Information Fusion Technique for Evaluating Radiation Embrittlement of Reactor Pressure Vessel Steels

2005 ◽  
Vol 127 (1) ◽  
pp. 98-104 ◽  
Author(s):  
J. A. Wang ◽  
S. Konduri ◽  
N. S. V. Rao
Keyword(s):  
Information Fusion ◽  
Pressure Vessel ◽  
Prediction Models ◽  
Irradiation Time ◽  
Irradiation Temperature ◽  
Reactor Pressure Vessel ◽  
Radiation Embrittlement ◽  
Fusion Technique ◽  
Information Fusion Technique ◽  
Reactor Pressure

A new approach that utilizes the information fusion technique was developed to predict the radiation embrittlement of reactor pressure vessel (RPV) steels. The Charpy transition temperature-shift data is used as the primary index of the 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 about 66% and 53% reductions, respectively, in the uncertainties for the update General Electric (GE) Boiling Water Reactor (BWR) plate and weld data compared to the Nuclear Regulatory Commission (NRC) Regulatory Guide 1.99, Rev. 2 (RG1.99/R2). The implications of irradiation temperature effects for the development of radiation embrittlement models are also discussed.

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.

Fracture Mechanics Assessment of Reactor Pressure Vessel Irradiated Structural Steel for Short Column Type Supports and Neutron Shield Tank

2020 ◽  
Author(s):  
Alexandria M. Carolan ◽  
J. Brian Hall ◽  
Stephen K. Longwell ◽  
F. Arzu Alpan ◽  
Gregory M. Imbrogno ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Structural Steel ◽  
Pressure Vessel ◽  
Radiation Effects ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Radiation Embrittlement ◽  
Short Column ◽  
License Renewal ◽  
Reactor Pressure

Abstract As plants apply for 80 year licensure (subsequent license renewal), the United States Nuclear Regulatory Commission (U.S. NRC) has queried the nuclear power plant industry to investigate the impact of neutron embrittlement (radiation effects) on the reactor pressure vessel (RPV) structural steel supports due to extended plant operation past 60 years. The radiation effects on RPV supports were previously investigated and resolved as part of Generic Safety Issue No. 15 (GSI-15) in NUREG-0933 Revision 3 [1], NUREG-1509 [2] (published in May 1996), and NUREG/CR-5320 [3] (published in January 1989) for design life (40 years) and for first license renewal (20 additional years). The conclusions in NUREG-0933, Revision 3 stated that there were no structural integrity concerns for the RPV support structural steels; even if all the supports were totally removed (i.e. broken), the piping has acceptable margin to carry the load of the vessel. Nevertheless, for plants applying for 80 year life licensure, the U.S. NRC has requested an evaluation to show structural integrity of the RPV supports by accounting for radiation embrittlement (radiation damage) for continued operation into the second license renewal period (i.e. 80 years). The RPV support designs in light water reactors are grouped into one of five categories or types of supports: (1) skirt; (2) long-column; (3) shield-tank; (4) short column; and (5) suspension. In this paper, two of these RPV support configurations (short column supports and neutron shield tank) will be investigated using fracture mechanics to evaluate the effect of radiation embrittlement of the structural steel supports for long term operations (i.e. 80 years). The technical evaluation of other support configurations will be provided in a separate technical publication at a future date.

scholarly journals Specific Features of Structural-Phase State and Properties of Reactor Pressure Vessel Steel at Elevated Irradiation Temperature

2017 ◽  
Vol 2017 ◽  
pp. 1-12
Author(s):  
E. A. Kuleshova ◽  
B. A. Gurovich ◽  
E. V. Krikun ◽  
A. S. Frolov ◽  
D. A. Maltsev ◽  
...  
Keyword(s):  
Pressure Vessel ◽  
Operating Temperature ◽  
Structural Phase ◽  
Irradiation Temperature ◽  
Reactor Pressure Vessel ◽  
Pressure Vessel Steel ◽  
Dislocation Loops ◽  
Radiation Induced ◽  
Higher Temperature ◽  
Reactor Pressure

This paper considers influence of elevated irradiation temperature on structure and properties of 15Kh2NMFAA reactor pressure vessel (RPV) steel. The steel is investigated after accelerated irradiation at 300°C (operating temperature of VVER-1000-type RPV) and 400°C supposed to be the operating temperature of advanced RPVs. Irradiation at 300°C leads to formation of radiation-induced precipitates and radiation defects-dislocation loops, while no carbide phase transformation is observed. Irradiation at a higher temperature (400°C) neither causes formation of radiation-induced precipitates nor provides formation of dislocation loops, but it does increase the number density of the main initial hardening phase—of the carbonitrides. Increase of phosphorus concentration in grain boundaries is more pronounced for irradiation at 400°C as compared to irradiation at 300°C due to influence of thermally enhanced diffusion at a higher temperature. The structural-phase changes determine the changes of mechanical properties: at both irradiation temperatures irradiation embrittlement is mainly due to the hardening mechanism with some contribution of the nonhardening one for irradiation at 400°C. Lack of formation of radiation-induced precipitates at T = 400°C provides a small ΔTK shift (17°C). The obtained results demonstrate that the investigated 15Kh2NMFAA steel may be a promising material for advanced reactors with an elevated operating temperature.

The Effect of Radiation-Induced Structural Changes under Accelerated Irradiation on the Behavior of Water-Cooled Reactor Pressure Vessel Steels

2013 ◽  
Vol 592-593 ◽  
pp. 573-576 ◽  
Author(s):  
Boris A. Gurovich ◽  
Evgenia A. Kuleshova ◽  
Dmitry A. Maltsev ◽  
Oleg Zabusov ◽  
Kirill Prikhodko ◽  
...  
Keyword(s):  
Pressure Vessel ◽  
Structural Changes ◽  
Structural Features ◽  
Reactor Pressure Vessel ◽  
Fast Neutron Flux ◽  
Radiation Embrittlement ◽  
Pressure Vessel Steels ◽  
Radiation Induced ◽  
Hardening Mechanisms ◽  
Reactor Pressure

In this paper the influence of fast neutron flux on the structural features and properties of VVER-1000 reactor pressure vessel steels was studied. It is shown that for high Ni steels the flux effect is due to hardening and non-hardening mechanisms of radiation embrittlement.

scholarly journals Radiation embrittlement of reactor pressure vessel materials of Rivne NPP unit 1 due to re-irradiation after recovery annealing

2019 ◽  
Vol 20 (3) ◽  
pp. 248-257
Author(s):  
M.G. Holiak ◽  
G.P. Grynchenko ◽  
V.M. Revka ◽  
O.V. Trygubenko ◽  
Yu.V. Chaykovskyi ◽  
...  
Keyword(s):  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Radiation Embrittlement ◽  
Recovery Annealing ◽  
Reactor Pressure

Numerical Simulation of Irradiation-Induced Grain-Boundary Phosphorous Segregation in Reactor Pressure Vessel Steels Using Rate Theory Model With First-Principles Calculations

2009 ◽  
Author(s):  
Ken-ichi Ebihara ◽  
Masatake Yamaguchi ◽  
Yutaka Nishiyama ◽  
Kunio Onizawa ◽  
Hiroshi Matsuzawa
Keyword(s):  
Grain Boundary ◽  
Dose Rate ◽  
Pressure Vessel ◽  
First Principles ◽  
Theory Model ◽  
Irradiation Temperature ◽  
Reactor Pressure Vessel ◽  
Rate Theory ◽  
First Principles Calculations ◽  
Reactor Pressure

The experimental results on neutron-irradiated reactor pressure vessel (RPV) steels have revealed grain boundary segregation of phosphorous (P) due to neutron irradiation, which may lead to intergranular fracture. Because of the lack of experimental database, however, the dependence of the segregation on variables such as dose, dose-rate, and temperature is not clear. Here, we incorporate the parameters determined by first-principles calculations into the rate theory model which was developed for bcc lattice on the basis of the fcc lattice model proposed by Murphy and Perks [1], and apply it to the simulation of irradiation-induced P segregation in bcc iron. We evaluate the grain boundary P coverage and discuss its dependence on dose-rate and irradiation temperature by comparing our results with previously reported results and experimental data. As results, we find that dose-rate does not affect the grain boundary P coverage within the range of our simulation condition and that the dependence on irradiation temperature differs remarkably from the previous results.

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