Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “Waste Repository”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 1,009 records · Page 56
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Development of local-power-free, remote α -particle detection using optical fibers

We demonstrate the application of fluorescence optical fiber coupled to a telecom grade fiber as a sensor for alpha particles using alpha-specific ZnS(Ag) scintillation materials whose wavelength is down-shifted into a low-loss region of the telecom grade fiber transmission band. Telecom-grade fiber optics offer a solution for sensing alpha radiation in deep repositories and cask storage for radioactive materials due to the stability of SiO 2 under normal environmental conditions and its relative radiation hardness at low radiation doses. Long-term nuclear waste storage facilities require sensors for the detection of leakage of radioactive materials that are maintenance-free, do not require power and can survive with no ‘wear out’ mechanisms for decades. By accomplishing the wavelength transformation, we maximize efficiencies in the detection of α-particles and signal transport and can detect alpha scintillation at distances on the order of >1 km with a sensor that is ~3% efficient and can be easily scaled as a sensor array. This paper describes the construction and testing of the sensor including manufacture of the controlled thickness films, verification of the wavelength shift from 450 to 620 nm and optimization of the sensitivity as a function of thickness. We also model the relative sensitivity of the film as a function of film thickness, and we demonstrate a signal-to-noise ratio of 10 at a range of greater than 1 km.

alpha particles↗
DOE OSTI2018

Proceedings of the 8th US/German Workshop on Salt Repository Research Design and Operation.

This document records the Proceedings of the 2017 gathering of salt repository nations. In a spirit of mutual support, technical issues are dissected, led capably by subject matter experts. As before, it is not possible to explore all contemporary issues regarding nuclear waste disposal in salt formations. Instead, the group focused on a few selected issues to be pursued in depth, while at the same time acknowledging and recording ancillary issues.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Generic Crystalline Disposal Reference Case

A generic reference case for disposal of spent nuclear fuel and high-level radioactive waste in crystalline rock is outlined. The generic cases are intended to support development of disposal system modeling capability by establishing relevant baseline conditions and parameters. Establishment of a generic reference case requires that the emplacement concept, waste inventory, waste form, waste package, backfill/buffer properties, EBS failure scenarios, host rock properties, and biosphere be specified. The focus in this report is on those elements that are unique to crystalline disposal, especially the geosphere representation. Three emplacement concepts are suggested for further analyses: a waste packages containing 4 PWR assemblies emplaced in boreholes in the floors of tunnels (KBS-3 concept), a 12-assembly waste package emplaced in tunnels, and a 32-assembly dual purpose canister emplaced in tunnels. In addition, three failure scenarios were suggested for future use: a nominal scenario involving corrosion of the waste package in the tunnel emplacement concepts, a manufacturing defect scenario applicable to the KBS-3 concept, and a disruptive glaciation scenario applicable to both emplacement concepts. The computational approaches required to analyze EBS failure and transport processes in a crystalline rock repository are similar to those of argillite/shale, with the most significant difference being that the EBS in a crystalline rock repository will likely experience highly heterogeneous flow rates, which should be represented in the model. The computational approaches required to analyze radionuclide transport in the natural system are very different because of the highly channelized nature of fracture flow. Computational workflows tailored to crystalline rock based on discrete transport pathways extracted from discrete fracture network models are recommended.

Environmental Protection↗
DOE OSTI2013

Exploratory Design of a Reactor/Fuel Cycle Using Spent Nuclear Fuel Without Conventional Reprocessing - 13579

General Atomics has started design of a waste to energy nuclear reactor (EM2) that can use light water reactor (LWR) spent nuclear fuel (SNF). This effort addresses two problems: using an advanced small reactor with long core life to reduce nuclear energy overnight cost and providing a disposal path for LWR SNF. LWR SNF is re-fabricated into new EM2 fuel using a dry voloxidation process modeled on AIROX/ OREOX processes which remove some of the fission products but no heavy metals. By not removing all of the fission products the fuel remains self-protecting. By not separating heavy metals, the process remains proliferation resistant. Implementation of Energy Multiplier Module (EM2) fuel cycle will provide low cost nuclear energy while providing a long term LWR SNF disposition path which is important for LWR waste confidence. With LWR waste confidence recent impacts on reactor licensing, an alternate disposition path is highly relevant. Centered on a reactor operating at 250 MWe, the compact electricity generating system design maximizes site flexibility with truck transport of all system components and available dry cooling features that removes the need to be located near a body of water. A high temperature system using helium coolant, electricity is efficiently produced using an asynchronous high-speed gas turbine while the LWR SNF is converted to fission products. Reactor design features such as vented fuel and silicon carbide cladding support reactor operation for decades between refueling, with improved fuel utilization. Beyond the reactor, the fuel cycle is designed so that subsequent generations of EM2 reactor fuel will use the previous EM2 discharge, providing its own waste confidence plus eliminating the need for enrichment after the first generation. Additional LWR SNF is added at each re-fabrication to replace the removed fission products. The fuel cycle uses a dry voloxidation process for both the initial LWR SNF re-fabrication and later for EM2 discharge reuse. The EM2 waste disposal profile is effectively only fission products, which reduces the mass (about 3% vs LWR), average half life, heat and long term radio-toxicity of the disposal. Widespread implementation of EM2 fuel cycle is highly significant as it would increase world energy reserves; the remaining energy in U.S. LWR SNF alone exceeds that in the U.S. natural gas reserves. Unlike many LWR SNF disposition concepts, the EM2 fuel cycle conversion of SNF produces energy and associated revenue such that the overall project is cost effective. By providing conversion of SNF to fission products the fuel cycle is closed and a non-repository LWR SNF disposition path is created and overall repository requirements are significantly reduced. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

A study of transfer of UNF from non-disposable canisters - 15388

A study was performed by AREVA and URS for the Department of Energy (DOE) [1] to establish optimal methods for the opening of dry storage canisters (DSC) and the subsequent repackaging of used nuclear fuel (UNF) from the DSCs into generic standard transportation, aging, and disposal canisters (STADs) in wet or dry environments at reactor, consolidated storage, or repository sites. Eight dry and wet cutting technologies were evaluated: (1) skiving; (2) milling; (3) arc gouging/plasma torch; (4) cutting wheel; (5) diamond cable cutting; (6) laser cutting; (7) sawing; and (8) electrical discharge machining. The evaluated UNF repackaging technologies options included: (1) wet transfer in a pool; (2) dry transfer in a dry transfer system (DTS); (3) dry transfer in a fixed hot cell (FHC); (4) dry transfer in a mobile hot cell (MHC), and (5) a hybrid approach utilizing wet transfer in a pool and dry transfer in a FHC. A systems engineering evaluation was utilized for ranking the eight viable cutting options and the five different repackaging approaches for each of the sites. The systems engineering evaluation took into account six criteria (i.e., life cycle costs, worker dose, operational efficiency, technical maturity, ease of licensing, and waste generation) and performed a qualitative pairwise ranking of their relative importance to one another. Then quantitative or qualitative factors associated with each criterion (e.g., cost in millions of dollars, relative ease of licensing, etc.) for each option at each location (reactor, consolidated storage facility [CSF], repository) were developed based on historical experience, existing design studies, current 'state-of-the art', and/or engineering judgment. Using this input, the systems engineering evaluation identified dry skiving as the preferred cutting technology, with the wheel cutting technology in either a dry or wet environment identified as a promising technology that, with some RD and D, could mature to a better option (depending also on the disposition of the used DSC). For a reactor and a CSF/repository site, the systems engineering evaluation ranked the repackaging option using the skiving dry cutting technique and wet transfer of UNF in a pool as the highest option. However for a CSF and a repository site, the team recommended the hybrid option using the skiving dry cutting technique and both dry and wet transfer of UNF in a pool and FHC as a potentially better option (providing specific benefits such as higher throughput and more flexibility in operation). (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Hydrologic Nuclide Transport Models in Cyder, A Geologic Disposal Software Library - 13328

Component level and system level abstraction of detailed computational geologic repository models have resulted in four rapid computational models of hydrologic radionuclide transport at varying levels of detail. Those models are described, as is their implementation in Cyder, a software library of interchangeable radionuclide transport models appropriate for representing natural and engineered barrier components of generic geology repository concepts. A proof of principle demonstration was also conducted in which these models were used to represent the natural and engineered barrier components of a repository concept in a reducing, homogenous, generic geology. This base case demonstrates integration of the Cyder open source library with the Cyclus computational fuel cycle systems analysis platform to facilitate calculation of repository performance metrics with respect to fuel cycle choices. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Spent Nuclear Fuel Storage and Radioactive Waste Disposal in the United States: A Law and Policy Analysis - 20302

While Congress plays political football, spent nuclear fuel continues to sit in de facto interim nuclear waste storage sites throughout the country. This ad hoc approach - if it can even be called an approach - to nuclear waste management is in no one's best interests, because of financial, safety, environmental, and national security concerns with the status quo. Senate leadership had asserted that an up-or-down vote on the Yucca Mountain Project would take place in September or October 2019, but that did not occur. Rather, a conference bill was agreed to, and it provided that, although the nuclear program should receive some funding, funding was not made available for centralized interim storage ('CIS') or the Yucca Mountain Project. The President's FY2021 Budget Request does not contain any proposed funding for the Yucca Mountain Project ('the Project'). This paper will discuss the implications of that process, both for the Project itself and for the U.S. nuclear waste program more generally. In order to provide context for the current situation, this paper will begin by chronicling the history of U.S. nuclear waste management, from the passage of the Nuclear Waste Policy Act of 1982 and to the present day. It will discuss the roles of the U.S. Department of Energy, the Nuclear Regulatory Commission, federal and state political delegations, and the nuclear energy industry. The paper will then identify key themes that can be expected to affect the future of nuclear waste management in the United States. For example, it will consider whether development of additional CIS facilities is a viable short- to medium-term solution. CIS facilities may relieve immediate pressure at decommissioned or soon-to-be-decommissioned nuclear power plants; but with that pressure partially alleviated, support for building a long-term repository may be reduced. It will also discuss opposition by the State of New Mexico and others to the proposed New Mexico CIS project, based not only on siting concerns but also whether it might become a de facto repository, and whether the State of Texas and others may also be opposed, for similar reasons, to the proposed CIS facility in Texas. Additionally, the paper will examine consent-based siting and its implications for Nevada and other states that are potential hosts for spent nuclear fuel storage and disposal facilities. Further, if the Project is not a politically viable solution, the paper will examine whether it is now time for serious consideration of an alternate location for a permanent repository, as well as consideration of other options entirely, such as deep geologic disposal. The paper will conclude by looking ahead to the 2020 elections. It will describe the positions of candidates (particularly Presidential candidates) on spent nuclear fuel storage and disposal options, without advocating for any particular candidate or political party. Perhaps in 2020, we will at least see a start to a solution to the seemingly intractable problem of nuclear waste management in the United States. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Salt International Collaborations FY23 Update

This report summarizes the international collaborations conducted by Sandia funded by the US Department of Energy Office (DOE) of Nuclear Energy (DOE-NE) Spent Fuel and Waste Science & Technology (SFWST) as part of the Sandia National Laboratories Salt R&D and Salt International work packages. This report satisfies the level-three milestone M3SF-23SN010303062. Several stand-alone sections make up this summary report, each completed by the participants. The sections discuss granular salt reconsolidation (KOMPASS), engineered barriers (RANGERS), numerical model comparison (DECOVALEX) and an NEA Salt Club working group on the development of scenarios as part of the performance assessment development process. Finally, we summarize events related to the US/German Workshop on Repository Research, Design and Operations.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Impact of Zr Content on Radionuclide Release Rates in Metal Waste Forms

Reference alloy waste forms (RAW) underwent hybrid corrosion/immersion testing to parameterize the ANL analytical oxidative-dissolution model to enable the calculation of fractional release rates. The waste forms tested at SRNL were variations of the RAW-3 formulation that uses HT9 as the main alloy component, and are meant to enable evaluation of the impact of Zr:U ratio on the release rates of actinides and Tc-99. The test solutions were air saturated alkaline and acidic brines, representing potential repositories with those conditions. The testing approach consisted of 4 major steps; 1) bare surface corrosion measurements, 2) hybrid potentiostatic hold/exposure measurements, 3) measurement of radionuclide concentrations and relations to anodic current from potentiostatic holds, and 4) identification of corroding phases using SEM/EDS of electrodes.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Experimental determination of solubilities of magnesium borates: Solubility constants of boracite [Mg 3 B 7 O 13 Cl(cr)] and aksaite [MgB 6 O 7 (OH) 6 ·2H 2 O(cr)]

In this work, solubility measurements regarding boracite [Mg 3 B 7 O 13 Cl(cr)] and aksaite [MgB 6 O 7 (OH) 6 ·2H 2 O(cr)] from the direction of supersaturation were conducted at 22.5 ± 0.5 °C. The equilibrium constant (log 10 K 0 ) for boracite in terms of the following reaction, Mg 3 B 7 O 13 Cl(cr) + 15H 2 O(l) ⇌ 3Mg 2+ + 7B(OH) 4 – + Cl – + 2H + is determined as -29.49 ± 0.39 (2σ) in this study. The equilibrium constant for aksaite according to the following reaction, MgB 6 O 7 (OH) 6 •2H 2 O(cr) + 9H 2 O(l) ⇌ Mg 2+ + 6B(OH) 4 – + 4H + is determined as -44.41 ± 0.41 (2σ) in this work. This work recommends a set of thermodynamic properties for aksaite at 25 °C and 1 bar as follows: ΔH$0\atop{f}$ =-6063.70 ± 4.85 kJ·mol -1 , ΔG =-5492.55 ± 2.32 kJ·mol -1 , and S 0 = 344.62 ± 1.85 J·mol -1 ·K -1 . Among them, ΔG$0\atop{f}$ is derived from the equilibrium constant for aksaite determined by this study; ΔH$0\atop{f}$ is from the literature, determined by calorimetry; and S 0 is computed in the present work from ΔG$0\atop{f}$ and ΔH$0\atop{f}$. This investigation also recommends a set of thermodynamic properties for boracite at 25 °C and 1 bar as follows: ΔH$0\atop{f}$ =-6575.02 ± 2.25 kJ·mol -1 , ΔG$0\atop{f}$ =-6178.35 ± 2.25 kJ·mol -1 , and S 0 = 253.6 ± 0.5 J·mol -1 ·K -1 . Among them, ΔG$0\atop{f}$ is derived from the equilibrium constant for boracite determined by this study; S 0 is from the literature, determined by calorimetry; and ΔH$0\atop{f}$ is computed in this work from ΔG$0\atop{f}$ and S 0 . The thermodynamic properties determined in this study can find applications in many fields. For instance, in the field of material science, boracite has many useful properties including ferroelectric and ferroelastic properties. The equilibrium constant of boracite determined in this work will provide guidance for economic synthesis of boracite in an aqueous medium. Similarly, in the field of nuclear waste management, iodide boracite [Mg 3 B 7 O 13 I(cr)] is proposed as a waste form for radioactive 129 I. Therefore, the solubility constant for chloride boracite [Mg 3 B 7 O 13 Cl(cr)] will provide the guidance for the performance of iodide boracite in geological repositories. Boracite/aksaite themselves in geological repositories in salt formations may be solubility-controlling phase(s) for borate. Finally, solubility constants of boracite and aksaite will enable researchers to predict borate concentrations in equilibrium with boracite/aksaite in salt formations.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2018

Sorption of Cationic Radionuclides on Chlorite and Sericite formed by the Hydrothermal Alteration - 18180

Chlorite and sericite formed by the hydrothermal alteration of granite distribute predominantly around the geological repository in deep underground. These minerals often cover the surface of a solid phase and fill the fracture of granite considered as the natural barrier to the migration of radionuclides leaching from radioactive waste. Therefore, to assess the practical performance of the natural barrier, it is necessary to clarify the interaction between radionuclides and these minerals. This study examined the sorption of cationic radionuclides by chlorite and sericite through batch experiments, whereby Cs{sup +}, Sr{sup 2+}, and Eu{sup 3+} were used as tracer ions. In these experiments, powder-form chlorite or sericite was mixed with the solutions of the tracer ions to a liquid/solid ratio of 10 ml/g, and the resulting mixture was allowed to incubate for 2-7 days. The concentration of tracer ions was set to 1 mM and that of NaCl to a specific value in the 0-0.6 M range. The solution pH was 3, 5, or 8. As a result, tracer ions were significantly adsorbed by chlorite and sericite, regardless of pH or NaCl concentration, except when NaCl concentration reached the high value of 0.6 M. The distribution coefficient for europium ions at pH 5 was estimated to be 2-7 mL/g in chlorite and 10{sup 3}-10{sup 4} mL/g in sericite. These values significantly exceeded that measured for feldspar (2-4 mL/g), the predominant component of granite before it undergoes any hydrothermal alteration. These results suggest that cationic radionuclides will be immobilized through sorption in chlorite and sericite filling granite fractures, even if these thermal-fluid-formed minerals cover the surface of granite. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

A Method for Understanding the Behaviour of Radioactive Decay Chains in Radiological Assessments - 19382

In many radiological groundwater pathway assessments, especially those where the assessment timescales of interest are relatively short (say a few thousand years or less), contributions to effective dose or radiological risk are dominated by radionuclides such as C-14, Cl-36 or I-129. Over longer timescales, additional radionuclides can also become key contributors to risk. In particular, members of the U-238 decay chain, notably Ra-226 and Pb-210, are likely to be key radionuclides. Other decay chains that can contribute significantly to dose or risk at times far into the future include Np-237 → U-233 → Th-229 and U-235 → Pa-231 → Ac-227. In this paper, a method is described for understanding decay chain behaviour in radiological assessments. The method uses a graphical approach in which the various modes of transport and decay in a system of connected compartments are represented as pathways between two nodes in a two-dimensional grid. Each pathway corresponds to a distinct transport and decay process. By so doing, the different transport and decay processes can be identified in a systematic way. The result is an algorithm for generating the contributions to effective dose or radiological risk without explicit solution of the transport equations. The method has been tested against numerical models and analytic solutions and found to give excellent agreement, while at the same time providing information about decay chain behaviour that would be difficult to obtain from a numerical assessment model. The power of the method is illustrated by applying it to some examples based on real repository concepts. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2015

The fourth decade of underground research at the Grimsel Test Site - what we have learned and where we go from here - 15075

At the Grimsel Test Site, NAGRA and its partners have celebrated thirty years' worth of essential contributions to the characterization of crystalline host rock, and the development and confirmation of safe geological disposal concepts. In operation since 1984, the underground research laboratory in the Swiss Alps shall not accept radioactive waste but can host experiments with radiotracers/radionuclides under strictly controlled conditions. The strong drive for international cooperation at the Grimsel Test Site is reflected in the more than two dozen organizations and research institutes from twelve countries and the European Union which have participated in the research program to date. The program has been designed dynamically to always mirror the current requirements of national programs for the long-term management of radioactive waste, as well as attempt to anticipate the next steps. Early work featured the development of methodologies for formation characterization. Then followed a period concentrating on performance assessment issues, before radionuclide migration and barriers were investigated intensely. The past 15 years have been dominated by engineering demonstration experiments. The research conducted at the Grimsel Test Site has complemented laboratory experiments, surface investigations and numerical modelling studies through the bridging of different scales and drawing off links between laboratory setups and natural analogues. The crystalline rock formations of the Aar Massif have provided an opportunity to do experiments in environments similar to those of future repositories. The facility also serves as a platform for continual knowledge transfer between generations of scientists and fosters synergies among disciplines. However, considering the ambitious schedule for repository implementation in several continuous, one may ask whether the era of generic underground rock laboratories is approaching its end. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Influence of Cracks in Cementitious Engineered Barriers in a Near-Surface Disposal System: Assessment Analysis of the Belgian Case

In large cement-based structures such as a near surface disposal facility for radioactive waste voids and cracks are inevitable. However, the pattern and nature of cracks are very difficult to predict reliably. Cracks facilitate preferential water flow through the facility because their saturated hydraulic conductivity is generally higher than the conductivity of the cementitious matrix. Moreover, sorption within the crack is expected to be lower than in the matrix and hence cracks in engineered barriers can act as a bypass for radionuclides. Consequently, understanding the effects of crack characteristics on contaminant fluxes from the facility is of utmost importance in a safety assessment. In this paper we numerically studied radionuclide leaching from a crack-containing cementitious containment system. First, the effect of cracks on radionuclide fluxes is assessed for a single repository component which contains a radionuclide source (i.e. conditioned radwaste). These analyses reveal the influence of cracks on radionuclide release from the source. The second set of calculations deals with the safety assessment results for the planned near-surface disposal facility for low-level radioactive waste in Dessel (Belgium); our focus is on the analysis of total system behaviour in regards to release of radionuclide fluxes from the facility. Simulation results are interpreted through a complementary safety indicator (radiotoxicity flux). We discuss the possible consequences from different scenarios of cracks and voids. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Salt International Collaborations FY24 Update

This report summarizes the international collaborations conducted by Sandia funded by the US Department of Energy Office (DOE) of Nuclear Energy (DOE-NE) Spent Fuel and Waste Science & Technology (SFWST) as part of the Sandia National Laboratories Salt R&D and Salt International work packages. This report satisfies the level-three milestone M3SF-24SN010303063. Several stand-alone sections make up this summary report, each completed by the participants. The sections discuss granular salt reconsolidation (KOMPASS/MEASURES), engineered barriers (RANGERS), numerical model comparison (DECOVALEX), an NEA Salt Club working group on the development of scenarios as part of the performance assessment development process, and progress on seal percolation experiments with German colleagues at Gesellschaft für Anlagen- und Reaktorsicherheit (GRS). Finally, we summarize events related to the US/German Workshop on Repository Research, Design and Operations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Impact of lump characteristics on self-shielding corrections of active neutron interrogation on fissile material in waste - 15405

Non-destructive active neutron assay techniques such as active neutron coincidence counting, differential die away analysis or the californium shuffler method rely on the interrogation of fissile nuclear materials by an external neutron source. Whether it is a spontaneous fission or (α,n) neutron source, like {sup 252}Cf or AmLi; or a fusion neutron generator, all of these techniques rely on induced secondary fission events to provide prompt or delayed neutron signals that serve as the basis for the fissile material quantification. The accuracy of these fissile mass measurements is therefore dependent on the neutron interrogation efficiencies determined for calibration reference standards and whether the measured items are comparable to the calibration condition. In such favorable conditions, it is possible to establish a reliable empirical non-linear relationship between apparent (uncorrected) and real or true fissile mass. However, in waste measurements, this is unlikely to be the case; thus a calibration approach that takes account of the neutron self-shielding effect is required. By incorporating the Self-Shielding Factor (SSF) of the calibration reference standards, a linear mass calibration factor [s{sup -1}. {sup 239}Pu{sub eq} g{sup -1}] can be found; this calibration is accurate for waste in which the fissile material is 'infinitely dilute' i.e. the waste exhibits no self-shielding; this is assumed to be representative of dispersed material. The self-shielding correction and its systematic uncertainty are then left to be treated as part of the Total Measurement Uncertainty (TMU) evaluation associated with the assay result. The derivation of the TMU contribution associated with the self-shielding effect in nuclear waste is a complex problem for which a reliable evaluation has to be performed. This is usually based on assuming 'worst case scenarios' for the distribution of material that can arise in a particular waste stream. Such an approach is conservative, but often very penalizing, especially if one considers larger lumps (those containing grams of fissile material); for such lumps, the self-shielding correction becomes very significant, potentially altering the mass of {sup 239}Pu{sub eq} by over an order of magnitude. A worst case 'nuclear safety value' may not be the most appropriate choice for building a consignment best estimate or repository sentencing value for an item. In the present work the authors describe a numerical approach to study the variability of neutron self-shielding factors associated with distributions of a wide range of lump characteristics in terms of size, shape, chemical composition, density and isotopic composition. The described novel calculation tool is capable of accurately predicting the real impact of random distribution of lumps in a waste material, and allowed to show that given reasonable, defensible, assumptions on the composition of the lumps in waste stream, the systematic error component of the DDA TMU can be significantly reduced. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Criticality Consequence Analysis Roadmap for Spent Nuclear Fuel Canister in a Repository

This report discusses an initial criticality consequence analysis approach of commercial spent fuel canisters in a repository that will be developed to support a performance assessment (PA) of a repository with criticality events. A repository PA with criticality events may be needed to support direct disposal of currently loaded dual-purpose canisters. The ultimate goal is to develop a criticality consequence analysis model that can interface with any repository PA model. The criticality consequence analysis approach describes multi-physics coupling between neutronics, thermal hydraulics, and mechanics analysis codes. This report also discusses the sensitivity and uncertainty analysis approach for quantifying conservatism in the criticality consequence outcomes, as well as a defensible multi-physics validation approach to support any future licensing effort. This criticality consequence analysis approach will be refined in the future to support specific regulatory frameworks for a specific repository PA if needed.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗