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At least 181 records · Page 10
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Performance Assessment of a Generic Repository for Defense-Related HLW/SNF in Fractured Crystalline Host Rock - 17059

Recently the U.S. made a policy decision to begin R and D for geologic disposal of defense-related HLW/SNF in a facility separate from commercially generated waste. A deep geologic repository for the disposal of DOE-managed HLW, and some thermally cooler DOE-managed SNF, arising from defense and DOE R and D activities, is part of a stepwise, phased approach for disposal of the Nation's nuclear waste, as recommended by the Blue Ribbon Commission on America's Nuclear Future. The work discussed here focuses on post-closure safety (or performance) assessment for such a Defense Waste Repository (DWR), which is divided into four major activities: (1) development of generic reference cases (i.e., knowledge or technical bases for 'generic' or 'non-site-specific' deep geologic repositories); (2) features, events, and processes (FEPs) analyses and screening to support the technical bases and the performance assessment (PA) model; (3) performance evaluation of alternative EBS design concepts; and (4) post-closure safety analyses of the repository system under consideration. Using the known inventory of defense-related SNF, as well as defense-related HLW stored at the Savannah River and Hanford sites, the Geologic Disposal Safety Assessment (GDSA) Framework modeling and software system has been applied to simulate the potential performance of a DWR in crystalline host rock, resulting in a suite of single-realization (i.e., deterministic) and multi-realization (i.e., probabilistic) 3-D post-closure system analyses, over a performance period of one million years. Two types of emplacement concepts are examined, including single-canister vertical-borehole emplacement for the hotter defense SNF waste (KBS-3V concept) and multi-canister horizontal emplacement for defense HLW (similar to Yucca Mountain co-disposal waste packages). Sensitivity analyses examine the effect of key uncertain parameters on repository performance, including the effects of fracture distribution, waste package degradation rate, buffer and disturbed rock zone (DRZ) properties, and sorption parameters. Initial results indicate that a crystalline host rock with a connected fracture system may require additional safety features to ensure robustness of the isolation safety function, such as a deep unsaturated zone, a sufficiently thick sedimentary overburden, and/or a disposal overpack with a very slow corrosion rate. None of these should pose an undue obstacle for successful disposal. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Probabilistic Performance Assessment for Deep Borehole Disposal of Cs/Sr Capsules - 17143

The US Department of Energy (DOE) performed an assessment of disposal options for spent nuclear fuel (SNF) and high-level radioactive waste (HLW) that recommended the consideration of deep borehole disposal (DBD) of smaller DOE-managed waste forms, such as cesium (Cs) and strontium (Sr) capsules. To further assess the safety and viability of the DBD concept, post-closure performance assessment (PA) analyses were performed for DBD of Cs/Sr capsules. The post-closure PA included a reference design concept for Cs/Sr capsule disposal, feature, event, and process (FEP) analysis, scenario development, and modeling using the PFLOTRAN code. PA simulations were run for both nominal and disturbed scenarios for long-term radionuclide transport away from the deep disposal borehole. The nominal (expected) post-closure release scenario included FEPs for short-duration (a few hundred years) thermally-induced upward fluid flux through the borehole seals and/or disturbed rock zone (DRZ) followed by longer-term slow diffusive transport. Simulations included a baseline deterministic run and a set of probabilistic realizations (using Latin Hypercube Sampling (LHS) from parameter distributions) to examine the sensitivity and importance of the long-term radionuclide transport to processes and parameters such as: waste package durability; seal porosity and permeability, disturbed rock zone (DRZ) porosity and permeability; and radionuclide sorption in the seal materials, DRZ, and crystalline host rock. The disturbed scenario included a waste package 'stuck' in the crystalline basement above the emplacement zone near a hypothetical borehole-intersecting fracture. Simulations included two deterministic runs to examine the sensitivity to the regional head gradient. The nominal and disturbed scenario PA results suggest that a favorable safety case can be developed for DBD of Cs/Sr capsules and helped to identify key areas of uncertainty, which in turn can inform future DBD research and development (R and D) activities, including a field demonstration project. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Decision and Modeling Strategy for a Radiological Risk Assessment at Area G, Los Alamos, New Mexico - 17258

The U.S. Department of Energy (DOE) owns a low-level radioactive waste (LLW) disposal site at Material Disposal Area G, in Los Alamos, New Mexico, USA. Area G has been the primary LLW disposal site for Los Alamos National Laboratory (LANL) since the 1960's, and is now approaching the end of its service life. In addition to LLW, Area G is host to a variety of other wastes, the disposition of which must be determined before closure of the site. A probabilistic Radiological Risk Assessment (RRA) for Area G is used in order to support decision making regarding some wastes that are not addressed in the extant Area G Performance Assessment (PA) and Composite Analysis (CA). Between 1979 and 1987, 33 special shafts were augered into the Bandelier Tuff at Area G. This tuff is present across Pajarito Plateau on the eastern slopes of the Jemez Mountains, and varies widely in its consistency, from weakly indurated non-welded layers to welded layers that uphold the mesa cliffs of the Plateau. These mesas are home to LANL, Area G, and the town of Los Alamos. The 33 Shafts were lined with steel casing, and contain remote-handled (RH) transuranic wastes (TRU) resulting from experiments and analysis performed in special glove boxes at the Chemistry and Metallurgy Research (CMR) facility at LANL. Some of these wastes originated as used ('spent') nuclear fuel. The purpose of the Area G RRA is to evaluate the potential future risk to humans and the environment from the RH TRU in the 33 Shafts in the context of the risk associated with the surrounding wastes at Area G. The analysis is responsive to expectations outlined in DOE Order 458.1, Radiation Protection of the Public and the Environment, and is informed by the Manual and Guidance accompanying DOE O 435.1, Radioactive Waste Management. Because the waste meets the definition of TRU, the regulatory context also takes into consideration the regulation governing the disposal of TRU from the U.S. Environmental Protection Agency (EPA): 40 CFR 191, Environmental Radiation Protection Standards for Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes. Given the broader regulatory context for the RRA, the analysis is subject to different assumptions from those made in the existing DOE O 435.1 PA and CA, such as allowing for future occupation of the site. The analysis begins with a comprehensive evaluation of features, events, processes, and exposure scenarios (FEPS) for Area G and the wastes it contains. These FEPS are screened to eliminate from further consideration those of extremely low probability and/or consequence, and a conceptual site model (CSM) is subsequently developed. The scope and structure of the Area G RRA Model is informed by this CSM, and the Area G RRA Model is developed using the GoldSim systems analysis modeling platform. This paper outlines the development of a defensible, transparent, and reasonably realistic model, which is based on the state of knowledge of the wastes, the site, and the FEPS that govern contaminant transport and human exposures. Probabilistic model input distributions represent uncertainties inherent in the real and modeled systems. The results of the Area G RRA Model will inform decision analysis regarding the disposition of the RH TRU in the 33 Shafts. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Statistical Methods for Effective Spatial and Temporal Scaling in Support of Probabilistic Performance Assessments - 17340

Performance assessments (PAs) for the disposal of low-level radioactive waste have traditionally been performed in a deterministic manner, with inputs specified as single values as if they carry no uncertainty. Modern PAs, however, explicitly incorporate uncertainty through use of probability distributions for inputs instead of single values. While this is clearly an improvement, challenges remain in specifying appropriate distributions, many of which relate to capturing the appropriate spatial and/or temporal scale given the PA model, and matching the scale(s) of data and information to that specified by the model. More careful attention to scaling will lead to more defensible characterizations of physical variability and uncertainty in model inputs, thus leading to more defensible information regarding outcomes of interest to be used in decision making, and hence more defensible decisions. This work introduces general strategies and considerations for explicitly incorporating considerations of scale into distribution development. The strategies are explained as a series of steps and then presented with an example that starts to illustrate the new statistical concepts. This work frames the problem in terms of hierarchical statistical modeling, providing important connections to current statistical approaches to formally and rigorously account for temporal and spatial scaling. The larger goal is to build a more formal and unified approach to distribution development to help teams of modelers, subject matter experts, and statisticians work together to develop distributions and justify the choices made in the process, ultimately leading to more informed decision-making. Without this rigor, statistical distributions are not specified correctly and can have unexpected consequences for modeling and sensitivity analysis. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Final Closure of the Maxey Flats Low-Level Radioactive Waste Disposal Site - 16088

This paper describes the final closure remedy for the former Maxey Flats Low-Level Radioactive Waste Disposal Site (MFDS), the challenges presented at the site due to the condition of the disposal facility and waste, and their resolution through comprehensive geo-synthetic design. The MFDS is an inactive landfill located in Kentucky on a ridge surrounded by a buffer zone. Historically, waste disposal occurred without reducing voids or without compacting the placed waste. Settlement and water management issues led to releases of radioactive contamination into surrounding waters, resulting in the closure of the facility in 1977. The final closure remedy consists of 1) development and implementation oversight of a sump abandonment design to ensure no direct pathways for infiltrating water to contact the waste and 2) development and implementation oversight of the Final Closure Period (FCP) cap design. Challenges successfully addressed by the FCP scope included: Abandon sumps to minimize storm water migrating into the waste; Address potential existing groundwater waste interaction; Maximum use of on-site cap construction materials; Design cap to fit very restricted site footprint; Assess and address potential isolated differential settlement; Maximize passive storm water management; Provide alternative site access during construction. The FCP remedy and resultant design satisfies these challenges through use of layers of geo-grid-reinforced soil and geo-synthetic materials, contours and slopes that minimize the quantities of construction materials while maximizing the use of available material, probabilistic subsidence modeling, an integrated storm water management system ensuring the slow release of storm water into receiving streams, and a dedicated construction haul road system. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Material Properties and Temperature Effects Quantified for Slurry Wear - 16141

Literature indicates that the hardness of solid particles relative to the target material hardness and temperature influence erosion. Investigations of temperature effects on erosion also show that erosion resistance is a function of target material's homologous temperature to melting point temperature. While relative hardness effects on erosion and temperature effects on erosion have been investigated independently, their combined effect has not been evaluated. This paper summarizes a review of relevant literature on relative hardness and temperature effects on erosion and via probabilistic calculations, quantifies the combined effect of relative hardness and temperature on liquid-particle induced wear, or slurry wear. These insights can assist material selection and plant operations at waste processing facilities. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Enhanced Performance Assessment Models for Generic Deep Geologic Repositories for High-Level Waste and Spent Nuclear Fuel - 16223

Development of an enhanced performance assessment (PA) capability for geologic disposal of spent nuclear fuel and high-level waste has been ongoing for several years in the U.S. repository program. The new Generic Disposal System Analysis (GDSA) modeling and software framework is intended to be flexible enough to evolve through the various phases of repository activities, beginning with generic PA activities in the current Concept Evaluation phase to site-specific PA modeling in the Repository Development phase. The GDSA Framework utilizes modern software and hardware capabilities by being based on open-source software architecture and being configured to run in a massively parallel, high-performance computing (HPC) environment. It consists of two main components, the open-source Dakota uncertainty sampling and analysis software and the PFLOTRAN reactive multi-phase flow and transport simulator. Reference cases or 'generic repositories' have been, and are being developed, based on typical properties for potential salt, clay, and granite host-rock formations and corresponding engineered design concepts for each medium. Past simulations have focused on a generic repository in bedded-salt host rock, while the most recent research has focused on a reference case for a typical clay/shale host rock. A variety of single-realization (i.e., deterministic) and multi-realization (probabilistic) results for the new clay reference case are presented, including an analysis of the effects of heat generation on repository performance, assuming a 100-year out-of-reactor commercial SNF waste form. Order-of-magnitude differences between predicted radionuclide concentrations in thermal versus isothermal simulations imply that mechanistic, coupled-process modeling in three-dimensional (3-D) domains can be important for building confidence in post-closure performance assessments. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

ZEM: Integrated Framework for Real-Time Data and Model Analyses for Robust Environmental Management Decision Making - 16438

Typically environmental management (EM) problems require analysis of large and complex data sets originating from concurrent data streams with different data collection frequencies and pedigree. These big data sets require on-the-fly analysis, assimilation and integration into a series of models with different complexity for various types of model-based analyses. This is needed to provide fast iterative model analyses based on the latest available data to guide decision-making. Furthermore, the data and model are associated with uncertainties. The uncertainties are probabilistic (e.g. measurement errors) and non-probabilistic (unknowns, e.g. alternative conceptual models characterizing site conditions). To address all of these issues, we have developed an integrated framework for real-time data and model analyses for environmental decision-making called ZEM. The framework allows for seamless and on-the-fly integration of data and modeling results for robust and scientifically-defensible decision-making applying advanced decision analyses tools such as Bayesian-Information-Gap Decision Theory (BIG-DT). The framework also includes advanced methods for optimization that are capable of dealing with a large number of unknown model parameters, and surrogate (reduced order) modeling capabilities based on support vector regression techniques. The framework is predominantly coded in Julia, a state-of-the-art high-performance programing language (http://julialang.org) and incorporates the MADS computational tools (Model Analysis and Decision Support; http://mads.lanl.gov). We demonstrate the application of ZEM for data and model analyses at the LANL Chromium site. The ZEM framework will be released as open-source code and can be applied to any environmental management site. (authors)

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

Long-term Prediction of Spent Fuel Storage Canisters - Issues and Approaches - 18112

The considerations involved in aging management of stainless steel canisters used in dry cask storage of spent fuel are discussed in this paper in a probabilistic framework. The two important modes of degradation in the long-term performance of the casks are localized corrosion and stress corrosion cracking initiating from the external surface. These degradation modes are determined by the evolution of the electrochemical behavior and the strain rate of the stainless steel. Laboratory data suggest that Type 304L/316L stainless steels should suffer from localized corrosion and SCC under salt deposits. However, field inspection of the canisters has not indicated extensive pitting or stress corrosion cracking. This discrepancy can be understood in terms of the joint probabilities of various factors that must come together to cause these failure modes. A Bayesian network approach is suggested as the framework for modeling the probability of failure of canisters and long-term management of aging. The gaps in knowledge to conduct a Bayesian network analysis are discussed. (authors)

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

Self-Navigating Robotic Assistant for Long-term Wide Area Floor Contamination Monitoring - 18213

For Routine Contamination Testing (RCT) of an operational nuclear facility or one undergoing Decontamination and Decommissioning (D and D), workers utilize specified travel corridors (enclosed hallways or areas designated with tape or rope in open spaces) to minimize the risk of spreading contamination. Radiation health technicians must continually monitor these areas for contamination to minimize worker dose and prevent its spread to uncontrolled areas. For many sites, it is impractical for the technician to adequately survey such areas. The Nuclear Robotics Group (NRG) at UT Austin has developed a self-navigating robotic platform equipped with radiation detectors to augment the technician to improve both survey coverage and frequency. While maintaining a controlled velocity and floor-to-sensor distance, the system takes radiation measurements to identify and/or localize contamination sources. Readings above background levels are reported to a remote workstation to allow technicians to then respond accordingly. The completion of these mundane tasks by a robot allows the trained technicians to focus instead on operational tasks, D and D, or responding to alarms. The robot generates a map using a Lidar range sensor with established Simultaneous Localization and Mapping (SLAM) techniques. Range data is then compared to previously generated maps for robot localization and autonomous navigation. To maximize coverage, the robot executes a Complete Coverage Path Planning (CCPP) algorithm using a static map that is discretized into cells to assure comprehensive yet efficient coverage. As the system executes the generated plan, it builds a radiation map of the area for the technician to validate and verify the results. If contamination is detected, the robot will minimize the spread of contamination by stopping, sounding an alarm, and alerting the remote technician. The proposed system allows for long-term deployment with minimal human intervention. It is robust to both highly dynamic obstacles (i.e. personnel) and low dynamic changes (i.e. temporally stored cart) in the environment. This is accomplished through continuous spatiotemporal updating of the static map, and the utilization of a supervisory controller that allows the system to reliably respond to unexpected events or states (i.e. low battery, closed door, etc.). While the supervisory controller can respond to temporary complications, the system also recognizes the environment can change slowly over time necessitating that the CCPP automatically be updated to account for observed 'permanent' changes to the environment. To do this, the robot uses intermediate probabilistic data to make intelligent planning decisions offline while still maintaining local reactive obstacle avoidance and safety features. The self-updating of the map allows the robot to run longer without the need to remake the map, leading to more consistent monitoring data and less maintenance from the human technician. This paper will review the current system hardware, the CCPP algorithm, the supervisory controller, the probabilistic algorithm for identifying permanent changes, and recent experiments at UT Austin to experimentally validate the reported sensor coverage. (authors)

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

Subsidence Risk Mitigation in LLW Final Covers: A Case History - 18286

Typical waste packaging and disposal practices at low-level radioactive waste (LLW) disposal facilities, while adequate for safe transportation and facility operations, introduce significant subsidence risk for the facilities' final cover systems. Subsidence is often caused by the collapse of voids in the waste packaging and surrounding cover soils that eventually propagate to the ground surface. This subsidence could result in the loss of final cover system integrity, stormwater ponding, concentrated flow of surface water and/or gas, increased leachate generation, transport of contaminants from the landfill, and potential adverse chemical reactions within the waste mass. For new LLW facilities, steps can be taken to minimize the subsidence potential of newly placed waste. However, there is a significant inventory of existing disposal facilities (trenches, pits, shafts, vaults, etc.), tank farms, and historic, subsurface burial grounds that require final covers - all of which present real subsidence risks. Final covers for these existing LLW disposal sites can be engineered to mitigate these risks. A case history of a 60-acre LLW final cover system design and construction project illustrates an innovative approach to this problem that offers superior risk mitigation compared to typical final cover design practice. This approach combines engineering simulations of the projected final cover mechanical performance, specific probabilistic performance targets, and targeted cover system improvements to manage the subsidence risk in the final design. The result is a successful final cover project that meets the needs of the community, state and federal regulators, and facility owner. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Post-Fukushima Safety Enhancement Activities in Canada

In the immediate aftermath of the common-cause seismic and tsunami events that triggered the Fukushima Daiichi nuclear accident on March 11, 2011, the nuclear regulator, the Canadian Nuclear Safety Commission (CNSC), requested that all Cass I nuclear facilities re-examine the safety cases for their nuclear facilities. Lessons-learned task forces were established by the CNSC, and the nuclear utilities, Ontario Power Generation (OPG), Bruce Power (BP), and New Brunswick Power (NBP). The purposes of these independent task forces was to develop relevant lessons learned issues for the Candu nuclear power plants operating in Canada and to identify actions that could enhance safety. Subsequently, a CNSC Integrated Action Plan was developed to address the recommendations arising from the CNSC Task Force Report, and the licensees aligned their actions with this plan. The actions were grouped into four categories, as follows. 1. Strengthening defence in depth; Actions included re-assessment of a site-specific external hazards using PSA; physical plant changes to add further cooling water make-up capability to heat sinks; development of new Canadian standards and updated regulatory documents that address requirements for Beyond Design Basis Accidents (BDBA), Probabilistic Safety Assessment (PSA) and Period Safety Review (PSR); development of a new and novel methodology for assessing multi-unit risk that is consistent with the IAEA construct of defence-in-depth (including a safety goal hierarchy and complementary risk analysis methods that address 'Black Swan' events); enhanced management of hydrogen combustion challenges; and enhanced Severe Accident Management Guidelines (SAMG). 2. Enhancing emergency response; Actions included provision of Emergency Mitigation Equipment (EME) such as transportable diesel-driven pumps or fire engines, transportable diesel generators for AC and instrument power; pre-installed connections to in-plant equipment via easily accessible and protected quick-connect points for water and electrical power; upgraded emergency response centres; enhanced communication systems and protocols; and enhanced emergency response programs, training and exercises. 3. Improving the regulatory framework and processes; 4. Enhancing international collaboration. Actions assigned under categories 3 and 4 were placed upon the regulator. This presentation will focus on the assigned actions in categories 1 and 2. The underlying motivation, objectives, and safety benefits associated with the actions will be described within the context of broader aspects of severe accident mitigation in water-cooled reactors. (author)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Explicitly Accounting for Multiple Spatial Scales in Data Informing Kd Distribution Development - 18414

This work provides an example of incorporating information about spatial scale into input parameter distribution development for probabilistic performance assessment (PA) models and highlights the importance of having an interdisciplinary team working together in the distribution development process. The novel approach is demonstrated for contaminant partitioning coefficients for water/rock (Kds), providing a specific context for discussing the approach and challenges, but the strategy is general and applies to many other PA model inputs. In this example, a PA model of a radioactive waste disposal site is assumed to have large cells representing spatial scales of up to thousands of square meters in area and several to tens of meters in depth. Spatial variability in Kds is expected to occur at scales much smaller than the units in the model domain over which values are applied, and therefore values are intended to represent possible averages over the modeled volume (i.e., spatial scale represented by a PA model cell). Information about possible Kd values for each material category comes from a variety of literature sources, and the reported results used to inform the distribution are available at a wide range of scales (e.g., individual batch experiment results from small-scale samples, averages based on samples collected over site-scales, averages based on samples from large regions, etc.). Effectively addressing the mismatch in spatial scale between available information and the PA Model is a challenge within PA distribution development and it takes effort from database managers, subject matter experts, modelers, and statisticians. (authors)

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

A Decision-Based Iterative Modeling Framework for Probabilistic Performance Assessments - 18441

Radiological probabilistic performance assessments (PPAs) employ highly complex computer models to simulate the release of radionuclides from wastes in an engineered disposal facility, migration of these radionuclides over time, and resulting radiological dose to potentially exposed individuals. An iterative approach to PPA modeling is described to help define the appropriate complexity of the computer model. An initial model is developed that addresses all relevant processes and media, but that employs preliminary statistical parameter distributions to minimize the initial effort expended. Where appropriate, transport and exposure submodels with simplifying assumptions are also used in the initial model. Different assumptions and boundaries may be explored with the initial model to assess their importance in constraining model results. Exploration of the results of the initial model are used to determine the benefits of model refinements, and to guide such refinements to focus on specific areas of the model that most influence the model outputs of interest. This approach promotes efficiency and model parsimony, but does not compromise defensibility when performed in a transparent manner. Key model boundaries for a PPA include the time frame of the modeling (temporal boundaries), where future human receptors may be located (spatial boundaries), and what types of land uses and activities may exist in the future (exposure scenarios). Simplified transport and exposure models can be useful in helping decision makers and stakeholders establish such boundaries by analyzing the consequences of selecting different bounds and understanding which assumptions and boundaries have the greatest influence on the magnitude and distribution of model results. Pathway analysis and parameter sensitivity analysis performed on PPA model results are key components of the iterative modeling approach. The objective of pathway analysis is to identify the critical radionuclides and pathways contributing to radiation dose during the modeling performance period. An environmental 'pathway' is described by the connections from a waste package or cell in the disposal facility to a future human exposure event. These connections are commonly visualized using graphical means. A simple example of a potentially complete pathway for radon-222 is: radium-containing waste → radon emanation → upwards diffusion in cap material → diffusion/advection into a future building → inhalation exposure to a building occupant Parameter sensitivity analysis is conducted on the results of a PPA to identify those specific model input parameters whose distribution of values contributes the most to the distribution of target model output, such as annual radiation dose. Pathway and sensitivity analyses on initial modeling results provide information for guiding subsequent model development, and possibly interim decisions, by informing questions tied to the modeling objectives and performance criteria. For example, questions related to radiological doses from existing disposals may include: 1. Is radiation dose from radiological exposures either obviously negligible or obviously unacceptable based on initial modeling results? 2. If dose is not obviously negligible or unacceptable relative to applicable performance criteria, what model refinements would reduce uncertainty in the results to better inform decisions? (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Advanced Computational Accident Scenario Modeling for Safety Margin Economic Analysis

The intent of this research is to establish a new tool in the Risk Informed Safety Margin Characterization (RISMC) tool-kit for the fiscal analysis of potential power plant upgrades. This is being done by providing proof-of-concept of the Safety Margin Economic Evaluation (SMEE) methodology of performing cost-benefit analyses of potential Nuclear Power Plant (NPP) upgrades to allow for more data-driven decision making with regards to nuclear safety. To this end, the RAVEN code developed by the Idaho National Laboratory is being used to perform a cost-benefit analysis of several potential power plant safety upgrades to compare the material cost of the upgrades with the probabilistic benefit of avoiding accident scenario consequences to find the most cost effective safety upgrades. Additionally, the MELCOR and MACCS2 codes developed by the Sandia National Laboratory are being used for accident scenario modeling. Evaluating potential NPP upgrades is concluded by doing a detailed cost-benefit analysis of the upgrades. The cost of the upgrade is the direct material cost of implementing, maintaining, and operating the upgrade at an NPP for the projected duration of its lifetime. The benefit of the upgrade is more complex, and is essentially the Risk avoided, where Risk is defined as Consequences multiplied by Probability. Here, Consequences is the economic impact of a severe accident and/or nuclear release and Probability is the likelihood that an NPP upgrade will prevent or mitigate a severe accident, reducing or even eliminating its Consequences. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

External Events Analysis for LWRS/RISMC Project: Methodology Development and Early Demonstration

The ultimate scope of Industrial Application no.2 (IA) of the LWRS/RISMC project is a realistic simulation of natural external hazards that impose threat to a NPP. This scope requires the development of a methodology and of a qualified set of tools able to perform advanced risk-informed safety analysis. In particular the methodology should be able to combine results from seismic, flooding and thermal-hydraulic (TH) deterministic calculations with dynamic PRA. A so-called Risk Informed Margin Management (RIMM) approach is being used to represent meaningful (i.e., realistic facility representation and simulation) event scenarios and consequences by linking probabilistic simulations to external events physics. Calculations performed so far demonstrated the capabilities of the proposed methodology in analyzing multi-hazards generated by External Events. The sample application described in this paper allowed identifying the main issues (code interfaces, computational and human resources needed, simplifications) connected to such kind of advanced analysis. The next steps will be the risk-informed earthquakes and flooding analysis of a system of a generic PWR.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Corrosion of Aluminum in the Chemical Environment after a Loss of Coolant Accident (LOCA)

Probabilistic assessment of Generic Safety Issue (GSI)- 191 generated further studies to improve the characterization of post-Loss of Coolant Accident (LOCA) aluminum release from corrosion sources, since the current industry approach of assessing aluminum release is only valid for deterministic assessments of GSI-191. The nuclear industry has long accepted that solution pH, solution temperature, sump chemical constituents and the phenomena of silicate or phosphate 'inhibition' affects the quantity of aluminum released and the quality of results obtained from models used to predict aluminum release. While improvements in the characterization of aluminum release has been made, the impacts of solution velocity, concurrent corrosion of other metals [6], and trace amount of chemical constituents such as chloride and sulfate ions on aluminum release in post-LOCA conditions have not been assessed. Those factors are well known to influence aluminum corrosion, and will be investigated in this work. The effect of flow on aluminum corrosion rate is characterized using different hydrodynamic system, which has not been investigated in GSI-191 related studies. The corrosion rate depends on the mass-transfer rate which is a function of flow velocity and flow pattern. The inhibition effect of silicon on aluminum corrosion is characterized, and the representative silicon concentration in post-LOCA condition has the potential to significantly inhibit the aluminum corrosion rate. The experiments that investigated silicon effects on aluminum corrosion also showed that calcium leached from Cal-sil can slow corrosion and that the silicon inhibition effect on aluminum corrosion became less pronounced at higher temperature and lower pH. Trace chloride and fluoride ions slightly increased aluminum corrosion rate, and the presence of sulfite ions could mitigate this harmful effect. The corrosion of zinc significantly retarded the aluminum corrosion rate, while presence of copper and carbon steel has little effect on aluminum corrosion in both galvanic and non-galvanic corrosion cases. This work is not complete and is currently evaluating high temperature (>95 C) test with all parameters mentioned and evaluation of the effects on variable boron concentrations at set conditions; Therefore an improved aluminum release model for probabilistic aluminum release model is not proposed. However, this work shows the important impact on aluminum corrosion of parameters not characterized by other studies. To obtain representative aluminum release results for use in GSI-191 probabilistic analysis, the model should be built on data obtained from experiments that account for all parameters that may impact aluminum corrosion and release. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

Development of SiC-based Cladding for Accident Tolerant Fuels

Fuels and core structures in current light water reactors (LWR's) are vulnerable to catastrophic failure in severe accidents. This vulnerability is attributed primarily to the rapid oxidation kinetics of zirconium alloys in a water vapor environment at very high temperatures. Alternative materials to zirconium alloys with reduced oxidation kinetics are sought to enable enhanced accident-tolerant fuels and cores. Among the candidate alternatives, silicon carbide (SiC) - based materials, in particular continuous SiC fiber-reinforced SiC matrix ceramic composites (SiC/SiC composites), are considered to offer outstanding safety benefits. The primary objective of this work is to provide with updated overviews of the concepts, development status, and future prospect for the SiC-based accident-tolerant fuel (ATF) claddings. A systematic technology evaluation program plan was developed and updated to review the technologies needed, to identify and analyze critical gaps, and to lay out a road-map toward establishing key technologies. The technology review and the gap analysis in this exercise were based on an early study and community workshops. Many of the technical gaps identified are related with the three key feasibility issues: hydrothermal corrosion, mechanical failure, and fuel compatibility. Additional performance issues including accident-tolerance features and fission product retention during the normal operation have also been identified. The program plan was designed to systematically address the key gap issues and is formulated in a work breakdown structure. Simultaneously, the plan is being established to set a technical program for advancing the technological readiness levels of concept non-specific essential technologies in three top-level categories of Design and Failure, Environmental Effects, and Off-normal Behavior. The execution of the program plan is in progress across the U.S. Department of Energy laboratories, industries, universities, and international collaborators. The critical issues for SiC composite-based fuel cladding concepts were found to be related to the performance in the normal operation. In particular, radiation-assisted hydrothermal corrosion and the gaseous fission product (FP) retention are identified to be the most critical issues. The primary ORNL concept pursues the use of dual-purpose coating on the SiC composite cladding wall to mitigate the corrosion and the gas permeation issues at the same time, whereas the alternative approaches such as the dense monolithic SiC protective layer combined with the water chemistry control may be feasible. Surface metallization through variations of physical vapor deposition and electroplating are being demonstrated. The FP gas retention is an issue associated with the cracking for the ceramic matrix composite cladding. To establish the probabilistic failure behavior, the relevant test methods were developed in ASTM Committee C28 on Advanced Ceramics. Toward the end-plugging technology development, various methods for joining SiC to SiC had been developed and tested for mechanical properties, irradiation effects, and environmental effects. Multiple methods including the pressureless transient eutectic SiC joining and CVD-based SiC joining were identified particularly promising. In addition, a multi-physics, three-dimensional finite element-based modeling is in progress toward an integrated simulation of the SiC composite-cladded fuel evolutions. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗