Swelling Characteristics of a Type 304SS Baffle Plate Irradiated up to 50 DPA in PWR and Validation of a Swelling Equation

2017 ◽  
Author(s):  
Yuichi Mogami ◽  
Toru Matsubara ◽  
Seiji Yaguchi ◽  
Tomohiro Tsuda ◽  
Koji Fujimoto
Keyword(s):  
Structural Integrity ◽  
Irradiation Temperature ◽  
Baffle Plate ◽  
Pressurized Water ◽  
Volumetric Expansion ◽  
Void Swelling ◽  
Swelling Characteristics ◽  
Similar Temperature ◽  
Water Reactors ◽  
Reactor Internals

When austenitic stainless steel is highly irradiated in a reactor under high temperature, voids will be created in the material, leading to volumetric expansion of structures. This phenomenon is known as void swelling. The deformation caused by the swelling possibly deteriorates the functionality of reactor internals in Pressurized Water Reactors (PWRs), especially baffle former assemblies. To evaluate the functionality of the internals against the swelling and to assure the structural integrity, simulation technologies play a key role, enabling the estimation of the swelling behavior of the internals through plant life. The simulation results strongly depend on inputs, especially on a swelling equation; however, it includes uncertainty to some extent because quite limited swelling data in PWR environment have been available, which are necessary for the validation of the equation. To enable the validation of swelling equations and improve the reliability on the simulations, the authors investigated the swelling characteristics of a type 304SS baffle plate removed from a decommissioned PWR plant. A total of nine swelling data were obtained with the variety in neutron dose (33 to 47 dpa) and irradiation temperature (299 to 327°C). The swelling ratios obtained are ranging from 0.02 to 0.08%, which corresponds well with the swelling equation, showing the similar temperature dependency with the equation. Since the irradiation temperature range of the obtained data, up to 327°C, covers major part of baffle former assemblies, swelling ratios of most part of them are expected to be small, which is probably too small to harm the functionality of the assemblies. The results contribute to the better confidence on swelling simulations and to assure the integrity of PWR reactor internals.

Plant Event Signatures on Neutron Noise Data

2014 ◽  
Author(s):  
Jianfeng Yang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Time History ◽  
Data Monitoring ◽  
Pressurized Water ◽  
Neutron Noise ◽  
Noise Data ◽  
Water Reactors ◽  
Reactor Internals

Neutron noise monitoring and analysis has been a valuable tool for in-service monitoring of the structural integrity of reactor internals. Several nuclear power plants have experienced structural degradation in their reactor internals. Significant signatures were noticed in the neutron noise data when the degradations occurred. This article briefly summarizes the findings based on these experiences. A significant amount of neutron noise time-history data has been obtained based on the continuous neutron noise data monitoring and analysis for pressurized water reactors (PWRs). Over the past decade, power plants have made many major component repairs and replacements. This article correlates the plant operating history with its corresponding signatures on the neutron noise data. This information is an extremely valuable reference for diagnosing the condition of the reactor internals through neutron noise data monitoring and analyses.

scholarly journals Experimental approach to investigate the dynamics of mixing coolant flow in complex geometry using PIV and PLIF techniques

2015 ◽  
Vol 19 (3) ◽  
pp. 989-1004 ◽  
Author(s):  
Ezddin Hutli ◽  
Valer Gottlasz ◽  
Dániel Tar ◽  
Gyorgy Ezsol ◽  
Gabor Baranyai
Keyword(s):  
Heat Transfer ◽  
Structural Integrity ◽  
Complex Geometry ◽  
Measurement Techniques ◽  
Coolant Flow ◽  
Temperature Fields ◽  
Head Part ◽  
Pressurized Water ◽  
Area Of Interest ◽  
Water Reactors

The aim of this work is to investigate experimentally the increase of mixing phenomenon in a coolant flow in order to improve the heat transfer, the economical operation and the structural integrity of Light Water Reactors-Pressurized Water Reactors (LWRs-PWRs). Thus the parameters related to the heat transfer process in the system will be investigated. Data from a set of experiments, obtained by using high precision measurement techniques, Particle Image Velocimetry and Planar Laser-Induced Fluorescence (PIV and PLIF, respectively) are to improve the basic understanding of turbulent mixing phenomenon and to provide data for CFD code validation. The coolant mixing phenomenon in the head part of a fuel assembly which includes spacer grids has been investigated (the fuel simulator has half-length of a VVER 440 reactor fuel). The two-dimensional velocity vector and temperature fields in the area of interest are obtained by PIV and PLIF technique, respectively. The measurements of the turbulent flow in the regular tube channel around the thermocouple proved that there is rotation and asymmetry in the coolant flow caused by the mixing grid and the geometrical asymmetry of the fuel bundle. Both PIV and PLIF results showed that at the level of the core exit thermocouple the coolant is homogeneous. The discrepancies that could exist between the outlet average temperature of the coolant and the temperature at in-core thermocouple were clarified. Results of the applied techniques showed that both of them can be used as good provider for data base and to validate CFD results.

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.

Validation and Benchmarking of Comprehensive Vibration Assessment Program for Prototype Reactor Internals

2012 ◽  
Author(s):  
Jianfeng Yang ◽  
Lixin Yu ◽  
Byounghoan Choi
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Boiling Water ◽  
Pressurized Water ◽  
Water Reactor ◽  
Assessment Program ◽  
Water Reactors ◽  
The U.S ◽  
Reactor Internals

Reactor internals important to nuclear power plant safety shall be designed to accommodate steady-state and transient vibratory loads throughout the service life of the reactor. Operating experience has revealed failures of reactor internals in both pressurized water reactors (PWRs) and boiling water reactors (BWRs) due to flow-induced vibrations (FIVs). U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide 1.20 presents a Comprehensive Vibration Assessment Program (CVAP) that the NRC staff considers acceptable for use in verifying the structural integrity of reactor internals for FIV prior to commercial operation. A CVAP supports the NRC reviews of applications for new nuclear reactor construction permits or operating licenses under 10 CFR Part 50, as well as design certifications and combined licenses that do not reference a standard design under 10 CFR Part 52. The overall CVAP should be implemented in conjunction with preoperational and initial startup testing. For prototype reactor internals, the comprehensive program should consist of a vibration and fatigue analysis, a vibration measurement program, an inspection program, and a correlation of their results. Validation and benchmarking processes should be integrated into the CVAP throughout each individual program. Based on the authors’ experiences in Advanced Boiling Water Reactor and AP1000® CVAPs and based on detailed reviews of the U.S. Evolutionary Power Reactor and the U.S. Advanced Pressurized Water Reactor CVAPs, this article summarizes the essential CVAP validation and benchmarking processes with proper consideration of bias errors and random uncertainties. This article provides guidance to a successful CVAP that satisfies the NRC requirements and ensures the reliability of the evaluation of potential adverse flow effects on nuclear power plant components.

The Sensitivity of Risk-Informed Reactor Structural Integrity Analysis Results to Various Interpretations of Warm Pre-Stress

2009 ◽  
Author(s):  
Mark T. EricksonKirk ◽  
Terry L. Dickson
Keyword(s):  
Structural Integrity ◽  
The United States ◽  
Integrated Analysis ◽  
Total Risk ◽  
Pressurized Water ◽  
Best Estimate ◽  
Water Reactors ◽  
Using Data ◽  
Accident Scenarios ◽  
Reactor Operating

Warm pre-stress, or WPS, is a phenomenon by which the apparent fracture toughness of ferritic steel can be elevated in the fracture mode transition if crack is first “pre-stressed” at an elevated temperature. Taking proper account of WPS is important to the accurate modeling of the postulated accident scenarios that, collectively, are referred to as pressurized thermal shock, and to the accurate modeling of routine cool-down transients. For both accident and routine cool-downs the transients begin at the reactor operating temperature (approximately 290°C for pressurized water reactors in the United States) and proceed to colder temperatures as time advances. The probabilistic fracture mechanics code FAVOR, which is being used by the NRC to provide the technical basis for risk-informed revisions of 10 CFR 50.61 and 10 CFR 50 Appendix G, adopts a model of WPS as part of its fracture driving force module. In this paper we assess the conservatism inherent to the FAVOR WPS model relative to a best-estimate WPS model constructed using data recently produced by the European Commission “SMILE” project and published by Moinereau and colleagues. Assessments of the conservatisms inherent to the so-called “conservative principle” WPS model, and also to a classic LEFM model that does not credit WPS are also made. The data presented herein demonstrate that, for an integrated analysis of PTS risk, the FAVOR and conservative principle WPS models both over-estimate the vessel failure risk by a factor of between 2 and 3× relative to the best estimate model. Our examination of the effect of WPS models on the predictions of individual transients reveals that for the severe transients that dominate risk there is little difference (usually less than 4×) between the conditional probabilities of crack initiation and of through wall cracking predicted by the different WPS models. There are considerable differences in the predictions of the various WPS and non-WPS models for low severity transients, however, the contribution of these transients to the total risk of vessel failure is small.

Coordinated Research Projects at IAEA Concerning RPV and Structural Integrity

2006 ◽  
Author(s):  
William Server ◽  
Timothy Hardin ◽  
Milan Brumovsky´
Keyword(s):  
Ferritic Steel ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Current Status ◽  
Research Projects ◽  
Energy Agency ◽  
Pressurized Water ◽  
Toughness Testing ◽  
Water Reactors

The International Atomic Energy Agency (IAEA) has had a series of reactor pressure vessel (RPV) structural integrity programs that started back in the 1970s. These Coordinated Research Projects most recently have focused on use of the Master Curve fracture toughness testing approach for RPV and other ferritic steel components and on the issue of pressurized thermal shock (PTS) in operating pressurized water reactors. This paper will provide the current status for these projects and discuss the implications for improved safety of key ferritic steel components in nuclear power plants (NPPs).

The Development of Radiation Embrittlement Models for U.S. Power Reactor Pressure Vessel Steels

2006 ◽  
Author(s):  
J. A. Wang ◽  
N. S. V. Rao ◽  
S. Konduri
Keyword(s):  
Pressure Vessel ◽  
Prediction Models ◽  
Irradiation Temperature ◽  
Reactor Pressure Vessel ◽  
Boiling Water ◽  
Radiation Embrittlement ◽  
Pressurized Water ◽  
Water Reactor ◽  
Water Reactors ◽  
Reactor Pressure

The information fusion technique is used to develop radiation embrittlement prediction models for reactor pressure vessel (RPV) steels from U.S. power reactors, including boiling water reactors and pressurized water reactors. The Charpy transition temperature-shift data is used as the primary index of RPV radiation embrittlement in this study. Six parameters—Cu, Ni, P, neutron fluence, irradiation time, and irradiation temperature—are used in the embrittlement prediction models. The results indicate that this new embrittlement predictor achieved reductions of about 49.5% and 52% in the uncertainties for plate and weld data, respectively, for pressurized water reactor and boiling water reactor data, compared with the Nuclear Regulatory Commission Regulatory Guide 1.99, Rev. 2. The implications of dose-rate effect and irradiation temperature effects for the development of radiation embrittlement models are also discussed.

Development of Structural Integrity Assessment Method for Flow-Induced Vibration of Reactor Internals in PWR

2018 ◽  
Author(s):  
Hirokazu Sugiura ◽  
Shigeyuki Watanabe ◽  
Akihisa Iwasaki ◽  
Hideyuki Morita ◽  
Hideyuki Sakata ◽  
...  
Keyword(s):  
Model Test ◽  
Structural Integrity ◽  
Pressure Fluctuations ◽  
Vibration Response ◽  
Scale Model ◽  
Empirical Equations ◽  
Cfd Analysis ◽  
Pressurized Water ◽  
Forcing Function ◽  
Reactor Internals

For verifying the structural integrity of the reactor internals (RIs) in a pressurized water reactor (PWR) plant, it is important to estimate the vibration response of the core barrel (CB) due to flow turbulence. Instead of scale model test, the computational fluid dynamics (CFD) has been expected as a method to predict the turbulence forcing function for the response analysis of the CB. In this article, a hybrid approach combining empirical equations based on flow test and CFD analysis is proposed in order to predict the turbulence forcing function. The scale model test of new RIs, which were developed by Mitsubishi Heavy Industries, Ltd., was conducted, and the pressure fluctuations for the turbulence forcing function and the vibration response of the CB were measured. The pressure fluctuations were calculated by CFD analysis, and the vibration analysis using the turbulence forcing function determined from the calculated pressure fluctuations was performed. This article provides the scale model test data and the empirical equations of the turbulence forcing functions, and validation results of the proposed method to predict the turbulence forcing function using CFD.

Mechanical Characterization for a Large Test Design of a Dissimilar Metals Welding With a Narrow Gap Nickel Alloy Weld: Experimental and Numerical Analysis on Specimens

2012 ◽  
Author(s):  
Myriam Bourgeois ◽  
Olivier Ancelet ◽  
Stéphane Marie ◽  
Stephane Chapuliot
Keyword(s):  
Nickel Alloy ◽  
Structural Integrity ◽  
Atomic Energy Commission ◽  
Dissimilar Metals ◽  
Narrow Gap ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Integrity Assessment ◽  
Piping Systems ◽  
Water Reactors

Dissimilar metal welds are a common feature of light water reactors in connections between ferritic components and austenitic stainless steel piping systems. Inspection difficulties, variability of material properties and residual stresses all combine to create problems for structural integrity assessment. Within the framework of European project STYLE, a fracture test on a pipe containing a through wall crack in a narrow gap Nickel alloy Dissimilar Metals (DMWinc) is under preparation. The work is focusing on the nickel alloy - ferrite steel interface which is the weakest area of such welded pipes in front of ductile tearing. The study temperature is 300°C, which covers typical temperatures in components like hot pipes in the primary coolant system of pressurized water reactors. This paper gives an overview of the program and the first results of works which is been carried out by the French Atomic Energy Commission and Alternative Energies (CEA) in order to study the mechanical properties and integrity of component of the DMWinc pipes provided and designed by AREVA France.

Helium Effects on 316L Austenitic Stainless Steel Fracture Mechanism

2016 ◽  
Vol 713 ◽  
pp. 228-231 ◽  
Author(s):  
I. Villacampa ◽  
Jia Chao Chen ◽  
Philippe Spätig ◽  
Hans Peter Seifert ◽  
F. Duval
Keyword(s):  
Fracture Mechanism ◽  
Nuclear Reactor ◽  
Pressurized Water ◽  
Term Operation ◽  
316L Austenitic Stainless Steel ◽  
Radiation Induced ◽  
Water Reactors ◽  
High Doses ◽  
Reactor Internals

The most common fracture mechanism of nuclear reactor internals is irradiation-assisted stress corrosion cracking (IASCC). Its susceptibility at relatively low dose is dominated by conventional mechanisms such as radiation-induced segregation and radiation hardening. However, the aging of the nuclear fleet combined with the increase of their life-span reveals other mechanisms that could play an important role on IASCC susceptibility. A large amount of helium (He) can be accumulated in reactor internal components of pressurized water reactors (PWR) after long term operation. This occurrence could significantly increase (or even dominate) the IASCC susceptibility at high doses. He has been homogeneously implanted in an especially designed miniaturized specimen at 300°C up to 1000 appm. Slow strain rate tests (SSRT) results in high temperature air and in simulated PWR conditions indicate that homogenized, as-implanted He does not have a significant effect on IASCC up to 1000 appm under these test conditions.

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