Development of Reactor Pressure Vessel Design, Neutron Fluence Calculation, and Material Specification to Minimize Irradiation Effects

2008 ◽  
pp. 130-130-15 ◽  
Author(s):  
C Leitz ◽  
J Koban
Keyword(s):  
Pressure Vessel ◽  
Neutron Fluence ◽  
Reactor Pressure Vessel ◽  
Irradiation Effects ◽  
Reactor Pressure

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.

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.

Investigation on Integrity of JMTR Reactor Pressure Vessel

2010 ◽  
Author(s):  
Hiroshi Ide ◽  
Akihiro Kimura ◽  
Hiroshi Miura ◽  
Yoshiharu Nagao ◽  
Naohiko Hori ◽  
...  
Keyword(s):  
Pressure Vessel ◽  
Neutron Fluence ◽  
Visual Observation ◽  
Reactor Pressure Vessel ◽  
Materials Testing ◽  
View Point ◽  
Componential Analysis ◽  
Underwater Camera ◽  
Reactor Pressure

Visual observation of inner side of a reactor pressure vessel of Japan Materials Testing Reactor (JMTR) was carried out using an underwater camera before the JMTR refurbishment work from the view point of its long term utilization, because the reactor pressure vessel of the JMTR will be used continuously after restart of the JMTR. As a result of the visual observation, the harmful wound was not confirmed. Moreover, there was no loosening of the bolts and the screws. On the other hand, adhesion materials which can be easily removed using the gauze were observed around nozzles in a top closure of the reactor pressure vessel. A major component of the adhesion materials is an iron as a result of the componential analysis. However, no significant problem affecting the integrity of the reactor pressure vessel was observed, and then the integrity of the reactor pressure vessel was confirmed. From view points of the stress corrosion cracking, fast neutron fluence and fatigue, it became clear that the reactor pressure vessel of the JMTR can be used for more than 20 years. The visual observation by the underwater camera is to be carried out periodically to confirm the integrity of the reactor pressure vessel in future.

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.

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