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Analysis of tank calibration data from several runs

Description: The experimentally determined relationship between the level of liquid and volume of liquid in a tank can be used to obtain volume estimates that correspond to liquid-level measurements. Several calibration experiments (runs) are made to estimate the calibration equation. The calibration equation is used to estimate the quantity of liquid transferred from a tank between measurement periods. Difficulties can arise when run-to-run differences are large relative to the precisions of liquid-level and volume measurements. This paper addresses the seldom-discussed but important problem of combining and analyzing data from two or more calibration runs. Emphasis is placed on exploratory methods such as diagnostic plots that are compatible with applicable statistical models. 10 references, 3 figures, 2 tables.
Date: January 1, 1984
Creator: Goldman, A.S. & Liebetrau, A.M.
Partner: UNT Libraries Government Documents Department

Modeling the uncertain impacts of climate change

Description: Human and earth systems are extremely complex processes. The modeling of these systems to assess the effects of climate change is an activity fraught with uncertainty. System models typically involve the linking of a series of computer codes, each of which is a detailed model of some physical or social process in its own right. In such system models, the output from one process model is the input to another. Traditional methods for dealing with uncertainty are inadequate because of the sheer complexity of the modeling effort: Monte Carlo methods and the exhaustive evaluation of what if '' scenarios estimate sensitivities fail because of the heavy computational burden. More efficient methods are required for learning about system models that are constructed from a collection of computer codes. A two-tiered modeling approach is being developed to estimate the distribution of outcomes from a series of nested models. The basic strategy is to develop a simplified executive, or simplified system code (SSC), that is analogous to the more complex underlying code. An essential feature of the SSC is that it uses information abstracted from the detailed underlying process codes in a manner that preserves their essential features and interactions among them. Of course, to be useful, the SSC must be much faster to run than its complex counterpart. The success of the SSC modeling strategy depends on the methods used to extract essential features of the complex underlying codes.
Date: August 1, 1992
Creator: Liebetrau, A.M.
Partner: UNT Libraries Government Documents Department

Analysis of Tank 241-AN-106 characterization and grout performance criteria

Description: This report provides an assessment of how well we can resolve the following issues concerning Tank 241-AN-106 at the Hanford Reservation, given the current state of information: How well we can characterize the contents of 241-AN-106; whether the degree of characterization is sufficient to use 241-AN-106 wastes to develop tests of grout adequacy. The wastes must be characterized not only to ensure grout adequacy but also to provide assurance that the wastes can be successfully and safely transferred. In this report, we evaluate the adequacy of characterization for transfer and tests of grout adequacy, and we evaluate the current status of acceptance criteria and grout formulation experiments.
Date: February 1, 1993
Creator: Liebetrau, A.M. & Anderson, C.M.
Partner: UNT Libraries Government Documents Department

The Analytical Repository Source-Term (AREST) model: Analysis of spent fuel as a nuclear waste form

Description: The purpose of this report is to assess the performance of spent fuel as a final waste form. The release of radionuclides from spent nuclear fuel has been simulated for the three repository sites that were nominated for site characterization in accordance with the Nuclear Waste Policy Act of 1982. The simulation is based on waste package designs that were presented in the environmental assessments prepared for each site. Five distinct distributions for containment failure have been considered, and the release for nuclides from the UO/sub 2/ matrix, gap (including grain boundary), crud/surface layer, and cladding has been calculated with the Analytic Repository Source-Term (AREST) code. Separate scenarios involving incongruent and congruent release from the UO/sub 2/ matrix have also been examined using the AREST code. Congruent release is defined here as the condition in which the relative mass release rates of a given nuclide and uranium from the UO/sub 2/ matrix are equal to their mass ratios in the matrix. Incongruent release refers to release of a given nuclide from the UO/sub 2/ matrix controlled by its own solubility-limiting solid phase. Release of nuclides from other sources within the spent fuel (e.g., cladding, fuel/cladding gap) is evaluated separately from either incongruent or congruent matrix release. 51 refs., 200 figs., 9 tabs.
Date: February 1, 1989
Creator: Apted, M.J.; Liebetrau, A.M. & Engel, D.W.
Partner: UNT Libraries Government Documents Department

Uncertainty and sampling issues in tank characterization

Description: A defensible characterization strategy must recognize that uncertainties are inherent in any measurement or estimate of interest and must employ statistical methods for quantifying and managing those uncertainties. Estimates of risk and therefore key decisions must incorporate knowledge about uncertainty. This report focuses statistical methods that should be employed to ensure confident decision making and appropriate management of uncertainty. Sampling is a major source of uncertainty that deserves special consideration in the tank characterization strategy. The question of whether sampling will ever provide the reliable information needed to resolve safety issues is explored. The issue of sample representativeness must be resolved before sample information is reliable. Representativeness is a relative term but can be defined in terms of bias and precision. Currently, precision can be quantified and managed through an effective sampling and statistical analysis program. Quantifying bias is more difficult and is not being addressed under the current sampling strategies. Bias could be bounded by (1) employing new sampling methods that can obtain samples from other areas in the tanks, (2) putting in new risers on some worst case tanks and comparing the results from existing risers with new risers, or (3) sampling tanks through risers under which no disturbance or activity has previously occurred. With some bound on bias and estimates of precision, various sampling strategies could be determined and shown to be either cost-effective or infeasible.
Date: June 1, 1997
Creator: Liebetrau, A.M.; Pulsipher, B.A. & Kashporenko, D.M.
Partner: UNT Libraries Government Documents Department

Characterization report for the ferrocyanide safety issue

Description: Recently PNNL was tasked by DOE to develop and demonstrate a risk-based strategic approach to characterizing Hanford`s Nuclear Waste Tanks. This strategic approach was documented in a report entitled ``A Risk-Based Focused Decision-Management Approach for Justifying Characterization of Hanford Tank Waste``. In support of the general approach, a specific strategy for addressing each of the several safety issues associated with the tanks was developed. This report documents the approach for the Ferrocyanide Safety Issue. The purpose of this report is to describe a structured logic diagram (SLD) for determining the risk associated with the ferrocyanide tank safety issue and provide the supporting information for the SLD. The SLD addresses the resolution of risks resulting from the presence of ferrocyanide layers within the Hanford tanks. The informational requirements for determining risk from any reaction stemming from ferrocyanide are outlined in the SLD. This report will describe the potential paths to a successful resolution of the ferrocyanide safety issue. Complete development of the intervention pathway is outside the scope of this current activity. General descriptions of the approach, key components of the SLD, and conclusions are provided in the body of this report. The complete SLD, descriptions of each box shown in the SLD, a discussion on how to fill data needs, and a list of contributors is provided in the appendices.
Date: June 1, 1997
Creator: Pulsipher, B.A.; Burger, L.L.; Liebetrau, A.M. & Scheele, R.D.
Partner: UNT Libraries Government Documents Department

Comparison of NDA and DA measurement techniques for excess plutonium powders at the Hanford Site: Statistical design and heterogeneity testing

Description: Quantitative physical measurements are a n component of the International Atomic Energy Agency (IAEA) nuclear material m&guards verification regime. In December 1994, LA.FA safeguards were initiated on an inventory of excess plutonium powder items at the Plutonium Finishing Plant, Vault 3, on the US Department of Energy`s Hanford Site. The material originl from the US nuclear weapons complex. The diversity of the chemical form and the heterogenous physical form of this inventory were anticipated to challenge the precision and accuracy of quantitative destructive analytical techniques. A sampling design was used to estimate the degree of heterogeneity of the plutonium content of a variety of inventory items. Plutonium concentration, the item net weight, and the {sup 240}Pu content were among the variables considered in the design. Samples were obtained from randomly selected location within each item. Each sample was divided into aliquots and analyzed chemically. Operator measurements by calorimetry and IAEA measurements by coincident neutron nondestructive analysis also were performed for the initial physical inventory verification materials and similar items not yet under IAEA safeguards. The heterogeneity testing has confirmed that part of the material is indeed significantly heterogeneous; this means that precautionary measures must be taken to obtain representative samples for destructive analysis. In addition, the sampling variability due to material heterogeneity was found to be comparable with, or greater than, the variability of the operator`s calorimetric measurements.
Date: June 1, 1995
Creator: Welsh, T.L.; McRae, L.P.; Delegard, C.H.; Liebetrau, A.M.; Johnson, W.C.; Theis, W. et al.
Partner: UNT Libraries Government Documents Department

The Analytical Repository Source-Term (AREST) model: Description and documentation

Description: The geologic repository system consists of several components, one of which is the engineered barrier system. The engineered barrier system interfaces with natural barriers that constitute the setting of the repository. A model that simulates the releases from the engineered barrier system into the natural barriers of the geosphere, called a source-term model, is an important component of any model for assessing the overall performance of the geologic repository system. The Analytical Repository Source-Term (AREST) model being developed is one such model. This report describes the current state of development of the AREST model and the code in which the model is implemented. The AREST model consists of three component models and five process models that describe the post-emplacement environment of a waste package. All of these components are combined within a probabilistic framework. The component models are a waste package containment (WPC) model that simulates the corrosion and degradation processes which eventually result in waste package containment failure; a waste package release (WPR) model that calculates the rates of radionuclide release from the failed waste package; and an engineered system release (ESR) model that controls the flow of information among all AREST components and process models and combines release output from the WPR model with failure times from the WPC model to produce estimates of total release. 167 refs., 40 figs., 12 tabs.
Date: October 1, 1987
Creator: Liebetrau, A.M.; Apted, M.J.; Engel, D.W.; Altenhofen, M.K.; Strachan, D.M.; Reid, C.R. et al.
Partner: UNT Libraries Government Documents Department

An evaluation and analysis of three dynamic watershed acidification codes (MAGIC, ETD, and ILWAS)

Description: The US Environmental Protection Agency is currently using the dynamic watershed acidification codes MAGIC, ILWAS, and ETD to assess the potential future impact of the acidic deposition on surface water quality by simulating watershed acid neutralization processes. The reliability of forecasts made with these codes is of considerable concern. The present study evaluates the process formulations (i.e., conceptual and numerical representation of atmospheric, hydrologic geochemical and biogeochemical processes), compares their approaches to calculating acid neutralizing capacity (ANC), and estimates the relative effects (sensitivity) of perturbations in the input data on selected output variables for each code. Input data were drawn from three Adirondack (upstate New York) watersheds: Panther Lake, Clear Pond, and Woods Lake. Code calibration was performed by the developers of the codes. Conclusions focus on summarizing the adequacy of process formulations, differences in ANC simulation among codes and recommendations for further research to increase forecast reliability. 87 refs., 11 figs., 77 tabs.
Date: January 1, 1989
Creator: Jenne, E.A.; Eary, L.E.; Vail, L.W.; Girvin, D.C.; Liebetrau, A.M.; Hibler, L.F. et al.
Partner: UNT Libraries Government Documents Department

An example postclosure risk assessment using the potential Yucca Mountain Site

Description: The risk analysis described in this document was performed for the US Department of Energy`s (DOE) Office of Civilian Radioactive Waste Management (OCRWM) over a 2-year time period ending in June 1988. The objective of Pacific Northwest Laboratory`s (PNL) task was to demonstrate an integrated, though preliminary, modeling approach for estimating the postclosure risk associated with a geologic repository for the disposal of high-level nuclear waste. The modeling study used published characterization data for the proposed candidate site at Yucca Mountain, Nevada, along with existing models and computer codes available at that time. Some of the site data and conceptual models reported in the Site Characterization Plan published in December 1988, however, were not yet available at the time that PNL conducted the modeling studies.
Date: May 1, 1992
Creator: Doctor, P.G.; Eslinger, P.W.; Elwood, D.M.; Engel, D.W.; Freshley, M.D.; Liebetrau, A.M. et al.
Partner: UNT Libraries Government Documents Department