• Nuclear Engineering and Technology
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• 제52권3호
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• pp.477-484
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• 2020

The control rod drive mechanism needs to be debugged after reactor fresh fuel loading. It is of great importance to monitor the subcriticality of this process accurately. A modified method was applied to the subcriticality monitoring process, in which only a single control rod cluster was fully withdrawn from the core. In order to correct the error in the results obtained by Neutron Source Multiplication Method, which is based on one point reactor model, Monte Carlo neutron transport code was employed to calculate the fission neutron distribution, the iterated fission probability and the neutron flux in the neutron detector. This article analyzed the effect of a coarse mesh and a fine mesh to tally fission neutron distributions, the iterated fission probability distributions and to calculate correction factors. The subcriticality before and after modification is compared with the subcriticality calculated by MCNP code. The modified results turn out to be closer to calculation. It's feasible to implement the modified NSM method in large local reactivity addition process using Monte Carlo code based on 3D model.

• 대한기계학회논문집A
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• 제38권6호
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• pp.637-647
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• 2014

국내 가압경수로형 원자로의 압력용기 상부헤드에는 많은 제어봉구동장치(CRDM) 노즐이 분포한다. 이들 노즐은 억지끼워맞춤(Shrink fitting) 방식으로 결합되어 용접 처리 된다. 용접에 의해 발생되는 인장잔류응력은 일차수응력부식균열을 발생시키는 주요 요인이다. 이러한 이유로 최근 15 여 년 동안 관통노즐 용접부 부위에서 균열 발생 사례가 증가하고 있으며, 이를 극복하기 위해 다양한 방안이 모색되고 있다. 또한 용접과정에서 발생되는 불필요한 결함은 일차수응력부식균열(PWSCC)을 가속화 시키는 원인이 되기도 한다. 원자로 제작과정에서 용접에 의한 결함은 보수용접에 의해 즉시 수리가 이루어 진다. 기존의 연구에서는 정상적인 용접과정에서 발생되는 잔류응력을 예측하였으나, 본 연구에서는 용접과정에서 발생되는 결함을 보수하기 위해 실시되는 보수용접이 용접잔류응력에 미치는 영향을 분석하였다.

• 대한기계학회논문집A
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• 제36권10호
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• pp.1235-1239
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• 2012

원자력발전소의 원자로압력용기 상부헤드에는 출력제어 및 정지용 제어봉이 통과하는 노즐이 있으며 이 노즐은 상부헤드 노즐과 J 형태의 홈으로서 용접 되어 있다. 최근 외국의 원자력발전소에서 이 용접영역 주변의 노즐 및 용접부에서 일차수응력부식 균열이 발생한 사례가 보고되고 있다. 본 논문에서는 이 용접부의 용접잔류응력과 운전 중에서의 응력상태를 유한요소해석을 이용하여 평가함으로써 고응력 위치를 확인하고 응력관점에서 균열발생 가능성이 높은 지역을 예측하고자 하였다. 해석결과 용접에 의해서 형성된 잔류응력이 수압시험과 운전조건에 의해 다소 변동되기는 하나 응력분포형태는 큰 변화가 없었다. 전반적으로 노즐내면에서는 용접이 시작되는 지점 주변에서 최대 인장응력이 형성되고 노즐외면에서는 용접이 끝나는 지점 주변에서 최대인장응력이 형성되는 것을 확인하였다.

• 한국자기학회지
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• 제26권5호
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• pp.154-158
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• 2016

원자력발전소의 원자로에는 노심 반응 속도를 제어하기 위하여 제어봉구동장치가 사용된다. 한국원자력연구원의 SMART 원자로는 원자로 가동 중 제어봉집합체의 위치를 확인하기 위하여 제어봉구동장치에 영구자석과 리드스위치로 구성되는 위치지시기가 설치된다. 원자로 가동 온도는 최대 $350^{\circ}C$로 고려되어 설계되며, 영구자석은 원자로 내에 설치된다. 반면에 리드스위치와 전기회로는 원자로 외부에 설치된다. Test coil은 리드스위치의 품질 검증을 위한 장비로서, 코일과 철심으로 구성되어 있다. 본 연구는 리드스위치에 미치는 Test coil과 영구자석의 자기 특성을 비교하고자 수행되었으며, 유한요소 전자기 시뮬레이션을 활용하였다.

• Nuclear Engineering and Technology
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• 제26권2호
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• pp.197-204
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• 1994

원자로운전정지시 사용되는 제어봉집합체는 제어봉구동장치에서 분리되어 핵연료집합체의 안내관으로 자유낙하한다. 이 제어봉집합체의 주요변수로는 낙하시간과 충격속도가 있는데, 낙하시간은 원자로 안전정지와 관계가 있으며, 충격속도는 핵연료집합체의 건전성과 관계가 있다. 따라서, 제어봉 낙하시간과 충격속도의 적절한 결정은 제어봉집합체와 핵연료집합체의 설계에 매우 중요하다. 제어봉집합체는 낙하도중 유체저항이나 마찰력 및 부력과 같은 여러 힘들에 의해 낙하시간이 감소하게 되는데, 이러한 여러가지 힘의 복잡한 결합으로 인해 낙하시간과 충격속도를 해석적으로 유추하는 것은 매우 어렵다. 본 논문에서는 국산핵연료집합체에 적용되는 해석적인 방정식을 포함하고 있는 프로그램을 개발하였고, 이 프로그램을 단일제어봉 낙하시험과 비교하였다. 비교결과 시험 및 해석결과가 잘일치하고 있음으로써 개발된 프로그램의 검증을 확인할 수 있었고, 따라서 이 프로그램이 제어봉및 안내관의 설계변경시 매우 유용하게 사용할 수 있게 되었다.

• 한국소음진동공학회논문집
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• 제20권7호
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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 Magnetics
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• 제21권4호
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• pp.586-592
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• 2016

The magnetic-jack type CRDM withdraws or inserts a control rod assembly from/to the reactor core to control the core reactivity. The CRDM housings form not only the path of the electromagnetic field but also the pressure boundary of a nuclear reactor, and a periodic in-service inspection should be carried out if there are welded or flange jointed parts on the pressure boundary. The in-service inspection is a time-consuming process during the reactor refueling, and moreover it is difficult to perform the inspection over the reactor head. A magnetic motor housing is applied for the current SMART CRDM and has several welding joints, however a nonmagnetic motor housing with fewer or no welding joints may improve the operational efficiency of the nuclear reactor by avoiding or simplifying the in-service inspection process. Prior to the development, the magnetic field transfer efficiency of the nonmagnetic housing was required to be assessed. It was verified and optimized by the electromagnetic analysis of the lifting force estimation. Magnetic flux rings were adopted to improve the efficiency. In this paper, the design and optimization process of a nonmagnetic motor housing with the magnetic flux rings for the SMART CRDM are introduced and the analyses results are discussed.

• 한국소음진동공학회논문집
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• 제25권12호
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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 $5.2{\mu}m$.

• Nuclear Engineering and Technology
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• 제48권5호
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• pp.1083-1095
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• 2016

The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR) has been developed and the validation and verification (V&V) activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1), produced satisfactory results, which were used for the computer codes V&V, and the performance test results of the model pump in sodiumshowed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs) have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.

• 대한용접접합학회:학술대회논문집
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• 대한용접접합학회 2005년도 춘계학술발표대회 개요집
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• pp.185-187
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• 2005

The reactor pressure vessel head of PWR has penetrations for control rod drive mechanism and instrumentation systems. The Primary coolant water and operating temperature can cause the stress-corrosion cracking of these nickel-based alloy penetrations. It is difficult to detect and size flaws such as SCC in the reactor head penetrations using conventional W methods because of complex geometry, Therefore, the utilities are using the TOFD technique for the detection and sizing of the flaw. This study shows the correlation between the ultrasonic wave direction and the orientation of the flaw and the range of flaw depth which can be detected by the TOFD techniques.