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.

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.

Mcc Corrosion Tests at Reference Testing Conditions for A27 Cast steel in Hanford Ground Water

1986 ◽  
Vol 84 ◽  
Author(s):  
M.D. Merz ◽  
F. Gerber ◽  
R. Wang
Keyword(s):  
Pacific Northwest ◽  
Pressure Vessel ◽  
Static Pressure ◽  
Cast Steel ◽  
Statistical Treatment ◽  
Testing Conditions ◽  
Corrosion Tests ◽  
Corrosion Rates ◽  
High Level Nuclear Waste ◽  
High Level

AbstractThe Materials Characterization Center (MCC) at Pacific Northwest Lab- oratory is performing three kinds of corrosion tests for the Basalt Waste Isolation Project (BWIP) to establish the interlaboratory reproducibility and uncertainty of corrosion rates of container materials for high-level nuclear waste. The three types of corrosion tests were selected to address two distinct conditions that are expected in a repository constructed in basalt. An air/steam test is designed to address corrosion during the operational period and static pressure vessel and flowby tests are designed to address corrosion under conditions that bound the condi ring the post-closure period of the repository.The results of tests at reference testing conditions, which were defined to facilitate interlaboratory comparison of data, are presented. Data are reported for the BWIP/MCC-105.5 Air/Steam Test, BWIP/MCC-105.1 Static Pressure Vessel, and BWIP/MC-105.4 Flowby Test. In those cases where data are available from a second laboratory, a statistical analysis of interlaboratory results is reported and expected confidence intervals for mean corrosion rates are given. Other statistical treatment of data include analyses of the effects of vessel-to-vessel variations, test capsule variations for the flowby test, and oven-to-oven variations for air/steam tests.

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.

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.

Incorporation of Risk Insights in the Regulatory Treatment of Nuclear Power Plant Structures, Systems, and Components

2002 ◽  
Author(s):  
Eugene Imbro ◽  
Thomas G. Scarbrough
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Special Treatment ◽  
Regulatory Commission ◽  
High Level

The U.S. Nuclear Regulatory Commission (NRC) has established an initiative to risk-inform the requirements in Title 10 of the Code of Federal Regulations (10 CFR) for the regulatory treatment of structures, systems, and components (SSCs) used in commercial nuclear power plants. As discussed in several Commission papers (e.g., SECY-99-256 and SECY-00-0194), Option 2 of this initiative involves categorizing plant SSCs based on their safety significance, and specifying treatment that would provide an appropriate level of confidence in the capability of those SSCs to perform their design functions in accordance with their risk categorization. The NRC has initiated a rulemaking effort to allow licensees of nuclear power plants in the United States to implement the Option 2 approach in lieu of the “special treatment requirements” of the NRC regulations. In a proof-of-concept effort, the NRC recently granted exemptions from the special treatment requirements for safety-related SSCs categorized as having low risk significance by the licensee of the South Texas Project (STP) Units 1 and 2 nuclear power plant, based on a review of the licensee’s high-level objectives of the planned treatment for safety-related and high-risk nonsafety-related SSCs. This paper discusses the NRC staff’s views regarding the treatment of SSCs at STP described by the licensee in its updated Final Safety Analysis Report (FSAR) in support of the exemption request, and provides the status of rulemaking that would incorporate risk insights into the treatment of SSCs at nuclear power plants.

Instrument Health Monitoring for Extending Calibration Intervals of Safety Related Instruments

2016 ◽  
Author(s):  
R. Steven Black ◽  
Aaron J. Hussey ◽  
Randall L. Bickford
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Critical Path ◽  
Maintenance Cost ◽  
Calibration Error ◽  
Or Efficiency ◽  
Active Monitoring ◽  
Plant Safety ◽  
Calibration Program ◽  
High Level

The ability to extend calibration intervals for nuclear plant instrumentation has multiple benefits for improving productivity and reducing operating costs at nuclear plants. Benefits include fewer calibrations inside containment during an outage and associated reduced critical path time and ALARA exposure, reduced risk of calibration error or instrument damage during removal and replacement, and reduced operations and maintenance cost for instrument removal, calibration and replacement. A good instrument calibration program ensures instruments are checked frequently enough to provide a high level of confidence that they are performing within acceptable limits, but no more frequently. Over-testing of plant instruments and equipment should be avoided for two reasons: valuable resources are expended on maintenance that might not measurably improve plant safety, reliability, or efficiency; and the potential exists for adjustment errors or equipment damage each time an instrument is removed from service for testing. Over-testing increases the risk of errors or damage being introduced without a justifiable improvement in reliability. This paper discusses the regulatory framework for extending calibration intervals of safety related instruments for U.S. based nuclear power plants. Necessary changes to licensing, plant processes and procedures, training, and configuration management are summarized. An example application of pattern recognition modeling is provided to highlight the analytical support for the processes provided by active monitoring to confirm on-going instrument heath. The paper concludes with a listing of recommended steps to implement a practical program for extending calibration intervals of safety related instruments within the U.S. nuclear regulatory environment.

Development Trends in Nuclear Technology and Related Safety Aspects

2002 ◽  
Author(s):  
B. Kuczera ◽  
P. E. Juhn ◽  
K. Fukuda
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Fuel Cycle ◽  
Nuclear Reactors ◽  
Nuclear Technology ◽  
User Requirements ◽  
Safety Standards ◽  
Safety Requirements ◽  
Fuel Cycles ◽  
High Level

The IAEA Safety Standards Series include, in a hierarchical manner, the categories of Safety Fundamentals, Safety Requirements and Safety Guides, which define the elements necessary to ensure the safety of nuclear installations. In the same way as nuclear technology and scientific knowledge advance continuously, also safety requirements may change with these advances. Therefore, in the framework of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) one important aspect among others refers to user requirements on the safety of innovative nuclear installations, which may come into operation within the next fifty years. In this respect, the major objectives of the INPRO subtask “User Requirements and Nuclear Energy Development Criteria in the Area of Safety” have been: a. to overview existing national and international requirements in the safety area, b. to define high level user requirements in the area of safety of innovative nuclear technologies, c. to compile and to analyze existing innovative reactor and fuel cycle technology enhancement concepts and approaches intended to achieve a high degree of safety, and d. to identify the general areas of safety R&D needs for the establishment of these technologies. During the discussions it became evident that the application of the defence in depth strategy will continue to be the overriding approach for achieving the general safety objective in nuclear power plants and fuel cycle facilities, where the emphasis will be shifted from mitigation of accident consequences more towards prevention of accidents. In this context, four high level user requirements have been formulated for the safety of innovative nuclear reactors and fuel cycles. On this basis safety strategies for innovative reactor designs are highlighted in each of the five levels of defence in depth and specific requirements are discussed for the individual components of the fuel cycle.

Corrosion of Ticode-12 in A Simulated Waste Isolation Pilot Project (Wipp) Brine

1982 ◽  
Vol 15 ◽  
Author(s):  
T. M. Ahn ◽  
B. S. Lee ◽  
J. Woodward ◽  
R. L. Sabatini ◽  
P. Soo
Keyword(s):  
Corrosion Rate ◽  
Pilot Project ◽  
Intensity Level ◽  
Uniform Corrosion ◽  
Static Loads ◽  
Waste Container ◽  
Simulated Waste ◽  
High Level Nuclear Waste ◽  
Acidic Conditions ◽  
High Level

ABSTRACTThe corrosion behavior of TiCode-12 (Ti-0.3 Mo-0.8 Ni) high level nuclear waste container alloy has been studied for a simulated WIPP brine at a temperature of 150°C or below. Crevice corrosion was identified as a potentially important failure mode for this material. Within a mechanical crevice, a thick oxide film was found and shown to be the rutile form of TiO2, with a trace of lower oxide also present. Acidic conditions were found to cause a breakdown of the passive oxide layer. Solution aeration and increased acidity accelerate the corrosion rate. In hydrogen embrittlement studies, it was found that hydrogen causes a significant decrease in the apparent stress intensity level in fracture mechanics samples. Hydride formation is thought to be responsible for crack initiation. Stress corrosion cracking under static loads was not observed. Attention has also been given to methods for extrapolating short term uniform corrosion rate data to extended times.

Life Time Management: A Practical Approach of Nuclear Power Plants

2002 ◽  
Author(s):  
Jaroslav Bartonicek ◽  
Klaus-Juergen Metzner ◽  
Friedrich Schoeckle
Keyword(s):  
Time Management ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Mechanical Systems ◽  
Life Time ◽  
Practical Approach ◽  
Degradation Mechanisms ◽  
High Level ◽  
Quality Status

A comprehensive life time management has to take care of all safety and availability relevant components in nuclear power plants, with different intensity, of course. For instance, mechanical systems and components can be basically classified/ranked into three different groups: (1): The quality status of the components in this group has to be guaranteed on a pre-defined (high) level. (2): The quality status of the components in this group has to be maintained on its actual level. (3): Other components with no specific quality demands. Regarding the first group, integrity has to be guaranteed. Therefore it is necessary to monitor the possible root causes of degradation mechanisms during plant operation; thus the degradation effects can be assessed and — more important — controlled to maintain the safety standard on the demanded high level without any compromise. The monitoring of consequences of degradation mechanisms is being performed as an additional redundant measure. The requirements to maintain the quality status of the second group of components can be fulfilled by monitoring of the consequences of operational degradation mechanisms to be performed by preventive maintenance activities, in terms of tests, inspections and repairs, using either time dependant procedures or component condition orientated methods. For the third group of components, no preventive action is necessary. However, failures and malfunctions have to be assessed statistically to avoid a reduction of the required basic component quality. In the first two groups all safety relevant components and systems are included. Generally, aging management programs cover these two groups of components; life time management covers all of above groups. This paper concentrates on mechanical systems and components; it summarizes the practical approach to life time management as it is realized in German nuclear power plants. The application is discussed using dedicated examples.

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