• Transactions of the Korean Society of Automotive Engineers
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• v.9 no.4
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• pp.176-183
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• 2001

It is difficult to find out the kinematic characteristics of a vehicle suspension without the usage of CAE software. The application of CAE software into suspension kinematics and dynamics yields the more precise knowledge on the chassis design. In this study, the influence of the suspension geometry on the handling performances of a large-sized bus is investigated using the DADS software. The front and rear suspension of a large-sized bus are a rigid axle suspension with the four control links. The elastokinematic analysis is performed to evaluate the roll characteristics of the front and rear suspension. The elastokinematic responses are evaluated in terms of the roll center height and roll steer for various geometric parameters. The roll center height is mainly dependent on the vertical displacement of a panhard rod and the vertical displacements of lower control links affect the roll steer of a rear suspension. The parameter study with the change of rear suspension geometry is conducted to investigate the vehicle handling performances. This parameter study shows that the vertical displacement and orientation of a panhard rod influence the handling performances of a large-sized bus significantly.

• Nuclear Engineering and Technology
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• v.50 no.7
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• pp.1037-1042
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• 2018

The present study aims to assess the excess induced reactivity in a Miniature Neutron Source Reactor (MNSR) for a Beyond Design Basis Accident (BDBA) scenario. The BDBA scenario as defined in the Safety Analysis Report (SAR) of the reactor involves sticking of the control rod and filling of the inner and outer irradiation sites with water. At the end of the MNSR core life, 10.95 cm of Beryllium is added to the top of the core as a reflector which affects some neutronic parameters such as effective delayed neutrons fraction (${\beta}_{eff}$), the reactivity worth of inner and outer irradiation sites that are filled with water and the reactivity worth of the control rod. Given those influences and changes, new neutronic calculations are required to be able to demonstrate the reactor safety. Therefore, a validated MCNPX model is used to calculate all neutronic parameters at the end of the reactor core life. The calculations show that the induced reactivity in the BDBA scenario increases at the end of core life to $7.90{\pm}0.01mk$ which is significantly higher than the induced reactivity of 6.80 mk given in the SAR of MNSR for the same scenario but at the beginning of the core's life. Also this value is 3.90 mk higher than the maximum allowable operational limit (i.e. 4.00 mk).

• Nuclear Engineering and Technology
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• v.54 no.3
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• pp.781-789
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• 2022

KHNP had licensed Dynamic Control rod Reactivity Measurement (DCRM) method using detector current signals of PWRs in 2006. The method has been applied to all PWRs in Korea for about 15 years successfully. However, the original method was inapplicable to PWRs using low-sensitivity integral fission chamber as ex-core detectors because of their pulse pile-up and the nonlinearity of the mean-square voltage at low power region. Therefore, to overcome this disadvantage, a modified method, DCRM-EK, was developed using kinetics behavior after equilibrium condition where the pulse counts maintain the maximum value before pulse pile-up. Overall measurement, analysis procedure, and related computer codes were changed slightly to reflect the site test condition. The new method was applied to a total of 15 control rods of 1000 MWe and 1400 MWe PWRs in Korea with worths in the range of 200 pcm -1200 pcm. The results show the average difference of -0.4% and the maximum difference of 7.1% compared to the design values. Therefore, the new DCRM-EK will be applied to PWRs using low sensitivity integral fission chambers, and also can replace the original DCRM when the evaluation fails by big noises present in current or voltage signals of uncompensated/compensated ion chambers.

• 전기의세계
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• v.15 no.5
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• pp.1-5
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• 1966

The stability and safety of operation of a reactor is determined mainly by the sign and magnitude of its reactivity responses to temperature changes. Reactors are subject to temperature fluctuations due to the changes in reactor power and ambient temperature. These temperature fluctuations cause reactivity disturbances through changes in the nuclear and physical properties of the core. Because of these important phenomena by the temperature effects, a large portion of study and testing on a reactor design has been conducted. In this experiment the overall temperature coefficient of the TRIGA MARK-II reactor is measured. The basic procedure is to change the tgemperature of the water moderator, and from the movements of a newly recalibrated control rod(this is necessary due to the effects of fuel burn-up and control rod depression) required to mintain criticality, the reactivity worth of the temperature change is determined. From this measurement, the overall temperature coefficient seems to be smoothly varying, almost a linear function of temperature, and a value of approximately -0.267${\c}$/$^{\circ}C$ can be obtained for an average temperature range from $17.6^{\circ}C$ to $32.5^{\circ}C$.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.3106-3116
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• 2022

The present work presents analysis results of the core design optimizations for an annular, prismatic High Temperature Gas-cooled Reactor (HTGR) with passive decay-heat removal features. Its thermal power is 100 MW_t and the operating temperature is 850 ℃ (1123 K). The neutronic calculations are done for the core with heterogeneous distribution of fuel and burnable poison particles (BPPs) to flatten the reactivity swing and power peaking factor (PPF) during the reactor operation as well as for control rod (CR) insertion into the core to restrain a small excess reactivity less than 1$. The next step of the study is done for evaluation of core reactivity coefficient of temperature.

• Nuclear Engineering and Technology
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• v.45 no.6
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• pp.721-730
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• 2013

Within the framework of the ASTRID project, core design studies are being conducted by the CEA with support from AREVA and EDF. The pre-conceptual design studies are being conducted in accordance with the GEN IV reactor objectives, particularly in terms of improving safety. This involves limiting the consequences of 1) a hypothetical control rod withdrawal accident (by minimizing the core reactivity loss during the irradiation cycle), and 2) an hypothetical loss-of-flow accident (by reducing the sodium void worth). Two types of cores are being studied for the ASTRID project. The first is based on a 'large pin/small spacing wire' concept derived from the SFR V2b, while the other is based on an innovative CFV design. A distinctive feature of the CFV core is its negative sodium void worth. In 2011, the evaluation of a preliminary version (v1) of this CFV core for ASTRID underlined its potential capacity to improve the prevention of severe accidents. An improved version of the ASTRID CFV core (v2) was proposed in 2012 to comply with all the control rod withdrawal criteria, while increasing safety margins for all unprotected-loss-of-flow (ULOF) transients and improving the general design. This paper describes the CFV v2 design options and reports on the progress of the studies at the end of pre-conceptual design phase 1 concerning: - Core performance, - Intrinsic behavior during unprotected transients, - Simulation of severe accident scenarios, - Qualification requirements. The paper also specifies the open options for the materials, sub-assemblies, absorbers, and core monitoring that will continue to be studied during the conceptual design phase.

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.75.2-75
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• 2002

• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.357.1-357
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.105.2-105
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• 1999

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.293-298
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• 1997

The H$_{\infty}$robust controller fur the reactor power control system is designed by use of the mixed weight sensitivity. The system is configured into the typical two-port model with which the weight functions are augmented. Since the solution depends on the weighting functions and the problem is of non-convex, the genetic algorithm is used to determine the weighting functions. The cost function applied in the genetic algorithm permits the direct control of the power tracking performances. In addition, the actual operating constraints such as rod velocity and acceleration can be treated as design parameters. Compared with the conventional approach, the controller designed by the genetic algorithm results in the better performances with the realistic constraints. Also, it is found that the genetic algorithm could be used as an effective tool in the robust design. robust design.