Effect of alternative conceptual models in a preliminary performance assessment for the waste isolation pilot plant (original) (raw)
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Summary discussion of the 1996 performance assessment for the Waste Isolation Pilot Plant
Reliability Engineering & System Safety, 2000
The Waste Isolation Pilot Plant (WIPP) is under development by the U.S. Department of Energy (DOE) for the geologic disposal of transuranic waste. The construction of complementary cumulative distribution fimctions (CCDFS) for total radionuclide release from the WIPP to the accessible environment is described. The resultant CCDFS (i) combine releases due to cuttings~d cavings, spallings, direct brine release, and long-term transportin flowing groundwater, (ii) fall substantially to the left of the boundary line specified by the U.S. Environmental Protection Agency's (EPA's) standard40 CFR 191 for the geologic disposal of radioactive waste, and (iii) constitute an important component of the DOE's successfid Compliance Certification Application to the EPA for %.heWIPP.
1990
agencythereof, norany oftheir employees, norany of their contractors, subcontractors, ortheir employees, makes any warranty, express or implied, or_ssumesany legal liability or responsibility fortheaccuracy, completeness, ol usefulness of any information, apparatus, product, or process disclosed, er represents thati_suse would not infringe privately ownedrights. Reference herein toar.y specific commercial product, process, or service by tradename, trademdrk, manufacturer, or otherwise, does not necessarily constitute orimpiyits endorsement, recommendation, orfavoring by the United StatesGovernment,any agency thereof or any of their contractors orsubcontractors. The viewsand opinions expressed hereindo notnecessarily state or reflect thoseoftheUnitedStates Government, any agencythereof or any oftheir contractors. Printed intheOnitedStates ofAmerica.Thisreport has beenreproduced directly fromthebestavailable copy.
A Source Release Model with Application to the LANL LLRW Disposal Site Performance Assessment
MRS Proceedings
A source release model was developed to quantify time dependent liquid phase releases of radioactive material to the vadose zone from a disposal site. The model has been implemented to evaluate the source terms in the Performance Assessment for the Los Alamos National Laboratory (LANL) Low Level Radioactive Waste (LLRW) Disposal Facility at Area G, an analysis of the long term post-closure impact of disposal operations required by the USDOE orders.Analytic solutions describe transport through the model compartments: the solid phase waste package, the liquid phase within each waste package and the liquid phase within the disposal unit. The model accounts for elemental solubility limits and retardation coefficients (R) separately in the waste package and in the disposal unit. Several parameters define the site specific aspects of the disposal unit. In our case for example, the disposal unit is waste buried with crushed volcanic tuff, and a small net infiltration rate is determined fro...
Conceptual structure of the 1996 performance assessment for the Waste Isolation Pilot Plant
Reliability Engineering & System Safety, 2000
The conceptual structure of the 1996 performance assessment (PA) for the Waste Isolation Pilot Plant (WIPP) is described. This structure involves three basic entities (EN1, EN2, EN3): (i) EN1, a probabilistic characterization of the likelihood of different fimmes occurring at the WIPP site over the next 10,000 yr, (ii) EN2, a procedure for estimating the radionuclide releases to the accessible environment associated with each of the possl%le futures that could occur at the WIPP site over the next 10,000 yr, and (iii) EN3, a probabilistic characterization of the uncertainty in the parameters used in the definition of EN1 and EN2. In the formal development of the 1996 WIPP PA EN 1 is characterized by a probability space {S~fi~~pJ for stochastic (i.e., aleatory) uncertainly; EN2 is characterized by a function J that corresponds to the models and associated computer programs used to estimate radionuclide release$ and EN3 is characterized by a probability space (S~u, d SW pm) for subjective (i.e., epistemic) uncertainty. A high-level overview of the 1996 WIPP PA and references to additional sources of information are given in the context of (S~t, Sp ps:),f md (Ssu,~Su, PJ
Release rates from partitioning and transmutation waste packages
1991
1 .Introduction 1 2. Need for Evaluating the Benefits of Partitioning and Transmutation 1 3. An Equal Energy Production Comparison 2 4. Waste Characteristics and Inventories 5. Calculation of Release Rates 5.3 Calculated Release Rates 6. Conclusions 49 References Partitioning the actinides in light-water reactor spent fuel and transmuting them in actinide-burning liquid-metal reactors has been proposed as a potential method of reducing the public risks from geologic disposal of nuclear waste. As a first step towards quantifying the benefits for waste disposal of actinide burning, we have calculated the release rates of key radionuclides from waste packages resulting from actinide burning, and compare them with release rates from LWR spent fuel destined for disposal at the potential repository at Yucca Mountain. The wet-drip water-contact mode has been used. Analytic methods and parameter values are very similar to those used for assessing Yucca Mountain as a potential repository. Once released, the transport characteristics of radionuclides Will be largely determined by site geology. For the most important nuclides such as 1-129 and Tc-99, which are undiminished by actinide-burning reactors, it is not surprising that actinide burning offers little reduction in releases. For important actinides such as Np-237 and PU isotopes, which are reduced in inventory, the releases are not reduced because the release rates are proportional to solubility, rather than inventory. ' ' I " " " ' ' I " " " ' '
Derived concentration guideline levels for Argonne National Laboratory's building 310 area
2011
The derived concentration guideline level (DCGL) is the allowable residual radionuclide concentration that can remain in soil after remediation of the site without radiological restrictions on the use of the site. It is sometimes called the single radionuclide soil guideline or the soil cleanup criteria. This report documents the methodology, scenarios, and parameters used in the analysis to support establishing radionuclide DCGLs for Argonne National Laboratory's Building 310 area.
Implementation of Requirements on Non-Radioactive Waste Package Constituents - 10449 rev
2010
In radioactive waste disposal, attention must be paid not only to radiological impacts but also to impacts of chemotoxic waste package constituents. Thus, in order to demonstrate the safety of a geological repository in the post-closure phase, the possible releases of non-radioactive organic and inorganic substances via the water pathway are to be investigated. Within the licensing procedure for the Konrad repository it was shown that such constituents may or may not reach the near-surface groundwater in a quantity so small as to obviate the danger of its harmful pollution or a detrimental change of its characteristics. Based on those findings, the licensing authority issued the permitted levels of waterrelated properties as Annex 4 to the Konrad license. This included, in particular, quantitative limitations on maximum masses of 94 non-radioactive substances (waste package constituents). The determination of these limits and their implementation into practicable measures for the wa...
In the long run, nuclear waste packages will fail gradually due to localized and general corrosion and radionuclides will be transported to the accessible environment by ground water. Two main failure scenarios are expected for the waste packages: Flow-through model and bathtub model; in the flow-through model water flows through the waste container, while in the bathtub model water pools inside the waste package. The Department of Energy, in their performance assessment of the proposed repository at Yucca Mountain, excluded the bathtub model from their analysis and assumed the diffusion and advection to be in the same direction in the flow-through model. In this research, a new conceptual model is introduced for radionuclide release from a flow-through category failed waste container. Due to the residual heat release of the nuclear waste, this model expects a bidirectional transport for radionuclides in the sheltered areas; advection toward the warmest region and diffusion in the o...
National Low-Level Waste Management Program Radionuclide Report Series, Volume 7
1994
The National Low Level Waste Management Program at the Idaho National Engineering and Environmental Laboratory has published a report containing key information about selected radionuclides that are most likely to contribute significantly to the radiation exposures estimated from a performance assessment of a low-level radioactive waste (LLW) disposal facility. The information includes physical and chemical characteristics, production means, waste forms, behavior of the radionuclide in soils, plants, groundwater, and air, and biological effects in animals and humans. The radionuclides included in this study comprise all of the nuclides specifically listed in 10CFR61.55,
II-G-22: Conceptual Models Third Supplementary Peer Review Report, April 1997
This report is a third supplement to a July 1996 report that presented the results of an independent technical peer review of the adequacy of 24 conceptual models representing features, events and processes involved in assessing the long-term performance of the Waste Isolation Pilot Plant (WIPP). The peer review was initially conducted from April through June 1996 at the U.S. Department of Energy's (DOE's) Sandia National Laboratories by a six-member interdisciplinary Review Panel having the requisite broad experience to address the range of issues associated with waste isolation over the 10,000-year regulatory time frame. The Panel selection process and the biographies of the Panel members are included in the July 1996 report. In its January 1997 second supplementary report, the Panel stated that it continued to find two of the models not adequate to represent the future states of the repository. For the two models found not adequate, Spallings and Chemical Conditions, the Panel identified its remaining issues. In this third supplementary report the Panel considers the DOE's April 1997 responses to these remaining issues. The Panel's evaluation of these responses is presented in Section 3 of this report. For the Spallings Model, the Panel concluded that the predicted volumes of spalled materials presented in the WIPP Compliance Certification Application (CCA) are reasonable based on additional consideration of processes that could lead to spalled releases. The Panel also concluded that the MgO backfill component of the Chemical Conditions Model will function as assumed in the CCA and that this model is adequate to represent the future states of the repository.