Fatigue Evaluation for Thermally Stratified DVI Piping

2012 ◽  
Author(s):  
Hwan Ho Lee ◽  
Joon Ho Lee ◽  
Dong Jae Lee ◽  
Seok Hwan Hur ◽  
Il Kwun Nam ◽  
...  
Keyword(s):  
Temperature Distribution ◽  
Thermal Stratification ◽  
Nuclear Power ◽  
Heat Removal ◽  
Local Stress ◽  
Piping System ◽  
Hydraulic Analysis ◽  
Fatigue Evaluation ◽  
Core Cooling ◽  
Thermal Hydraulic Analysis

A numerical analysis has been performed to estimate the effect of thermal stratification in the safety injection piping system. The Direct Vessel Injection (DVI) system is used to perform the functions of Emergency Core Cooling and Residual Heat Removal for an APR1400 nuclear power plant (Korea’s Advanced Power Reactor 1400 MW-Class). The thermal stratification is anticipated in the horizontally routed piping between the DVI nozzle of the reactor vessel and the first isolation valve. Non-axisymmetric temperature distribution across the pipe diameter induced by the thermal stratification leads to differential thermal growth of the piping causing the global bending stress and local stress. Thermal hydraulic analysis has been performed to determine the temperature distribution in the DVI piping due to the thermal stratification. Piping stress analysis has also been carried out to evaluate the integrity of the DVI piping using the thermal hydraulic analysis results. This paper provides a methodology for calculating the global bending stresses and local stresses induced by the thermal stratification in the DVI piping and for performing fatigue evaluation based on Subsection NB-3600 of ASME Section III.

Assessment of a Piping System Subjected to Thermal Stratification and Thermal Striping

2010 ◽  
Author(s):  
Somnath Chattopadhyay
Keyword(s):  
Thermal Stratification ◽  
Nuclear Power ◽  
Power Plants ◽  
Thermal Stresses ◽  
Temperature Gradients ◽  
Low Flow ◽  
Piping System ◽  
Thermal Transients ◽  
Fatigue Evaluation ◽  
Thermal Striping

Piping systems in nuclear power plants are often designed for pressure and mechanical loadings (including seismic loads) and operating thermal transients. In the last few decades a number of failures have occurred due to thermal stratification caused by the mixing of hot and cold fluids under certain low flow conditions. Such stratified temperature fluid profiles give rise to circumferential metal temperature gradients through the pipe leading to high stresses causing fatigue damage. In this work, thermal stresses due to such temperature gradients have been calculated using a finite element method. The peak stresses calculated by this method has been used for fatigue evaluation. In addition the stresses due to thermal striping associated with stratification have also been independently assessed for high cycle fatigue. The method outlined in this paper is a simplified conservative procedure to obtain stratification stresses.

scholarly journals Research on Influence of Different Simulation Methods of Bypass Flow in Thermal Hydraulic Analysis on Temperature Distribution in HTR-10

2020 ◽  
Vol 2020 ◽  
pp. 1-8
Author(s):  
Shiyan Sun ◽  
Youjie Zhang ◽  
Yanhua Zheng
Keyword(s):  
Temperature Distribution ◽  
Reactor Core ◽  
Flow Distribution ◽  
Maximum Temperature ◽  
Flow Ratio ◽  
Hydraulic Analysis ◽  
Bypass Flow ◽  
Pebble Bed ◽  
The Core ◽  
Thermal Hydraulic Analysis

In pebble-bed high temperature gas-cooled reactor, gaps widely exist between graphite blocks and carbon bricks in the reactor core vessel. The bypass helium flowing through the gaps affects the flow distribution of the core and weakens the effective cooling of the core by helium, which in turn affects the temperature distribution and the safety features of the reactor. In this paper, the thermal hydraulic analysis models of HTR-10 with bypass flow channels simulated at different positions are designed based on the flow distribution scheme of the original core models and combined with the actual position of the core bypass flow. The results show that the bypass coolant flowing through the reflectors enhances the heat transfer of the nearby components efficiently. The temperature of the side reflectors and the carbon bricks is much lower with more side bypass coolant. The temperature distribution of the central region in the pebble bed is affected by the bypass flow positions slightly, while that of the peripheral area is affected significantly. The maximum temperature of the helium, the surface, and center of the fuel elements rises as the bypass flow ratio becomes larger, while the temperature difference between them almost keeps constant. When the flow ratio of each part keeps constant, the maximum temperature almost does not change with different bypass flow positions.

Steady-State Thermal-Hydraulic Analysis of a Particle-Bedded Boiling Water Reactor

2009 ◽  
Author(s):  
Xiaoyu Cai ◽  
Suizheng Qiu ◽  
Guanghui Su ◽  
Changyou Zhao
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Superheated Steam ◽  
Nuclear Reactor ◽  
Hydraulic Analysis ◽  
Light Water ◽  
Safety Systems ◽  
Fuel Elements ◽  
Thermal Hydraulic Analysis

The current Light Water Reactors both BWR and PWR have extensive nuclear reactor safety systems, which provide safe and economical operation of Nuclear Power Plants. During about forty years of operation history the safety systems of Nuclear Power Plants have been upgraded in an evolutionary manner. The cost of safety systems, including large containments, is really high due to a capital cost and a long construction period. These conditions together with a low efficiency of steam cycle for LWR create problems to build new power plants in the USA and in the Europe. An advanced Boiling Water Reactor concept with micro-fuel elements (MFE) and superheated steam promises a radical enhancement of safety and improvement of economy of Nuclear Power Plants. In this paper, a new type of nuclear reactor is presented that consists of a steel-walled tube filled with millions of TRISO-coated fuel particles (Micro-Fuel Elements, MFE) directly cooled by a light-water coolant-moderator. Water is used as coolant that flows from bottom to top through the tube, thereby fluidizing the particle bed, and the moderator water flows in the reverse direction out of the tube. The fuel consists of spheres of about 2.5 mm diameter of UO2 with several coatings of different carbonaceous materials. The external coating of steam cycle the particles is silicon carbide (SiC), manufactured with chemical vapor deposit (CVD) technology. Steady-State Thermal-Hydraulic Analysis aims at providing heat transport capability which can match with the heat generated by the core, so as to provide a set of thermal hydraulic parameters of the primary loop. So the temperature distribution and the pressure losses along the direction of flow are calculated for equilibrium core in this paper. The calculation not only includes the liquid region, but the two phase region and the superheated steam region. The temperature distribution includes both the temperature parameters of micro-fuel elements and the coolant. The results show that the maximum fuel temperature is much lower than the limitation and the flow distribution can meet the cooling requirement in the reactor core.

Thermal-hydraulic analysis of an innovative decay heat removal system for lead-cooled fast reactors

2016 ◽  
Vol 305 ◽  
pp. 168-178 ◽  
Author(s):  
Fabio Giannetti ◽  
Damiano Vitale Di Maio ◽  
Antonio Naviglio ◽  
Gianfranco Caruso
Keyword(s):  
Heat Removal ◽  
Hydraulic Analysis ◽  
Fast Reactors ◽  
Decay Heat ◽  
Heat Removal System ◽  
Thermal Hydraulic Analysis ◽  
Removal System

The Effect of IRWST Simulation Method on the Analysis of Small Break LOCA Using RELAP5

2017 ◽  
Author(s):  
Caihong Xu ◽  
Guobao Shi ◽  
Kemei Cao ◽  
Xiaoyu Cai ◽  
Zhanfei Qi
Keyword(s):  
Nuclear Power ◽  
Heat Removal ◽  
Simulation Method ◽  
Horizontal Direction ◽  
Water Storage Tank ◽  
Initiation Phase ◽  
Injection Flow ◽  
Loss Of Coolant Accident ◽  
Perfect Mixing ◽  
Core Cooling

The In-containment Refueling Water Storage Tank (IRWST) provides low-pressure safety injection flow for passive CAP1400 Nuclear Power Plant (NPP) during Loss-Of-Coolant-Accident (LOCA) and subsequent Long Term Core Cooling (LTCC). The Passive Residual Heat Removal Heat Exchanger (PRHR HX) and the spargers of Automatic Depressurization System (ADS) stage 1∼3 are submerged in the IRWST. During small break LOCA, heat and mass are delivered through PRHR HX and ADS spargers to IRWST, and IRWST is heated up before its safety injection. However, numerical and experimental investigation has shown that IRWST is not perfect mixing, and thermal stratification exists. During ADS-4/IRWST initiation phase, the temperature of IRWST injection flow is of great importance, and is affected greatly by IRWST simulation method when modeling with system code like RELAP5. In this paper, two different IRWST simulation methods where one use multi channels in horizontal direction while the other use only one, are analyzed for CAP1400 SBLOCA with RE-LAP5, and their effects are compared. Finally, the better method which uses only one channel in horizontal direction is recommended.

Investigation of Thermal Fatigue Characteristics of Charging Line of RCV in PWR

2017 ◽  
Author(s):  
Naeem Ahmad ◽  
XiangBin Li ◽  
Iftikhar Ahmad ◽  
Nan Li ◽  
Shahroze Ahmed ◽  
...  
Keyword(s):  
Fatigue Damage ◽  
Thermal Fatigue ◽  
Nuclear Power ◽  
Thermal Loading ◽  
Structural Model ◽  
Fatigue Cracks ◽  
Volume Control ◽  
Piping System ◽  
Thermal Transients ◽  
Fatigue Evaluation

Nuclear Power Plant (NPP) components need to tolerate thermal constraints, internal pressure and thermal transients. These thermal transients being repeated again and again can lead to thermal fatigue of the component. It has significant effect on the degradation of the NPP components in long term. Studies of thermal fatigue on different NPP components such as mixing tees and valves have been carried out before but the charging line in the chemical and volume control system (RCV) of the NPP seems to have been ignored for thermal fatigue analysis. Charging Line is the connection from RCV towards Reactor Coolant System (RCP). To enhance the safety of the charging line, thermal fatigue evaluation of piping system was performed using the Fluid Structure Interaction (FSI) analysis. Temperature distributions in the pipes were determined via thermal hydraulic analysis (CFX) and the results were applied to the structural model of the piping system to determine the thermal stress (Transient Structural). Results revealed the location of fatigue cracks. Types of stress were identified that caused the fatigue damage. The CFD analysis enabled us to clarify the role of turbulence with respect to the thermal loading of the structure. The study will provide valuable information for establishing a permanent methodology to help minimize thermal fatigue damage in NPP components.

Thermal-Hydraulic Analysis of the Fort St. Vrain Modular Vault Dry Store System

1991 ◽  
Vol 113 (4) ◽  
pp. 522-529 ◽  
Author(s):  
S. M. Cho ◽  
A. H. Seltzer ◽  
M. Blackbourn
Keyword(s):  
Nuclear Power ◽  
Hydraulic Characteristics ◽  
Spent Fuel ◽  
Hydraulic Analysis ◽  
Power Station ◽  
Limit Values ◽  
Hydraulic Design ◽  
Loading And Unloading ◽  
Design Analyses ◽  
Thermal Hydraulic Analysis

A passive, natural thermosyphon, air-cooled modular vault dry store (MVDS) system is being constructed for the storage of nuclear spent fuel for the Fort St. Vrain (FSV) Nuclear Power Station. In support of this FSV-MVDS system, thermal-hydraulic design analyses have been performed. The objective of the analyses is to determine flow and temperature distributions within the system and thus to ensure that the maximum fuel element temperatures shall not exceed specified design limit values under various loading and unloading conditions. This paper presents the method of analysis and discusses the resulting thermal-hydraulic characteristics of the MVDS system.

CFD Analysis of Thermal Stratification Phenomena in the RNS Suction Line of a Gen-III Nuclear Power Plant

2013 ◽  
Author(s):  
Yan Li ◽  
Daogang Lu ◽  
Zhigang Wang ◽  
Jian Wu ◽  
Fengyun Yu
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Thermal Stratification ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Heat Removal ◽  
Radial Temperature ◽  
Piping Systems ◽  
Suction Line ◽  
Power Operation

Thermal stratification phenomena in piping systems of nuclear power plant would threaten the structural integrity of pipes, which are caused by the significant change of water density with temperature. To provide temperature gradients for the stress analysis of Normal heat Removal System (RNS) suction line of a Gen-III nuclear power plant, the relevant thermal stratification phenomena are analyzed by CFD in this paper. Cases without leakage (normal power operation) and with leakage are both studied. The results show that the first portion of pipe (one meter or so) near the hot leg is isothermal for normal power operation due to the penetrating flow. In the remaining portion, the radial temperature drops are of the order of 20∼27 K for no leakage case. For the leakage case, the radial temperature drops are 23 K or less, which are relatively smaller than those for the no leakage case due to the net hot flow from the hot leg to the valve.

Fatigue Effect of RCS Branch Line by Thermal Stratification

2003 ◽  
Author(s):  
Hag-Ki Youm ◽  
Kwang-Chu Kim ◽  
Man-Heung Park ◽  
Tea-Eun Jin ◽  
Sun-Ki Lee ◽  
...  
Keyword(s):  
Thermal Stratification ◽  
Nuclear Power ◽  
Power Plants ◽  
Evaluation Methodology ◽  
Element Analysis ◽  
Branch Line ◽  
Reactor Coolant ◽  
Stratification Effect ◽  
Fatigue Evaluation ◽  
Coolant System

Recent events reported at a number of nuclear power plants worldwide have shown that thermal stratification, cycling, and striping in piping can cause excessive thermal stress and fatigue on the piping material. These phenomena are diverse and complicated because of the wide variety of geometry and thermal hydraulic conditions encountered in reactor coolant system. Thermal stratification effect of re-branched lines is not yet considered in the fatigue evaluation. To evaluate the thermal load due to turbulent penetration, this paper presents a fatigue evaluation methodology for a branch line of reactor coolant system with the re-branch line. The locations of fatigue monitoring and supplemented inspections are discussed as a result of fatigue evaluations by Interim Fatigue Management Guideline (ITFMG) and detail finite element analysis. Although the revised CUF was increased less than 50 %, the CUF values for some locations was greater than the ASME Code limits.

Sign in / Sign up

Export Citation Format

Share Document