The purpose of this study is to provide technical issues in upgrade and modification of fuel handling equipment at operating nuclear power plants. The improvement for safety function and performance enhancement of fuel handling equipment has been going on for 20 years since the early 2000’s. This improvement is recently focused on the replacement of components through the performance analysis and the operation and maintenance plan based on replacement cycle of its component. Additionally, it is required to secure spare parts so that it can be operated at all times with compatibility and standardization to other domestic nuclear power plants. The fuel handling equipment is consisted of refueling machine, upender and carriage of fuel transfer system, spent fuel handling machine, new fuel elevator and various tools, and the equipment are linked in systematic interlocks. Fuel handling is a critical task during a nuclear power plant refueling outage. Even minor component defects may stop operation of the whole system and have a significant impact on the overall system process. To achieve this goal, major components that are expected to be replaced for reliable operation are summarized as follows; 1) motor assembly with AC servomotors and driver for bridge, trolley and hoist of refueling machine and spent fuel handling machine, 2) winch motor and drive for upender and carriage of fuel transfer system, 3) operator control console with a HMI PC base PLC (Programmable Logic Controller) control system, 4) positioning and load weighing sensors such as an encoder and a load cell with its support for periodic calibration and maintenance, 5) main power drapped style festoon cable assembly for bridge of refueling machine, 6) pneumatic control assembly for gripper operation of refueling machine, 7) active components (e. g., air motor, hydraulic cylinder and limit switch) to be removable and reinstallable without requiring the water level to be lowered. It is advisable to utilize such various information as it can help to improve reliability of fuel handling as a critical path in upgrade and modification of fuel handling equipment at operating nuclear power plants.

For deep geological repository of the spent nuclear fuel, the fuel assemblies loaded in the storage cask are transferred to the disposal cask and the operation is performed in the fuel handling hot cell at the fuel re-packaging facility. As the fuel handling hot cell shielding is accomplished by the concrete wall and the viewing glass window, the required shielding thickness was evaluated for both materials. The ordinary concrete is applied to hot cell wall and two kinds of glasses, i.e., single layer of lead glass and double layer of lead glass and borosilicate glass, are considered for the viewing glass window. A bare spent PWR fuel assembly exposed to the environment in the hot cell was considered as the neutron and gamma radiation sources. The neutron and gamma transport calculations were performed using the MAVRIC program of the SCALE code system for the dose rate evaluation. The dose limit of 10 μSv/h is applied as the target dose to establish the required shielding thickness. The concrete wall of 94 cm thickness reduces the total dose rate to 6.9 μSv/h, which is the sum of neutron dose and gamma dose. Penetrating the concrete wall, both of the neutron dose and the gamma dose decrease constantly with shield thickness and the gamma dose is always dominant through whole penetrating distance. Single layer lead glass of 74 cm thickness reduces total dose rate to 6.2 μSv/h. Applying double layer shield glass combined of lead glass and borosilicate glass, the total dose rate reduces to 3.6 μSv/h at same shield thickness of 74 cm. Through the shield glass, gamma dose decreases rapidly and neutron dose decreases slowly compared with those for concrete wall. In result, neuron dose becomes dominant on the window glass shielding. The more efficient dose reduction of double layer glass is achieved by the borosilicate glass’s superior neutron shielding power. Thus, the use of double layer glass of lead glass and borosilicate glass is recommended for the viewing glass of the fuel handling hot cell. Finally, it is concluded that about 1 m thick concrete wall and 75 cm thick viewing glass window are sufficient for the radiation shielding of the hot cell at the spent fuel repackaging facility.

The purpose of this study is to provide lessons learned in the experience of improvement work of fuel handling equipment at operating nuclear power plants. The upgrade of fuel handling equipment for safety enhancement and performance improvement has been going on for 15 years since the early 2000’s. The main goal is to increase fuel loading/unloading capability of the equipment from about 2.5 fuel assemblies per hour to more than six (6). It is achieved with sequential operations of three (3) fuel handling equipment, which consists of the refueling machine, the fuel transfer system and the spent fuel handling machine. The scope of the upgrade for fuel handling equipment is summarized as follows. The PC data control system based on PLC for interlocks and high speed motor drive system is introduced for better operating efficiency. The motors and drives for bridge, trolley, and hoist are replaced with AC servomotors and drivers, respectively. The fuel transfer system has an auto-initiation feature operating from the refueling machine or the spent fuel handling machine. The winch motor and drive for the carriage of fuel transfer system is also replaced with AC servomotors and drivers. And some of HPU (hydraulic power units) equipment for each building (reactor containment building and fuel handling building) are replaced to improve their function. The considerations for improvement of fuel handling equipment are as belows. 1) Fuel handling should be consistent with the instructions provided by the fuel designer and/or manufacturer, which are for Standard type fuel and Westinghouse type fuel, used in domestic nuclear power plants. Each fuel has unique design characteristics, which are PLC setpoints for overload and underload, slow speed zones for a bridge, trolley and hoist, allowable acceleration/deceleration value in handling, hoist elevation and manual speed in off-index operation at reactor. 2) The interlock system should be designed in accordance with design criteria specified by the utilities of nuclear power plant. 3) Performance should be improved according to the operating characteristics of the fuel handling equipment. High-speed and optimization of FTS upender and carriage are essential for operating performance so that its modification should be considered first. And the low speed and range in the operation mechanism of the hoist should be designed to comply with guidelines. 4) The accident analysis through self-diagnosis function and operation history in modification at domestic operating nuclear power plants would be good lessons learned. It is advisable to utilize such various information as it can help to improve reliability of nuclear fuel handling operation and power plant operation rate.

The design of nuclear fuel storage and handling area includes the activities related to the storage and inspection before fuel loading, transfer into the reactor, removal of irradiated fuel to the spent fuel storage rack, underwater handling and storage, and handling into a shipping cask. The purpose of this study is to provide the design requirements for the spent fuel pool to be prevented from the loss of cooling water and for heavy load control to prevent any load drop resulting in damage to safetyrelated systems during heavy load handling in accordance with the regulatory guidelines. And another purpose is to review the sizing of minimum wet storage capacity in the spent fuel pool based on the maximum refueling batch from the core during refueling plus a full core off-load of fuel assemblies and the minimum discharge burnup spent fuel storage during the design life of plant requested by the utility. As the results of this study, the current general arrangement for the spent fuel storage and handling area and the minimum storage capacity are evaluated. These can be good recommendations to enhance more safe and efficient if implemented to the new nuclear power plants.

This study provides technical information about the nuclear fuel handling process, which consists of various subprocesses starting from new fuel receipt to spent fuel shipment at a nuclear power plant and the design requirements of fuel handling equipment. The fuel handling system is an integrated system of equipment, tools, and procedures that allow refueling, handling and storage of fuel assemblies, which comprise the fuel handling process. The understanding and reaffirming of detailed code requirements are requested for application to the design of the fuel handling and storage facility. We reviewed the design requirements of the fuel handling equipment for its adequate cooling, prevention of criticality, its operability and maintainability, and for the prevention of fuel damage and radiological release. Furthermore, we discussed additional technical issues related to upgrading the current code requirements based on the modification of the fuel handling equipment. The suggested information provided in this paper would be beneficial to enhance the safety and the reliability of the fuel handling equipment during the handling of new and spent fuel.

본 논문에서는 MDO기법에 의한 핵연료교환장치의 구조해석 단계 중 핵연료교환장치의 휨 변형을 구하는 재료역학해석을 수행하였다. 이는 액체 금속로(LMR) 핵연료교환장치의 기본설계를 위하여 매우 중요하다. 해석대상 핵연료교환장치의 정적구조는 기 수행한 핵연료교환장치의 기구 동역 학 해석 결과를 활용하였다. 네 가지 핵연료교환동작에 대하여 핵연료 봉의 무게를 100㎏에서 500㎏까지 100㎏씩 증가시켜 휨 변형의 크기를 구하였다. 그 결과 회전 중심 축에서 가장 멀리 있는 핵연료 봉을 교환하는 핵연료교환동작에서 최대 휨 변형이 발생함이 밝혀졌다. 또한 이 최대 휨 변형이 발생하는 핵연료교환장치구조에 대하여 부재의 단면두께를 축소하면서, 또 단면형상을 여러 가지로 바꾸면서 휨 변형크기를 구하여 비교하였다. 비교결과 비교대상 단면형상 중에서 중공직사각형 단면이 최소 휨 변형이 발생하는 최적단면형상임이 밝혀졌다.

액체 금속로(LMIR) 핵연료교환장치의 기본설계를 위해서는 여러 분야(예를 들면, 기구학, 동역 학, 재료역학 등)의 해석을 동시에 수행해야 한다. 그러나 이와 같은 해석들은 각각 별개로 연속적으로 수행되는 것이 아니라, 상호 유기적인 연관을 갖고 수행되어야 한다. 이와 같은 해석에 적합한 기법이 MDO 기법이다. 본 논문에서는 MDO기법에 의한 핵연료교환장치 구조해석의 한 단계로 핵연료교환장치의 기구 동역 학 해석을 수행하여 핵연료 교환장치 작동에 대한 기구운동학적 특성 및 동역학적 특성을 분석하였다. 분석결과 해석대상 핵연료교환장치는 예상한대로 원활하게 작동됨이 확인되었다. 아울러 이 분석 결과를 토대로 핵연료교환장치의 정적 휨 변형을 구하기 위한 재료역학해석에서 요구되는 정적구조를 결정하였다.