Manufacturing and material properties of forgings for the reactor pressure vessel of the high temperature engineering test reactor

1997 ◽  
Vol 171 (1-3) ◽  
pp. 45-56 ◽  
Author(s):  
Ikuo Sato ◽  
Komei Suzuki
Keyword(s):  
High Temperature ◽  
Pressure Vessel ◽  
Material Properties ◽  
Reactor Pressure Vessel ◽  
Test Reactor ◽  
Reactor Pressure

Reactor pressure vessel design of the high temperature engineering test reactor

2004 ◽  
Vol 233 (1-3) ◽  
pp. 103-112 ◽  
Author(s):  
Yukio Tachibana ◽  
Shigeaki Nakagawa ◽  
Tatsuo Iyoku
Keyword(s):  
High Temperature ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Test Reactor ◽  
Reactor Pressure

Structural Integrity Research for Reactor Pressure Vessel Under In-Vessel Retention of a Core Melt

2016 ◽  
Author(s):  
Yongjian Gao ◽  
Yinbiao He ◽  
Ming Cao ◽  
Yuebing Li ◽  
Shiyi Bao ◽  
...  
Keyword(s):  
Pressure Vessel ◽  
Material Properties ◽  
Power Plants ◽  
Structural Integrity ◽  
Plastic Material ◽  
Plastic Region ◽  
Reactor Pressure Vessel ◽  
Elastic Plastic ◽  
Elastic Plastic Material ◽  
Reactor Pressure

In-Vessel Retention (IVR) is one of the most important severe accident mitigation strategies of the third generation passive Nuclear Power Plants (NPP). It is intended to demonstrate that in the case of a core melt, the structural integrity of the Reactor Pressure Vessel (RPV) is assured such that there is no leakage of radioactive debris from the RPV. This paper studied the IVR issue using Finite Element Analyses (FEA). Firstly, the tension and creep testing for the SA-508 Gr.3 Cl.1 material in the temperature range of 25°C to 1000°C were performed. Secondly, a FEA model of the RPV lower head was built. Based on the assumption of ideally elastic-plastic material properties derived from the tension testing data, limit analyses were performed under both the thermal and the thermal plus pressure loading conditions where the load bearing capacity was investigated by tracking the propagation of plastic region as a function of pressure increment. Finally, the ideal elastic-plastic material properties incorporating the creep effect are developed from the 100hr isochronous stress-strain curves, limit analyses are carried out as the second step above. The allowable pressures at 0 hr and 100 hr are obtained. This research provides an alternative approach for the structural integrity evaluation for RPV under IVR condition.

Manufacturing and Properties of Closure Head Forging Integrated With Flange for PWR Reactor Pressure Vessel

2004 ◽  
Author(s):  
Komei Suzuki ◽  
Etsuo Murai ◽  
Yasuhiko Tanaka ◽  
Iku Kurihara ◽  
Tomoharu Sasaki ◽  
...  
Keyword(s):  
Weld Joint ◽  
Pressure Vessel ◽  
Material Properties ◽  
Impact Test ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Conventional Design ◽  
Manufacturing Technologies ◽  
Reactor Pressure

Closure head forging (SA508, Gr.3 Cl.1) integrated with flange for PWR reactor pressure vessel has been developed. This is intended to enhance structural integrity of closure head resulted in elimination of ISI, by eliminating weld joint between closure head and flange in the conventional design. Manufacturing procedures have been established so that homogeneity and isotropy of the material properties can be assured in the closure head forging integrated with flange. Acceptance tensile and impact test specimens are taken and tested regarding the closure head forging integrated with flange as very thick and complex forgings. This paper describes the manufacturing technologies and material properties of the closure head forging integrated with flange.

Design and Fabrication of Reactor Pressure Vessel for High Temperature Engineering Test Reactor (HTTR)

2004 ◽  
Author(s):  
Yukio Tachibana ◽  
Shigeaki Nakagawa ◽  
Tatsuo Iyoku
Keyword(s):  
Pressure Vessel ◽  
Temper Embrittlement ◽  
Reactor Pressure Vessel ◽  
The Body ◽  
Inlet Temperature ◽  
Irradiation Embrittlement ◽  
Fabrication Procedure ◽  
Test Reactor ◽  
Service Inspection ◽  
Reactor Pressure

The reactor pressure vessel (RPV) of the HTTR is 5.5 m in inside diameter, 13.2 m in inside height, and 122 mm and 160 mm in wall thickness of the body and the top head dome, respectively. Because the reactor inlet temperature of the HTTR is higher than that of LWRs, 2 1/4Cr-1Mo steel is chosen for the RPV material. Fluence of the RPV is estimated to be less than 1×1017 n/cm2 (E>1 MeV), and so irradiation embrittlement is presumed to be negligible, but temper embrittlement is not. For the purpose of reducing embrittlement, content of some elements is limited on 2 1/4 Cr-1 Mo steel for the RPV using embrittlement parameters, J-factor and X. In this paper design, fabrication procedure, and in-service inspection technique of the RPV for the HTTR are described.

Fracture Toughness Characterization of Low Upper-Shelf Linde 80 Weld Using Mini-C(T) Specimens

2018 ◽  
Author(s):  
Michael R. Ickes ◽  
J. Brian Hall ◽  
Robert G. Carter
Keyword(s):  
Fracture Toughness ◽  
Direct Measurement ◽  
Pressure Vessel ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Weld Metals ◽  
Surveillance Programs ◽  
Test Reactor ◽  
Reactor Pressure

The Charpy V-notch specimen is the most commonly used specimen geometry in reactor pressure vessel irradiation surveillance programs and there is an extensive stored inventory of irradiated broken Charpy specimens. The advantage of the mini-C(T) (4mm thick C(T)) specimen technique is that multiple specimens can be machined from each half of broken irradiated Charpy specimens. Fracture toughness specimens that can be machined from broken halves of standard Charpy specimens enable the direct measurement of fracture toughness which can be used for engineering evaluation of reactor pressure vessels. Work to validate the mini-C(T) specimens has been performed mostly on unirradiated reactor pressure vessel base and weld metals . In this study, mini-C(T) specimens were tested providing fracture toughness characterization of an irradiated low upper-shelf Linde 80 weld (WF-70). This weld was utilized in the Midland beltline and has been previously well characterized at ORNL with various types and sizes of fracture toughness specimens. The mini-C(T) specimens were machined from broken previously tested Charpy V-notch size specimens which were irradiated in a material test reactor. The effect of different methods of measuring the displacement on the results is assessed. The ASTM E1921 results are compared to previous test data produced from larger fracture toughness specimens. In addition, the sensitivity of T0 to the ASTM E1921 censoring value is discussed.

Investigation on the Effects of Carbon Macro-Segregation of Mechanical Properties Using a Large-Scale Forged Low Alloy Steel for a BWR Reactor Pressure Vessel

2020 ◽  
Author(s):  
Takahiro Hayashi ◽  
Takuya Ogawa ◽  
Rie Sumiya ◽  
Tetsushi Yamaoka ◽  
Shigeaki Tanaka ◽  
...  
Keyword(s):  
Mechanical Properties ◽  
Carbon Content ◽  
Pressure Vessel ◽  
Material Properties ◽  
Large Scale ◽  
Structural Reliability ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Steel Making ◽  
Reactor Pressure

Abstract Control of carbon macro-segregation in the steel-making process for large steel forgings is of great importance in order to achieve the material properties and structural reliability required for the pressure vessels of nuclear power plant components. It is well known that high carbon content due to carbon macro-segregation can affect the mechanical properties of steels, leading to decreases in ductility and fracture toughness. In this study, possible effects of carbon macro-segregation have been examined using a large-scale forged steel “bottom head dome” of a reactor pressure vessel (RPV) manufactured for a recent BWR. Material testing conducted included chemical analyses, tensile tests and Charpy impact tests. In the center part of the concave disk-shaped forged material, carbon content varied slightly in the material thickness direction within the range of carbon content requirement, as expected from the relationship between the solidification and the resultant segregation process in the cast ingot material and the forging process from the ingot to the dome material. The results of each mechanical test also showed full compliance with the properties required in the code regardless of the carbon content at each of the thickness locations examined. All the tests results demonstrated that with the steel-making technology and practice employed, carbon macro-segregation is well controlled to achieve the required material properties even in large-scale forged materials used in BWRs.

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