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.

Verifying Calculation of Reactor Pressure Vessel Fast Neutron Fluence

2014 ◽  
Vol 986-987 ◽  
pp. 985-989
Author(s):  
Qiao Feng Liu ◽  
Jing Ru Han ◽  
Hai Ying Chen ◽  
Chun Ming Zhang
Keyword(s):  
Standard Deviation ◽  
Fast Neutron ◽  
Radiation Protection ◽  
Pressure Vessel ◽  
Neutron Fluence ◽  
Reactor Pressure Vessel ◽  
Light Water Reactor ◽  
Light Water ◽  
Water Reactor ◽  
Reactor Pressure

The reactor pressure vessel is an unchangeable component of the light water reactor. To some extent, the life of the pressure vessel depends on the fast neutron fluence. In addition, the fast neutron fluence is an important parameter for radiation protection. So, the fast neutron fluence is one of the main parameters which should be verifying calculated by the reviewers. The verifying calculation of the fast neutron fluence of one reactor pressure vessel is presented in this paper, and the standard deviation between the verifying and designing calculations is lower than 10%. The reasons for the deviation 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.

Fracture Toughness Characterization of Midland Beltline Weld After High Dose of Irradiation

2006 ◽  
Author(s):  
Mikhail A. Sokolov ◽  
Randy K. Nanstad
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Neutron Fluence ◽  
National Laboratory ◽  
Reactor Pressure Vessel ◽  
High Dose ◽  
Pressure Vessel Steel ◽  
Vessel Steel ◽  
Oak Ridge ◽  
Reactor Pressure

The Heavy-Section Steel Irradiation (HSSI) Program at Oak Ridge National Laboratory includes a task to investigate the shape of the fracture toughness master curve for reactor pressure vessel steel highly embrittled as a consequence of irradiation exposure, and to examine the ability of the Charpy 41-J shift to predict the fracture toughness shift. As part of this task, a low upper-shelf WF-70 weld obtained from the beltline region of the Midland Unit 1 reactor pressure vessel was characterized in terms of static initiation and Charpy impact toughness in the unirradiated and irradiated conditions. Irradiation of this weld was performed at the University of Michigan Ford Reactor at 288°C to neutron fluence of 3.4×1019 neutron/cm2 in the HSSI irradiation-anneal-reirradiation facility. This reusable facility allowed the irradiation of either virgin or previously irradiated material in a well-controlled temperature regime, including the ability to perform in-situ annealing. This was the last capsule irradiated in this facility before reactor shut down. Thus, the Midland beltline weld was irradiated within the HSSI Program to three fluences — 0.5×1019; 1.0×1019; and 3.4×1019 neutron/cm2. It was anticipated that it would provide an opportunity to address fracture toughness curve shape and Charpy 41-J shift compatibility issues at different levels of embrittlement, including the highest dose considered to be in the range of the current end of life fluence. It was found that the Charpy 41-J shift practically saturated after neutron fluence of 1.0×1019 neutron/cm2. The transition fracture toughness shift after 3.4×1019 neutron/cm2 was only slightly higher than that after 1.0×1019 neutron/cm2. In all cases, transition fracture toughness shifts were lower than predicted by the Regulatory Guide 1.99, Rev. 2 equation.

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.

Conservative estimations of irradiation embrittlement of reactor pressure vessel steels for WWER-1000 lifetime prediction

2019 ◽  
Vol 15 (1) ◽  
pp. 246-257
Author(s):  
Nikolai Petrovich Anosov ◽  
Vladimir Nikolaevich Skorobogatykh ◽  
Lyubov’ Yur’yevna Gordyuk ◽  
Vasilii Anatol’evich Mikheev ◽  
Egor Vasil’yevich Pogorelov ◽  
...  
Keyword(s):  
Transition Temperature ◽  
Brittle Fracture ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Lifetime Prediction ◽  
Critical Brittleness Temperature ◽  
Irradiation Embrittlement ◽  
Content Type ◽  
Brittleness Temperature ◽  
Reactor Pressure

Purpose The purpose of this paper is to consider a procedure of water-water energetic reactor (WWER) reactor pressure vessel (RPV) lifetime prediction at the stages of design and lifetime extension using the standard irradiation embrittlement parameters as defined in regulatory documents. A comparison is made of the brittle fracture resistance (BFR) values evaluated using two criteria: shift in the critical brittleness temperature ΔTc or shift in the brittle-to-ductile transition temperature ΔTp and without shifts (Tc and Tp). Design/methodology/approach The radiation resistance was determined using the following three approaches: calculation based on standard values ΔTc and Tc0 or ΔTp and Tp0 (a level of excessive conservatism); calculation based on standard value ΔTc and actual value Tc0 or actual values ΔTp and Tp0 (the level of realistic conservatism); or calculation based on actual values of Tc and Tc0 or Tp and Tp0 (the level of actual conservatism). The BFR was evaluated based on the results of testing the specimens subjected to irradiation in research reactors as well as surveillance specimens subjected to irradiation immediately under operating conditions. Findings The excessive conservatism in determining the actual lifetime of nuclear reactor vessel materials can be eliminated by using the immediate values of critical brittleness temperature and ductile-to-brittle transition temperature. Originality/value Obtained results can be applied to extend WWER vessel operating time at the stages of designing and operation due to substantiated decrease in conservatism. And it will allow carrying out a statistical substantiated assessment of the resistance to brittle fracture of the RPV steels.

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.

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