Suzhou Nuclear Power Research Institute

Suzhou, China

Suzhou Nuclear Power Research Institute

Suzhou, China

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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.

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.

Zhu W.,Nanjing University of Science and Technology | Wang R.,Suzhou Nuclear Power Research Institute | Shu G.,China Nuclear Power Engineering Co. | Wu P.,Institute of High Performance Computing of Singapore | Xiao H.,Nanjing University of Science and Technology
Journal of Physical Chemistry C | Year: 2010

We present a detailed study on the electronic structure, mechanical properties, phase stability, and thermodynamic properties of four polymorphs of crystalline zirconium hydride by using density functional theory within the generalized gradient approximation. An analysis of electronic structure shows that the zirconium hydrides retain metallic bonding over the whole hydrogen composition range. The calculated mechanical properties indicate that ζ-Zr2H and γ-ZrH are ductile, while δ-ZrH 1.5 and ε-ZrH2 are brittle compared with α-Zr. The hydrides change from ductile to brittle in the order of ζ-Zr 2H, γ-ZrH, ε-ZrH2, and δ-ZrH 1.5. The formation enthalpies are negative for the four hydrides at the ambient pressure indicating that they are thermodynamically stable. Only the δ phase is thermodynamically stable in the whole pressure range. As the temperature increases, the decomposition reactions of the four hydrides are more and more favorable thermodynamically. The δ phase is thermodynamically easier to decompose than the others in the whole temperature range. © 2010 American Chemical Society.

Yang C.,Suzhou Nuclear Power Research Institute
Advanced Materials Research | Year: 2013

The effective methods of the ageing and life assessment for large and medium-sized power transformers used in nuclear power plants are analyzed and described, including the thermal ageing life assessment method for transformer solid insulation, the gas analysis method of CO and CO2 in the transformer oil, the average degree of polymerization method, furfural content analysis method, and the analysis method based on the insulation ageing-related electrical parameters. The analysis results show that the methods used can reasonably assess the remaining life of the transformers. These methods have important reference value to the ageing and life management for the large and medium-sized power transformers in nuclear power plants. © (2013) Trans Tech Publications, Switzerland.

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.

Chen M.,Suzhou Nuclear Power Research Institute | Lu F.,Suzhou Nuclear Power Research Institute | Wang R.,Suzhou Nuclear Power Research Institute | Ren A.,Suzhou Nuclear Power Research Institute
Nuclear Engineering and Design | Year: 2014

Fracture mechanics analysis of pressurized thermal shock (PTS) is the key element of the integrity evaluation of the nuclear reactor pressure vessel (RPV). While the regulation of 10 CFR 50.61 and the ASME Code provide the guidance for the structural integrity, the guidance has been prepared under conservative assumptions. In this paper, the effects of conservative assumptions involved in the PTS analysis were investigated. The influence of different parameters, such as crack size, cladding effect and neutron fluence, were reviewed based on 3-D finite element analyses. Also, the sensitivity study of elastic-plastic approach, crack type and cladding thickness were reviewed. It was shown that crack depth, crack type, plastic effect and cladding thickness change the safety margin (SM) significantly, and the SM at the deepest point of the crack is not always smaller than that of the surface point, indicating that both the deepest and surface points of the crack front should be considered. For the reference transient, deeper cracks always give more conservative prediction. So compared to the prescribed analyses of a set of postulated defects with varying depths in the ASME code, it only needs to assess the crack with maximum depth in the code for the reference transient according to the conclusions. © 2014 Elsevier B.V.

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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