Corrosion Behavior of Pre-oxidized High Burnup Spent Fuel in Salt Brine

2003 ◽  
Vol 807 ◽  
Author(s):  
Andreas Loida ◽  
Bernhard Kienzler ◽  
Horst Geckeis

ABSTRACTDuring long-term interim storage of spent fuel, pre-oxidation of the UO2-matrix may not be ruled out completely. This can happen if air could find access to the fuel in the case of cladding failure. The aim of this work is to study the impact of pre-oxidation of the fuel surface on the UO2 matrix dissolution rate and the associated mobilization or retention of radionuclides in highly concentrated salt solutions. The tests were performed with samples that suffered pre-oxidation during up to seven years. The dissolution rate of a fuel sample contacted by small quantities of air-oxygen was found to be roughly a factor of 10 higher in comparison to non oxidized samples, but concentrations of radionuclides, especially Pu and U were hardly affected. The majority of dissolved radionuclides, especially Pu, U appear to have been reimmobilized on the fuel sample itself.

2006 ◽  
Vol 985 ◽  
Author(s):  
Andreas Loida ◽  
Volker Metz ◽  
Bernhard Kienzler

AbstractRecent studies have shown that in the presence of H2 overpressure, which forms due to the corrosion of the Fe based container, the dissolution rate of the spent fuel matrix is slowed down by a factor of about 10, associated with a distinct decrease of concentrations of important ra-dionuclides. However, in a natural salt environment as well as in geological formations with chloride rich groundwater the presence of radiation chemically active impurities such as bro-mide must be taken in consideration. Bromide is known to react with β/γ radiolysis products, thus counteracting the protective H2 effect. In the present experiments using high burnup spent fuel it is observed that during 212 days the matrix dissolution rate was enhanced by a factor of about10 in the presence of up to 10-3 M bromide and 3.2 bar H2 overpressure. However, concen-trations of matrix bound actinides were found at the same level or below as found under identical conditions, but in the absence of bromide. In the long-term it is expected that the effect of bro-mide becomes less important, because the decrease of β/γ-activity results in a decrease of oxida-tive radicals, which react with bromide, while α activity will dominate the radiation field.


2000 ◽  
Vol 663 ◽  
Author(s):  
A. Loida ◽  
B. Grambow ◽  
H. Geckeis

ABSTRACTWith respect to the assessment of the long-term behavior of the waste form spent fuel it is of high importance to study in particular the UO2 matrix dissolution behavior and the associated release/retention of radionuclides in contact with groundwater. During long term fuel storage, fuel oxidation may not be avoided. Main issue of this work is to identify the impact on the corrosion of partly oxidized fuel of environmental conditions such as (1) the nature of solution contacting the matrix, the (2) presence/absence of CO2, (3) fixed pH values within a range between pH 7- pH 11, and (4) the presence/absence of corroding container material (Fe-powder). Dissolution tests with powdered oxidized spent fuel in various granite waters, and NaCl-brine resulted in matrix dissolution rates in the same order of magnitude for all investigated media (ca.5×10−4/d). The presence of CO2 and fixed pH values (pH 5 – 11) was without a distinct effect. The independence of the dissolution rate of the oxidized fuel matrix upon the nature of solution, pCO2, fixed pH values (5-11) can probably be explained by a masking effect of radiolysis. In presence of Fe powder the matrix dissolution rate was found to be slowed down by a factor of ca. 20, associated with strong retention effects of radionuclides.


2009 ◽  
Vol 1193 ◽  
Author(s):  
D. Roudil ◽  
C. Jegou ◽  
V. Broudic ◽  
M. Tribet

AbstractTo assess the long-term behavior of spent fuel in a nuclear waste repository, the chemical reactions between the fuel and possible intruding water must be understood and the resulting radionuclide release must be quantified. The instant release fraction (IRF) source term assumed to be instantaneously accessible to water after the failure of the waste container. Some IRF values for different kinds of spent fuel are available in the literature. However, the possible contribution the rim restructured zone for high-burnup UOX fuels, was not necessary taken into account. A specific study of the leaching behavior of the rim zone has been carried out on a UOX fuel sample with a burnup of 60 GWd·t-1 and 2.8% FGR. The 134/137Cs and 90Sr rim IRF are effectively higher than the gap and grain boundary inventories (respectively 4.4 wt% and 0.3 wt% of the total Cs and Sr inventories). Nevertheless because of the relative small volume of this zone in the pellet, the impact of the rim inventory appears to be limited and the complete Cs and Sr IRF (gap, grain boundaries and rim), were estimated at 1.85 and 0.3 wt%, respectively


2002 ◽  
Vol 757 ◽  
Author(s):  
Andreas Loida ◽  
Bernhard Kienzler ◽  
Horst Geckeis

ABSTRACTWith respect to the assessment of the long-term behavior of the waste form spent fuel it is of high importance to study the fuel alteration in contact with groundwater and near field materials. The aim of this work is to evaluate the impact of candidate backfill materials hydroxylapatite and magnetite on the overall corrosion behavior of this waste form in salt brine; both materials are used in corrosion tests together with spent fuel. The instant releases and the matrix dissolution rates appear to be similar in presence and in absence of any backfill material under study. However, Am,Np,Pu,U and Sr are retained at different ratios on the hydroxylapatite, on the magnetite and on the fuel sample, indicating possibly the formation of different radionuclide containing new solid phases.


2005 ◽  
Vol 346 (1) ◽  
pp. 24-31 ◽  
Author(s):  
Andreas Loida ◽  
Volker Metz ◽  
Bernhard Kienzler ◽  
Horst Geckeis

2021 ◽  
pp. 0958305X2110513
Author(s):  
Adam J. Mallette ◽  
Aparajita Datta ◽  
Ramanan Krishnamoorti

Over the last 50 years, nuclear energy has reduced US energy-related CO2 emissions by over 30 gigatons compared to if the same electricity were produced by fossil fuels such as coal and natural gas. However, many kilotons of spent nuclear fuel have accumulated at different sites across the country, and sociopolitical factors have frustrated efforts to address the challenge of nuclear waste disposal. Presently, a consolidated interim storage facility in Andrews, Texas, provides a promising temporary solution. In this paper, we compare the technical and policy risks of the project to continued storage at independent spent fuel storage installations. Our results indicate that the cost of the radiological risk is low (<$30,000) for both scenarios. However, policy and societal considerations will impact the viability of the proposed consolidated interim storage facility. The safety and suitability of this interim storage facility will be affected by when a permanent repository becomes available, whether insurance for offsite waste storage is available, and the impact of climate risks. Although a consolidated interim storage facility at Andrews can potentially serve as a safe and economically advantageous solution, we highlight why these concerns must be addressed for the successful implementation of this facility, and more broadly for the future of the US nuclear industry.


Author(s):  
Lorimer Fellingham ◽  
Philippe Michou ◽  
Bruno Alquier

The Murmansk and Arkhangelsk regions of north-west Russia produce large amounts of spent nuclear fuel. These arise from the Kola Nuclear Power Plant, nuclear propulsion units in icebreaker and container ships of the Murmansk Shipping Company, but mostly from the submarines of the Russian Navy’s Northern Fleet. Many marine vessels have been withdrawn from service, but retain their final fuel charges. There are more than 300 reactors and 57500 spent fuel assemblies. Some fuel has been sent to the RT-1 reprocessing plant at the Mayak. However, most marine fuel remains stored in old temporary and effectively full surface or floating facilities around the Kola Peninsula. Damaged, experimental and liquid metal reactor fuel cannot be reprocessed. This creates special problems for handling, transportation and final disposal. It is against this background that the study reported was undertaken. It was part of the European Commission’s TACIS programme and was aimed at improving the safety of radioactive waste management in north-west Russia. Its prime objectives were to identify the factors restricting spent fuel transport from the region to Mayak and potentially suitable storage and reprocessing facilities. Options were to be developed and costed for restoring effective transportation. Their implementation was to lead to safe interim storage of a limited amount of spent fuel in the north-west of Russia. The north-west region is a large, remote area with a harsh terrain and climate. Fuel is stored in two main areas: the Kola Peninsula and the Arkhangelsk region, which are approximately 1,000 km apart. There is a limited transport infrastructure, but the marine facilities have sea access. Hence fuel movement is predominantly by sea to railheads and then rail to Mayak. Road transport is limited, but important for local linking. Routine transportation of spent fuel to Mayak has been restricted by several factors. These include technical, financial and organisational issues. However, the main ones are a lack of available transportation means in both capability and capacity, problems created by the poor state of some fuel, the inadequate safety of the current storage facilities and inadequate interim and buffer storage capacity. Three main types of shipment option were identified: A) regular shipments with storage at existing sites; B) shipments immediately upon arising; and C) regular shipments to Mayak with optimised construction of additional storage capacity in line with demand. Each option was judged on whether it could provide real improvements in radiological and nuclear safety and/or aid the rapid and cost effective defuelling of inadequate existing storage facilities. An optimisation study was performed considering different defuelling, shipping and rail movement rates, and interim and buffer storage capacities, utilisation and locations. The conclusion was that two options could provide similar good solutions. These were: i) Option A.2/C.2 — regular shipment with interim storage of spent fuel at three key node locations: Kola, Murmansk and Severodvinsk; ii) Option B — immediate shipment upon arising. The final choice depends on the capacity of the Mayak plant to receive and reprocess the fuel and the public acceptability of constructing large new, spent fuel stores in north-west Russia. Given the major uncertainties over Mayak’s capacity to store and reprocess submarine fuel, options A.2/C.2 may be optimum.


Author(s):  
Ulrich Knopp

Abstract The CASTOR® BR3 cask has been designed and manufactured to accomodate irradiated fuel (U and MOX) from the BR3 test reactor at the nuclear research centre SCK/CEN in Dessel near Mol, Belgium, which is currently being dismantled. The CASTOR® BR3 is designed as a Type B(U)F package for transport and will be licensed in Belgium. In addition, the CASTOR® BR3 needs a license as a storage cask to be operated in an interim cask storage facility. To obtain these licenses, the cask design has to observe the international regulations for the safe transport of radioactive material as well as the special requirements for the cask storage. The CASTOR® BR3 is a member of the CASTOR® family of spent fuel casks, delivered by the German company GNB. In this way, the cask has such typical features as the following: • monolithic cask body made of ductile cast iron; • double-lid system consisting of primary and secondary lid for long-term interim storage of the fuel. This family of casks has been used for over 20 years for transport and storage of spent fuel. In this paper, the IAEA regulatory requirements for transport casks are summarized and it is shown by selected examples how these requirements have been converted into the cask design and the analyses performed for the cask. Finally, the cask features for an interim storage period of up to 50 years will be spotlighted. Main topics are the evaluation of the long term behaviour of selected cask components and the cask monitoring system for the surveillance of the leak tightness of the cask during the storage period.


Author(s):  
SangWon Shin ◽  
JaMin Lee ◽  
SuHong Lee ◽  
ChangYeal Baeg ◽  
JeongHyoun Yoon

The simulation tool of spent fuel transportation plan is being developed to support the development of core technology for transporting and storing spent fuel developed by the Korea Radioactive Waste Management Corporation (KRMC). So we extracted the major activities involved in spent fuel transport — such as setting-up a transport volume, infrastructure, processes and systems. And then we analyzed the components and properties of each activity. The impact factors related to spent fuel transport were determined to be maximum loading capacity of the INF ship and interim storage facility location, and it was around these that the preliminary scenario was generated. This will be used as a conceptual data model for the development of a simulation tool for spent fuel transportation plans.


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