Transient Diffusional Release from Waste Packages in a Repository in Basalt

1985 ◽  
Vol 50 ◽  
Author(s):  
M. S. Bensky ◽  
D. L. Oliver
Keyword(s):  
Release Rate ◽  
Host Rock ◽  
Near Field ◽  
Temporal Variations ◽  
Waste Form ◽  
Spent Fuel ◽  
Constant Concentration ◽  
Release Rates ◽  
Total Inventory ◽  
System Characteristics

AbstractAnalyses of diffusional release of several typical radionuclides in spent fuel from waste packages emplaced in a repository in basalt were conducted to assess the effects of system characteristics and boundary conditions on computed release rates. Radionuclide releases, including spatial and temporal variations that may be present, represent the source term for transport in the geohydrologic setting and are therefore critical to the assessment of repository acceptability.Two mathematical approaches were utilized to determine radionuclide release rate versus time characteristics; (1) an analytical solution for one-deimensional diffusion based upon a Dirichlet (constant-concentration) boundary at the waste form surface; and (2) a finite-element numerical solution based upon a Neumann (zero-flux boundary at the waste form surface. The latter method is suitable for radionuclides such as 129I, whose total inventory in spent fuel could be quickly depleted from the waste form and dissolved in the pore spaces of the packing material surrounding the waste form and which, therefore, cannot be adequately represented by a constant concentration at the waste form (i.e., container) surface.The analysis revealed several system characteristics that are not intuitively obvious. For example, strong sorption in the near-field host rock behaves like a strong mass sink and can yield calculated transient release rates exceeding allowable limits. Similarly, a short half-life effectively removes the radionuclide from the host rock, which induces a steep concentration gradient at the host rock/packing interface and thereby increases the diffusional release rate at that boundary.Typical results for 79Se and 129I are presented to illustrate these effects. The effects of perturbations to key assumptions are shown to indicate the importance of (1) formulating models that accurately represent the physical system and (2) interpreting analytical results carefully to ensure understanding of the capability of the system.

The Effect of Nearfield Constraints on the Corrosion Behavior of High Burnup Spent Fuel

2006 ◽  
Vol 932 ◽  
Author(s):  
Andreas Loida ◽  
Manfred Kelm ◽  
Bernhard Kienzler ◽  
Horst Geckeis ◽  
Andreas Bauer
Keyword(s):  
Host Rock ◽  
Near Field ◽  
Experimental Studies ◽  
Spent Fuel ◽  
Mutual Support ◽  
Significant Extent ◽  
Dissolution Rates ◽  
The Matrix ◽  
Container Material

ABSTRACTThe long-term immobilization for individual radioelements released from the waste form “spent fuel” in solid phases upon groundwater contact depends strongly on the (geo)chemical constraints prevailing in the repository. Related experimental studies comprise effects induced by the presence of Fe based container material, and near field materials other than Fe for a rock salt environment. The effect of the presence of an argillaceous host rock containing organic matter and pyrite on fuel alteration was studied in addition. The results have shown that oxidative radio-lysis products were found to be consumed at a significant extent by the metallic Fe and by the argillaceous host rock. Under these conditions a decrease at a factor of ca.100 for both the matrix dissolution rates and the solution concentrations of U and Pu was found. There is mutual support between the matrix dissolution rates, the solution concentrations and the amounts of oxygen encountered during the experiments under various conditions controlled by the presence of near field materials under study.

scholarly journals The NNWSI Waste form Testing Program

1983 ◽  
Vol 26 ◽  
Author(s):  
V. M. Oversby
Keyword(s):  
Release Rate ◽  
Chemical Elements ◽  
Performance Testing ◽  
Interactive Effects ◽  
Waste Form ◽  
Spent Fuel ◽  
Testing Program ◽  
Reactor Pool ◽  
Reaction Vessels ◽  
Topopah Spring Tuff

ABSTRACTA waste form testing program has been developed to ensure that the release rate of radionuclides from the engineered barrier system will meet NRC and EPA regulatory requirements. Waste form performance testing will be done under unsaturated, low water availability conditions which represent the expected repository conditions. Testing will also be done under conditions of total immersion of the waste form in repository-type water to cover the possibility that localized portions of the repository might contain standing water. Testing of reprocessed waste forms for CHLW and DHLW will use reaction vessels fabricated from Topopah Spring tuff. Chemical elements which are expected to show the highest release rates in the mildly oxidizing environment of the Topopah Spring tuff horizon at Yucca Mountain are Np and Tc. To determine the effect of residual canister material and of corrosion products from the canister/overpack, waste form testing will be done in the presence of these materials. The release rate of all radionuclides which are subject to NRC and EPA regulations will be measured, and the interactive effects of the released radionuclides and the rock reaction vessels will be determined. The testing program for spent fuel will determine the release rate from bare spent fuel pellets and from Zircaloy clad spent fuel where the cladding contains minor defects. A metal testing program for Zircaloy will establish the expected lifetime of the cladding material. Estimation of the state of cladding for fuel presently in reactor pool storage will provide baseline data for Zircaloy containment credit.

Mobilization/Retention of Radionuclides during Corrosion of High Burnup Spent Fuel and Backfill Materials in Salt Brines

2002 ◽  
Vol 757 ◽  
Author(s):  
Andreas Loida ◽  
Bernhard Kienzler ◽  
Horst Geckeis
Keyword(s):  
Near Field ◽  
Waste Form ◽  
Spent Fuel ◽  
The Matrix ◽  
Backfill Materials ◽  
Salt Brine ◽  
Solid Phases ◽  
Long Term Behavior ◽  
The Impact

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.

Radionuclide Transport Calculations in the Safety Assessment SR 97

2000 ◽  
Vol 663 ◽  
Author(s):  
M. Lindgren ◽  
F. Lindström
Keyword(s):  
Input Data ◽  
Safety Assessment ◽  
Spent Nuclear Fuel ◽  
Near Field ◽  
Radionuclide Transport ◽  
Spent Fuel ◽  
Release Rates ◽  
Dose Limits ◽  
The Impact ◽  
Different Parts

ABSTRACTThis study treats radionuclide transport calculations for a canister defect scenario in the safety assessment SR 97, which concerns a deep repository for spent nuclear fuel of the KBS-3 type in Sweden. The aims of the calculations are to:Quantitatively describe the radionuclide transport.Show the impact of uncertainty in input data and show which parameters govern the calculated release rates.Compare three different real sites in Sweden (Aberg, Beberg and Ceberg) with each other and with dose limits given in Swedish regulations (none of the sites is considered in the on-going localization process). Only briefly described in this paper.Illustrate the impact of the different barriers in the system.Deterministic calculations illustrate the radionuclide transport for reasonable conditions. Uncertainty cases show the influence of the uncertainty for data related to different parts of the repository system by systematically giving them pessimistic values while all others are reasonable. Simplified probabilistic calculations have also been performed.The analysis shows that the most important parameters in the near field are the number of defective canisters and the instant release fraction. In the far field the most important uncertainties affecting release and retention are connected to permeability and connectivity of the fractures in the rock. The dose rate in the biosphere is essentially controlled by the possibilities of dilution.The calculated maximum doses for the hypothetical repositories are well below the dose limits, and hence they meet the acceptance criteria for a deep repository for spent fuel.

Host Phases for Actinide Elements in the Metallic Waste Form

2002 ◽  
Vol 757 ◽  
Author(s):  
D. E. Janney
Keyword(s):  
Stainless Steel ◽  
Electron Microscope ◽  
Electron Diffraction ◽  
Transmission Electron Microscope ◽  
Waste Form ◽  
Dissolution Behavior ◽  
Release Rates ◽  
Waste Forms ◽  
Transmission Electron ◽  
Simulated Waste

ABSTRACTArgonne National Laboratory has developed an electrometallurgical process for conditioning spent sodium-bonded metallic reactor fuel prior to disposal. A waste stream from this process consists of stainless steel cladding hulls that contain undissolved metal fission products such as Tc, Ru, Rh, Pd, and Ag; a small amount of undissolved actinides (U, Np, Pu) also remains with the hulls. These wastes will be immobilized in a waste form whose baseline composition is stainless steel alloyed with 15 wt% Zr (SS-15Zr). Scanning electron microscope (SEM) observations of simulated metal waste forms (SS-15Zr with up to 11 wt% actinides) show eutectic intergrowths of Fe-Zr-Cr-Ni intermetallic phases with steels. The actinide elements are almost entirely in the intermetallics, where they occur in concentrations ranging from 1–20 at%. Neutron- and electron-diffraction studies of the simulated waste forms show materials with structures similar to those of Fe2Zr and Fe23Zr6.Dissolution experiments on simulated waste forms show that normalized release rates of U, Np, and Pu differ from each other and from release rates of other elements in the sample, and that release rates for U exceed those for any other element (including Fe). This paper uses transmission electron microscope (TEM) observations and results from energy-dispersive X-ray spectroscopy (EDX) and selected-area electron-diffraction (SAED) to characterize relationships between structural and chemical data and understand possible reasons for the observed dissolution behavior.Transmission electron microscope observations of simulated waste form samples with compositions SS-15Zr-2Np, SS-15Zr-5U, SS-15Zr-11U-0.6Rh-0.3Tc-0.2Pd, and SS-15Zr-10Pu suggest that the major actinide-bearing phase in all of the samples has a structure similar to that of the C15 (cubic, MgCu2-type) polymorph of Fe2Zr, and that materials with this structure exhibit significant variability in chemical compositions. Material whose structure is similar to that of the C36 (dihexagonal, MgNi2-type) polymorph of Fe2Zr was also observed, and it exhibits less chemical variability than that displayed by material with the C15 structure. The TEM data also demonstrate a range of actinide concentrations in materials with the Fe23Zr6 (cubic, Mn23Th6-type) structure.Microstructures similar to those produced during experimental deformation of Fe-10 at% Zr alloys were observed in intermetallic materials in all of the simulated waste form samples. Stacking faults and associated dislocations are common in samples with U, but rarely observed in those with Np and Pu, while twins occurred in all samples. The observed differences in dissolution behavior between samples with different actinides may be related to increased defect-assisted dissolution in samples with U.

A Description of the Ceramic Waste Form Production Process from the Demonstration Phase of the Electrometallurgical Treatment of EBR-II Spent Fuel

2001 ◽  
Vol 134 (3) ◽  
pp. 263-277 ◽  
Author(s):  
Michael F. Simpson ◽  
K. Michael Goff ◽  
Stephen G. Johnson ◽  
Kenneth J. Bateman ◽  
Terry J. Battisti ◽  
...  
Keyword(s):  
Production Process ◽  
Waste Form ◽  
Spent Fuel ◽  
Ceramic Waste

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 RADIONUCLIDE CHARACTERISTICS OF RDE SPENT FUELS

2018 ◽  
Vol 20 (2) ◽  
pp. 69 ◽  
Author(s):  
Ihda Husnayani ◽  
Pande Made Udiyani
Keyword(s):  
Thermal Power ◽  
Neutron Production ◽  
Cooling Time ◽  
Spent Fuel ◽  
Release Rates ◽  
Radiation Dose Rate ◽  
The Core ◽  
Spent Fuel Storage ◽  
Fuel Storage

Reaktor Daya Eksperimental (RDE) is a 10 MWth pebble-bed High Temperature Gas-cooled Reactor that is planned to be constructed by National Nuclear Energy Agency of Indonesia (BATAN) in Puspiptek complex, Tangerang Selatan. RDE utilizes low enriched UO2 fuel coated by TRISO layers and loaded into the core by means of multipass loading scheme. Determination of radionuclide characteristics of RDE spent fuel; such as activity, thermal power, neutron and photon release rates; are very important because those characteristics are crucial to be used as a base for evaluating the safety of spent fuel handling system and storage tank. This study is aimed to investigate the radionuclide characteristics of RDE spent fuel at the end of cycle and during the first 5 years cooling time in spent fuel storage. The method used to investigate the radionuclide characteristics is burnup calculation using ORIGEN2.1 code. In performing the ORIGEN2.1 calculation, one pebble fuel was assumed to be irradiated in the core for 5 cycles and then decayed for 5 years. At the end of the fifth cycle, it is obtained that the total activity, thermal power, neutron production, and photon release rates from all radionuclides inside one spent fuel are approximately 105.68 curies, 0.41 watts, 2.65 x 103 neutrons/second, and 1.79 x 104 photons/second, respectively. The results for the radionuclides characteristics during the first 5 years cooling time in the spent fuel storage show that the radioactivity characteristics from all radionuclides are rapidly decreasing at the first year and then slowly decreasing at the second until the fifth year of cooling time. The results obtained in this study can provide data for safety evaluation of fuel handling and spent fuel storage, such as the calculation of sourceterm, radiation dose rate, and the determination of radiation shielding.Keywords: RDE, spent fuel, radionuclide activity, thermal power, neutron production, photon releaserates KARAKTERISTIK RADIONUKLIDA DI DALAM BAHAN BAKAR RDE. Reaktor Daya Eksperimental (RDE) adalah reaktor tipe Reaktor Temperatur Tinggi Berpendingin Gas dengan daya termal 10MW yang akan dibangun oleh BadanTenagaNuklirNasional (BATAN) di kawasanPuspiptek, Tangerang Selatan. RDE menggunakan bahan bakar UO2 yang dilapisi dengan lapisan TRISO dan dimasukkan ke dalam teras RDE menurut skema multipass (5 siklus). Penentuan karakteristik radionuklida di dalam bahan bakar RDE; seperti aktivitas, daya termal, laju produksi neutron dan pelepasan foton; adalah sangat penting karena informasi karakteristik ini diperlukan sebagai dasar untuk melakukan evaluasi keselamatan system penanganan dan penyimpanan bahan bakar bekas. Penelitian ini bertujuan untuk menganalisis karakteristik radionuklida bahanbakar RDE setelah 5 siklus dan pada 5 tahun pertama pendinginan ditempat penyimpanan bahan bakar bekas. Metode yang digunakan dalam menghitung karakteristik radionuklida adalah menggunakan program ORIGEN2.1. Satu bola bahan bakar RDE diasumsikan diiradiasi selama 5 siklus dan kemudian meluruh selama 5 tahun. Pada akhir siklus, diperoleh hasil aktivitas total, daya termal, laju produksi neutron dan pelepasan foton dari seluruh radionuklida di dalam satu bola bahan bakar RDE sebesar 105,68 curies, 0,41 watts, 2,65 x 103 neutron/detik, dan 1,79 x 104 foton/detik. Hasil untuk karakteristik radionuklida selama 5 tahun penyimpanan menunjukkan bahwa karakteristik radioktivitas radionuklida menurun dengan cepat pada tahun pertama dan kemudian menurun lebih lambat pada tahun kedua hingga tahun kelima. Hasil perhitungan karakteristik radionuklida dari penelitian ini dapat digunakan sebagai basis untuk analisis keselamatan penanganan dan penyimpanan bahan bakarbekas RDE.Kata kunci:RDE, bahan bakar bekas, aktivitas radionuklida, daya termal, produksi neutron, laju foton

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.

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