• Proceedings of the KSME Conference
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• 2001.06d
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• pp.921-926
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• 2001

The reliability and accuracy of the information on control rod position are very important to the reactor safety and the design of the core protection system. In this study, a thorough investigation on the RSPT(Reed Switch Position Transmitter) type control rod position indication system and its actual implementation in the exiting nuclear power plants in Korea was performed first. A design of the control rod position indication system using reed switch for the CEDM on the system-integrated reactor SMART was developed based on the position indicator technology identified through the investigation. The feasibility of the design was evaluated by test of manufactured control rod position indicator using reed switch for SMART.

• Nuclear Engineering and Technology
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• v.54 no.3
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• pp.826-833
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• 2022

A two-phase natural circulation test using SMART integral test loop (SMART-ITL) was conducted to explore thermo-hydraulic phenomena of two-phase natural circulation in the SMART reactor. Specifically, the test examined the natural circulation in the primary loop under a stepwise coolant inventory loss while keeping the core power constant at 5% of the scaled full power. Based on the test results, three flow regimes were observed: single-phase natural circulation (SPNC), two-phase natural circulation (TPNC), and boiler-condenser natural circulation (BCNC). The flow rate remained steady in the SPNC, slightly increased in the TPNC, and dropped abruptly and maintained in the BCNC. Using a natural circulation flow map, the natural circulation characteristic in the SMART-ITL was compared with those in pressurized water reactor simulators. In the SMART-ITL, a BCNC regime appeared instead of siphon condensation and reflux condensation regimes because of the use of once-through steam generators.

• 한국전산유체공학회:학술대회논문집
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• 2001.10a
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• pp.118-125
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• 2001

The performance prediction of SMART MCP was performed using a computational fluid dynamics code. General capacity-head performance curve of MCP, which is provided to other design branches as design input, was obtained and it showed the typical type of axial pump performance curve. When four MCPs operate in parallel and one of them stops while the others continue to operate, SMART requires reduced power operation. A procedure for predicting the performance of SMART MCP for that case was developed and verified with available experimental data. An analysis based on the developed procedure was performed for two cases; the impeller of sloped MCP is fixed or free to rotate in reverse direction. According to the results, $73\%$ flow rate of normal operation enters the reactor core in the case of the locked impeller. In case of the impeller free rotation, the flow rate entering the reactor core is $62.8\%$.

• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.741-751
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• 2021

There has been a deep interest in trying to find better-performing fuel clad motivated by the desire to decrease the likelihood of the reactor barrier failure like what happened in Fukushima in recent years. In this study, the effect of move towards accident tolerant fuel (ATF) cladding as the most attracting concept for improving reactor safety is investigated for SMART modular reactor. These reactors have less production cost, short construction time, better safety and higher power density. The SiC and FeCrAl materials are considered as the most potential candidate for ATF cladding, and the results are compared with Zircaloy cladding material from reactor physics point of view. In this paper, the calculations are performed by generating PMAX library by DRAGON lattice physics code to be used for further reactor core analysis by PARCS code. The differential and integral worth of control and safety rods, reactivity coefficient, power and temperature distributions, and boric acid concentration during the cycle are analyzed and compared from the conventional fuel cladding. The rod ejection accident (REA) is also performed to study how the power changed in response to presence of the ATF cladding in the reactor core. The key quantitative finding can be summarized as: 20 ℃ (3%) decrease in average fuel temperature, 33 pcm (3%) increase in integral rod worth and cycle length, 1.26 pcm/℃ (50%) and 1.05 pcm/℃ (16%) increase in reactivity coefficient of fuel and moderator, respectively.