• Nuclear Engineering and Technology
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• v.52 no.8
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• pp.1732-1741
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• 2020

The structural integrity of welded joints in the reactor pressure vessel (RPV) is directly related to the safety of nuclear power plants. The RPV is made from SA508-III steel in a pressurized water reactor. In this study, we investigated the effects of temperature on the tensile and fracture toughness properties of Chinese SA508-III welded joint in different sampling areas in order to provide reference data for structural integrity assessments of RPVs. The specimens used in tensile and fracture toughness tests were fabricated from the base metal (BM), weld metal (WM), and the heat-affected zone (HAZ) in the welded joint. The representative testing temperatures included the ambient temperature (20 ℃), upper shelf temperature (100 ℃), and service temperature (320 ℃). The results showed that temperature greatly affected the fracture toughness (JIC) values for the SA508-III welded joint. The JIC values for BM and HAZ both decreased remarkably from 20 ℃ to 320 ℃. The fracture morphologies showed that the BM and HAZ in the welded joint exhibited fully ductile fracture at 20 ℃, whereas partial cleavage fracture was mixed in ductile fracture mode at 100 ℃ and 320 ℃. The WM exhibited the ductile and cleavage fracture mixed mode at various temperatures, and the JIC values showed slight changes.

• Nuclear Engineering and Technology
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• v.27 no.3
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• pp.336-343
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• 1995

The Self Tuning Regulator(STR) method which is an approach of adaptive control theory, is ap-plied to design the fully automatic power controller of the nonlinear reactor model. The adaptive control represent a proper approach to design the suboptimal controller for nonlinear, time-varying stochastic systems. The control system is based on a third­order linear model with unknown, time-varying parameters. The updating of the parameter estimates is achieved by the recursive extended least square method with a variable forgetting factor. Based on the estimated parameters, the output (average coolant temperature) is predicted one-step ahead. And then, a weighted one-step ahead controller is designed so that the difference between the output and the desired output is minimized and the variation of the control rod position is small. Also, an integral action is added in order to remove the steady­state error. A nonlinear M plant model was used to simulate the proposed controller of reactor power which covers a wide operating range. From the simulation result, the performances of this controller for ramp input (increase or decrease) are proved to be successful. However, for step input this controller leaves something to be desired.

• Nuclear Engineering and Technology
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• v.51 no.7
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• pp.1738-1748
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• 2019

Previous research has shown that the inclined condition has an impact on the natural circulation (natural circulation) mode operation of Floating Nuclear Power Plant (FNPP) mounted on the movable marine platform. Due to its compact structure, small volume, strong maneuverability, the Integral Pressurized Water Reactor (IPWR) is adopted as marine reactor in general. The OTSGs of IPWR are symmetrically arranged in the annular region between the reactor vessel and core support barrel in this paper. Therefore, many parallel natural circulation loops are built between the core and the OTSGs primary side when the main pump is stopped. and the inclined condition would lead to discrepancies of the natural circulation drive head among the OTSGs in different locations. In addition, the flow rate and temperature nonuniform distribution of the core caused by inclined condition are coupled with the thermal hydraulics parameters maldistribution caused by OTSG group operating mode on low power operation. By means of the RELAP5 codes were modified by adding module calculating the effect of inclined, heaving and rolling condition, the simulation model of IPWR in inclined condition was built. Using the models developed, the influences on natural circulation operation by inclined angle and OTSG position, the transitions between forced circulation (forced circulation) and natural circulation and the effect on natural circulation operation by different OTSG grouping situations in inclined condition were analyzed. It was observed that a larger inclined angle results the temperature of the core outlet is too high and the OTSG superheat steam is insufficient in natural circulation mode operation. In general, the inclined angle is smaller unless the hull is destroyed seriously or the platform overturn in the ocean. In consequence, the results indicated that the IPWR in the movable marine platform in natural circulation mode operation is safety. Selecting an appropriate average temperature setting value or operating the uplifted OTSG group individually is able to reduce the influence on natural circulation flow of IPWR by inclined condition.

• Nuclear Engineering and Technology
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• v.54 no.4
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• pp.1253-1260
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• 2022

This paper reports the control method for the core power of the China initiative Accelerator Driven System (CiADS) facility. In the CiADS facility, an intense external neutron source provided by a proton accelerator coupled to a spallation target is used to drive a sub-critical reactor. Without any control rod inside the sub-critical reactor, the core power is controlled by adjusting the proton beam intensity. In order to continuously change the beam intensity, an adjustable aperture is considered to be used at the Low Energy Beam Transport (LEBT) line of the accelerator. The aperture size is adjusted based on the Proportional Integral Derivative (PID) controllers, by comparing either the setting beam intensity or the setting core power with the measured value. To evaluate the proposed control method, a CiADS core model is built based on the point reactor kinetics model with six delayed neutron groups. The simulations based on the CiADS core model have indicated that the core power can be controlled stably by adjusting the aperture size. The response time in the adjustment of the core power depends mainly on the adjustment time of the beam intensity.

• Tribology and Lubricants
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• v.19 no.1
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• pp.21-25
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• 2003

The 440C stainless steel is used for ball bearings and bevel gears in the control rod drive mechanism for the integral reactor, SMART. The friction characteristics of 400C stainless steel a investigated in sliding motion using the reciprocating tribometer which can simulate the operating conditions of the control rod drive mechanism. Highly purified water is used as lubricant, and the water is heated and pressurized in the autoclave. Friction force on the reciprocating specimens is measured by the load cells and transformed into friction coefficient. It is verified that frictional characteristic of the 440c steel is not drastically changed up to operating temperature and variation of friction coeffcient at operating temperature from room temperature to 160$^{\circ}C$ is within 5%.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.11a
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• pp.263-269
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• 1996

A comprehensive scaling method is proposed for a scaled-down facility simulating SBLOCA in the CARR passive reactor (CP-1300). The present method consists of two stages: scaling methodology, and validation of scaling methodology and code. The present scaling methodology is based on the integral response scaling method. Through sensitivity study, the condensation of the top of the CMT is identified as one of the little-known phenomenon with high importance which should be addressed for the applicability of the code. Using the similarity of the derived scaling parameters, the major component geometries of the scaled-down facility are determined. In the case of 1/4 height and 1/100 area ratio scaling, it is found out that the power ratio is the same as the area ratio, and the present scaling methodology generates the design parameters of the scaled-down facility without any distortion.

• Journal of Korea Technology Innovation Society
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• v.2 no.3
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• pp.118-128
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• 1999

This study attempts to provide an efficient method for managing the Nuclear R&D projects. It should be noted that the technology and experiences to develop a well-balanced plan and mon-itoring for the R&D project are not yet well established when compared with the commercial con-structionand engineering projects. This study recommends that the R&D project manager utilize the scheduling techniques to establish an integral management system taking in to account the un-certainties in the research environment. Based upon the results of a case study for the Liquid Metal Reactor Design Technology Development Project which is led by the KAERI this study also con-firms that it is possible to run the national nuclear R&D projects with the Scheduling & Planning.

• Journal of the Korean Society of Safety
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• v.21 no.4 s.76
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• pp.49-54
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• 2006

The evaluation of elastic-plastic fracture characteristic was investigated in ferrite steel SA 516- Gr70 used for reactor coolant piping elbow and support skirt of pressure vessels. This paper describes the effect of temperature on J-R curve characteristic of this material. The elastic-plastic fracture mechanics parameter J is obtained with unloading compliance method. The test method were analyzed according to ASTM E 813-89 and E 1152-89. Unloading compliance $J_{IC}$ tests were performed on 1 CT specimens at varied temperatures from $25^{\circ}C$ to about $400^{\circ}C$ using a high temperature extensometer. At all temperature, valid $J_{IC}$ measurements could be made and $J_{IC}$ decreased with increasing temperature. SEM fractography schematically illustrates microvoid initiation, growth and coalescence at the tip of a preexisting crack.

• Nuclear Engineering and Technology
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• v.26 no.3
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• pp.420-424
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• 1994

This paper presents an improved water level control scheme for Pressurized Water Reactor(PWR) Steam Generator(S/G) at the low power operation and transient states. To reduce fluctuations of the water level by the swell and shrink phenomena, the scheme adds feedforward terms considering S/G pressure and the feedwater temperature into the conventional proportional-integral feedback controller. The simulation results using the Compact Nuclear Simulator show that smaller level errors and much faster settling time than those of the conventional scheme can be obtained. The proposed algorithm is easily implementable and has a potential for the real applications.

• Nuclear Engineering and Technology
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• v.52 no.8
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• pp.1834-1846
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• 2020

The SMART (System-integrated Modular Advanced ReacTor) is an integral-type small modular reactor developed by KAERI (Korea Atomic Energy Research Institute). This paper discusses the development of a dynamic equivalent model of the SMART control rod drive mechanism that can be efficiently utilized for complicated analysis during the design of the SMART. A semi-empirical approach is used to develop the equivalent model; that is, the equivalent model is defined analytically and verified empirically. Two types of tests, dynamic characteristics tests and seismic loading tests, are conducted for the development and verification of the dynamic equivalent model, respectively. Acceleration response spectra from the seismic analysis based on the developed equivalent model show good agreement with those from the seismic loading tests.