Application of Negligible Creep Criteria to Candidate Materials for HTGR Pressure Vessels

2011 ◽  
Vol 133 (2) ◽  
Author(s):  
R. I. Jetter ◽  
T.-L. Sham ◽  
R. W. Swindeman
Keyword(s):  
Wall Thickness ◽  
Pressure Vessel ◽  
Failure Modes ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Normal Operation ◽  
Allowable Stress ◽  
Regulatory Review ◽  
Advantages And Disadvantages ◽  
Reactor Pressure

Two of the proposed high temperature gas reactors (HTGRs) under consideration for a demonstration plant have the design object of avoiding creep effects in the reactor pressure vessel during normal operation. This work addresses the criteria for negligible creep in subsection NH, Division 1 of the ASME Boiler and Pressure Vessel Code, Sec. III, other international design codes, and some currently suggested criteria modifications and their impact on permissible operating temperatures for various reactor pressure vessel materials. The goal of negligible creep could have different interpretations depending on what failure modes are considered and associated criteria for avoiding the effects of creep. It is shown that for the materials of this study, consideration of localized damage due to cycling of peak stresses results in a lower temperature for negligible creep than consideration of the temperature at which the allowable stress is governed by the creep properties. In assessing the effect of localized cyclic stresses, it is also shown that consideration of cyclic softening is an important effect that results in a higher estimated temperature for the onset of significant creep effects than would be the case if the material were cyclically hardening. There are other considerations for the selection of vessel material besides avoiding creep effects. Of interest for this review are (1) the material’s allowable stress level and impact on the wall thickness (the goal being to minimize the required wall thickness) and (2) ASME code approval (inclusion as a permitted material in the relevant section and subsection of interest) to expedite regulatory review and approval. The application of negligible creep criteria to two of the candidate materials, SA533 and Mod 9Cr–1Mo (also referred to as Grade 91), and to a potential alternate, normalized and tempered 214 Cr–1Mo, is illustrated, and the relative advantages and disadvantages of the materials are discussed.

scholarly journals The Impact of Material and Design Criteria on the Assessment of Negligible Creep

2009 ◽  
Author(s):  
Robert I. Jetter ◽  
T.-L. Sam Sham ◽  
Robert W. Swindeman
Keyword(s):  
Wall Thickness ◽  
Failure Modes ◽  
Normal Operation ◽  
Regulatory Review ◽  
Advantages And Disadvantages ◽  
Asme Code ◽  
Vessel Material ◽  
The Impact ◽  
Reactor Pressure ◽  
Selection Of

Two of the proposed High Temperature Gas Reactors (HTGRs) under consideration for a demonstration plant have the design object of avoiding creep effects during normal operation. The goal of negligible creep could have different interpretations depending upon what failure modes are considered and associated criteria for avoiding the effects of creep. This paper addresses the criteria for negligible creep in Subsection NH of Section III of the ASME B&PV Code, other international design codes and some currently suggested criteria modifications and their impact on permissible operating temperatures for various reactor pressure vessel materials. There are a number of other considerations for the selection of vessel material besides avoiding creep effects. Of particular interest for this review are (1) the material’s allowable stress level and impact on wall thickness (the goal being to minimize required wall thickness) and (2) ASME Code approval (inclusion as a permitted material in the relevant Section and Subsection of interest) to expedite regulatory review and approval. The application of negligible creep criteria to two of the candidate materials, SA533 and Mod 9Cr-1Mo, and to a potential alternate, normalized and tempered 2 1/4 Cr-1Mo, are illustrated and the relative advantages and disadvantages are discussed.

Failure Probability Assessment for a Boiling Water Reactor Pressure Vessel Under Low Temperature Over-Pressure Event

2012 ◽  
Author(s):  
Hsoung-Wei Chou ◽  
Chin-Cheng Huang ◽  
Bo-Yi Chen ◽  
Hsien-Chou Lin ◽  
Ru-Feng Liu
Keyword(s):  
Low Temperature ◽  
Failure Probability ◽  
Pressure Vessel ◽  
Fracture Probability ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Water Reactor ◽  
Reactor Pressure

The fracture probability of a boiling water reactor pressure vessel for a domestic nuclear power plant in Taiwan has been numerically analyzed using an advanced version of ORNL’s FAVOR code. First, a model of the vessel beltline region, which includes all shell welds and plates, is built for the FAVOR code based on the plant specific parameters of the reactor pressure vessel. Then, a novel flaw model which describes the flaw types of surface breaking flaws, embedded weld flaws and embedded plate flaws are simulated along both inner and outer vessel walls. When conducting the fracture probability analyses, a transient low temperature over-pressure event, which has previously been shown to be the most severe challenge to the integrity of boiling water reactor pressure vessels, is considered as the loading condition. It is found that the fracture occurs in the fusion-line area of axial welds, but with only an insignificant failure probability. The low through-wall cracking frequency indicates that the analyzed reactor pressure vessel maintains sufficient stability until either the end-of-license or for doubling of the present license of operation.

Measurements of Stress During Thermal Shock in Clad Reactor Pressure Vessel Material Using Time-Resolved In-Situ Synchrotron X-Ray Diffraction

2018 ◽  
Author(s):  
Sam Oliver ◽  
Chris Simpson ◽  
Andrew James ◽  
Christina Reinhard ◽  
David Collins ◽  
...  
Keyword(s):  
Thermal Shock ◽  
Pressure Vessel ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Pressure Vessel Steel ◽  
X Ray Diffraction ◽  
Time Resolved ◽  
Vessel Steel ◽  
X Ray ◽  
Reactor Pressure

Nuclear reactor pressure vessels must be able to withstand thermal shock due to emergency cooling during a loss of coolant accident. Demonstrating structural integrity during thermal shock is difficult due to the complex interaction between thermal stress, residual stress, and stress caused by internal pressure. Finite element and analytic approaches exist to calculate the combined stress, but validation is limited. This study describes an experiment which aims to measure stress in a slice of clad reactor pressure vessel during thermal shock using time-resolved synchrotron X-ray diffraction. A test rig was designed to subject specimens to thermal shock, whilst simultaneously enabling synchrotron X-ray diffraction measurements of strain. The specimens were extracted from a block of SA508 Grade 4N reactor pressure vessel steel clad with Alloy 82 nickel-base alloy. Surface cracks were machined in the cladding. Electric heaters heat the specimens to 350°C and then the surface of the cladding is quenched in a bath of cold water, representing thermal shock. Six specimens were subjected to thermal shock on beamline I12 at Diamond Light Source, the UK’s national synchrotron X-ray facility. Time-resolved strain was measured during thermal shock at a single point close to the crack tip at a sample rate of 30 Hz. Hence, stress intensity factor vs time was calculated assuming K-controlled near-tip stress fields. This work describes the experimental method and presents some key results from a preliminary analysis of the data.

Accounting for the Effects of Inservice Inspection Flaw Depth Sizing Errors in Developing Flaw Size Distribution Tables

2009 ◽  
Author(s):  
S. R. Gosselin ◽  
F. A. Simonen
Keyword(s):  
Pressure Vessel ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Expert Elicitation ◽  
Flaw Size ◽  
Physical Models ◽  
Asme Code ◽  
Valid Comparison ◽  
Vessel Material ◽  
Reactor Pressure

Probabilistic fracture mechanics studies have addressed reactor pressure vessels that have high levels of material embrittlement. These calculations have used flaw size and density distributions determined from precise and optimized laboratory measurements made and validated with destructive methods as well as from physical models and expert elicitation. The experimental data were obtained from reactor vessel material samples removed from cancelled plants (Shoreham and the Pressure Vessel Research Users Facility (PVRUF)). Consequently, utilities may need to compare the numbers and sizes of reactor pressure vessel flaws identified by the plant’s inservice inspection program to the numbers and sizes of flaws assumed in prior failure probability calculations. This paper describes a method to determine whether the flaws in a particular reactor pressure vessel are consistent with the assumptions regarding the number and sizes of flaws used in other analyses. The approach recognizes that ASME Code Section XI examinations suffer from limitations in terms of sizing errors for very small flaws. Direct comparisons of a vessel specific flaw distribution with other documented flaw distributions would lead to pessimistic conclusions. This paper provides a method for a valid comparison that accounts for flaw sizing errors present in ASME Code Section XI examinations.

Ways for Extending the Reactor Pressure Vessel Lifetime

2008 ◽  
Author(s):  
Milan Brumovsky
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Integrity Assessment ◽  
Low Leakage ◽  
Vessel Integrity ◽  
Direct Use ◽  
Reactor Pressure ◽  
Lifetime Extension

Reactor pressure vessels are components that usually determine the lifetime of the whole nuclear power plant and thus also its efficiency and economy. There are several ways how to ensure conditions for reactor pressure vessel lifetime extension, mainly: - pre-operational, like: • optimal design of the vessel; • proper choice of vessel materials and manufacturing technology; - operational, like: • application of low-leakage core; • increase of water temperature in ECCS; • insertion of dummy elements; • vessel annealing; • decrease of conservatism during reactor pressure vessel integrity assessment e.g. using direct use of fracture mechanics parameters, like “Master Curve” approach. All these ways are discussed in the paper and some qualitative as well as quantitative evaluation is given.

Recent Progress Toward an Integrated Multiscale-Multiphysics Model of Reactor Pressure Vessel Embrittlement

2001 ◽  
Vol 677 ◽  
Author(s):  
B. D. Wirth ◽  
G. R. Odette ◽  
R. E. Stoller
Keyword(s):  
Radiation Damage ◽  
Cross Sections ◽  
Pressure Vessel ◽  
Production Cross ◽  
Vacancy Cluster ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Rate Theory ◽  
Late Time ◽  
Reactor Pressure

ABSTRACTThe continued safe operation of nuclear reactors and their potential for lifetime extension depends on ensuring reactor pressure vessel integrity. Reactor pressure vessels and structural materials used in nuclear energy applications are exposed to intense neutron fields that create atomic displacements and ultimately change material properties. The physical processes involved in radiation damage are inherently multiscale, spanning more than 15 orders of magnitude in length and 24 orders of magnitude in time. This paper reports our progress in developing an integrated, multiscale-multiphysics (MSMP) model of radiation damage for the prediction of reactor pressure vessel embrittlement. Key features of the fully integrated MSMP model include: i) combined molecular dynamics (MD) and kinetic lattice Monte Carlo (KMC) simulations of cascade defect production and cascade aging to produce cross-sections for vacancy, self- interstitial and vacancy-solute cluster size classes for times on the order of seconds; ii) an integrated reaction rate theory and thermodynamic code to predict the evolution of nanostructural and nanochemical features for times on the order of decades; iii) a micromechanics model to calculate the resulting mechanical property changes. This paper will focus on the combined use of MD and KMC to simulate the long-term rearrangement (aging) of defects in displacement cascades and thus, produce late-time production cross-sections for vacancy and vacancy cluster features.

Assessment of a Stress-Free Temperature Model for Residual Stresses in Surface Cladding of a Reactor Pressure Vessel

2017 ◽  
Author(s):  
B. Richard Bass ◽  
Paul T. Williams ◽  
Terry L. Dickson ◽  
Hilda B. Klasky
Keyword(s):  
Residual Stress ◽  
Finite Element ◽  
Thermal Expansion ◽  
Residual Stresses ◽  
Base Metal ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Normal Operation ◽  
Inner Surface ◽  
Reactor Pressure

This paper describes further results from an ongoing study of a simplified engineering model that is intended to account for effects of clad residual stresses on the propensity for initiation of preexisting inner-surface flaws in a commercial nuclear reactor pressure vessel (RPV). The deposition of stainless steel cladding during fabrication of an RPV generates residual stresses in the cladding and the heat affected zone of the under-lying base metal. In addition to residual stress, thermal strains are generated by the differential thermal expansion (DTE) of the cladding and base material due to temperature changes during normal operation. A simplified model used in the ORNL-developed FAVOR probabilistic fracture mechanics (PFM) code accounts for the clad residual stress by incorporating a stress-free temperature (SFT) approach. At the stress-free temperature (Ts-free), the model assumes there is no thermal strain, i.e., the thermal expansion stresses and clad residual stresses offset each other. For normal cool-down transients applied to the RPV, interactions of the latter stresses generate additional crack driving forces on shallow, internal surface-breaking flaws near the clad/base metal interface; those flaws tend to dominate the RPV failure probability computed by FAVOR. In a previous report from this study (PVP2015-45086), finite element analysis was used to compare the stresses and stress-intensity factors (SIF) during a cool-down transient for two cases: (1) the existing SFT model of FAVOR, and (2) directly applied RPV clad residual stress (CRS) distribution obtained from empirical (hole-drilling) measurements made at room temperature on an RPV that was never put into service. However, those analyses were limited in scope and focused on a single flaw orientation. In this updated study, effects of CRS on the SIF histories computed for both circumferential and axial flaw orientations subjected to a cool-down transient were determined from an extended set of finite element analyses. Specifically, comparisons were made between results from applying CRS experimental data to ABAQUS two-dimensional, inner-surface flaw models and those generated by the FAVOR SFT model. It is demonstrated that the FAVOR-recommended SFT value of 488 °F produces conservatively high values of SIF relative to the use of CRS profiles in the ABAQUS models. For the vessel and flaw geometry and transient under study, the circumferential flaw (360° continuous) required a decrease of SFT down to 390 °F to match the CRS SIF histories. For the infinite axial flaw model, a decrease down to 300 °F matched the CRS SIF histories. Future plans are described to develop more general conclusions regarding the FAVOR model.

MAI Benchmark Campaign of International Software for Reactor Pressure Vessel Integrity Assessment

2014 ◽  
Author(s):  
Emilie Dautreme ◽  
Emmanuel Remy ◽  
Roman Sueur ◽  
Jean-Philippe Fontes ◽  
Karine Aubert ◽  
...  
Keyword(s):  
United States ◽  
Pressure Vessel ◽  
Software Tool ◽  
The United States ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Pressurized Water ◽  
Integrity Assessment ◽  
Benchmark Study ◽  
Reactor Pressure

Nuclear Reactor Pressure Vessel (RPV) integrity is a major issue concerning plant safety and this component is one of the few within a Pressurized Water Reactor (PWR) whose replacement is not considered as feasible. To ensure that adequate margins against failure are maintained throughout the vessel service life, research engineers have developed and applied computational tools to study and assess the probability of pressure vessel failure during operating and postulated loads. The Materials Ageing Institute (MAI) sponsored a benchmark study to compare the results from software developed in France, Japan and the United States to compute the probability of flaw initiation in reactor pressure vessels. This benchmark study was performed to assess the similarities and differences in the software and to identify the sources of any differences that were found. Participants in this work included researchers from EDF in France, CRIEPI in Japan and EPRI in the United States, with each organization using the probabilistic software tool that had been developed in their country. An incremental approach, beginning with deterministic comparisons and ending by assessing Conditional Probability of crack Initiation (CPI), provided confirmation of the good agreement between the results obtained from the software used in this benchmark study. This conclusion strengthens the confidence in these probabilistic fracture mechanics tools and improves understanding of the fundamental computational procedures and algorithms.

scholarly journals A Data Scatter for a Shift of the Ductile to Brittle Transition Temperature for WWER-1000 Reactor Pressure Vessel Materials

2018 ◽  
pp. 27-30
Author(s):  
V. Revka
Keyword(s):  
Transition Temperature ◽  
Standard Deviation ◽  
Pressure Vessel ◽  
Neutron Fluence ◽  
Safety Margin ◽  
Temperature Shift ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Data Scatter ◽  
Reactor Pressure

In the most countries that operate the nuclear power plants with reactor pressure vessels a safety margin accounting a data scatter is applied for a conservative evaluation of a radiation shift of the ductile to brittle transition temperature for RPV metal. This scatter is to a significant extent due to material inhomogeneity and errors in determining the temperature shift and neutron fluence. In the regulatory practice of Ukraine, the obsolete approaches are used that can lead to an underestimation or overestimation of the transition temperature shift depending on the number of test data points. In order to use the updated regulatory approaches that will be consistent with international practice, it is necessary to know the magnitude of the data scatter on the transition temperature shift which is characterized by a standard deviation. Therefore, the aim of the research work was to estimate the data scatter for WWER reactor pressure vessel materials using statistical methods. The paper presents the results of a statistical analysis for a large array of surveillance test data for WWER-1000 reactor pressure vessels of NPP units which are operated in Ukraine. The data scatter for RPV base and weld metal has been estimated using a statistical treatment for the dependencies of a transition temperature shift, ΔTF, on the fast (Е > 0,5 MeV) neutron fluence. The ΔTF values have been derived from the Charpy impact tests. The Charpy V-notch specimens have been irradiated in the nuclear power reactors within a neutron fluence range of (3,0 ÷ 92,2)·1022 m-2 in the frame of a national surveillance program. The analysis has shown the data scatter relative to the average regression line for RPV materials is characterized by a standard deviation of 5,5 °С. Based on the results obtained, it was suggested to use a double standard deviation of 11 °С as a safety margin to provide a conservative estimate for the radiation shift of the transition temperature of the WWER-1000 reactor pressure vessel materials.

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