Session: 04-03: SMRs, Advanced Reactors and Fusion

Paper Number: 130895

130895 - Study on Parameters Design Methodology for Rolling Test of Control Rod Hydraulic Drive System 

Abstract: 

Control rods are the most important power regulation components in nuclear reactors. The startup, shutdown, and the reactivity control of reactors can be achieved by adjusting the position of control rods. The function of control rod drive mechanism (CRDM) is to adjust the position of control rods. Therefore, the reliability of the CRDM is important to the safety operation of reactors. According to different installation locations of the CRDM, it can be divided into two types, as the built-in type CRDM and the built-out type CRDM. Compared with the built-out type CRDM, when the built-in type CRDM is used by reactors, the volume of the reactors can be reduced and the rod ejection accidents can be avoided. Based on the actual requirements of the small modular reactors (SMRs), the built-in type control rod hydraulic drive system (CRHDS) is designed by Institute of Nuclear and New Energy Technology (INET), Tsinghua University. The CRHDS is mainly made up of six components, as the combination valve, the spring box, the rod position indicator, the hydraulic drive mechanism, the control rods, and the buffer. The CRHDS can be used by many kinds of SMRs, including the 5 MW and 200 MW nuclear heating reactors designed by Tsinghua University. In the field of naval architecture and marine engineering, the status of SMRs is irreplaceable. Compared with land-based nuclear power plants, when reactors are used by the marine nuclear power platforms, the responses and the characteristics of the fluid and some key equipment in the reactors should be studied in detail under rolling conditions, for the aims that the ocean environment suitability of the reactors can be evaluated. For the CRHDS, a series of rolling tests are planned by INET, aiming to verify the performance of the CRHDS under the rolling conditions that are defined by the relevant standards. In the rolling test, the structure of the CRHDS can be equivalently represented by a slender straight rod, with the length of 5800 mm and the diameter of 300 mm. It should be noted that the rolling center is located at the bottom surface of the rod in the rolling test. However, the actual rolling center does not coincide with the rolling center of the rolling test. Therefore, based on the locations of the actual rolling center and the rolling center of the rolling test, the input parameters of the rolling test should be determined, aiming to ensure that the load at the peril point of the CRHDS provided by the rolling test can reflect the conditions defined by the standards.

In this paper, the load analysis of the CRHDS in the rolling conditions is conducted. The test parameters are designed based on the equivalent overload coefficients. The parameter combinations used by the rolling test are determined. The results in this paper can provide a theoretical guidance for the rolling test of the CRHDS.

Presenting Author: Yanlin Li Tsinghua University

Presenting Author Biography: I obtained a bachelor's degree in energy and power systems and automation from the school of energy and power engineering of Xi'an Jiaotong University in 2016, and a master's degree in China Institute of Atomic Energy in 2019, during this period, I was engaged in the safety analysis of annular nuclear fuel elements. In September 2019, I entered the Institute of nuclear energy and new energy technology of Tsinghua University to pursue a doctoral degree, At present, I mainly engaged in the research and development of rod position measurement sensor for control rod drive mechanism of nuclear reactor.

Authors:

Yanlin Li Tsinghua University
Benke Qin Tsinghua University
Hanliang Bo Tsinghua University