Suzhou Nuclear Power Research Institute

Suzhou, China

Suzhou Nuclear Power Research Institute

Suzhou, China
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Patent
CGN Inspection Technology Company, Suzhou Nuclear Power Research Institute, China General Nuclear Power Group and CGN Power Co. | Date: 2017-04-05

The present invention discloses a nondestructive inspection robot for a pressure vessel (1) of a nuclear reactor and an inspection method thereof. The robot comprises a plurality of support legs (22), a plurality of mechanical arms (25a, 25b, 25c, 25d), and a main rotary joint (23); two ends of each support leg (22) are respectively connected onto a stand column assembly (21) and the pressure vessel (1) so that the central axis of the stand column assembly (21) and the axis line of the pressure vessel (1) coincide with each other; the plurality of mechanical arms (25a, 25b, 25c, 25d) are provided with a plurality of probe assemblies (24a, 24b, 24c, 24d) for scanning respective components; the main rotary joint (23) is rotatably connected to a lower part of the stand column assembly (21) about the axial direction of the stand column assembly (21); a second mechanical arm (25b) is rotatably connected to the main rotary joint (23) about a direction perpendicular to the axial direction of the stand column assembly (21) by means of a swinging joint (26), and other mechanical arms are detachably connected to the main rotary joint (23).


Wang L.,Suzhou Nuclear Power Research Institute
Hedongli Gongcheng/Nuclear Power Engineering | Year: 2017

Based the theoretical derivation and the canned motor RCP design, operation and trouble shooting experience, the submerged flywheel friction losses is analyzed and calculated. According to the Double-Acting Pivoted Pad Thrust Bearing tribology design principles, the impact of the submerged flywheel on the AP1000 reactor coolant pump performance is analyzed. © 2017, Editorial Board of Journal of Nuclear Power Engineering. All right reserved.


Patent
CGNPC Inspection Technology Co., Suzhou Nuclear Power Research Institute and China General Nuclear Power Group | Date: 2016-08-24

An intelligent testing method of a non-destructive testing robot based on virtual reality technology, the method comprises the following steps:(1) a non-destructive testing robot for a non-destructive test is installed to a preset position inside the reactor pressure vessel;(2) the position calibration of each degree-of-freedom movement axis is performed after each degree-of-freedom movement axis of the non-destructive testing robot is recovered to an initial state, and a global coordinate system and an axle coordinate system of each degree-of-freedom movement axis are constructed;(3) a corresponding relationship between the simulation model and the actual device;(4) the non-destructive testing robot simulation model transform the position and posture in three-dimension virtual environment on basis of the real-time position and posture information feedback values of each degree-of-freedom movement axle of the non-destructive testing robot acquired thereby, and synchronous movement of the non-destructive testing robot is virtually displayed and controlled for a non-destructive test.


Yang D.,Suzhou Nuclear Power Research Institute | Liu H.,Suzhou Nuclear Power Research Institute
AIP Conference Proceedings | Year: 2017

THERP (Technique for Human Error Rate Prediction) and HCR (Human Cognitive Reliability) are generally analytic methods in HEFA (Human factor event analysis). The two methods have their own features. THERP HCR is two method's combination. After analysis the general overhaul steps of safety shell isolated valve in nuclear power plant, this paper utilize THERP HCR to calculate the error probability of every step and the valve overhaul. At the same time, the error probability of three levels including experts, common workers and novices is analyzed and compared. © 2017 Author(s).


Li G.D.,Suzhou Nuclear Power Research Institute | Li N.,Suzhou Nuclear Power Research Institute
IOP Conference Series: Materials Science and Engineering | Year: 2017

High-strength bolts have been widely used in power plants. However, the high-strength bolts which being employed in pumping station, steel structure and pipeline anti-whip structure have been found delayed fracture for many times in a power plant, this will affect the reliability of steel fracture and bring blow risk caused by falling objects. The high-strength bolt with delayed fracture was carried out fracture analysis, metallurgical analysis, chemical analysis, mechanical analysis, as well as bolts installation analysis, it can be comprehensively confirmed that the direct cause of high-strength bolts delayed fracture is the stress corrosion, and the root cause of high-strength bolts delayed fracture should be the improper installation at the initial and the imperfect routine anti-corrosion maintenance. © Published under licence by IOP Publishing Ltd.


Zhe Y.,Suzhou Nuclear Power Research Institute
IOP Conference Series: Materials Science and Engineering | Year: 2017

There are often mechanical problems of emergency power generation units in nuclear power plant, which bring a great threat to nuclear safety. Through analyzing the influence factors caused by mechanical failure, the existing defects of the design of mechanical support system are determined, and the design idea has caused the direction misleading in the field of maintenance and transformation. In this paper, research analysis is made on basic support design of diesel generator set, main pipe support design and important components of supercharger support design. And this paper points out the specific design flaws and shortcomings, and proposes targeted improvement program. Through the implementation of improvement programs, vibration level of unit and mechanical failure rate are reduced effectively. At the same time, it also provides guidance for design, maintenance and renovation of diesel generator mechanical support system of nuclear power plants in the future. © Published under licence by IOP Publishing Ltd.


Zhan Z.G.,Suzhou Nuclear Power Research Institute
Applied Mechanics and Materials | Year: 2014

This paper introduces the major indexes of the coating performance evaluation (Thickness, elastic modulus, bonding strength, surface quality, residual stress, etc.), and it introduces the application of nondestructive testing methods in order to evaluate each index. © (2014) Trans Tech Publications, Switzerland.


Shan Z.,Suzhou Nuclear Power Research Institute
PSAM 2014 - Probabilistic Safety Assessment and Management | Year: 2014

This paper presents the safety management system designed for CGNPC. It's developed based on the idea adopted from NRC, the Reactor Oversight Process (ROP). A system independent from the classic one is planned to be built with which system, risk management in three level including Unit level, Site level and multi-site level is provided in risk-informed manner using performance indicators and risk significances evaluated from internal and licensee events. The reports are provided both monthly and quarterly to the risk management committee to support decision-making.


Lu F.,Suzhou Nuclear Power Research Institute | Chen M.-Y.,Suzhou Nuclear Power Research Institute
Gongcheng Lixue/Engineering Mechanics | Year: 2013

The improvement of the pressure-temperature (P-T) limit curve methodologies given by the ASME code Ver.2010 and the RCCM code Ver.2007 are presented, and a new P-T curve calculation procedure is proposed according to the RCCM code Ver.2007. The effect of crack size, cooling rate, stress intensity factor plasticity correction and cladding on the P-T curve are studied by performing 3D finite element analysis. The result indicates that crack size and cladding have significant impact on P-T curve, while stress intensity factor plasticity correction has limited influence; the cooling rate effect depends on which standard code is used.


Yang C.F.,Suzhou Nuclear Power Research Institute
Advances in Energy Science and Equipment Engineering - Proceedings of International Conference on Energy Equipment Science and Engineering, ICEESE 2015 | Year: 2015

In order to ensure that the reliability and safety of nuclear power plants remain within acceptable limits and to achieve cost-effective investment, based on technical and economic analysis method, an economic analysis model of Life Cycle Management (LCM) for the generators and exciters at nuclear power plants is established. A nuclear power plant in operation is made as an example, the 40-year and 60-year alternative LCM plans for the plant are given. According to the proposed economic analysis model and various economic data input, the preventive maintenance costs, corrective maintenance costs, loss production costs, the net present value of total benefit and total investment for the LCM alternatives are output, and ultimately the net present value index is calculated to determine the advantages and disadvantages of various alternatives and to identify the best economic LCM alternative. The results provide a basis for decision making for the LCM and life extension plans in nuclear power plants. © 2015 Taylor & Francis Group, London.

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