Status report of the Finnish spent fuel geologic repository programme and ongoing corrosion studies

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.

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Centralization of Canada’s Spent Nuclear Fuel

Volume 6: Beyond Design Basis Events; Student Paper Competition ◽

10.1115/icone21-16511 ◽

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.

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A Comparison of the Performance of Nuclear Waste Glasses by Modeling

MRS Proceedings ◽

10.1557/proc-112-713 ◽

1987 ◽

Vol 112 ◽

Cited By ~ 8

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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Comparison and Modeling of Aqueous Dissolution Rates of Various Uranium Oxides

MRS Proceedings ◽

10.1557/proc-465-557 ◽

1996 ◽

Vol 465 ◽

Cited By ~ 7

Author(s):

S. A. Steward ◽

E. T. Mones

Keyword(s):

Dissolution Rate ◽

Atmospheric Oxygen ◽

Fold Increase ◽

Stable Phase ◽

Uranium Oxides ◽

Spent Fuel ◽

Geologic Repository ◽

Dissolution Rates ◽

Intrinsic Dissolution ◽

Carbonate Concentration

ABSTRACTThe purpose of this work has been to measure and model the intrinsic dissolution rates of uranium oxides under a variety of well-controlled conditions that are relevant to a geologic repository. When exposed to air at elevated temperature, spent fuel may form the stable phase U3O8. Dehydrated schoepite, UO3H2O, has been shown to exist in drip tests on spent fuel.Equivalent sets of U3O8 and UO3H2 dissolution experiments allowed a systematic examination of the effects of temperature (25–75°C), pH (8–10) and carbonate (2–200×10−4 molar) concentrations at atmospheric oxygen conditions.Results indicate that UO3H2O has a much higher dissolution rate (at least ten-fold) than U3O8 under the same conditions. The intrinsic dissolution rate of unirradiated U3O8 is about twice that of UO2. Dissolution of both U3O8 and UO3.H2O shows a very high sensitivity to carbonate concentration. Present results show a 25 to 50-fold increase in room-temperature UO3H2O dissolution rates between the highest and lowest carbonate concentrations.As with the UO2 dissolution data the classical observed chemical kinetic rate law was used to model the U3O8 dissolution rate data. The pH did not have much effect on the models, in agreement with the earlier analysis of the UO2 and spent fuel dissolution data,. However, carbonate concentration, not temperature, had the strongest effect on the U3O8 dissolution rate. The U3O8 dissolution activation energy was about 6000 cal/mol, compared with 7300 and 8000 cal/mol for spent fuel and UO2 respectively.

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The Effect of Fuel Burnup and Dispersed Water Intrusion on the Criticality of Spent High-Level Nuclear Fuel in a Geologic Repository

MRS Proceedings ◽

10.1557/proc-333-445 ◽

1993 ◽

Vol 333 ◽

Author(s):

William G. Culbreth ◽

Paige R. Zielinski

Keyword(s):

Fuel Burnup ◽

Spent Fuel ◽

Geologic Repository ◽

Chain Reaction ◽

Worst Case ◽

Mixture Ratio ◽

Waste Package ◽

Radiological Data ◽

A Chain ◽

High Level

ABSTRACTStudies of the spent fuel waste package have been conducted through the use of a Monte-Carlo neutron simulation program to determine the ability of the fuel to sustain a chain reaction. These studies have included fuel burnup and the effect of water mists on criticality. Results were compared with previous studies.In many criticality studies of spent fuel waste packages, fresh fuel with an enrichment as high as 4.5% is used as the conservative (worst) case. The actual spent fuel has a certain amount of “burnup” that decreases the concentration of fissile uranium and increases the amount of radionuclides present. The LWR Radiological Data Base from OCRWM has been used to determine the relative radionuclide ratios and KENO 5.a was used to calculate values of the effective multiplication factor, keff.1Spent fuel is not capable of sustaining a chain reaction unless a suitable moderator, such as water, is present. A completely flooded container has been treated as the worst case for criticality. Results of a previous report that demonstrated that keff actually peaked at a water-to-mixture ratio of 13% were analyzed for validity. In the present study, these results did not occur in the SCP waste package container.

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The Importance of Criticality in the Safety Analysis of the Spent-Fuel Waste Container

MRS Proceedings ◽

10.1557/proc-294-73 ◽

1992 ◽

Vol 294 ◽

Author(s):

William G. Culbreth ◽

Paige Zielinski

Keyword(s):

Thermal Effects ◽

Federal Regulations ◽

Spent Fuel ◽

Geologic Repository ◽

Waste Container ◽

Neutron Multiplication ◽

Accident Conditions ◽

High Level ◽

Internal Configuration ◽

Selection Of

ABSTRACTThe storage of high-level spent reactor fuel in a proposed national geologic repository will require the construction of containers to be placed in boreholes drilled into the host rock. Federal regulations require that the fuel be maintained subcritical under normal or accident conditions. This is determined through the calculation of a neutron multiplication factor, keff, that must remain below 0.95. Criticality will play an important role in the container design, the internal configuration of the fuel, and the selection of neutron poisons. An analysis of keff should be a normal step in the conceptualization of new waste container designs. Unlike thermal effects in a proposed repository, criticality will remain a problem long after the 10,000 year lifetime of the facility.

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Leaching of Spent Fuel under Anaerobic and Reducing Conditions

MRS Proceedings ◽

10.1557/proc-757-ii9.8 ◽

2002 ◽

Vol 757 ◽

Cited By ~ 2

Author(s):

Yngve Albinsson ◽

Arvid Ödegaard-Jensen ◽

Virginia M. Oversby ◽

Lars O. Werme

Keyword(s):

Spent Nuclear Fuel ◽

Fission Products ◽

Bentonite Clay ◽

Fluid Phase ◽

Inert Atmosphere ◽

Dissolution Behavior ◽

Spent Fuel ◽

Geologic Repository ◽

Reducing Conditions ◽

Reaction Vessels

ABSTRACTSweden plans to dispose of spent nuclear fuel in a deep geologic repository in granitic rock. The disposal conditions allow water to contact the canisters by diffusion through the surrounding bentonite clay layer. Corrosion of the canister iron insert will consume oxygen and provide actively reducing conditions in the fluid phase. Experiments with spent fuel have been done to determine the dissolution behavior of the fuel matrix and associated fission products and actinides under conditions ranging from inert atmosphere to reducing conditions in solutions. Data for U, Pu, Np, Cs, Sr, Tc, Mo, and Ru have been obtained for dissolution in a dilute NaHCO3 groundwater for 3 conditions: Ar atmosphere, H2 atmosphere, and H2 atmosphere with Fe(II) in solution. Solution concentrations forU, Pu, and Mo are all significantly lower for the conditions that include Fe(II) ions in the solutions together with H2 atmosphere, while concentrations of the other elements seem to be unaffected by the change of atmospheres or presence of Fe(II). Most of the material that initially dissolved from the fuel has reprecipitated back onto the fuel surface. Very little material was recovered from rinsing and acid stripping of the reaction vessels.

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Lead Slowing-Down Spectrometry for Spent Fuel Assay: FY11 Status Report

10.2172/1025687 ◽

2011 ◽

Cited By ~ 1

Author(s):

Glen A. Warren ◽

Andrew M. Casella ◽

R. C. Haight ◽

Kevin K. Anderson ◽

Yaron Danon ◽

...

Keyword(s):

Status Report ◽

Spent Fuel ◽

Slowing Down

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Lead Slowing-Down Spectrometry Time Spectral Analysis for Spent Fuel Assay: FY11 Status Report

10.2172/1027699 ◽

2011 ◽

Author(s):

Jonathan A. Kulisek ◽

Kevin K. Anderson ◽

Sonya M. Bowyer ◽

Andrew M. Casella ◽

Christopher J. Gesh ◽

...

Keyword(s):

Spectral Analysis ◽

Status Report ◽

Spent Fuel ◽

Slowing Down

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Spent Fuel Storage in the Post-Yucca Mountain Paradigm

Volume 7: Operations, Applications and Components ◽

10.1115/pvp2012-78642 ◽

2012 ◽

Author(s):

Earl Easton ◽

Christopher Bajwa ◽

Zhian Li ◽

Matthew Gordon

Keyword(s):

Spent Nuclear Fuel ◽

Data Availability ◽

Spent Fuel ◽

Geologic Repository ◽

Safety Standards ◽

Dose Rates ◽

Fuel Storage ◽

High Burnup ◽

High Level ◽

Time Frames

The current uncertainty surrounding the licensing and eventual opening of a long term geologic repository for the nation’s civilian and defense spent nuclear fuel (SNF) and high level radioactive waste (HLW) has shifted the window for the length of time spent fuel could be stored to periods of time significantly longer than the current licensing period of 40 years for dry storage. An alternative approach may be needed to the licensing of high-burnup fuel for storage and transportation based on the assumption that spent fuel cladding may not always remain intact. The approach would permit spent fuel to be retrieved on a canister basis and could lessen the need for repackaging of spent fuel. This approach is being presented as a possible engineering solution to address the uncertainties and lack of data availability for cladding properties for high burnup fuel and extended storage time frames. The proposed approach does not involve relaxing current safety standards for criticality safety, containment, or permissible external dose rates.

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