• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.4111-4124
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• 2022

The control rod drive system is critical to the reactor's reliable operation. The performance of its control system and mechanical system will gradually deteriorate because of operational and environmental stresses, thus increasing the reactor's operational risk. Currently there are few researches on the aging-related degradation of the entire control rod drive system. Because it is difficult to quantify the effect of various environmental stresses and establish an accurate physical model when multiple mechanisms superimposed in the degradation process. Therefore, this paper investigates the aging-related degradation of a control rod drive system by integrating Dynamic Object-Oriented Bayesian Network and Hidden Markov Model. Uncertainties in the degradation of the control system and mechanical system are addressed by using fuzzy theory and the Hidden Markov Model respectively. A system which consists of eight control rod drive mechanisms divided into two groups is used to demonstrate the method. The aging-related degradation of the control rod drive system is analyzed by the Bayesian inference algorithm based on the accelerated life test data, and the impact of different operating schemes on the system performance is also investigated. Meanwhile, the components or units that have major impact on the system's performance are identified at different operational phases. Finally, several essential safety measures are suggested to mitigate the risk caused by the system degradation.

• Geotechnical Engineering
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• v.13 no.1
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• pp.169-182
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• 1997

It is known that some amount of loss in impact energy takes place due to some limitations and problems during the performance of the field SPT. Actual energy level tractsferred to the rod should be measured to correct the SPT-N values tested in the field In this paper, the ratio of energy transferred to the rod through the anvil to impact energy is measured by using sharpy impact test equipment and also analysed by using GRL-WEAP This result is certified and compared with that of field SPT As the results of this study, the average rod energy ratio of the R-P hammer and the Trip hammer is calculated at 0.726 and 0.728 respectively, but it is suggested that 0.72 should be used. By using the hammer energy ratio 64.2% and 75.0% obtained from field measurement, the average energy ratio of the SPT for the R-P hammer is calculated at 46.7% and 54.5% for the Trip hammer.

• Nuclear Engineering and Technology
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• v.26 no.2
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• pp.197-204
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• 1994

In a PWR rod cluster control assembly(RCCA) for shutdown is released upon action of control rod drive mechanism and falls down through the guide thimble by its weight. Drop time and impact velocity of the RCCA are two key parameters with respect to reactivity insertion time and the mechanical integrity of fuel assembly. Therefore, the precise control of drop time and impact velocity is prerequisite to modifying the existing design features of the RCCA and guide thimble or newly designing them. During its falling down into the core, the RCCA is retarded by various forces acting on it such as fluid resistance caused by the RCCA movement, buoyance and mechanical friction caused by contacting inner surface of the guide thimble, etc. However, complicated coupling of the various forces makes it difficult to derive an analytical dynamic equation for the drop time and impact velocity. This paper deals with the development of a computer program containing an analytical dynamic equation applicable to the Korean Fuel Assembly(KOFA). The computer program is benchmarked with an available single control rod drop tests. Since the predicted values are in good agreement with the test results, the computer program developed in this paper can be employed to modify the exiting design features of the RCCA and guide thimble and to develope their new design features for advanced nuclear reactors.

• Proceedings of the KSME Conference
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• 2003.04a
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• pp.101-106
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• 2003

In order to investigate the mechanical behavior of newly developed materials, the evaluation of mechanical properties using small-size specimen is essential. For those purposes, an instrumented impact testing apparatus, which provides the load-displacement curve under impact loading without oscillations, was devised. To develop the test procedure with the setup, the impact behaviors of various kinds of structural materials such as S45C, SCM4, Ti alloys (Ti-6V-4Al) and Zr-based bulk amorphous metal, were investigated through the instrumented Charpy V-notch impact tests. The calibrations of the dynamic load and displacement that was calculated based on the Newton' second law were carried out through the quasi-static load test and the comparison of a directly measured value using a laser displacement meter. Satisfactory results could be obtained. The crack initiation and propagation processes during impact fracture could be well divided on the curve, depending on the intrinsic characteristic of specimen tested; ductile or brittle. The absorbed impact energy in Zr-basd BAM was largely used for crack initiation not for crack propagation process. The fracture surfaces under impact loading showed different feature when compared with the static cases.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.14 no.6
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• pp.512-519
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• 2004

This paper described the free vibration characteristics of Optimized H Type (OHT) spacer grids (SG) supporting the PWR fuel rod. The vibration test and the finite element (FE) analysis are performed under the free boundary condition and the clamped at two points (or three points) in the bottom which is the same one as the experimental condition for the dummy rod continuously supported by spacer grids. A modal test is conducted by the impulse excitation method using an impulse hammer and an accelerometer, and the TDAS module of the I-DEAS software is used to acquire and analyze the sensor signals. The softwares related to the FE analysis are the I-DEAS for the geometrical shape modeling and meshing, and the ABAQUS for solving. The fundamental frequency of the OHT SG by experiment under a clamped condition at two points is 175.18 Hz, and shows a bending mode. We think there is no resonance between the fuel rod and the SG because the SG's frequency is higher than that of the fuel rod existing in the range from 30 to 120 Hz. The fundamental frequency of the SG under the free boundary condition is 349.2 Hz showing a bending mode, and the results between the test and the analysis have a good agreement with maximum 7 ％ in error It is also found that the FE analysis model of the OHT SGs to analyze an impact, a buckling and vibration et al. has been generated with reliability.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.20 no.7
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• pp.658-664
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• 2010

A control element drive mechanism(CEDM) is a reactor regulating system, which inserts, withdraws or maintains a control rod containing a neutron absorbing material within a reactor core to control the reactivity of the core. The ball-screw type CEDM for the integral reactor has a spring-damper system to reduce the impact force due to the scram of the CEDM. This paper describes the experimental results to obtain the drop and damping characteristics of the CEDM. The drop tests are performed by using a drop test rig and a facility. A drop time and a displacement after an impact are measured using a LVDT. The influences of the rod weight, the drop height and the flow area of hydraulic damper on the drop and damping behavior are also estimated on the basis of test results. The drop time of the control element is within 4.5s to meet the design requirement, and the maximum displacement is measured as 15.6 mm. It is also found that the damping system using a spring-hydraulic damper plays a good damper role in the CEDM.

• Journal of KSNVE
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• v.5 no.1
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• pp.95-106
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• 1995

Flow-induced vibration in heat exchanger (or fuel rod) in nuclar power plant can cause dynamic interactions between tubes and tube supports resulting in fretting-wear. To increase the reliability and design life of heat exchanger components, design criteria that establish acceptable limits of vibration and minimize fretting wear are necessary. The fretting-wear rate is dependent upon material combination, contact configuration, environmental conditions and tube-to tube support dynamic interaction. It is demostrated that the fretting -wear rate correlates well with tube-to-support contact force or work rate. The tube-to-support dynamic interaction, which consists of dynamic contact forces and tube motion, is used to relate single-span wear data to real heat exchanger configurations consisting of multi-span tube bundles. This paper describes the test facility to measure tube-to-support dynamic impact force and reports its dynamic characteristics through the four impact tests - a force transduces independent and external impact tests, central ring inside impact test and additional cylinder impact test. Through the tests the impact parameter change dependent upon the material difference of impacting ball is studied, and the impact parameters of Force Transducer Assembly components are measured. And also the dynamic behavior of Force Transducer Assembly is analyzed. The force measurement technique herein is shown to provide a reasonable measure of dynamic contact forces.

• Proceedings of the Korean Society for Technology of Plasticity Conference
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• 1999.08a
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• pp.172-179
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• 1999

As the martensite structure cause fracture failure during drawing from 5.5mm rod to 3.05mm dia. wire without additional heat treatment, the optium cooling condition to inhibit the occurrence of martensite was investigated. In order to get SAE9254+V quality, the effects of alloying element, vanadium on the mechanical properties were investigated. Based upon CCT and TTT curves and the results form cooling test in mill, optimun cooling was found in the condition of the laying head temp of 780$^{\circ}C$ and of the conveyor speed at 0.15m/sec with the whole cover closed. The wire rods produced under the condition showed the best mechanical properties of 120kg/$\textrm{mm}^2$ in TS and 50% in RA, having an excellent drawability. In vanadium added steels, tensile strength was improved without degrading elongation and charpy impact value. That means the strengthening by vanadium is mainly due to the grain refinement by the fine precipitates during tempering process.

• Journal of the Korean Society of Manufacturing Technology Engineers
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• v.7 no.1
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• pp.75-81
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• 1998

We have studied internal quality including chemical compositions, microscopic structure and nonmetallic inclusion of test materials. We have analyzed tensile strength value, hardness value, impact value etc. In analyzing internal quality, all of the test materials showed typical ferrite+pearlite structure. But nonmetallic inclusion showed oxide and sulfide inclusions in medium carbon steels, and sulfide inclusion is S-free cutting steels. In ca+ S-free cutting steels, the calcium aluminate and sulfide complex inclusion had low-melting points as deformation of sulfide and oxide inclusion is existed. It was found that tensile strength and hardness give maximum value in medium carbon steels, where as minimum in Ca + -free cutting steels. But values of elongation, reduction of area impact are reverse. Fracture surface of impact specimen is ductile in free cutting steels but brittle in medium cabon steels.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.25 no.12
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• pp.807-814
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• 2015

A control rod drive mechanism(CRDM) is a reactor regulating system, which inserts, withdraws or maintains a control rod containing a neutron absorbing material within a reactor core to control the reactivity of the core. The top-mounted CRDM for Jordan Research and Training Reactor(JRTR) with 5 MW power has been designed and fabricated based on the HANARO's experience through KAERI and DAEWOO consortium project. This paper describes the performance qualification test results to demonstrate the operability of a prototype and four production CRDMs during the reactor lifetime. The driving performance, the drop performance and the endurance tests for CRDM are carried out at a test rig simulating the actual reactor conditions. A vibration of internal components due to the coolant flow is also measured using a laser vibrometer. As a result, the CRDMs are driven having a good driving performance without a malfunction between command and output signals for the stepping motor. Also, the pure drop time and the impact acceleration are within 0.72 s and 4.2 g to meet the design requirements, and the vibrational displacement of control rod is measured as maximum $5.2{\mu}m$.