Repair Welding of Irradiated Reactor Pressure Vessels Steel and Current Direction of Revising JSME Fitness-for-Service Code

2017 ◽  
Author(s):  
Hiroyuki Adachi ◽  
Ryuji Kimura ◽  
Yusuke Kono
Keyword(s):  
Power Plants ◽  
Primary Repair ◽  
Pressure Vessels ◽  
Light Water Reactors ◽  
Weld Repair ◽  
Asme Code ◽  
Neutron Transmutation ◽  
Water Reactors ◽  
Reactor Pressure ◽  
Reactor Internals

A primary repair option of Light Water Reactors (LWR) components is welding. However, it is known that welding on steels that have been exposed to neutron irradiation [i.e. Reactor Pressure Vessels (RPV) in PWR and Reactor Internals in BWR] can result in Helium Induced Cracking (HeIC). Helium forms from neutron transmutation reactions of Boron (B) and Nickel (Ni) during operation of the plant. In order to address this issue and establish verified methods for weld repair of irradiated RPV and Reactor Internals materials in Japanese power plants, an investigation denominated WIM (Welding of Irradiated Materials) Project was conducted; the WIM project was carried out between the years of 1997 and 2004 in an analytical conservative manner, correlating the results of weld repaired irradiated materials with the concentration of helium and the heat input used while welding. It was concluded that, under determined conditions, the irradiated materials were able to be successfully welded in accordance with the requirements established in both the JSME and ASME Code Cases. In the light of such discovery, the necessity of establishing and a new code case and revising the standard JSME Rules on Fitness-for-Service concerning the weld repair of irradiated RPV and Reactor Internals steels is currently under investigation.

Quantification of the Conservatisms in the Flaw Acceptability Assessment of Doel 3 and Tihange 2 RPVs Containing Hydrogen Flakes

2019 ◽  
Author(s):  
Valéry Lacroix ◽  
Pierre Dulieu
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Pressure Vessels ◽  
Assessment Methodology ◽  
Operating Modes ◽  
Safety Case ◽  
Asme Code ◽  
Component Manufacturing ◽  
Accident Conditions ◽  
Reactor Pressure

Abstract During the 2012 outage at Doel 3 and Tihange 2 Nuclear Power Plants, a large number of quasi-laminar indications were detected in the lower and upper core shells of the reactor pressure vessels (RPVs). The observed indications could subsequently be attributed to hydrogen flaking induced during the component manufacturing process. As a consequence, both units remained core unloaded pending the elaboration of an extensive Safety Case demonstrating that they can be safely operated. One of the most challenging parts of this demonstration was the Flaw Acceptability Assessment, aiming at demonstrating that the identified indications do not jeopardize the integrity of the reactor vessel in all operating modes, transients and accident conditions. This analysis was done by using a methodology: innovative, in line with existing ASME Code Section XI requirements, specific, sufficiently wide to be accepted and, first and foremost, conservative. Through a brief reminder of the Flaw Acceptability Assessment methodology, the paper presents the main hypotheses done for the calculation and quantifies the conservatism related to each of them. This quantification clearly highlights the reliability of final result i.e., the demonstration of the Fitness-for-Service for continued operation of both Doel 3 and Tihange 2 RPVs.

Ensuring the Performance of Nuclear Reactor Pressure Vessels for Long-Time Service

2010 ◽  
Author(s):  
Randy K. Nanstad ◽  
G. Robert Odette ◽  
Mikhail A. Sokolov
Keyword(s):  
Pressure Vessels ◽  
High Dose ◽  
Critical Element ◽  
Light Water Reactors ◽  
Time Service ◽  
Light Water ◽  
Time Operation ◽  
Long Time ◽  
Water Reactors ◽  
Reactor Pressure

Structural integrity of the reactor pressure vessel is a critical element in demonstrating the capability of light water reactors for operation to at least 80 y. The Light Water Reactor Sustainability Program Plan is a collaborative program between the U.S. Department of Energy and the private sector directed at extending the life of the present generation of nuclear power plants to enable such long-time operation. Given that the current generation of light water reactors were intended to operate for 40 y, there are significant issues that need to be addressed to reduce the uncertainties in regulatory application. The neutron dose to the vessel will at least double, and the database for such high dose levels under the low flux conditions in the vessel is nonexistent. Associated with this factor are uncertainties regarding flux effects, effects of relatively high nickel content, uncertainties regarding application of fracture mechanics, thermal annealing and reirradiation. The issue of high neutron fluence/long irradiation times and flux effects is the highest priority. Both data and mechanistic understanding are needed to enable accurate, reliable embrittlement predictions at high fluences. This paper discusses the major issues associated with long-time operation of existing RPVs, the LWRSP plans to address those issues, and recent relevant results.

Verification of Probabilistic Fracture Mechanics Analysis Code for Reactor Pressure Vessel

2021 ◽  
Vol 143 (4) ◽  
Author(s):  
Yinsheng Li ◽  
Genshichiro Katsumata ◽  
Koichi Masaki ◽  
Shotaro Hayashi ◽  
Yu Itabashi ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Working Group ◽  
Power Plants ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Irradiation Embrittlement ◽  
Energy Agency ◽  
Probabilistic Fracture Mechanics ◽  
Integrity Assessments ◽  
Reactor Pressure

Abstract Nowadays, it has been recognized that probabilistic fracture mechanics (PFM) is a promising methodology in structural integrity assessments of aged pressure boundary components of nuclear power plants, because it can rationally represent the influencing parameters in their inherent probabilistic distributions without over conservativeness. A PFM analysis code PFM analysis of structural components in aging light water reactor (PASCAL) has been developed by the Japan Atomic Energy Agency to evaluate the through-wall cracking frequencies of domestic reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and pressurized thermal shock (PTS) transients. In addition, efforts have been made to strengthen the applicability of PASCAL to structural integrity assessments of domestic RPVs against nonductile fracture. A series of activities has been performed to verify the applicability of PASCAL. As a part of the verification activities, a working group was established with seven organizations from industry, universities, and institutes voluntarily participating as members. Through one-year activities, the applicability of PASCAL for structural integrity assessments of domestic RPVs was confirmed with great confidence. This paper presents the details of the verification activities of the working group, including the verification plan, approaches, and results.

scholarly journals Reactor Vessel Internals Segmentation Experience using Mechanical Cutting Tools

2013 ◽  
Vol 10 (2) ◽  
pp. 6-10 ◽  
Author(s):  
Petr Pospíšil
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Cutting Tools ◽  
Reactor Vessel ◽  
Boiling Water Reactors ◽  
Abrasive Waterjet ◽  
Pressurized Water ◽  
Water Reactors ◽  
Mechanical Cutting ◽  
Reactor Internals

Abstract Some commercial nuclear power plants have been permanently shut down to date and decommissioned using dismantling methods. Other operating plants have decided to undergo an upgrade process that includes replacement of reactor internals. In both cases, there is a need to perform a segmentation of the reactor vessel internals with proven methods for long term waste disposal. Westinghouse has developed several concepts to dismantle reactor internals based on safe and reliable techniques, including plasma arc cutting (PAC), abrasive waterjet cutting (AWJC), metal disintegration machining (MDM), or mechanical cutting. Mechanical cutting has been used by Westinghouse since 1999 for both Pressurized Water Reactors (PWR’s) and Boiling Water Reactors (BWR’s) and its process has been continuously improved over the years. The complexity of the work requires well designed and reliable tools. Different band saws, disc saws, tube cutters and shearing tools have been developed to cut the reactor internals. All of those equipments are hydraulically driven which is very suitable for submerged applications. Westinghouse experience in mechanical cutting has demonstrated that it is an excellent technique for segmentation of internals. In summary, the purpose of this paper will be to provide an overview of the Westinghouse mechanical segmentation process, based on actual experience from the work that has been completed to date.

Advanced Structural Materials and Cladding

MRS Bulletin ◽  
2009 ◽  
Vol 34 (1) ◽  
pp. 20-27 ◽  
Author(s):  
T. Allen ◽  
H. Burlet ◽  
R.K. Nanstad ◽  
M. Samaras ◽  
S. Ukai
Keyword(s):  
High Temperature ◽  
Structural Materials ◽  
Ceramic Composites ◽  
Pressure Vessels ◽  
Austenitic Steels ◽  
High Temperature Strength ◽  
Light Water Reactors ◽  
Modeling Tools ◽  
Light Water ◽  
Water Reactors

AbstractAdvanced nuclear energy systems, both fission- and fusion-based, aim to operate at higher temperatures and greater radiation exposure levels than experienced in current light water reactors. Additionally, they are envisioned to operate in coolants such as helium and sodium that allow for higher operating temperatures. Because of these unique environments, different requirements and challenges are presented for both structural materials and fuel cladding. For core and cladding applications in intermediate-temperature reactors (400–650°C), the primary candidates are 9–12Cr ferritic–martensitic steels (where the numbers represent the weight percentage of Cr in the material, i.e., 9–12 wt%) and advanced austenitic steels, adapted to maximize high-temperature strength without compromising lower temperature toughness. For very high temperature reactors (>650°C), strength and oxidation resistance are more critical. In such conditions, high-temperature metals as well as ceramics and ceramic composites are candidates. For all advanced systems operating at high pressures, performance of the pressure boundary materials (i.e., those components responsible for containing the high-pressure liquids or gases that cool the reactor) is critical to reactor safety. For some reactors, pressure vessels are anticipated to be significantly larger and thicker than those used in light water reactors. The properties through the entire thickness of these components, including the effects of radiation damage as a function of damage rate, are important. For all of these advanced systems, optimizing the microstructures of candidate materials will allow for improved radiation and high-temperature performance in nuclear applications, and advanced modeling tools provide a basis for developing optimized microstructures.

Sign in / Sign up

Export Citation Format

Share Document