scholarly journals Carbon-14 release and speciation from carbon steel in highly alkaline conditions

Radiocarbon ◽  
2018 ◽  
Vol 60 (6) ◽  
pp. 1683-1690
Author(s):  
Frank Druyts ◽  
Sébastien Caes ◽  
Peter Thomas
Keyword(s):  
Carbon Steel ◽  
Corrosion Rate ◽  
Nuclear Power ◽  
Power Plants ◽  
Carbon 14 ◽  
Alkaline Conditions ◽  
High Level Nuclear Waste ◽  
High Level ◽  
Reactor Pressure ◽  
Static Corrosion

ABSTRACTThe release and the speciation of carbon species from irradiated JRQ carbon steel samples, representative of the reactor pressure vessel of Belgian nuclear power plants, were studied in a saturated portlandite aqueous solution, relevant for the Belgian Supercontainer design, as perceived for the geological disposal of high-level nuclear waste. To achieve this, we performed simple immersion and potentiostatic corrosion tests. In addition, the corrosion rate (which determines the 14C release) was estimated by measuring the release of 60Co. Gas chromatography showed that during the static corrosion test, the carbonaceous species methane, carbon dioxide, ethene, and ethane were produced. Under the hypothesis that all the carbon released from the JRQ steel was transformed into carbon-base gaseous compounds, this corresponds to a corrosion rate of approximately 100 nm/yr, which is in good agreement with literature data.

scholarly journals Release and speciation of carbon from Zircaloy-4 in anaerobic and highly alkaline conditions: Comparison of simple immersion and potentiostatic corrosion tests

Radiocarbon ◽  
2018 ◽  
Vol 60 (6) ◽  
pp. 1787-1796
Author(s):  
Sebastien Caes ◽  
Frank Druyts ◽  
Peter Thomas
Keyword(s):  
Corrosion Rate ◽  
Nuclear Power ◽  
Power Plants ◽  
Corrosion Potential ◽  
Fuel Cladding ◽  
Corrosion Tests ◽  
Potentiodynamic Polarization Curves ◽  
Alkaline Conditions ◽  
High Level Nuclear Waste ◽  
High Level

ABSTRACTThe gas release and speciation of carbon species from irradiated and unirradiated Zircaloy-4 samples, representative for the fuel cladding as used in Belgian nuclear power plants, were studied in a saturated Ca(OH)2 solution in anaerobic conditions. This environment is relevant for the Belgian Supercontainer design, as perceived for the geological disposal of high-level nuclear waste. To achieve this, we performed simple immersion and potentiostatic corrosion tests. Potentiodynamic polarization curves, recorded prior to the potentiostatic tests, revealed that irradiation seems to induce changes on the Zircaloy-4 corrosion behavior, such as a shift of the corrosion potential. Potentiostatic corrosion tests on unirradiated Zircaloy-4 provided a corrosion rate of ~54 nm/yr over a 7 day-experiment, whilst a corrosion rate of only ~4 nm/yr was calculated for the irradiated sample. Gas chromatography revealed that during simple immersion tests, which lasted 195 days, hydrogen, methane, ethane, and CO2 were produced, with methane being the major compound. Assuming that all carbon released from the metal was transformed into gaseous compounds, this yields to a corrosion rate ranging from 57 to 84 nm/yr for the irradiated sample. However, caution has to be taken on these corrosion rate and more tests should be performed to confirm these results.

The method for calculating the probability of brittle destruction of NPP equipment in different operating modes with postulated defectiveness

2021 ◽  
Vol 14 (1) ◽  
pp. 34-39
Author(s):  
D. A. Kuzmin ◽  
A. Yu. Kuz’michevskiy
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Reactor Pressure Vessel ◽  
Practical Significance ◽  
Reactor Unit ◽  
Operating Modes ◽  
Brittle Destruction ◽  
Reactor Pressure

The destruction of equipment metal by a brittle fracture mechanism is a probabilistic event at nuclear power plants (NPP). The calculation for resistance to brittle destruction is performed for NPP equipment exposed to neutron irradiation; for example, for a reactor plant such as a water-water energetic reactor (WWER), this is a reactor pressure vessel. The destruction of the reactor pressure vessel leads to a beyond design-basis accident, therefore, the determination of the probability of brittle destruction is an important task. The research method is probabilistic analysis of brittle destruction, which takes into account statistical data on residual defectiveness of equipment, experimental results of equipment fracture toughness and load for the main operating modes of NPP equipment. Residual defectiveness (a set of remaining defects in the equipment material that were not detected by non-destructive testing methods after manufacturing (operation), control and repair of the detected defects) is the most important characteristic of the equipment material that affects its strength and service life. A missed defect of a considerable size admitted into operation can reduce the bearing capacity and reduce the time of safe operation from the nominal design value down to zero; therefore, any forecast of the structure reliability without taking into account residual defectiveness will be incorrect. The application of the developed method is demonstrated on the example of an NPP reactor pressure vessel with a WWER-1000 reactor unit when using the maximum allowable operating loads, in the absence of load dispersion in different operating modes, and taking into account the actual values of the distributions of fracture toughness and residual defectiveness. The practical significance of the developed method lies in the possibility of obtaining values of the actual probability of destruction of NPP equipment in order to determine the reliability of equipment operation, as well as possible reliability margins for their subsequent optimization.

NUCLEAR WASTE MANAGEMENT IN SPAIN: ANALYSIS OF THE CURRENT SITUATION AND ALTERNATIVE STRATEGIES

10.6036/10156 ◽  
2021 ◽  
Vol 96 (4) ◽  
pp. 355-358
Author(s):  
Pablo Fernández Arias ◽  
DIEGO VERGARA RODRIGUEZ
Keyword(s):  
Nuclear Waste ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Spent Fuel ◽  
Nuclear Waste Storage ◽  
Alternative Strategies ◽  
Temporary Storage ◽  
Long Term Storage ◽  
High Level

Centralized Temporary Storage Facility (CTS) is an industrial facility designed to store spent fuel (SF) and high level radioactive waste (HLW) generated at Spanish nuclear power plants (NPP) in a single location. At the end of 2011, the Spanish Government approved the installation of the CTS in the municipality of Villar de Cañas in Cuenca. This approval was the outcome of a long process of technical studies and political decisions that were always surrounded by great social rejection. After years of confrontations between the different political levels, with hardly any progress in its construction, this infrastructure of national importance seems to have been definitively postponed. The present research analyzes the management strategy of SF and HLW in Spain, as well as the alternative strategies proposed, taking into account the current schedule foreseen for the closure of the Spanish NPPs. In view of the results obtained, it is difficult to affirm that the CTS will be available in 2028, with the possibility that its implementation may be delayed to 2032, or even that it may never happen, making it necessary to adopt an alternative strategy for the management of GC and ARAR in Spain. Among the different alternatives, the permanence of the current Individualized Temporary Stores (ITS) as a long-term storage strategy stands out, and even the possibility of building several distributed temporary storage facilities (DTS) in which to store the SF and HLW from several Spanish NPP. Keywords: nuclear waste, storage, nuclear power plants.

Load Factor for Evaluating Non-Planar Flaws in Carbon Steel Piping Using Limit Load Methodology

2016 ◽  
Author(s):  
Phuong H. Hoang
Keyword(s):  
Carbon Steel ◽  
Nuclear Power ◽  
Power Plants ◽  
Pipe Wall ◽  
Local Strain ◽  
Limit Load ◽  
Evaluation Methodology ◽  
Load Factor ◽  
Wall Thinning ◽  
Steel Piping

Non-planar flaw such as local wall thinning flaw is a major piping degradation in nuclear power plants. Hundreds of piping components are inspected and evaluated for pipe wall loss due to flow accelerated corrosion and microbiological corrosion during a typical scheduled refueling outage. The evaluation is typically based on the original code rules for design and construction, and so often that uniformly thin pipe cross section is conservatively assumed. Code Case N-597-2 of ASME B&PV, Section XI Code provides a simplified methodology for local pipe wall thinning evaluation to meet the construction Code requirements for pressure and moment loading. However, it is desirable to develop a methodology for evaluating non-planar flaws that consistent with the Section XI flaw evaluation methodology for operating plants. From the results of recent studies and experimental data, it is reasonable to suggest that the Section XI, Appendix C net section collapse load approach can be used for non-planar flaws in carbon steel piping with an appropriate load multiplier factor. Local strain at non-planar flaws in carbon steel piping may reach a strain instability prior to net section collapse. As load increase, necking starting at onset strain instability leads to crack initiation, coalescence and fracture. Thus, by limiting local strain to material onset strain instability, a load multiplier factor can be developed for evaluating non-planar flaws in carbon steel piping using limit load methodology. In this paper, onset strain instability, which is material strain at the ultimate stress from available tensile test data, is correlated with the material minimum specified elongation for developing a load factor of non-planar flaws in various carbon steel piping subjected to multiaxial loading.

scholarly journals Occupational ALARA Planning for Reactor Pressure Vessel Dismantling at Kori Unit 1

2020 ◽  
Vol 17 (15) ◽  
pp. 5346
Author(s):  
Juyoul Kim ◽  
Batbuyan Tseren
Keyword(s):  
Radiation Exposure ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Occupational Safety ◽  
Health Management ◽  
Primary Component ◽  
Reactor Pressure Vessel ◽  
Safety And Health ◽  
Reactor Pressure

Assessing workers’ safety and health during the decommissioning of nuclear power plants (NPPs) is an important procedure in terms of occupational radiation exposure (ORE). Optimizing the radiation exposure through the “As Low As Reasonably Achievable (ALARA)” principle is a very important procedure in the phase of nuclear decommissioning. Using the VISIPLAN 3D ALARA planning tool, this study aimed at assessing the radiological doses to workers during the dismantling of the reactor pressure vessel (RPV) at Kori NPP unit 1. Fragmentation and segmentation cutting processes were applied to cut the primary component. Using a simulation function in VISIPLAN, the external exposure doses were calculated for each work operation. Fragmentation involved 18 operations, whereas segmentation comprised 32 operations for each fragment. Six operations were additionally performed for both hot and cold legs of the RPV. The operations were conducted based on the radioactive waste drum’s dimensions. The results in this study indicated that the collective doses decreased as the components were cut into smaller segments. The fragmentation process showed a relatively higher collective dose compared to the segmentation operation. The active part of the RPV significantly contributed to the exposure dose and thus the shielding of workers and reduced working hours need to be considered. It was found that 60Co contained in the stainless steel of the reactor vessel greatly contributed to the dose as an activation material. The sensitivity analysis, which was conducted for different cutting methods, showed that laser cutting took a much longer time than plasma cutting and contributed higher doses to the workers. This study will be helpful in carrying out the occupational safety and health management of decommissioning workers at Kori NPP unit 1 in the near future.

Upgrading the Operational Safety of Nuclear Power Plants Through the TACIS Nuclear Safety Assistance Programme

2004 ◽  
Author(s):  
M. Bie`th ◽  
R. Ahlstrand ◽  
C. Rieg ◽  
P. Trampus
Keyword(s):  
European Commission ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Safety ◽  
Reactor Pressure Vessel ◽  
Nuclear Safeguards ◽  
Integrity Assessment ◽  
Reactor Pressure

The European Union’ TACIS programme was established for the New Independent States since 1991. One priority for TACIS funding is nuclear safety. The European Commission has made available a total of € 944 million for nuclear safety programmes covering the period 1991–2003. The TACIS nuclear safety programme is devoted to the improvement of the safety of Soviet designed nuclear installations in providing technology and safety culture transfer. The Joint Research Center (JRC) of the European Commission is carrying out works in the following areas: • On-Site Assistance for TACIS Nuclear Power Plants; • Design Safety and Dissemination of TACIS results; • Reactor Pressure Vessel Embrittlement for VVER in Russia and Ukraine; • Regulatory Assistance; • Industrial Waste Management and Nuclear Safeguards. This paper gives an overview of the Scientific and Technical support that JRC is providing for the programming and the implementation of the TACIS nuclear safety programmes. In particular, two new projects are being implemented to get an extensive understanding of the VVER reactor pressure vessel embritttlement and integrity assessment.

The Position Design for M310 Reactor Vessel’s Instrumentation Penetration Based on None-Alignment Analysis

2013 ◽  
Author(s):  
Xin Xia ◽  
Hua Du ◽  
Ning Li ◽  
Bin Xu ◽  
Yan Li ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Economic Losses ◽  
Design Value ◽  
Alignment Analysis ◽  
Reactor Pressure ◽  
Reactor Internals

The position of the reactor pressure vessel’s instrumentation penetration is the most important factor that influences the none-alignment between the instrumentation penetration and the instrumentation tube of reactor internals. The more the none-alignment, the more wear the flux thimble will suffer, which will damage the flux detector and cause economic losses. This paper analyzes the none-alignment between the instrumentation penetration and the instrumentation tube, and combines the feedback of the experience in nuclear power plants having been built or being build, puts forward the reasonable design value for the instrumentation penetration’s position.

Underground Siting of Nuclear Power Plants: Insights From Fukushima

2012 ◽  
Author(s):  
Jay F. Kunze ◽  
James M. Mahar ◽  
Kellen M. Giraud ◽  
C. W. Myers
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Light Water Reactors ◽  
Light Water ◽  
Waste Storage ◽  
Small Modular Reactor ◽  
Fukushima Daiichi ◽  
Water Reactors ◽  
High Level

Siting of nuclear power plants in an underground nuclear park has been proposed by the authors in many previous publications, first focusing on how the present 1200 to 1600 MW-electric light water reactors could be sited underground, then including reprocessing and fuel manufacturing facilities, as well as high level permanent waste storage. Recently the focus has been on siting multiple small modular reactor systems. The recent incident at the Fukushima Daiichi site has prompted the authors to consider what the effects of a natural disaster such as the Japan earthquake and subsequent tsunami would have had if these reactors had been located underground. This paper addresses how the reactors might have remained operable — assuming the designs we previously proposed — and what lessons from the Fukushima incident can be learned for underground nuclear power plant designs.

Centralization of Canada’s Spent Nuclear Fuel

2013 ◽  
Author(s):  
Krista Nicholson ◽  
John McDonald ◽  
Shona Draper ◽  
Brian M. Ikeda ◽  
Igor Pioro
Keyword(s):  
Nuclear Fuel ◽  
Site Selection ◽  
Nuclear Power ◽  
Power Plants ◽  
Spent Nuclear Fuel ◽  
Selection Process ◽  
Spent Fuel ◽  
Geologic Repository ◽  
Candu Reactors ◽  
High Level

Currently in Canada, spent fuel produced from Nuclear Power Plants (NPPs) is in the interim storage all across the country. It is Canada’s long-term strategy to have a national geologic repository for the disposal of spent nuclear fuel for CANada Deuterium Uranium (CANDU) reactors. The initial problem is to identify a means to centralize Canada’s spent nuclear fuel. The objective of this paper is to present a solution for the transportation issues that surround centralizing the waste. This paper reviews three major components of managing and the transporting of high-level nuclear waste: 1) site selection, 2) containment and 3) the proposed transportation method. The site has been selected based upon several factors including proximity to railways and highways. These factors play an important role in the site-selection process since the location must be accessible and ideally to be far from communities. For the containment of the spent fuel during transportation, a copper-shell container with a steel structural infrastructure was selected based on good thermal, structural, and corrosion resistance properties has been designed. Rail has been selected as the method of transporting the container due to both the potential to accommodate several containers at once and the extensive railway system in Canada.

PERFECT—Prediction of Irradiation Damage Effects in Reactor Components: Update of Progress in RPV Mechanics Sub-Project

2007 ◽  
Author(s):  
David Lidbury ◽  
Ste´phane Bugat ◽  
Olivier Diard ◽  
Elisabeth Keim ◽  
Bernard Marini ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Fracture Behavior ◽  
Current Practice ◽  
Irradiation Damage ◽  
Multi Scale ◽  
Scale Modeling ◽  
Integrated Project ◽  
Main Components ◽  
Reactor Pressure

The EURATOM 6th Framework Integrated Project PERFECT (Prediction of Irradiation Damage Effects in Reactor Components) addresses irradiation damage in RPV materials and components by multi-scale modeling. This approach offers many potential advantages over the conventional empirical methods used in current practice of nuclear plant lifetime management. Launched in January 2004, PERFECT is a 48-month project focusing on two main components of nuclear power plants which are subject to irradiation damage: the ferritic steel reactor pressure vessel (RPV), and the austenitic steel internals. It is the purpose of the present paper to provide an update of progress of work being carried out in the Mechanics Sub-project of PERFECT to predict the fracture behavior of RPVs in PWR and WWER systems.

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