Source Trends for Performance Assessment of HLW Glass and Spent Fuel as Waste Forms

1997 ◽  
Vol 506 ◽  
Author(s):  
B. Grambow
Keyword(s):  
Mass Transfer ◽  
Performance Assessment ◽  
Source Term ◽  
Reaction Rates ◽  
Dissolution Mechanism ◽  
Experimental Measurements ◽  
Source Terms ◽  
Spent Fuel ◽  
Fuel Oxidation ◽  
Waste Forms

ABSTRACTwith respect to the state of validation for source term development. Consequences of the various mechanism on mass half lives of the waste forms are calculated with analytical equations. For glass the largest uncertainty stems from the yet unclear dissolution mechanism under silica saturated conditions. Source terms based on silica solubility coupled to Si-mass transfer are probably neither conservative nor realistic. For spent fuel the largest uncertainty is in the extrapolation of radiolytic fuel oxidation for long periods of time. Considering the uncertainties involved, reaction rates cannot yet be extrapolated reliably to values much lower than the lowest reliable experimental measurements.

Alternate Source Term Models for Yucca Mountain Performance Assessment Based on Natural Analog Data and Secondary Mineral Solubility

1999 ◽  
Vol 556 ◽  
Author(s):  
William M. Murphy ◽  
Richard B. Codell
Keyword(s):  
Performance Assessment ◽  
Release Rate ◽  
Source Term ◽  
Yucca Mountain ◽  
Solution Chemistry ◽  
Base Case ◽  
Total System ◽  
Spent Fuel ◽  
Natural Analog ◽  
Mineral Solubility

AbstractPerformance assessment calculations for the proposed high level radioactive waste repository at Yucca Mountain, Nevada, were conducted using the Nuclear Regulatory Commission Total-System Performance Assessment (TPA 3.2) code to test conceptual models and parameter values for the source term based on data from the Peña Blanca, Mexico, natural analog site and based on a model for coprecipitation and solubility of secondary schoepite. In previous studies the value for the maximum constant oxidative alteration rate of uraninite at the Nopal I uranium body at Peña Blanca was estimated. Scaling this rate to the mass of uranium for the proposed Yucca Mountain repository yields an oxidative alteration rate of 22 kg yr−1, which was assumed to be an upper limit on the release rate from the proposed repository. A second model was developed assuming releases of radionuclides are based on the solubility of secondary schoepite as a function of temperature and solution chemistry. Releases of uranium are given by the product of uranium concentrations at equilibrium with schoepite and the flow of water through the waste packages. For both models, radionuclides other than uranium and those in the cladding and gap fraction were modeled to be released at a rate proportional to the uranium release rate, with additional elemental solubility limits applied. Performance assessment results using the Peña Blanca oxidation rate and schoepite solubility models for Yucca Mountain were compared to the TPA 3.2 base case model, in which release was based on laboratory studies of spent fuel dissolution, cladding and gap release, and solubility limits. Doses calculated using the release rate based on natural analog data and the schoepite solubility models were smaller than doses generated using the base case model. These results provide a degree of confidence in safety predictions using the base case model and an indication of how conservatism in the base case model may be reduced in future analyses.

scholarly journals MELCOR Demonstration Analysis of Accident Scenarios at a Spent Nuclear Reprocessing Plant

2020 ◽  
Author(s):  
Kenneth C. Wagner ◽  
David L. Y. Louie
Keyword(s):  
Source Term ◽  
Model Development ◽  
Source Terms ◽  
Spent Fuel ◽  
Nuclear Facilities ◽  
Demonstration Analysis ◽  
Nuclear Reprocessing ◽  
Probabilistic Risk Assessments ◽  
Accident Scenarios ◽  
Reprocessing Plant

Abstract The work presented in this paper applies the MELCOR code developed at Sandia National Laboratories to evaluate the source terms from potential accidents in non-reactor nuclear facilities. The present approach provides an integrated source term approach that would be well-suited for uncertainty analysis and probabilistic risk assessments. MELCOR is used to predict the thermal-hydraulic conditions during fires or explosions that includes a release of radionuclides. The radionuclides are tracked throughout the facility from the initiating event to predict the time-dependent source term to the environment for subsequent dose or consequence evaluations. In this paper, we discuss the MELCOR input model development and the evaluation of the potential source terms from the dominated fire and explosion scenarios for a spent fuel nuclear reprocessing plant.

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.

Retrograde Solubilities of Source Term Phases

1996 ◽  
Vol 465 ◽  
Author(s):  
William M. Murphy
Keyword(s):  
Heat Dissipation ◽  
Source Term ◽  
Spent Nuclear Fuel ◽  
Radioactive Decay ◽  
Source Area ◽  
Reaction Rates ◽  
Mineral Dissolution ◽  
Spent Fuel ◽  
Natural Analog ◽  
Using Data

ABSTRACTNatural analog, experimental, and thermodynamic studies indicate that the properties of secondary uranyl minerals are likely to control the source term for U and other radioéléments incorporated in these phases in the proposed Yucca Mountain repository. Thermodynamic calculations using data from the 1992 NEA data base indicate an increase in the equilibrium constant for schoepite dissolution from 103.1 to 104.8 with decreasing temperature from 100° to 25°C, i.e., retrograde solubility. Enthalpies for mineral transformation reactions that consume protons and release cations are typically negative, suggesting that solubilities of other uranyl phases such as uranophane increase more than that of schoepite with decreasing temperature. Solubilities of mineral phases associated with spent nuclear fuel will be initially relatively low under the elevated repository temperature regime. As the temperature of the repository decreases due to radioactive decay and heat dissipation, source term mineral solubilities increase. The rate of release of U and other species is controlled by a series of processes: transport of oxidants and flux of water; oxidative dissolution of spent fuel; uranyl mineral precipitation; uranyl mineral dissolution or transformation; and radionuclide transport. Decreasing diffusion and reaction rates and increasing uranyl mineral solubilities with decreasing temperature may lead to a change with time from solubility to transport or reaction rate as a source term controlling mechanism. Preservation of large quantities of uranyl minerals formed by oxidation of uraninite and radiometrie ages of secondary uranophane at the natural analog site at Peña Blanca indicate that oxidative alteration of uraninite was fast relative to transport of U away from the deposit. The successive formation of schoepite and uranophane in natural settings where uraninite has been oxidized may represent a paragenesis reflecting increasing temperature or increasing incorporation of environmental components. In contrast, diminishing temperature conditions in a repository source area could lead to the reverse sequence of mineral formation.

Long Term Modelling of Spent Fuel Oxidation/Dissolution Under Repository Conditions

2001 ◽  
Author(s):  
Juan Merino ◽  
Esther Cera ◽  
Jordi Bruno ◽  
Trygve Eriksen ◽  
Javier Quiñones ◽  
...  
Keyword(s):  
Source Term ◽  
Redox System ◽  
Water Interface ◽  
Spent Fuel ◽  
Fuel Oxidation ◽  
The Subject ◽  
The Stability ◽  
Thermodynamic Processes

Abstract A model to study the stability of the spent fuel under repository conditions has been developed. The fuel-water interface is a dynamic redox system, where oxidising conditions due to the radiolysis of water can lead to the release of the uranium and the radionuclides embedded in the fuel matrix. Both kinetic and thermodynamic processes have been taken into account. Special attention is given to the unit rate of matrix oxidation/dissolution, which has been the subject of a specific radiolytic model. The findings of this work have important implications for the applicability of solubility limits in establishing source term models.

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 CP-violating transport theory for electroweak baryogenesis with thermal corrections

2021 ◽  
Vol 2021 (11) ◽  
pp. 042
Author(s):  
Kimmo Kainulainen
Keyword(s):  
Source Term ◽  
Transport Theory ◽  
Finite Width ◽  
Source Terms ◽  
Boltzmann Equations ◽  
Current Divergence ◽  
The One ◽  
Infrared Divergences ◽  
Correct Implementation ◽  
Divergence Equations

Abstract We derive CP-violating transport equations for fermions for electroweak baryogenesis from the CTP-formalism including thermal corrections at the one-loop level. We consider both the VEV-insertion approximation (VIA) and the semiclassical (SC) formalism. We show that the VIA-method is based on an assumption that leads to an ill-defined source term containing a pinch singularity, whose regularisation by thermal effects leads to ambiguities including spurious ultraviolet and infrared divergences. We then carefully review the derivation of the semiclassical formalism and extend it to include thermal corrections. We present the semiclassical Boltzmann equations for thermal WKB-quasiparticles with source terms up to the second order in gradients that contain both dispersive and finite width corrections. We also show that the SC-method reproduces the current divergence equations and that a correct implementation of the Fick's law captures the semiclassical source term even with conserved total current ∂μ j μ = 0. Our results show that the VIA-source term is not just ambiguous, but that it does not exist. Finally, we show that the collisional source terms reported earlier in the semiclassical literature are also spurious, and vanish in a consistent calculation.

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