A New Method to Weaken Thermal Stratification Phenomena in Pipes of Nuclear Power Plant

2013 ◽  
Author(s):  
Shengfei Wang ◽  
Yuxin Pang ◽  
Xiaojing Li ◽  
Dandan Fu ◽  
Yang Li ◽  
...  
Keyword(s):  
Thermal Stratification ◽  
Nuclear Power ◽  
Thermal Stresses ◽  
Heat Pipes ◽  
New Method ◽  
Simple Experiment ◽  
Pressurized Water ◽  
Piping Systems ◽  
Surge Line ◽  
Water Reactors

Thermal stratification phenomena are observed in piping systems of pressurized water reactors, especially in the pressurizer surge line. As a result of the thermal stratification induced thermal stresses, fatigue problems can occur in the pipework. US NRC requirements have also identified flow stratification in surge lines as a phenomenon that must be considered in the design basis of surge lines. In this paper, a new method to reduce thermal stratification is proposed. As we all know, heat pipe is a simple device with no moving parts and can transfer large quantities of heat over fairly large distance. The new method is that using heat pipes to weaken the thermal stratification. In order to validate the new method, a simple experiment and theoretical analysis was taken. The results show that, the temperature difference of thermal stratification with heat pipes is smaller than the stratification without heat pipes. A design scheme was also given at the end of paper.

The Surge Line Stress Analysis Model Setup With the Consideration of Thermal Stratification

2014 ◽  
Author(s):  
Jianjun Wang ◽  
Zengfang Ge ◽  
Zhongning Sun ◽  
Changqi Yan
Keyword(s):  
Stress Analysis ◽  
Transient Process ◽  
Thermal Stratification ◽  
Nuclear Power ◽  
Transient Temperature ◽  
Analysis Model ◽  
Pressurized Water ◽  
Line Model ◽  
Transient Temperature Distribution ◽  
Surge Line

In this paper, we deal with a typical pressurizer surge line in a conventional pressurized water reactor (PWR). This study is performed to develop an understanding of thermal stratification phenomenon, which may occur in the surge line during either normal condition or transient process, in the pressurizer surge line. The pressurizer surge line model of Daya Bay nuclear power plant is used as base analysis model, in which the hot leg is taken into account. The transient temperature distribution required to assess the phenomenon along the pressurizer surge line is obtained through CFD analysis technology using ANSYS FLUENT. The temperature loads are transferred to ANSYS Mechanical for stress evaluation for the heat up transient process. Subsequently, the usage factor is calculated on the basis of ASME Section-III design curve. The possible mitigation scheme for the thermal stratification phenomenon of changing the layout angles is also simulated and analyzed in detail. The results show that the thermal stratification phenomenon will occur both in normal operating condition and in heat up transient process. The circumfluent effect makes the thermal stratification phenomenon exhibit unique profile due to the introduction of the hot leg. The continuous spray mass flow rate may influence both the temperature difference and the occurrence range for the thermal stratification phenomenon. The stress analysis incorporating both temperature load and pressure load is performed for pressurizer surge line model with hot leg for the conservative and complete heat up case.

Computational Fluid Dynamics Modeling of an Experimental Thermal-Stratification Flow Case Using ABAQUS/CFD

2019 ◽  
Author(s):  
Daniel Franken ◽  
Subhasish Mohanty
Keyword(s):  
Thermal Stratification ◽  
Cfd Simulation ◽  
The United States ◽  
System Level ◽  
Dynamics Modeling ◽  
Pressurized Water ◽  
Line Pipe ◽  
Surge Line ◽  
Water Reactors ◽  
Abaqus Code

Abstract As the fleet of Pressurized Water Reactors (PWRs) in the United States begin to reach the end of their original lifespan many of them are undergoing assessment to extend their use. In order to investigate the potential for extending the life of the plant, a system level analysis of components needs to be performed in order to ensure that age and degradation of the system will not lead to a potential safety hazard. An area in which this system level investigation is particularly important is in the surge line of the pressurizer. One possible concern is that over the life of the reactor, the surge line pipe will experience thermal stratification many times. Thermal stratification can lead to significant stresses induced on the piping and over time may result in a less than ideal safety standard. Commercially available code Abaqus CFD was used to model the thermal stratification in a pipe. The corresponding experimental results, available in literature were compared. We found there is a good correlation between the experimental and computational results. However, the results discussed in this paper are based on our preliminary effort to study the capability of ABAQUS code for CFD simulation. A detailed parametric study is one of our future work.

A Validation of RELAP on Predicting Nuclear Power Plant Phenomena

2018 ◽  
Author(s):  
Ji Soo Ahn ◽  
Michael Bluck ◽  
Matthew Eaton ◽  
Chris Jackson
Keyword(s):  
Thermal Stratification ◽  
Nuclear Power ◽  
Power Plants ◽  
Feed Water ◽  
Multiple Channels ◽  
Horizontal Pipe ◽  
One Dimensional ◽  
Piping Systems ◽  
Surge Line ◽  
Input Model

In this study, RELAP5’s capability to simulate thermal stratification under different conditions is assessed. In nuclear power plants (NPPs), thermal stratification can occur in the following locations: pressurizer, piping systems such as hot legs, cold legs, surge lines, and cooling tanks if available. In general, thermal stratification in a horizontal pipe could not be simulated by RELAP5 due to the inherent one-dimensional setting. Moreover, RELAP5 failed to simulate turbulent penetration which was often a pre-requisite prior to thermal stratification in a pipe. This type of situation could arise in connection between hot leg and surge line, spray lines, feed water lines, etc. It is recommended that for this type of problem CFD be used. In the literature, it was found that RELAP5 was capable of simulating thermal stratification in a pool or a tank-like component if multiple channels and crossflow junctions were used. However, due to uncertainties associated with the input model, the current RELAP5 model failed to reproduce experimental data and therefore further investigation would be required to identify the sources of error.

14C Distribution in Atmospheric and Aquatic Environments Around Qinshan Nuclear Power Plant, China

Radiocarbon ◽  
2014 ◽  
Vol 56 (3) ◽  
pp. 1107-1114 ◽  
Author(s):  
Zhongtang Wang ◽  
Dan Hu ◽  
Hong Xu ◽  
Qiuju Guo
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Water Samples ◽  
Nuclear Power ◽  
Dissolved Inorganic Carbon ◽  
Tap Water ◽  
Surface Seawater ◽  
Pressurized Water ◽  
Specific Activities ◽  
Water Reactors

Atmospheric CO2 and aquatic water samples were analyzed to evaluate the environmental 14C enrichment due to operation of the Qinshan nuclear power plant (NPP), where two heavy-water reactors and five pressurized-water reactors are employed. Elevated 14C-specific activities (2–26.7 Bq/kg C) were observed in the short-term air samples collected within a 5-km radius, while samples over 5 km were close to background levels. The 14C-specific activities of dissolved inorganic carbon (DIC) in the surface seawater samples ranged from 196.8 to 206.5 Bq/kg C (average 203.4 Bq/kg C), which are close to the background value. No elevated 14C level in surface seawater was found after 20 years of operation of Qinshan NPP, indicating that the 14C discharged was well diffused. The results of the freshwater samples show that excess 14C-specific activity (average 17.1 Bq/kg C) was found in surface water and well water samples, while no obvious 14C increase was found in drinking water (groundwater and tap water) compared to the background level.

scholarly journals 14C Release in Various Chemical Forms With Gaseous Effluents From the Paks Nuclear Power Plant

Radiocarbon ◽  
1989 ◽  
Vol 31 (03) ◽  
pp. 754-761 ◽  
Author(s):  
Ede Hertelendi ◽  
György Uchrin ◽  
Peter Ormai
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Liquid Scintillation ◽  
Mean Value ◽  
Scintillation Counting ◽  
Pressurized Water ◽  
The Mean ◽  
Water Reactors

We present results of airborne 14C emission measurements from the Paks PWR nuclear power plant. Long-term release of 14C in the form of carbon dioxide or carbon monoxide and hydrocarbons were simultaneously measured. The results of internal gas-proportional and liquid scintillation counting agree well with theoretical assessments of 14C releases from pressurized water reactors. The mean value of the 14C concentration in discharged air is 130Bqm-3 and the normalized release is equal to 740GBq/GWe · yr. > 95% of 14C released is in the form of hydrocarbons, ca 4% is apportioned to CO2, and <1% to CO. Tree-ring measurements were also made and indicated a minute increase of 14C content in the vicinity of the nuclear power plant.

Evaluation of Representative Piping Systems Designed by Implicit Fatigue Concept

2008 ◽  
Author(s):  
Shin-Beom Choi ◽  
Sun-Hye Kim ◽  
Yoon-Suk Chang ◽  
Jae-Boong Choi ◽  
Young-Jin Kim ◽  
...  
Keyword(s):  
Fatigue Life ◽  
Nuclear Power ◽  
Heat Removal ◽  
Environmental Water ◽  
Lessons Learned ◽  
Metal Fatigue ◽  
Aging Effects ◽  
Age Related ◽  
Piping Systems ◽  
Surge Line

NUREG-1801 provides generic aging lessons learned to manage aging effects that may occur during continued operation beyond the design life of nuclear power plant. According to this report, the metal fatigue, among several age-related degradation mechanisms, is identified as one of time-limited aging analysis item. The objective of this paper is to introduce fatigue life evaluation of representative surge line and residual heat removal system piping which was designed by implicit fatigue concept. For the back-fitting evaluation employing explicit fatigue concept, detailed parametric CFD as well as FE analyses results are used. The well-known ASME Section III NB-3600 procedure is adopted for the metal fatigue and NUREG/CR-5704 procedure is further investigated to deal with additional environmental water effects. With regard to the environmental effect evaluation, two types of fatigue life correction factors are considered, such as maximum Fen and individual Fen. As a result, it was proven that a thermal stratification phenomenon is the governing factor in metal fatigue life of the surge line and strain rate is the most important parameter affecting the environmental fatigue life of both piping. The evaluation results will be used as technical bases for continued operation of OPR 1000 plant.

scholarly journals Reactor Vessel Internals Segmentation Experience using Mechanical Cutting Tools

2013 ◽  
Vol 10 (2) ◽  
pp. 6-10 ◽  
Author(s):  
Petr Pospíšil
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Cutting Tools ◽  
Reactor Vessel ◽  
Boiling Water Reactors ◽  
Abrasive Waterjet ◽  
Pressurized Water ◽  
Water Reactors ◽  
Mechanical Cutting ◽  
Reactor Internals

Abstract Some commercial nuclear power plants have been permanently shut down to date and decommissioned using dismantling methods. Other operating plants have decided to undergo an upgrade process that includes replacement of reactor internals. In both cases, there is a need to perform a segmentation of the reactor vessel internals with proven methods for long term waste disposal. Westinghouse has developed several concepts to dismantle reactor internals based on safe and reliable techniques, including plasma arc cutting (PAC), abrasive waterjet cutting (AWJC), metal disintegration machining (MDM), or mechanical cutting. Mechanical cutting has been used by Westinghouse since 1999 for both Pressurized Water Reactors (PWR’s) and Boiling Water Reactors (BWR’s) and its process has been continuously improved over the years. The complexity of the work requires well designed and reliable tools. Different band saws, disc saws, tube cutters and shearing tools have been developed to cut the reactor internals. All of those equipments are hydraulically driven which is very suitable for submerged applications. Westinghouse experience in mechanical cutting has demonstrated that it is an excellent technique for segmentation of internals. In summary, the purpose of this paper will be to provide an overview of the Westinghouse mechanical segmentation process, based on actual experience from the work that has been completed to date.

Comparison of US and UK Practices and Philosophies for Nuclear Plant Control Room Design

1982 ◽  
Vol 26 (7) ◽  
pp. 659-663
Author(s):  
Les Ainsworth
Keyword(s):  
Nuclear Power ◽  
Plant Design ◽  
Control Room ◽  
Pressurized Water ◽  
Basic Task ◽  
Room Design ◽  
Power Plant Control ◽  
Gas Cooling ◽  
Water Reactors ◽  
Plant Control

The American and British nuclear power programmes have in the past taken divergent routes, with the Americans choosing pressurized water reactors, whilst the British have opted for gas cooling. Although the technology and plant design for these two systems encompasses many fundamental differences, some of which have ramifications for the controllers, the basic task of monitoring and controlling a reactor holds many similarities in both countries. It is therefore instructive to compare and contrast the approaches which have been taken to human factors in nuclear power plant control room design on both sides of the Atlantic.

scholarly journals Plant Performance History of an Innovative Gate Valve in Critical Service Applications

2017 ◽  
Author(s):  
M. S. Kalsi ◽  
Patricio Alvarez ◽  
Thomas White ◽  
Micheal Green
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Gate Valve ◽  
Plant Performance ◽  
Flow Loop ◽  
Performance History ◽  
Pressurized Water ◽  
Maintenance Costs ◽  
History Of ◽  
Water Reactors

A previous paper [1] describes the key features of an innovative gate valve design that was developed to overcome seat leakage problems, high maintenance costs as well as issues identified in the Nuclear Regulatory Commission (NRC) Generic Letters 89-10, 95-07 and 96-05 with conventional gate valves [2,3,4]. The earlier paper was published within a year after the new design valves were installed at the Pilgrim Nuclear Plant — the plant that took the initiative to form a teaming arrangement as described in [1] which facilitated this innovative development. The current paper documents the successful performance history of 22 years at the Pilgrim plant, as well as performance history at several other nuclear power plants where these valves have been installed for many years in containment isolation service that requires operation under pipe rupture conditions and require tight shut-off in both Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs). The performance history of the new valve has shown to provide significant performance advantage by eliminating the chronic leakage problems and high maintenance costs in these critical service applications. This paper includes a summary of the design, analysis and separate effects testing described in detail in the earlier paper. Flow loop testing was performed on these valves under normal plant operation, various thermal binding and pressure locking scenarios, and accident/pipe rupture conditions. The valve was designed, analyzed and tested to satisfy the requirements of ANSI B16.41 [9]; it also satisfies the requirements of ASME QME 1-2012 [10]. The results of the long-term performance history including any degradation observed and its root cause are summarized in the paper. Paper published with permission.

3D Residual Stress Simulation of an Excavate and Weld Repair Mockup

2016 ◽  
Author(s):  
Francis H. Ku ◽  
Pete C. Riccardella ◽  
Steven L. McCracken
Keyword(s):  
Residual Stresses ◽  
Weld Metal ◽  
Nuclear Power ◽  
Three Dimensional ◽  
Weld Repair ◽  
Element Analysis ◽  
Pressurized Water ◽  
Piping Systems ◽  
Asme Code ◽  
Original Construction

This paper presents predictions of weld residual stresses in a mockup with a partial arc excavate and weld repair (EWR) utilizing finite element analysis (FEA). The partial arc EWR is a mitigation option to address stress corrosion cracking (SCC) in nuclear power plant piping systems. The mockup is a dissimilar metal weld (DMW) consisting of an SA-508 Class 3 low alloy steel forging buttered with Alloy 182 welded to a Type 316L stainless steel plate with Alloy 82/182 weld metal. This material configuration represents a typical DMW of original construction in a pressurized water reactor (PWR). After simulating the original construction piping joint, the outer half of the DMW is excavated and repaired with Alloy 52M weld metal to simulate a partial arc EWR. The FEA performed simulates the EWR weld bead sequence and applies three-dimensional (3D) modeling to evaluate the weld residual stresses. Bi-directional weld residual stresses are also assessed for impacts on the original construction DMW. The FEA predicted residual stresses follow expected trends and compare favorably to the results of experimental measurements performed on the mockup. The 3D FEA process presented herein represents a validated method to evaluate weld residual stresses as required by ASME Code Case N-847 for implementing a partial arc EWR, which is currently being considered via letter ballot at ASME BPV Standards Committee XI.

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