Studies on Spent Fuel Alterations During Storage and Radiolysis Effects on Corrosion Behaviour Using Alpha-Doped UO2

2003 ◽  
Author(s):  
V. V. Rondinella ◽  
T. Wiss ◽  
J.-P. Hiernaut ◽  
J. Cobos
Keyword(s):  
Corrosion Behaviour ◽  
Alpha Decay ◽  
Lattice Parameter ◽  
Dislocation Loops ◽  
Spent Fuel ◽  
Geologic Repository ◽  
Experimental Conditions ◽  
Dose Rate Effects ◽  
Radiolytic Process ◽  
Microstructure Examination

UO2 containing different fractions of short-lived alpha-emitters, the so-called alpha-doped UO2 simulates the level of activity of spent fuel after different storage times, and can be used to study the effects of radiolysis on the corrosion behaviour of aged spent fuel exposed to groundwater in a geologic repository. Furthermore, the integral over time of the alpha-decay in alpha-doped UO2 can simulate the decay damage accumulated in spent fuel during storage. This allows investigating property modifications occurring to the fuel during storage periods of interest (e.g. in view of spent fuel retrieval or in view of final disposal) within a laboratory-acceptable timescale. Periodical measurements of lattice parameter are performed on high activity alpha-doped UO2 to investigate the build-up of radiation damage and evaluate possible dose rate effects. Additionally, annealing tests combined with He-release measurements using a Knudsen cell and with microstructure examination using TEM are performed to establish a correlation among the annealing of damage in the microstructure (mainly characterized by dislocation loops) and the release behaviour of He. The effects on the microstructure due to the accumulation of He and α-decay damage are of interest as they may considerably affect the mechanical integrity of the fuel rods, by causing e.g. swelling or cracking in the fuel and/or overpressurization of the cladding. Alpha-doped UO2 with specific activities spanning over three orders of magnitude and undoped UO2 were used in static leaching experiments at room temperature in deionized water under nominally anoxic conditions. Under these experimental conditions (single effect studies) a clear dissolution enhancing effect of alpha-radiolysis was observed coupled with the establishment of higher redox potential due to the radiolytic process. An alpha-activity dependence of the dissolution behaviour was observed.

Dose Rate Effects on the Accumulation of Radiation Damage

2007 ◽  
Author(s):  
V. V. Rondinella ◽  
T. Wiss ◽  
J.-P. Hiernaut ◽  
D. Staicu
Keyword(s):  
Dose Rate ◽  
Radiation Damage ◽  
Alpha Decay ◽  
Lattice Parameter ◽  
Time Interval ◽  
Spent Fuel ◽  
Dose Rates ◽  
Dose Rate Effects ◽  
Doped Materials ◽  
Interim Storage

During storage, spent fuel and other waste forms accumulate alpha-decay damage (and He). The dose rates and the temperatures experienced during storage are lower than during in-pile operation: however, the duration of the storage is much longer (of the order of up to a few hundred years if extended interim storage concepts are considered); if final disposal in the repository is considered, the time interval in which radiation damage accumulates is open-ended. In order to simulate within timeframes suitable for laboratory experiments long-term accumulation of alpha-decay damage, the so-called alpha-doped materials can be used, i.e. materials loaded with short-lived alpha-emitters (like e.g. Pu-238, U-233, etc.). The question is often posed if the accelerated accumulation of decay damage and He obtained using alpha-doped materials does cause some artefact related to the rate of accumulation rather than by the integrated dose. This work presents evidence that, at least within the range of alpha-activities considered, there is no dose rate effect. By comparing property evolution as a function of accumulated dpa for alpha-doped materials with activities of ∼1010 and ∼108 Bq/g, respectively, the same trends and levels of alteration are observed. In particular, macroscopic properties like hardness (measured by Vickers indentation) or swelling (evolution of lattice parameter derived from XRD), and microstructural formation and accumulation of defects in the lattice of the alpha-doped material are investigated, showing a remarkable similarity of behaviour vs. dpa independently not only from the dose rate, but also from the composition (namely, Pu and U are considered).

Leaching Behavior and α-Decay Damage Accumulation of UO2 Containing Short-Lived Actinides

2000 ◽  
Vol 663 ◽  
Author(s):  
V.V. Rondinella ◽  
J. Cobos ◽  
Hj. Matzke ◽  
T. Wiss ◽  
P. Carbol ◽  
...  
Keyword(s):  
Redox Potential ◽  
Sample Surface ◽  
Spent Fuel ◽  
Geologic Repository ◽  
Experimental Conditions ◽  
Α Decay ◽  
Level Of Activity ◽  
Congruent Dissolution ◽  
High Concentrations ◽  
Α Activity

ABSTRACTUO2 containing short-lived α-emitters, the so-called α-doped UO2, can simulate type (i.e. α- decay) and level of activity of spent fuel at the time when it might become exposed to groundwater in a geologic repository during storage. This allows studying α-radiolysis effects on the dissolution of the fuel matrix. Additionally, UO2 with high concentrations of α-emitters accumulate, during experimentally acceptable short times, the amount of decays, hence of property modifications, corresponding to long storage times for spent fuel. UO2 containing ∼10 and ∼0.1 wt% 238Pu was fabricated and tested. Leaching experiments in deionized water under unaerated conditions, with continuous monitoring of the evolution of the redox potential and pH in the leaching solutions, were performed. The Eh measurements showed a fast increase of the redox potential in the case of the material with the highest α-activity, while the UO2 containing ∼0.1 wt% 238Pu increased its potential more slowly. The redox potential for undoped UO2 decreased steadily during the experiment. As previously observed, higher fractions of U were released in the case of α-doped UO2 compared to undoped UO2. The fractions of U and Pu released during leaching from the α-doped materials were very similar, suggesting that congruent dissolution occurred. After leaching times longer than 10 h, only dissolved species were present in the solutions. Under these experimental conditions, characterized by relatively low values of the ratio sample surface/leachant volume, a dependence of the released amounts on the α-activity of the samples was observed. Periodical measurements of parameters like hardness, showed a rapid buildup of radiation damage in the material with the high α-activity. After more than two years, noticeable changes, namely an increase of the hardness, have begun to be observed also for the material with the low concentration of 238Pu.

Testing of Uranium Dioxide Containing Different Levels of Alpha Activity Under Anaerobic and Reducing Conditions

2008 ◽  
Vol 1104 ◽  
Author(s):  
Arvid Ödegaard-Jensen ◽  
Virginia Oversby
Keyword(s):  
Uranium Dioxide ◽  
Alpha Decay ◽  
Sample Surface ◽  
Hydrogen Atmosphere ◽  
Nitrogen Atmosphere ◽  
Alpha Activity ◽  
Dissolution Behavior ◽  
Spent Fuel ◽  
Geologic Repository ◽  
Radiolysis Products

AbstractSweden plans to dispose of spent nuclear reactor fuel in a deep geologic repository in granitic rock. The conditions in the repository in the long term will be reducing and water is not expected to contact the fuel until after more than 1000 years. At that time, most of the beta- and gamma-active nuclides will have decayed away and the radiation will be dominated by alpha decay. In order to simulate the radiolysis field for dissolution of spent fuel with age more than 1000 years we have used uranium dioxide containing 5% U-235 and 0, 5, or 10% U-233. The 10% U-233 gives an alpha activity appropriate to about 3000 years after disposal. Samples were testied in a synthetic groundwater with low ionic strength and with the chemical composition dominated by sodium bicarbonate and calcium chloride. Tests were run in triplicate using an atmosphere of nitrogen (1atm), hydrogen (10 bar), hydrogen (10 bar) plus an iron strip in the solution, nitrogen (1 atm) plus an iron strip in the solution, hydrogen (10 bar) plus an iron strip in the solution, hydrogen (10 bar) without the iron strip. Each of these test conditions was run for 2 consecutive periods of at least 21 days. The results showed that the dissolution behavior of the samples was the same for both nitrogen atmosphere and hydrogen atmosphere. The amount of U dissolved under these conditions clearly showed the enhancement of dissolution due to oxidation of the sample surface by radiolysis products. When an iron strip was added to the solution, the amount of dissolution decreased dramatically indicating that the Fe(II) ions released from the corroding iron were able to react with most of the radiolysis products before they could oxidize the uranium dioxide surface.

Alpha-Decay Effects in 241Am-Doped Gadolinium Zirconate

2012 ◽  
Vol 1444 ◽  
Author(s):  
S.V. Stefanovsky ◽  
A.G. Ptashkin ◽  
S.V. Yudintsev ◽  
B.F. Myasoedov
Keyword(s):  
Heat Treatment ◽  
Relative Intensity ◽  
Alpha Decay ◽  
Fluorite Structure ◽  
Lattice Parameter ◽  
Oxide Mixture ◽  
Xrd Pattern ◽  
Gadolinium Zirconate ◽  
Main Reflection ◽  
Higher Doses

ABSTRACTSample of zirconate ceramic with a composition corresponding to formula Gd1.7241Am0.3Zr2O7 was synthesized by heat-treatment of mechanically activated and compacted in pellet oxide mixture at 1500 °C for 30 min. The d values on XRD pattern of the sample soon after synthesis (D = 7.9×1015 α-decays/g or 0.001 dpa) demonstrated fluorite structure with the most intensive peak with d111 =3.042 Å (a = 5.269 Å) and very weak diffuse reflections due to d-pyrochlore. At a dose of 7.9×1017 α-decays/g or 0.11 dpa the reflections were broadened by approximately 20% and their relative intensity slightly reduced. At higher doses all the weak superstructure reflections disappeared and the growth in intensity and narrowing of the main reflection occurred. Lattice parameter a increased with the dose and reached 5.343 Å (d111 = 3.085 Å) at a dose of 4.6×1018 α-decays/g or 0.42 dpa. At a dose of 5.5×1018 α-decays/g or 0.78 dpa positions of reflections were shifted to lower d-spaces (d111 value reduced to 3.071 Å) and the half-width of the major reflection was 67% of initial. For the 241Am-doped Gd-zirconate the structure recovery rate exceeds disordering rate and no amorphization occurred at doses higher than ∼0.2-0.3 dpa.

The effects of morphology of ferrite and non-metallic inclusions on corrosion behaviour of as-cast 304 stainless steel

CORROSION ◽  
10.5006/3763 ◽  
2021 ◽  
Author(s):  
Danbin Jia ◽  
Liangcai Zhong ◽  
Jingkun Yu ◽  
Zhaoyang Liu ◽  
Yuting Zhou ◽  
...  
Keyword(s):  
Stainless Steel ◽  
Corrosion Resistance ◽  
Corrosion Behaviour ◽  
Solidification Process ◽  
Inclusion Size ◽  
304 Stainless Steel ◽  
Experimental Conditions ◽  
Quenching Temperature ◽  
Δ Ferrite ◽  
Metallic Inclusions

The effects of morphology of ferrite and non-metallic inclusions on corrosion resistance of as-cast 304 stainless steel (304 SS) were investigated. With the decrease in quenching temperature from 1723 K to 1648 K, the different microstructures of the as-cast 304 SS were obtained as the following series: austenitic-lathy δ ferrite, austenitic-colony δ ferrite and austenitic-blocky δ ferrite, and the average inclusion size increased. The electrochemical results show that the sample with the microstructure of austenitic- lathy δ ferrite and smaller size inclusions had a higher corrosion tendency and the lower pitting resistance. Furthermore, the effect of morphology and content of ferrite on corrosion resistance was greater than that of inclusion size under the current experimental conditions. Therefore, a promising method was developed to improve the corrosion resistance of as-cast 304 SS by changing the solidification process.

scholarly journals Dislocation Loop Generation Differences between Thin Films and Bulk in EFDA Pure Iron under Self-Ion Irradiation at 20 MeV

Metals ◽  
2021 ◽  
Vol 11 (12) ◽  
pp. 2000
Author(s):  
Marcelo Roldán ◽  
Fernando José Sánchez ◽  
Pilar Fernández ◽  
Christophe J. Ortiz ◽  
Adrián Gómez-Herrero ◽  
...  
Keyword(s):  
Thin Films ◽  
Computational Models ◽  
Defect Formation ◽  
Ion Irradiation ◽  
Damage Zone ◽  
High Energy ◽  
Model Material ◽  
Dislocation Nucleation ◽  
Dislocation Loops ◽  
Experimental Conditions

In the present investigation, high-energy self-ion irradiation experiments (20 MeV Fe+4) were performed on two types of pure Fe samples to evaluate the formation of dislocation loops as a function of material volume. The choice of model material, namely EFDA pure Fe, was made to emulate experiments simulated with computational models that study defect evolution. The experimental conditions were an ion fluence of 4.25 and 8.5 × 1015 ions/cm2 and an irradiation temperature of 350 and 450 °C, respectively. First, the ions pass through the samples, which are thin films of less than 100 nm. With this procedure, the formation of the accumulated damage zone, which is the peak where the ions stop, and the injection of interstitials are prevented. As a result, the effect of two free surfaces on defect formation can be studied. In the second type of experiments, the same irradiations were performed on bulk samples to compare the creation of defects in the first 100 nm depth with the microstructure found in the whole thickness of the thin films. Apparent differences were found between the thin foil irradiation and the first 100 nm in bulk specimens in terms of dislocation loops, even with a similar primary knock-on atom (PKA) spectrum. In thin films, the most loops identified in all four experimental conditions were b ±a0<100>{200} type with sizes of hundreds of nm depending on the experimental conditions, similarly to bulk samples where practically no defects were detected. These important results would help validate computational simulations about the evolution of defects in alpha iron thin films irradiated with energetic ions at large doses, which would predict the dislocation nucleation and growth.

Phase Relations of the Uranyl Oxide Hydrates and their Relevance to the Disposal of Spent Fuel

1990 ◽  
Vol 212 ◽  
Author(s):  
R. J. Finch ◽  
R. C. Ewing
Keyword(s):  
General Formula ◽  
Phase Relations ◽  
Molecular Water ◽  
Spent Fuel ◽  
Geologic Repository ◽  
Structural Sheet ◽  
Natural Analogues ◽  
Ph Conditions ◽  
Image Position

ABSTRACTUranyl oxide hydrates, formed by the alteration of uraninite, are natural analogues for the long-term corrosion products of spent fuel in a geologic repository under oxidizing conditions. The uranyl oxide hydrates may be represented by the general formula:Pb-bearing hydrates require the addition of a neutral uranyl group into the structural sheet (UO2(OH)2) for each interlayer Pb ion. Distortion of the structure associated with the additional uranyl groups is reduced by replacing two structural hydroxyls with a structural oxygen and a molecular water. The general formula for the Pb-uranyl oxide hydrates is:This hypothesis explains the paragenetic sequences:1) schoepite ➛ billietite ➛ protasite ➛ bauranoite2) schoepite ➛ vandendriesscheite ➛ fourmarierite ➛ masuyite ➛ wölsendorfite3) schoepite ➛ vandendriesscheite ➛ fourmarierite ➛ ± masuyite ➛ sayrite ➛ curite, and indicates that, under relatively high pH conditions, schoepite will not be the long-term solubility-controlling phase for uranium in uranium-rich groundwaters.

High Efficiency Gamma-Beta Blind Alpha Spectrometry for Nuclear Energy Applications

2014 ◽  
Author(s):  
Jeffrey A. Webster ◽  
Alexander Hagen ◽  
Brian C. Archambault ◽  
Nicholas Hume ◽  
Rusi Taleyarkhan
Keyword(s):  
Nuclear Energy ◽  
Nuclear Reactor ◽  
High Efficiency ◽  
Detection Efficiency ◽  
Power Level ◽  
Bubble Chamber ◽  
Alpha Decay ◽  
Sensor Technology ◽  
Spent Fuel ◽  
Selective Sensitivity

A novel, Centrifugally Tensioned Metastable Fluid Detector (CTMFD) sensor technology has been developed over the last decade to demonstrate high selective sensitivity and detection efficiency to various forms of radiation for wide-ranging conditions (e.g., power level, safeguards, security, and health physics) relevant to the nuclear energy industry. The CTMFD operates by tensioning a liquid with centrifugal force to weaken the bonds in the liquid to the point whereby even a femto-scale nuclear particle interactions can break the fluid and cause a detectable vaporization cascade. The operating principle has only peripheral similarity to the superheated bubble chamber based superheated droplet detectors (SDDs); instead, CTMFDs utilize mechanical “tension pressure” instead of thermal superheat offering a lot of practical advantages. CTMFDs have been used to detect a variety of alpha and neutron emitting sources in near real-time. The CTMFD is selectively blind to gamma photons and betas allowing for detection of alphas and neutrons in extreme gamma/beta background environments such as spent fuel reprocessing plants or under full power conditions within an operating nuclear reactor itself. The selective sensitivity allows for differentiation between alpha emitters including the isotopes of Plutonium. Mixtures of Plutonium isotopes have been measured in ratios of 1:1, 2:1, and 3:1 Pu-238:Pu-239 with successful differentiation. Due to the lack of gamma-beta background interference, the CTMFD’s LLD can be effectively reduced to zero and hence, is inherently more sensitive than scintillation based alpha spectrometers or SDDs and has been proven capable to detect below femtogram quantities of Plutonium-238. Plutonium is also easily distinguishable from Neptunium making it easy to measure the Plutonium concentration in the NPEX stream of a UREX reprocessing facility. The CTMFD has been calibrated for alphas from Americium (5.5 MeV) and Curium (∼6 MeV) as well. The CTMFD has furthermore, recently also been used to detect spontaneous and induced fission events which can be differentiated from alpha decay allowing for detection of fissionable material in a mixture of isotopes. This paper discusses these transformational developments which are also being entered for real-world commercial use.

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.

A Comparison of the Performance of Nuclear Waste Glasses by Modeling

1987 ◽  
Vol 112 ◽  
Author(s):  
B. Grambow ◽  
D. M. Strachan
Keyword(s):  
Nuclear Waste ◽  
Nuclear Reactors ◽  
Liquid Waste ◽  
Spent Fuel ◽  
High Level Liquid Waste ◽  
Geologic Repository ◽  
Waste Forms ◽  
The Earth ◽  
Soluble Solid ◽  
High Level

The reprocessing of spent fuel from nuclear reactors and processing of fuels for defense purposes have generated large volumes of high-level liquid waste that need to be immobilized prior to final storage. For immobilization, the wastes must be converted to a less soluble solid, and, although other waste forms exist, glass currently appears to be the choice for the transuranic-containing portion of the reprocessed waste. Once produced, this glass will be sent in canisters to a geologic repository located some 200 to 500 m below the surface of the earth.

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