scholarly journals Elements Present in Leach Solutions from Unsaturated Spent Fuel Tests

1993 ◽  
Vol 333 ◽  
Author(s):  
P. A. Finn ◽  
J. K. Bates ◽  
J. C. Hoh ◽  
J. W. Emery ◽  
L D. Hafenrichter ◽  
...  
Keyword(s):  
Performance Assessment ◽  
Radionuclide Transport ◽  
Spent Fuel ◽  
Major Fraction ◽  
Transport Models ◽  
Preliminary Results ◽  
J 13

ABSTRACTPreliminary results for the composition of the leachate from unsaturated tests at 90°C with spent fuel for two successive periods of ~60 days each with pretreated J-13 groundwater are reported. The pH of the leachate solutions ranged from 4 to 7. The americium concentration was 104 to 105 greater than that reported for saturated spent fuel tests in which the leachate pH was 8. The major fraction of material in the leachate was present as colloids containing both americium and curium. The presence of actinides in a form not currently directly included in repository radionuclide transport models provides information that can be used in spent fuel reaction modeling, the performance assessment of the repository and the design of the engineering barrier system.

Static Dissolution of α-doped UO2 in Boom Clay Conditions: Preliminary Results

2003 ◽  
Vol 807 ◽  
Author(s):  
C. Cachoir ◽  
K. Lemmens
Keyword(s):  
Corrosion Rate ◽  
Performance Assessment ◽  
Dissolution Rate ◽  
Near Field ◽  
Spent Fuel ◽  
Boom Clay ◽  
Preliminary Results ◽  
Icp Ms ◽  
Radiochemical Measurement ◽  
Α Activity

ABSTRACTTo support the performance assessment (PA) calculations for the possible final disposal of spent fuel in the Boom Clay, leach experiments are performed with α-doped UO2 in clay media, simulating various near field ages. The experiments allow to measure the dissolution rate of the UO2 and to determine the assumed relationship between dissolution rate and α-activity. Tests are performed at six different α-activities, simulating various fuel ages, at 25–30°C, for durations ranging from 90 to 720 days, in a glove box with Ar/0.4%CO2 atmosphere. The solutions and solids are analyzed for U isotopes and 238Pu by radiochemical measurement and by ICP-MS. The dissolution rates of the α-doped UO2 are presented for different durations. The resulting corrosion rate is around 300 μgU.m−2.d−1. This is up to 100 times higher than found for similar conditions in the literature. In the presence of clay, there appears to be no correlation between the α-activity and the corrosion rate of the α-doped UO2.

Release of Radionuclides from Spent Fuel Under Repository Conditions: Mathematical Modelling and Preliminary Results

2000 ◽  
Vol 663 ◽  
Author(s):  
Esther Cera ◽  
Juan Merino ◽  
Jordi Bruno
Keyword(s):  
Numerical Model ◽  
Mathematical Modelling ◽  
Performance Assessment ◽  
Source Term ◽  
Relative Importance ◽  
Spent Fuel ◽  
Oxidative Dissolution ◽  
Water Turnover ◽  
Preliminary Results ◽  
The Matrix

ABSTRACTIn the framework of the Enresa 2000 PA exercise and as a continuation of the developments made during SR 97, we have developed a conceptual and numerical model to calculate the release of radionuclides from spent fuel under repository conditions. The model includes both thermodynamic and kinetic considerations. Hence, although certain radionuclides are solubility controlled, for other radionuclides their release is governed by kinetic processes such as radiolytically promoted oxidative dissolution of the matrix and the associated water turnover inthe gap. The fluxes of selected radionuclides are calculated as an indication of the relative importance of the various processes considered to define source term concentrations in the performance assessment of the spent fuel repository.

Engineered Components for Spent Fuel Radioactive Waste Isolation Systems—are They Technically Justified?

1981 ◽  
Vol 11 ◽  
Author(s):  
H. C. Burkholder
Keyword(s):  
United States ◽  
Radioactive Waste ◽  
Release Rate ◽  
The United States ◽  
Radioactive Waste Disposal ◽  
Radionuclide Transport ◽  
Spent Fuel ◽  
Isolation System ◽  
Waste Forms ◽  
Isolation Systems

In response to draft radioactive waste disposal standards, R&D programs have been initiated in the United States which are aimed at developing and ultimately using radionuclide transport-delaying (e.g., long-lived waste containers) and radionuclide transport-controlling (e.g., very low release rate waste forms) engineered components as part of the isolation system. Before these programs proceed significantly, it seems prudent to evaluate the technical justification for development and use of sophisticated engineered components in radioactive waste isolation.

scholarly journals Spent Fuel as a Waste Form - Data Needs to Allow Long Term Performance Assessment Under Repository Disposal Conditions.

1986 ◽  
Vol 84 ◽  
Author(s):  
V. M. Oversby
Keyword(s):  
Performance Assessment ◽  
Waste Form ◽  
High Level Waste ◽  
Detailed Knowledge ◽  
Spent Fuel ◽  
Expected Performance ◽  
High Level ◽  
Term Performance ◽  
Physical And Chemical ◽  
Waste Repositories

AbstractPerformance assessment calculations are required for high level waste repositories for a period of 10,000 years under NRC and EPA regulations. In addition, the Siting Guidelines (IOCFR960) require a comparison of sites following site characterization and prior to final site selection to be made over a 100,000 year period. In order to perform the required calculations, a detailed knowledge of the physical and chemical processes that affect waste form performance will be needed for each site. While bounding calculations might be sufficient to show compliance with the requirements of IOCFR60 and 40CFRI91, the site comparison for 100,000 years will need to be based on expected performance under site specific conditions. The only case where detailed knowledge of waste form characteristics in the repository would not be needed would be where radionuclide travel times to the accessible environment can be shown to exceed 100,000 years. This paper will review the factors that affect the release of radionuclides from spemt fuel under repository conditions, summarize our present state of knowledge, and suggest areas where more work is needed in order to support the performance assessment calculations.

REPRO: Through Diffusion Experiment (TDE) - Diffusion and Porosity Properties of Rock Matrix in Stress Field of Repository Level

2020 ◽  
Author(s):  
Kalle Rahkola ◽  
Antti Poteri ◽  
Lasse Koskinen ◽  
Peter Andersson ◽  
Kersti Nilsson ◽  
...  
Keyword(s):  
Stress Field ◽  
Performance Assessment ◽  
Spent Nuclear Fuel ◽  
Diffusion Experiment ◽  
Spent Fuel ◽  
Tracer Transport ◽  
Rock Matrix ◽  
Safety Case ◽  
Through Diffusion

<p>Radionuclides usually migrate slower than the flowing water due to sorption and matrix diffusion. The performance assessment assumes that retention takes place mostly in the vicinity of the deposition holes. REPRO (<em>REtention Properties of ROck matrix</em>) experiments analyzed the matrix retention properties of the rock matrix under realistic conditions deep in the bedrock in ONKALO underground characterization facility at Olkiluoto, Finland. The objective was to investigate tracer transport in the rock matrix, which was representative to the near-field of the final disposal repository of the spent nuclear fuel, and to demonstrate that the assumptions made in the safety case of the deep geological spent fuel repository were in line with site evidence.</p><p>REPRO is composed of several supporting laboratory and <em>in-situ</em> experiments which investigate the retention properties under different experimental configurations. The first <em>in-situ</em> experiments were water phase diffusion experiments performed 2012-2013. Through Diffusion Experiment (TDE) studies diffusion and porosity properties of rock matrix in stress field of repository level and sorption properties of nuclides in intact rock circumstances.</p><p>The TDE experiment has been performed in three parallel drillholes drilled near to each other. Breakthrough of the radioactive tracer is monitored with on-line measurements and samplings along and perpendicular to the foliation. The non-sorbing radioactive isotope traces of HTO and <sup>36</sup>Cl, as well as slightly sorbing <sup>22</sup>Na and strongly sorbing <sup>133</sup>Ba and <sup>134</sup>Cs were used. TDE was designed to control advective flow, as it had caused problems in previous <em>in-situ</em> tests.</p><p>Supporting laboratory studies were performed for drillcore samples sampled from the experimental drillholes. In these laboratory experiments, i.e. porosity, permeability and diffusion coefficients of the drillcores were determined using different methods.</p><p>The TDE experiment was carried out from 2016 to 2019. A breakthrough was seen in the timeframe predicted by scoping calculations carried out. REPRO has produced data and knowledge to the safety case and the performance assessment. According to the preliminary results, values measured in the laboratory are applicable also in larger scale and <em>in-situ</em> conditions.</p>

Building Confidence in Radionuclide Transport Models for Fractured Rock: The Nagra/JNC Radionuclide Retardation Programme

2000 ◽  
Vol 663 ◽  
Author(s):  
K. Ota ◽  
W.R. Alexander ◽  
P.A. Smith ◽  
A. Möri ◽  
B. Frieg ◽  
...  
Keyword(s):  
Fractured Rock ◽  
Test Site ◽  
Process Models ◽  
Swiss Alps ◽  
Radionuclide Transport ◽  
Transport Models ◽  
North East ◽  
Host Rocks ◽  
Experimental Technology

ABSTRACTThe joint Nagra/JNC Radionuclide Retardation Programme has now been ongoing for 15 years with the main aim of direct testing of radionuclide transport models in as realistic a manner as possible. A large programme of field, laboratory and natural analogue studies has been carried out at the Grimsel Test Site in the central Swiss Alps and the Kamaishi In Situ Test Site in north-east Japan. The understanding and modelling of both the processes and the structures influencing radionuclide transport/retardation in fractured host rocks have matured as has the experimental technology, which has contributed to develop confidence in the applicability of the underlying research models in a repository performance assessment. In this paper, the successes and set-backs of this programme are discussed as is the general approach to the thorough testing of the process models and of model assumptions. In addition, a set of key findings is presented, involving discussions on the enhancement of confidence through the program.

What really happens to the “vanishing overpack”?

2012 ◽  
Vol 1475 ◽  
Author(s):  
I. G. McKinley ◽  
F. B. Neall ◽  
E. M. Scourse ◽  
H. Kawamura
Keyword(s):  
Radionuclide Transport ◽  
Spent Fuel ◽  
Geological Disposal ◽  
Failure Times ◽  
Bentonite Buffer ◽  
Physical Barriers ◽  
Safety Assessments ◽  
High Level ◽  
Radionuclide Release

ABSTRACTConcepts for the disposal of high-level radioactive waste (HLW) and spent fuel (SF) in several countries include a massive steel overpack within a bentonite buffer. In past conservative safety assessments to demonstrate feasibility of geological disposal, overpacks are assumed to provide complete containment for a given lifetime, after which all fail simultaneously. After failure, they are ignored as physical barriers to radionuclide transport. In order to compare different repository designs for specific sites, however, a more realistic treatment of overpack failure and its subsequent behaviour is needed. In addition to arguing for much longer lifetimes before mechanical failure and a distribution of overpack failure times, such assessment indicates that the presence of the failed overpack greatly constrains radionuclide release from the waste matrix and subsequent migration through the engineered barrier system. It also emphasises the key role of the bentonite buffer and the need to be able to assure its performance over relevant timescales.

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