Research on Constant Velocity Extruding Process Control for 36,000-Ton Vertical Extrusion Press

2013 ◽  
Vol 135 (4) ◽  
Author(s):  
Wanzhou Li ◽  
Tao Sun ◽  
Yuechen Hu ◽  
Wei Li
Keyword(s):  
Constant Velocity ◽  
Hydraulic System ◽  
Nuclear Power ◽  
Power Plants ◽  
Large Diameter ◽  
Hot Extrusion ◽  
Extrusion Process ◽  
Complex Object ◽  
Loop Control ◽  
Extrusion Press

Metal extrusion is one of the most significant methods in the field of plastic deformation. The 36,000-ton vertical extrusion press uses one-piece hot extrusion molding method to produce high-pressure, large-diameter, thick-walled, seamless, alloy-steel pipe, which is required by ultrasupercritical power generator units and the third generation of nuclear power plants. The main table (extruding platform) is driven by 36,000-ton thrust. The extrusion process is a typical nonlinear multivariable strong-coupling finite-time system. In this paper, we analyze and solve the large-scaled engineering control problems, including (1) discussion of the mechanical structure of the controlled object and the feature of multivariable strongly coupling for large-scaled parallel-driving hydraulic system, (2) research on the engineering practical simplified control algorithm for multivariable system and propose a way to solve the coupling problem of the hydraulic system, and (3) design of double closed-loop control of velocity and pressure for the new technology of constant velocity extrusion. In our paper, we have proven through practice that the traditional proportional intergral (PI) controlling method, with proper controlling strategy, is still the most efficient and practicable way for a large-dimension complex object in industry.

Aging Management Program of Large Diameter Safety Class Piping

2005 ◽  
Author(s):  
Claude Faidy
Keyword(s):  
Thermal Aging ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Large Diameter ◽  
Management Program ◽  
Degradation Mechanisms ◽  
Ageing Management ◽  
Fatigue Corrosion ◽  
Systematic Program

Ageing management of Nuclear Power Plants is an essential issue for utilities, in term of safety and availability and corresponding economical consequences. Practically all nuclear countries have developed a systematic program to deal with ageing of components on their plants. This paper presents the ageing management program developed by EDF and that are compared with different other approaches in other countries (IAEA guidelines and GALL report). The paper presents an example of application to large diameter safety class piping. Different degradation mechanisms are considered like fatigue, corrosion or thermal aging. Maintenance and surveillance actions are discussed in the paper.

scholarly journals Study on the Influence of Processing Parameters on Piercing Extrusion Process of Large Diameter Cupronickel Alloy Pipes Using 3D FEM Analysis

2017 ◽  
Vol 2017 ◽  
pp. 1-10 ◽  
Author(s):  
Jun Cai ◽  
Kuaishe Wang ◽  
Bing Zhang ◽  
Wen Wang
Keyword(s):  
High Performance ◽  
Rapid Development ◽  
Large Diameter ◽  
Fem Analysis ◽  
Hot Extrusion ◽  
Processing Parameters ◽  
Extrusion Process ◽  
Ingot Casting ◽  
3D Fem ◽  
Deformation Parameters

With the rapid development of the shipping and the power industry, the demand for high-performance cupronickel alloy pipes is greatly increasing. The main processing methods of this alloy include semisolid ingot casting and deformation by hot extrusion. Many defects appear during the hot extrusion process for large diameter cupronickel alloy pipes, which results in considerable problems. Therefore, numerical simulation of hot extrusion for cupronickel alloy pipes before the practical production is of vital significance to properly determine the deformation parameters. In order to obtain the influence of processing parameters on the piercing extrusion process of large diameter cupronickel alloy pipe, metal flowing law under different deformation conditions was simulated and analyzed via employing a 3D FEM code. The results showed that piercing speed had no obvious influence on the cupronickel alloy billet. However, the friction had significant influence on the piercing process of cupronickel alloy billet: with the increase of friction coefficient, the temperature and the load increased.

Experimental Investigation of Steam Bubble Condensation in Flowing Subcooled Water With Two Different Injection Nozzle Geometries

2013 ◽  
Author(s):  
Suleiman Al Issa ◽  
Patricia B. Weisensee
Keyword(s):  
Test Section ◽  
High Speed ◽  
Nuclear Power ◽  
Power Plants ◽  
Bubble Diameter ◽  
Large Diameter ◽  
Low Pressure ◽  
Test Facility ◽  
Experimental Investigations ◽  
Subcooled Water

A multiphase flow test facility was built at the Department of Nuclear Engineering at the Technical University Munich. The main goal of this facility is to investigate the condensation of steam bubbles injected into a vertical large diameter pipe (104 mm) with flowing subcooled water (6–15 K) at low pressure conditions (1.1–1.45 bar). Current experimental investigations will contribute to a better understanding of subcooled boiling at low pressures, accidental conditions in nuclear power plants and low-pressure research reactors and correlations for the validation of CFD codes. The test section is a 1 m long transparent pipe that is surrounded by an 18×18 cm rectangular “aquarium” filled with distilled water for refraction correction. High-speed camera (HSC) recording was used to gather data about condensing bubbles including bubble diameter, shape and rising velocity. Steam was injected via two different vertical injection nozzles with an inner diameter of 4 and 6 mm, respectively, directly into the center of the test section. The present experiments were carried out at three different steam superficial velocities, water superficial velocities and water temperatures leading to bubble diameters up to 50 mm and bubble relative velocities around 1 m/s. The measurements enabled the calculation of bubble Reynolds and Nusselt numbers and comparison with correlations given in literature. Even though significant differences could be observed between the two injection nozzles with respect to the bubble’s diameter and velocity, the Nusselt and Reynolds numbers are in the same range of values. The bigger bubbles of the 6 mm with respect to the 4 mm nozzle are almost neutralized by the lower rising velocities.

Aging Management Program of Large Diameter Safety Class Piping

2006 ◽  
Author(s):  
Claude Faidy
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Large Diameter ◽  
Thermal Ageing ◽  
Management Program ◽  
Lessons Learned ◽  
Energy Agency ◽  
Ageing Management ◽  
Fatigue Corrosion ◽  
Systematic Program

Ageing management of Nuclear Power Plants is an essential issue for utilities, in term of safety and availability and corresponding economical consequences. Practically all nuclear countries have developed a systematic program to deal with ageing of components on their plants. This paper presents the ageing management program developed by EDF (Electricite´ de France) that are compared with different approaches in other countries, like IAEA (International Atomic Energy Agency)guidelines and GALL (Generic Ageing Lessons Learned) report. The paper presents an example of application to large diameter safety class piping. Different degradation mechanisms are considered fatigue, corrosion and thermal ageing. Maintenance and surveillance actions are also discussed in the paper.

ANALYSIS OF RESIDUAL STRESS FOR NARROW GAP WELDING USING FINITE ELEMENT METHOD

2010 ◽  
Vol 24 (15n16) ◽  
pp. 2797-2802 ◽  
Author(s):  
CHOON YEOL LEE ◽  
JAE KEUN HWANG ◽  
JOON WOO BAE
Keyword(s):  
Finite Element Method ◽  
Residual Stress ◽  
Finite Element ◽  
Nuclear Power ◽  
Power Plants ◽  
Large Diameter ◽  
Narrow Gap ◽  
Narrow Gap Welding ◽  
Welding Processes ◽  
Element Method

Reactor coolant loop (RCL) pipes circulating the heat generated in a nuclear power plant consist of so large diameter pipes that the installation of these pipes is one of the major construction processes. Conventionally, a shield metal arc welding (SMAW) process has been mainly used in RCL piping installations, which sometimes caused severe deformations, dislocation of main equipments and various other complications due to excessive heat input in welding processes. Hence, automation of the work of welding is required and narrow-gap welding (NGW) process is being reviewed for new nuclear power plants as an alternative method of welding. In this study, transient heat transfer and thermo-elastic-plastic analyses have been performed for the residual stress distribution on the narrow gap weldment of RCL by finite element method under various conditions including surface heat flux and temperature dependent thermo-physical properties.

Investigation on Modeling Thermal-Hydraulic System of CPR1000 NPP Based on RELAP5

2014 ◽  
Author(s):  
Cong Wang ◽  
Danmei Xie ◽  
Peng Zhang ◽  
Xinggang Yu ◽  
Xiuqun Hou
Keyword(s):  
Hydraulic System ◽  
Nuclear Power ◽  
Power Plants ◽  
Thermodynamic System ◽  
Pressure Variation ◽  
Actual Situation ◽  
Hydraulic Systems ◽  
System A ◽  
Variation Trends ◽  
Steam Parameter

Based on the best-estimate program RELAP5/MOD4.0, a full-scope thermal-hydraulic model with reference to CPR1000 nuclear power plant is established in this paper, which includes the thermal-hydraulic systems of conventional island as well as the primary nuclear island which has already been researched in traditional safety analysis. Therefore, this paper mainly details the numerical model of the turbine and other parts of the conventional island thermodynamic system. A comparison between the calculated results in steady-state and the actual data of reactor demonstrates a fine consistency, thus verifying the accuracy and reliability of the model. In addition, the steam parameter changes are numerically simulated during the steam turbine’s off-design operating condition such as back pressure variation and the variation trends are the same as the actual situation of nuclear power plants.

Ultrasonic Inspection of Large Diameter Polyethylene Pipe Used in Nuclear Power Plant

2014 ◽  
Author(s):  
Jianfeng Shi ◽  
Dongsheng Hou ◽  
Weican Guo ◽  
Yaoda Zhou ◽  
Xia Chen ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Large Diameter ◽  
Ultrasonic Inspection ◽  
Water System ◽  
Cooling Water ◽  
Operating Pressure ◽  
Traditional Natural

Polyethylene (PE) pipe has many advantages such as good flexibility, corrosion resistance and long service life. It has been introduced into nuclear power plants for transportation of cooling water both in U.S. and Europe. Recently, one Chinese nuclear power plant in Zhejiang Province also introduced four polyethylene pipelines in essential cooling water system with operating pressure of 0.6MPa and operating temperature of no more than 60°C. The PE pipes used in this nuclear power plant are DN762 SDR9 (30in OD, 3.3in wall), which are much larger and thicker than traditional natural gas PE pipe. As the pipe wall is so thick that the ultrasonic phased array instrument used in inspection of PE pipe with diameter less than 400mm has been improved. Results of field inspection in the Sanmen nuclear plant are reported, and the presented ultrasonic inspection technique proves to be effective for high density polyethylene (HDPE) pipe of large size in nuclear power plant.

Development of an Evaluation Method for Seismic Isolation Systems of Nuclear Power Facilities: Part 10 — Evaluation of Seismic Isolator Design

2014 ◽  
Author(s):  
Hiroyuki Asano ◽  
Tsutomu Hirotani ◽  
Takashi Nakayama ◽  
Takemi Norimono ◽  
Yuji Aikawa ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Seismic Isolation ◽  
Power Plants ◽  
Evaluation Method ◽  
Large Diameter ◽  
Seismic Isolator ◽  
Long Duration ◽  
Seismically Isolated ◽  
Isolation Systems ◽  
Seismic Isolation Systems

This paper provides a part of series of “Development of an Evaluation Method for Seismic Isolation Systems of Nuclear Power Facilities”. This part shows an evaluation of seismic isolator design established in this project where several methods are newly developed. The major four accomplishments are as follows. One: establishment of design earthquake specially considered for seismically isolated nuclear power facilities. The design earthquakes are made to fit multiple target spectra with different damping factors considering a building, equipment and seismic isolators for more precise response analyses. Two: design and development of a high-performance seismic isolator. Against the large design earthquakes, a seismic isolator is newly developed which has a large diameter lead plug for more damping; the isolators were actually manufactured and tested. Three: seismic response analyses for seismically isolated nuclear power plants. Light water reactors are designed where the structural characteristics of the seismic isolation system is reflected. Four: evaluation of thermal effects on seismic isolators by a long-duration earthquake. Considering a long-duration earthquake, the heat generation phenomenon in the lead plug is analytically evaluated to ensure the lead plug’s damping performance. By introducing these accomplishments, the realistic design of a seismically isolated nuclear power plant is achieved.

scholarly journals Thermal-Hydraulic System Codes in Nulcear Reactor Safety and Qualification Procedures

2008 ◽  
Vol 2008 ◽  
pp. 1-16 ◽  
Author(s):  
Alessandro Petruzzi ◽  
Francesco D'Auria
Keyword(s):  
Hydraulic System ◽  
Nuclear Power ◽  
Power Plants ◽  
Initial Conditions ◽  
Fluid Model ◽  
Balance Equations ◽  
Two Phase ◽  
Imperfect Knowledge ◽  
Two Fluid Model ◽  
Degree Of Maturity

In the last four decades, large efforts have been undertaken to provide reliable thermal-hydraulic system codes for the analyses of transients and accidents in nuclear power plants. Whereas the first system codes, developed at the beginning of the 1970s, utilized the homogenous equilibrium model with three balance equations to describe the two-phase flow, nowadays the more advanced system codes are based on the so-called “two-fluid model” with separation of the water and vapor phases, resulting in systems with at least six balance equations. The wide experimental campaign, constituted by the integral and separate effect tests, conducted under the umbrella of the OECD/CSNI was at the basis of the development and validation of the thermal-hydraulic system codes by which they have reached the present high degree of maturity. However, notwithstanding the huge amounts of financial and human resources invested, the results predicted by the code are still affected by errors whose origins can be attributed to several reasons as model deficiencies, approximations in the numerical solution, nodalization effects, and imperfect knowledge of boundary and initial conditions. In this context, the existence of qualified procedures for a consistent application of qualified thermal-hydraulic system code is necessary and implies the drawing up of specific criteria through which the code-user, the nodalization, and finally the transient results are qualified.

Use of Polyethylene (PE) Pipe in Safety-Related, Class 3, Service-Water Piping

2008 ◽  
Author(s):  
Prabhat Krishnaswamy ◽  
Eric M. Focht ◽  
Do-Jun Shim ◽  
Tao Zhang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Elevated Temperatures ◽  
Slow Crack Growth ◽  
Large Diameter ◽  
Modes Of Failure ◽  
Service Water ◽  
Division 1 ◽  
Piping Systems ◽  
Technical Issues

The ASME Boiler and Pressure Vessel Code Committee (BPVC) has recently published Code Case N-755 that describes the requirements for the use of Polyethylene (PE) pipe for the construction of Section III, Division 1 Class 3 buried piping systems for service water applications in nuclear power plants. The code case was developed by Special Working Group–Polyethylene Pipe (SWG-PP) within Section III (Design) of the BPVC. This paper provides a critical review of the design requirements described in CC N-755 from pressure boundary integrity considerations. The various technical issues that need to be addressed for safety-critical PE piping are discussed in this paper. Specifically, the premise of allowing defects in pipe that are 10% of the wall thickness has been reviewed especially for cases involving large diameter piping [> 304.8 mm (12 inches)] that is to be operated at elevated temperatures as high as 60°C (140°F). One of the common modes of failure in PE piping under sustained internal pressure is due to slow crack growth (SCG) from manufacturing or installation defects in the pipe wall. The effect of pipe diameter and stresses on the crack driving force for a 10% deep flaw is calculated for comparison with the material resistance to SCG at elevated temperatures.

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