The Bandsaw: A Highly Sophisticated Dismantling Technique for the Karlsruhe Multi Purpose Research Reactor

2003 ◽  
Author(s):  
U. Leffrang ◽  
D. Stanke ◽  
E. Prechtl ◽  
W. Suessdorf
Keyword(s):  
Pressure Vessel ◽  
Research Reactor ◽  
Reactor Pressure Vessel ◽  
Reactor Pressure

For the dry dismantling and segmentation of dedicated components within the reactor pressure vessel (RPV) of the Karlsruhe Multi Purpose Research Reactor (MZFR) a bandsaw has been applied. The general decommissioning concept for the MZFR as well as constructional details of the bandsaw itself and the experience gained so far during the segmentation of the RPV closure head and the upper spacer are given with the present paper.

CAREM-25: Energy and Safety for the Future

2002 ◽  
Author(s):  
La´zaro E. Pomier Ba´ez
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Research Reactor ◽  
Reactor Pressure Vessel ◽  
Water Desalination ◽  
Pressurized Water ◽  
Management Actions ◽  
Reactor Pressure

The CAREM-25 is an advanced small nuclear power plant being jointly developed by CNEA (Comisio´n Nacional de Energi´a Ato´mica) and INVAP S. E. in Argentina. CAREM is a modular 100 MWt pressurized water reactor with integral steam generators designed to be used for electricity generation (25 MWe) or as a research reactor or for water desalination. The plant has its entire primary coolant system within the reactor pressure vessel (hence ‘integral’ PWR), self-pressurized and relying entirely on natural convection. Fuel is standard 3.4% enriched PWR fuel, with burnable poison, and is refueled annually. It is a mature design, which could be deployed within a decade. The CAREM reactor shows itself as a design capable to fill the existing gap between a research reactor and a full-scale commercial nuclear power plant. From this point of view, CAREM is a very attractive option for countries who have not yet developed technical capabilities, human resources, and an industrial infrastructure needed for a medium or large size nuclear units. Advanced nuclear reactor designs, such as the CAREM reactor, include several improvements related to safety issues either enhancing the passive safety functions or allowing plant operators more time to undertake different management actions against radioactive releases to the environment The enhancement and development of safety measures for this type of system are a top priority of designers. In the development of the nuclear power plant CAREM, the possibility of including a passive metallic in-vessel container in its design is being considered, to arrest the reactor pressure vessel meltdown sequence during a core damaging event, and thereof prevent its failure. The present paper will provide an overview of the CAREM-25 technology as a general purpose generation unit, and highlight the work on safety measure’s enhancements, including design modifications aiming to improve overall safety targets of the plant.

Study of the Evolution of Defects in the Structure of Reactor Pressure Vessel by Rate Theory

2014 ◽  
Vol 10 (1) ◽  
pp. 123-127 ◽  
Author(s):  
Gyeong-Geun Lee ◽  
Yong-Bok Lee ◽  
Min-Chul Kim ◽  
Junhyun Kwon
Keyword(s):  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Rate Theory ◽  
Reactor Pressure

The effects of thermal stress on the crack propagation in AP1000 reactor pressure vessel

2020 ◽  
Vol 110 ◽  
pp. 102798
Author(s):  
KaiTai Liu ◽  
Mei Huang ◽  
JunJie Lin ◽  
HaiPeng Jiang ◽  
BoXue Wang ◽  
...  
Keyword(s):  
Thermal Stress ◽  
Crack Propagation ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Ap1000 Reactor ◽  
Reactor Pressure

scholarly journals Analysis of the Technological Evolution of Materials Requirements Included in Reactor Pressure Vessel Manufacturing Codes

2021 ◽  
Vol 13 (10) ◽  
pp. 5498
Author(s):  
Alvaro Rodríguez-Prieto ◽  
Mariaenrica Frigione ◽  
John Kickhofel ◽  
Ana M. Camacho
Keyword(s):  
Sustainable Development ◽  
Pressure Vessel ◽  
Nuclear Regulatory Commission ◽  
Green Energy ◽  
Reactor Pressure Vessel ◽  
Technical Literature ◽  
Pressurized Water ◽  
Water Reactors ◽  
Regulatory Commission ◽  
Reactor Pressure

The growth of green energy technologies within the frame of the 7th Sustainable Development Goal (SDG) along with the concern about climatic changes make nuclear energy an attractive choice for many countries to ensure energy security and sustainable development as well as to actively address environmental issues. Unlike nuclear equipment (immovable goods), which are often well-catalogued and analyzed, the design and manufacturing codes and their standardized materials specifications can be considered movable and intangible goods that have not been thoroughly studied based on a detailed evaluation of the scientific and technical literature on the reactor pressure vessel (RPV) materials behavior. The aim of this work is the analysis of historical advances in materials properties research and associated standardized design codes requirements. The analysis, based on the consolidated U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.99 Rev.2 model, enables determination of the best materials options, corresponding to some of the most widely used material specifications such as WWER 15Kh2MFAA (used from the 1970s and 1980s; already in operation), ASME SA-533 Grade B Cl.1 (used in pressurized water reactor-PWR 2nd–4th; already in operation), DIN 20MnMoNi55 and DIN 22NiMoCr37 (used in PWR 2nd–4th) as well as ASTM A-336 Grade F22V (current designs). Consequently, in view of the results obtained, it can be concluded that the best options correspond to recently developed or well-established specifications used in the design of pressurized water reactors. These assessments endorse the fact that nuclear technology is continually improving, with safety being its fundamental pillar. In the future, further research related to the technical heritage from the evolution of materials requirements for other clean and sustainable power generation technologies will be performed.

The ageing behavior and the correlation between hardness and coercivity of Cu-rich reactor pressure vessel model steels

2021 ◽  
Vol 527 ◽  
pp. 167698
Author(s):  
Xuejiao Wang ◽  
Wenjiang Qiang ◽  
Guogang Shu ◽  
Junwei Qiao ◽  
Yucheng Wu
Keyword(s):  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Reactor Pressure

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.

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