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.

Features and Results of Several Programs of Radiation-Damage Surveillance of Power Reactor Pressure Vessels

1967 ◽  
Vol 89 (1) ◽  
pp. 221-232 ◽  
Author(s):  
Charles Z. Serpan ◽  
L. E. Steele ◽  
J. R. Hawthorne
Keyword(s):  
Radiation Damage ◽  
Pressure Vessel ◽  
Power Reactor ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Surveillance Program ◽  
Surveillance Programs ◽  
Meaning And Purpose ◽  
Reactor Pressure

The meaning and purpose of reactor pressure vessel surveillance is briefly discussed. Features of the surveillance programs in the Yankee, Big Rock Point, SM-1, and SM-1A reactors are briefly described along with results of testing metallurgical specimens from these programs. Additionally, the surveillance program to be effected in the Army MH-1A reactor is described. Certain problems which have occurred in the course of these programs are discussed as well as the proposed ASTM recommendations for radiation-damage surveillance programs. The value of these programs to reactor operators is reviewed with relation to the results obtained to date.

Modeling Cascade Aging in Dilute Fe-Cu Alloys

2001 ◽  
Vol 701 ◽  
Author(s):  
B. D. Wirth ◽  
G. R. Odette
Keyword(s):  
Cross Sections ◽  
Production Cross ◽  
Three Dimensional ◽  
Continuous Distribution ◽  
Pressure Vessels ◽  
Rate Theory ◽  
Irradiation Embrittlement ◽  
Cu Alloys ◽  
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 highly non-equilibrium defect concentrations, consisting of a shell of self-interstitial atom and clusters surrounding a vacancy-rich core, over picosecond time scales. This spatially correlated defect production initiates a long chain of events responsible for microstructure evolution and hence irradiation embrittlement. In this paper, we describe the combined use of molecular dynamics (MD) and kinetic lattice Monte Carlo (KMC) to simulate the long-term rearrangement (aging) of displacement cascades in dilute Fe-Cu alloys. The simulations reveal the formation of a continuous distribution of three dimensional cascade vacancy-Cu cluster complexes and demonstrate the critical importance of spatial, as well as short and long-time correlated processes that mediate the effective production of primary defects. Finally, this approach can generate production cross-sections for vacancy-Cu clusters that can then be used in rate theory type models of long term global micro and microstructural evolution.

Study of the Evolution of Defects in the Structure of Reactor Pressure Vessel by Rate Theory

2014 ◽  
Vol 10 (1) ◽  
pp. 123-127 ◽  
Author(s):  
Gyeong-Geun Lee ◽  
Yong-Bok Lee ◽  
Min-Chul Kim ◽  
Junhyun Kwon
Keyword(s):  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Rate Theory ◽  
Reactor Pressure

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.

Attenuation of Radiation Damage and Neutron Field in Reactor Pressure Vessel Wall

2009 ◽  
pp. 190-190-6
Author(s):  
Milan Brumovsky ◽  
Milan Marek ◽  
Ladislav Zerola ◽  
Ladislav Viererbl ◽  
Victor N. Golovanov ◽  
...  
Keyword(s):  
Radiation Damage ◽  
Pressure Vessel ◽  
Vessel Wall ◽  
Neutron Field ◽  
Reactor Pressure Vessel ◽  
Reactor Pressure

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.

Sign in / Sign up

Export Citation Format

Share Document