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DOE OSTI2015

Framework for Modeling High-Impact, Low-Frequency Power Grid Events to Support Risk-Informed Decisions

Natural and man-made hazardous events resulting in loss of grid infrastructure assets challenge the electric power grid’s security and resilience. However, the planning and allocation of appropriate contingency resources for such events requires an understanding of their likelihood and the extent of their potential impact. Where these events are of low likelihood, a risk-informed perspective on planning can be problematic as there exists an insufficient statistical basis to directly estimate the probabilities and consequences of their occurrence. Since risk-informed decisions rely on such knowledge, a basis for modeling the risk associated with high-impact low frequency events (HILFs) is essential. Insights from such a model can inform where resources are most rationally and effectively expended. The present effort is focused on development of a HILF risk assessment framework. Such a framework is intended to provide the conceptual and overarching technical basis for the development of HILF risk models that can inform decision makers across numerous stakeholder sectors. The North American Electric Reliability Corporation (NERC) 2014 Standard TPL-001-4 considers severe events for transmission reliability planning, but does not address events of such severity that they have the potential to fail a substantial fraction of grid assets over a region, such as geomagnetic disturbances (GMD), extreme seismic events, and coordinated cyber-physical attacks. These are beyond current planning guidelines. As noted, the risks associated with such events cannot be statistically estimated based on historic experience; however, there does exist a stable of risk modeling techniques for rare events that have proven of value across a wide range of engineering application domains. There is an active and growing interest in evaluating the value of risk management techniques in the State transmission planning and emergency response communities, some of this interest in the context of grid modernization activities. The availability of a grid HILF risk model, integrated across multi-hazard domains which, when interrogated, can support transparent, defensible and effective decisions, is an attractive prospect among these communities. In this report, we document an integrated HILF risk framework intended to inform the development of risk models. These models would be based on the systematic and comprehensive (to within scope) characterization of hazards to the level of detail required for modeling risk, identification of the stressors associated with the hazards (i.e., the means of impacting grid and supporting infrastructure), characterization of the vulnerability of assets to these stressors and the probabilities of asset compromise, the grid’s dynamic response to the asset failures, and assessment of subsequent severities of consequence with respect to selected impact metrics, such as power outage duration and geographic reach. Specifically, the current framework is being developed to;1. Provide the conceptual and overarching technical paradigms for the development of risk models; 2. Identify the classes of models required to implement the framework - providing examples of existing models, and also identifying where modeling gaps exist; 3. Identify the types of data required, addressing circumstances under which data are sparse and the formal elicitation of informed judgment might be required; and 4. Identify means by which the resultant risk models might be interrogated to form the necessary basis for risk management.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2015

Enhanced Soundings for Local Coupling Studies: 2015 ARM Climate Research Facility Field Campaign

Matching observed diurnal cycles is a fundamental yet extremely complex test for models. High temporal resolution measurements of surface turbulent heat fluxes and boundary layer properties are required to evaluate the daytime evolution of the boundary layer and its sensitivity to land-atmosphere coupling. To address this need, (12) one-day intensive observing periods (IOP) with enhanced radiosonding will be carried out at the ARM Southern Great Plains (SGP) Central Facility (CF) during summer 2015. Each IOP will comprise a single launch to correspond with the nighttime overpass of the A-Train of satellites (~0830 UTC) and hourly launches during daytime beginning from 1130 UTC and ending at 2130 UTC. At 3-hourly intervals (i.e., 1140 UTC, 1440 UTC, 1740 UTC, and 2040 UTC) a duplicate second radiosonde will be launched 10 minutes subsequent to launch of the on-hour radiosonde for the purpose of assessing horizontal atmospheric variability. In summary, each IOP will have a 14-sounding supplement to the 6-hourly operational sounding schedule at the ARM-SGP CF. The IOP days will be decided before sunset on the preceding day, according to the judgment of the PI’s and taking into consideration daily weather forecasts and the operability of complimentary ARM-SGP CF instrumentation. An overarching goal of the project is to address how ARM could better observe land-atmosphere coupling to support the evaluation and refinement of coupled weather and climate models.

Ferguson, CR↗

DOE OSTI2016

A Report on the Validation of Beryllium Strength Models

This report discusses work on validating beryllium strength models with flyer plate and Taylor rod experimental data. Strength models are calibrated with Hopkinson bar and quasi-static data. The Hopkinson bar data for beryllium provides strain rates up to about 4000 per second. A limitation of the Hopkinson bar data for beryllium is that it only provides information on strain up to about 0.15. The lack of high strain data at high strain rates makes it difficult to distinguish between various strength model settings. The PTW model has been calibrated many different times over the last 12 years. The lack of high strain data for high strain rates has resulted in these calibrated PTW models for beryllium exhibiting significantly different behavior when extrapolated to high strain. For beryllium, the α parameter of PTW has recently been calibrated to high precision shear modulus data. In the past the α value for beryllium was set based on expert judgment. The new α value for beryllium was used in a calibration of the beryllium PTW model by Sky Sjue. The calibration by Sjue used EOS table information to model the temperature dependence of the heat capacity. Also, the calibration by Sjue used EOS table information to model the density changes of the beryllium sample during the Hopkinson bar and quasi-static experiments. In this paper, the calibrated PTW model by Sjue is compared against experimental data and other strength models. The other strength models being considered are a PTW model calibrated by Shuh- Rong Chen and a Steinberg-Guinan type model by John Pedicini. The three strength models are used in a comparison against flyer plate and Taylor rod data. The results show that the Chen PTW model provides better agreement to this data. The Chen PTW model settings have been previously adjusted to provide a better fit to flyer plate data, whereas the Sjue PTW model has not been changed based on flyer plate data. However, the Sjue model provides a reasonable fit to flyer plate and Taylor rod data, and also gives a better match to recently analyzed Z-machine data which has a strain of about 0.35 and a strain rate of 3e5 s -1 .

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗

DOE OSTI2016

Corrective Action Investigation Plan for Corrective Action Unit 413: Clean Slate II Plutonium Dispersion (TTR) Tonopah Test Range, Nevada (Rev. 1)

Corrective Action Unit (CAU) 413 is located on the Tonopah Test Range, which is approximately 130 miles northwest of Las Vegas, Nevada, and approximately 40 miles southeast of Tonopah, Nevada. The CAU 413 site consists of the release of radionuclides to the surface and shallow subsurface from the conduct of the Clean Slate II (CSII) storage–transportation test conducted on May 31, 1963. CAU 413 includes one corrective action site (CAS), TA-23-02CS (Pu Contaminated Soil). The known releases at CAU 413 are the result of the atmospheric deposition of contamination from the 1963 CSII test. The CSII test was a non-nuclear detonation of a nuclear device located inside a reinforced concrete bunker covered with 2 feet of soil. This test dispersed radionuclides, primarily plutonium, on the ground surface. The presence and nature of contamination at CAU 413 will be evaluated based on information collected from a corrective action investigation (CAI). The investigation is based on the data quality objectives (DQOs) developed on June 17, 2015, by representatives of the Nevada Division of Environmental Protection; the U.S. Air Force; and the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Field Office. The DQO process was used to identify and define the type, amount, and quality of data needed to develop and evaluate appropriate corrective actions for CAU 413. The CAI will include radiological surveys, geophysical surveys, collection and analyses of soil samples, and assessment of investigation results. The collection of soil samples will be accomplished using both probabilistic and judgmental sampling approaches. To facilitate site investigation and the evaluation of DQO decisions, the releases at CAU 413 have been divided into seven study groups.

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2016

Los Alamos National Laboratory SAVY-4000 Field Surveillance Plan Update for 2016

The Packaging Surveillance Program section of the Department of Energy (DOE) Manual 441.1-1, Nuclear Material Packaging Manual (DOE 2008), requires DOE contractors to “ensure that a surveillance program is established and implemented to ensure the nuclear material storage package continues to meet its design criteria.” The Los Alamos National Laboratory (LANL) SAVY-4000 Field Surveillance Plan was first issued in FY 2013 (Kelly et al. 2013). The surveillance plan is reviewed annually and updated as necessary based on SAVY-4000 surveillance and other surveillance findings, as well as results of the lifetime extension studies (Blair et al. 2012, Weis et al. 2015a). The LANL SAVY-4000 Field Surveillance Plan Update was issued in 2014 (Kelly et al. 2014). This 2016 update reflects changes to the surveillance plan resulting from restrictions on handling residue materials greater than 500 g, the addition of specific engineering judgment containers, and 2015 surveillance findings. The SAVY-4000 container has a design life of five years, which was chosen as a conservative estimate of the functional properties of the materials used in the construction of the SAVY 4000 when exposed to the potential insults including temperature, corrosive materials and gases, and radiation. The SAVY-4000 container design basis is described in a safety analysis report (Anderson et al. 2013). In the National Nuclear Security Administration's (NNSA’s) approval of the safety analysis report, it was recommended that the design life clock begin on March 2014 (Nez et al. 2014). However, it is expected that a technical basis can be developed to extend the design life of the SAVY-4000 containers to approximately 40 years (Blair et al. 2012, Weis et al. 2015a). This surveillance plan update covers five years (2015–2019) and is developed to ensure SAVY-4000 containers meet their design criteria over the current five-year design life and to gather data that can be used in developing the technical basis for a 40-year design life. The technical basis for the longer design life will be developed based on the results of this surveillance plan and the concurrent lifetime extension studies (Blair et al. 2012, Weis et al. 2015a). A long-term surveillance plan will be developed by the completion of this plan and will be based on the results of the first several years of this five-year surveillance plan and the lifetime extension studies. In addition, the current Hagan container population is so large (approximately 3,500 LANL-wide, approximately 3,000 in PF-4) that it is anticipated Hagan containers will be in use for many years to come. Therefore, this plan includes surveillance activities for the Hagan container system during the time it takes to replace the Hagan containers with manually-compliant containers such as the SAVY 4000. These Hagan examinations not only provide information to determine if additional Hagan surveillance is needed, but also support SAVY lifetime-extension studies because some of the container components are made from similar material (e.g., stainless steel and Viton).

42 ENGINEERING↗

DOE OSTI2016

Sensitivity-Uncertainty Based Nuclear Criticality Safety Validation

These are slides from a seminar given to the University of Mexico Nuclear Engineering Department. Whisper is a statistical analysis package developed to support nuclear criticality safety validation. It uses the sensitivity profile data for an application as computed by MCNP6 along with covariance files for the nuclear data to determine a baseline upper-subcritical-limit for the application. Whisper and its associated benchmark files are developed and maintained as part of MCNP6, and will be distributed with all future releases of MCNP6. Although sensitivity-uncertainty methods for NCS validation have been under development for 20 years, continuous-energy Monte Carlo codes such as MCNP could not determine the required adjoint-weighted tallies for sensitivity profiles. The recent introduction of the iterated fission probability method into MCNP led to the rapid development of sensitivity analysis capabilities for MCNP6 and the development of Whisper. Sensitivity-uncertainty based methods represent the future for NCS validation – making full use of today’s computer power to codify past approaches based largely on expert judgment. Validation results are defensible, auditable, and repeatable as needed with different assumptions and process models. The new methods can supplement, support, and extend traditional validation approaches.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

Plutonium Critical Mass Curve Comparison to Mass at Upper Subcritical Limit (USL) Using Whisper

Whisper is computational software designed to assist the nuclear criticality safety analyst with validation studies with the MCNP ® Monte Carlo radiation transport package. Standard approaches to validation rely on the selection of benchmarks based upon expert judgment. Whisper uses sensitivity/uncertainty (S/U) methods to select relevant benchmarks to a particular application or set of applications being analyzed. Using these benchmarks, Whisper computes a calculational margin. Whisper attempts to quantify the margin of subcriticality (MOS) from errors in software and uncertainties in nuclear data. The combination of the Whisper-derived calculational margin and MOS comprise the baseline upper subcritical limit (USL), to which an additional margin may be applied by the nuclear criticality safety analyst as appropriate to ensure subcriticality. A series of critical mass curves for plutonium, similar to those found in Figure 31 of LA-10860-MS, have been generated using MCNP6.1.1 and the iterative parameter study software, WORM_Solver. The baseline USL for each of the data points of the curves was then computed using Whisper 1.1. The USL was then used to determine the equivalent mass for plutonium metal-water system. ANSI/ANS-8.1 states that it is acceptable to use handbook data, such as the data directly from the LA-10860-MS, as it is already considered validated (Section 4.3 4) “Use of subcritical limit data provided in ANSI/ANS standards or accepted reference publications does not require further validation.”). This paper attempts to take a novel approach to visualize traditional critical mass curves and allows comparison with the amount of mass for which the k eff is equal to the USL (calculational margin + margin of subcriticality). However, the intent is to plot the critical mass data along with USL, not to suggest that already accepted handbook data should have new and more rigorous requirements for validation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

DYNA3D/ParaDyn Regression Test Suite Inventory

The following table constitutes an initial assessment of feature coverage across the regression test suite used for DYNA3D and ParaDyn. It documents the regression test suite at the time of preliminary release 16.1 in September 2016. The columns of the table represent groupings of functionalities, e.g., material models. Each problem in the test suite is represented by a row in the table. All features exercised by the problem are denoted by a check mark (√) in the corresponding column. The definition of “feature” has not been subdivided to its smallest unit of user input, e.g., algorithmic parameters specific to a particular type of contact surface. This represents a judgment to provide code developers and users a reasonable impression of feature coverage without expanding the width of the table by several multiples. All regression testing is run in parallel, typically with eight processors, except problems involving features only available in serial mode. Many are strictly regression tests acting as a check that the codes continue to produce adequately repeatable results as development unfolds; compilers change and platforms are replaced. A subset of the tests represents true verification problems that have been checked against analytical or other benchmark solutions. Users are welcomed to submit documented problems for inclusion in the test suite, especially if they are heavily exercising, and dependent upon, features that are currently underrepresented.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2016

NSRD-11: Computational Capability to Substantiate DOE-HDBK-3010 Data.

Safety basis analysts throughout the U.S. Department of Energy (DOE) complex rely heavily on the information provided in the DOE Handbook, DOE - HDBK - 3010, Airborne Release Fractions/Rates and Respirable Fractions for Nonreactor Nuclear Facilities, to determine radionuclide source terms. In calculating source terms, analysts tend to use the DOE Handbook's bounding values on airborne release fractions (ARFs) and respirable fractions (RFs) for various categories of insults (representing potential accident release categories). This is typically due to both time constraints and the avoidance of regulatory critique. Unfortunately, these bounding ARFs/RFs represent extremely conservative values. Moreover, they were derived from very limited small-scale bench/laboratory experiments and/or from engineered judgment. Thus, the basis for the data may not be representative of the actual unique accident conditions and configurations being evaluated. The goal of this research is to develop a more accurate and defensible method to determine bounding values for the DOE Handbook using state-of-art multi-physics-based computer codes. This enables us to better understand the fundamental physics and phenomena associated with the types of accidents in the handbook. In this year, this research included improvements of the high-fidelity codes to model particle resuspension and multi-component evaporation for fire scenarios. We also began to model ceramic fragmentation experiments, and to reanalyze the liquid fire and powder release experiments that were done last year. The results show that the added physics better describes the fragmentation phenomena. Thus, this work provides a low-cost method to establish physics-justified safety bounds by taking into account specific geometries and conditions that may not have been previously measured and/or are too costly to perform.

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2017

Analyst-to-Analyst Variability in Simulation-Based Prediction

This report describes findings from the culminating experiment of the LDRD project entitled, "Analyst-to-Analyst Variability in Simulation-Based Prediction". For this experiment, volunteer participants solving a given test problem in engineering and statistics were interviewed at different points in their solution process. These interviews are used to trace differing solutions to differing solution processes, and differing processes to differences in reasoning, assumptions, and judgments. The issue that the experiment was designed to illuminate -- our paucity of understanding of the ways in which humans themselves have an impact on predictions derived from complex computational simulations -- is a challenging and open one. Although solution of the test problem by analyst participants in this experiment has taken much more time than originally anticipated, and is continuing past the end of this LDRD, this project has provided a rare opportunity to explore analyst-to-analyst variability in significant depth, from which we derive evidence-based insights to guide further explorations in this important area.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2016

Light Water Reactor Sustainability Program: Risk-Informed Safety Margins Characterization (RISMC) Pathway Technical Program Plan

Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). As the current Light Water Reactor (LWR) NPPs age beyond 60 years, there are possibilities for increased frequency of Systems, Structures, and Components (SSCs) degradations or failures that initiate safety-significant events, reduce existing accident mitigation capabilities, or create new failure modes. Plant designers commonly “over-design” portions of NPPs and provide robustness in the form of redundant and diverse engineered safety features to ensure that, even in the case of well-beyond design basis scenarios, public health and safety will be protected with a very high degree of assurance. This form of defense-in-depth is a reasoned response to uncertainties and is often referred to generically as “safety margin.” Historically, specific safety margin provisions have been formulated, primarily based on “engineering judgment.”

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Advanced Mechanistic 3D Spatial Modeling and Analysis Methods to Accurately Represent Nuclear Facility External Event Scenarios

Probabilistic risk assessment of nuclear power plants initially focused on events initiated by internal faults at the plant, rather than external hazards including earthquakes and flooding. Although the importance of external hazards risk analysis is now well recognized, the methods for analyzing low probability external hazards rely heavily on subjective judgment of specialists, often resulting in substantial conservatism. This research developed a framework to integrate the risk of seismic and flooding events using realistic structural models and simulation of response of nuclear structures. The results of four application case studies are presented.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2018

NSRD-15:Computational Capability to Substantiate DOE-HDBK-3010 Data

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2016

Deep Borehole Disposal Concept: Development of Universal Canister Concept of Operations

This report documents key elements of the conceptual design for deep borehole disposal of radioactive waste to support the development of a universal canister concept of operations. A universal canister is a canister that is designed to be able to store, transport, and dispose of radioactive waste without the canister having to be reopened to treat or repackage the waste. This report focuses on the conceptual design for disposal of radioactive waste contained in a universal canister in a deep borehole. The general deep borehole disposal concept consists of drilling a borehole into crystalline basement rock to a depth of about 5 km, emplacing WPs in the lower 2 km of the borehole, and sealing and plugging the upper 3 km. Research and development programs for deep borehole disposal have been ongoing for several years in the United States and the United Kingdom; these studies have shown that deep borehole disposal of radioactive waste could be safe, cost effective, and technically feasible. The design concepts described in this report are workable solutions based on expert judgment, and are intended to guide follow-on design activities. Both preclosure and postclosure safety were considered in the development of the reference design concept. The requirements and assumptions that form the basis for the deep borehole disposal concept include WP performance requirements, radiological protection requirements, surface handling and transport requirements, and emplacement requirements. The key features of the reference disposal concept include borehole drilling and construction concepts, WP designs, and waste handling and emplacement concepts. These features are supported by engineering analyses.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

Risk-Informed Safety Margins Characterization (RISMC) Pathway Technical Program Plan

Safety is central to the design, licensing, operation, and economics of nuclear power plants (NPPs). As the current Light Water Reactor (LWR) NPPs age beyond 60 years, there are possibilities for increased frequency of systems, structures, and components (SSCs) degradations or failures that initiate safety-significant events, reduce existing accident mitigation capabilities, or create new failure modes. Plant designers commonly “over-design” portions of NPPs and provide robustness in the form of redundant and diverse engineered safety features to ensure that, even in the case of well-beyond design basis scenarios, public health and safety will be protected with a very high degree of assurance. This form of defense-in-depth is a reasoned response to uncertainties and is often referred to generically as “safety margin.” Historically, specific safety margin provisions have been formulated, primarily based on “engineering judgment.” The ability to better characterize (i.e., describe and quantify) safety margin is important to improved decision making about LWR design and long-term operation. A systematic approach to characterizing safety margins and the subsequent risk informed margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. In addition, as research and development (R&D) in the LWRS Program and other collaborative efforts yield new data and improved scientific understanding of physical processes that govern the aging and degradation of plant SSCs (and concurrently support technological advances in nuclear reactor fuels and plant instrumentation and control systems), needs and opportunities to better optimize plant safety, economics, and performance will become known. The purpose of the Risk-Informed Safety Margins Characterization (RISMC) Pathway R&D is to support plant decisions for risk-informed margins management with the aim to improve economics, reliability, and sustain safety of current NPPs over periods of extended plant operations. The goals of the RISMC Pathway are twofold: (1) develop and demonstrate a risk-assessment method that is coupled to safety margin quantification that can be used by NPP decision makers as part of risk-informed margin management strategies; (2) create an advanced RISMC Toolkit that enables more accurate representation of NPP safety margins. The methods and tools provided by RISMC are essential to a comprehensive and integrated risk-informed margin management approach that supports effective preservation of margin for both active and passive SSCs.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2018

Failure Rate Estimates for Passive Mechanical Components

Fusion design projects that reach the stage of design reviews typically require reliability, availability, maintainability and inspectability (RAMI) assessments. While many kinds of reliability and maintainability data are needed for a wide variety of fusion components to quantify a RAMI assessment, one of the types of components requiring data is the passive mechanical component. Passive mechanical components can be items such as metal plates, rods, shafts, brackets, bolts, and other items. Generally these passive components tend to have low failure rates since they are simple designs, have no or few moving parts, and do not need control signals to perform their design function. However, there are no widely recognized datasets for such components, so analysts must use multiple datasets and judgment for failure rates. This report gives recommended failure rates for various failure modes of these components, and gives suggested failure rate multipliers to apply to these failure rates if they reside in harsh environments. If the component resides far from the tokamak then typical industrial failure rates would suffice. Components near the tokamak find conditions can be demanding; similar to those in a fission reactor containment building. Some component failure rates can be modified if the components reside inside the tokamak.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2020

An Overview of the State-of-the-Art Reactor Consequence Uncertainty Assessment Accident Progression Insights

The U.S. Nuclear Regulatory Commission (NRC) with Sandia National Laboratories (Sandia) have completed three uncertainty analyses (UAs) as part of the State-of-the-Art Reactor Consequence Analyses (SOARCA) program. The SOARCA UAs included an integrated evaluation of uncertainty in accident progression, radiological release, and offsite health consequence projections. The UA for Peach Bottom, a boiling-water reactor (BWR) with a Mark I containment located in the State of Pennsylvania, analyzed the unmitigated long-term station blackout SOARCA scenario. The UA for Sequoyah, a 4-loop Westinghouse pressurized-water reactor (PWR) located in the State of Tennessee, analyzed the unmitigated short-term station blackout SOARCA scenario, with a focus on issues unique to the ice condenser containment and the potential for early containment failure due to hydrogen deflagration. The UA for Surry, a 3-loop Westinghouse PWR with a sub-atmospheric large dry containment located in the State of Virginia, analyzed the unmitigated short-term station blackout SOARCA scenario including the potential for thermally-induced steam-generator tube rupture. These three UAs are currently documented in three NUREG/CR reports. This report provides input to planned NRC documentation on the insights and findings from the SOARCA UA program. The purpose of the summary report is to provide a useful reference for regulatory applications that require the evaluation of offsite consequence risk from beyond design basis event severe accidents. This report focuses on the accident progression and source term insights developed from the MELCOR analyses. MELCOR is the NRC's best-estimate, severe accident computer code used in the SOARCA UAs. In anticipation of the SOARCA UA insights work, NRC and Sandia benchmarked the response of the Peach Bottom model to selected reference calculations from the Peach Bottom SOARCA UA. Peach Bottom was the first SOARCA UA performed and was completed in 2015 using the MELCOR 1.8.6 code. The PWR SOARCA UAs evolved the original methodology and utilized the updated MELCOR 2.2 computer code. The Peach Bottom model has been systematically updated for other NRC research efforts and has been updated to MELCOR 2.2. computer code. The findings from the new reference calculations using the updated model with the MELCOR 2.2 code are also integrated into the report. A second objective is an assessment of the applicability of the results to the other nuclear reactors in the U.S. As the key findings are reviewed, judgments are presented on the applicability of the results to other U.S. nuclear power plants. An important objective of the SOARCA program relied on high- fidelity plant-specific modeling. However, the nature of the insights and conclusions allowed judgements to be made on the applicability of the various insights to the same general classification of plant (i.e., BWR or PWR) or the entire fleet of plants. Finally, the results from the SOARCA UA accident progression calculations contain a wealth of information not previously documented in the NUREG/CRs. This report includes new but related information that can be used to benchmark past or support future regulatory decisions related to severe accidents. The new work includes a benchmark of the NUREG-1465 licensing source term definitions, the variability of key accident progression events and timing to radionuclide release, and an improved understanding of the timing and source terms from consequential steam generator tube ruptures. iii ACKNOWLEDGEMENTS The Sandia authors gratefully acknowledge the significant technical and programmatic contributions from the NRC SOARCA team which are reflected throughout the report. Dr. Tina Ghosh has been involved throughout the SOARCA UAs, providing the primary managerial and technical oversight. The long lists of NRC and Sandia contributors from the SOARCA UAs are cited in the three NUREG/CRs and are also gratefully acknowledged by the small team of authors compiling the results of their efforts. Significant technical contributions, advice, and reviews were provided by Dr. Hossein Esmaili, Dr. Alfred Hathaway, and Dr. Edward Fuller (retired) of the NRC. Dr. Randal Gauntt (retired), Mr. Patrick Mattie, Mr. Joseph Jones (retired), and Dr. Doug Osborn from Sandia are recognized as the SOARCA UA managers guiding the past efforts. There is a comparable list of project managers at the NRC including Ms. Patricia Santiago, Dr. Salman Haq, and Mr. Jon Barr. Sadly, we have lost Mr. Charlie Tinkler and Mr. Robert Prato, who were important contributors to the original SOARCA project. Finally, Mr. Kyle Ross and Mr. Mark Leonard have also retired but were significant technical contributors. Mr. Kyle Ross was the technical lead on all three SOARCA UAs and the original pressurized water reactor SOARCA study. Mr. Leonard was the technical lead on the original boiling water reactor SOARCA study and a key contributor to the first Peach Bottom SOARCA UA. iv

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2020

Group Structure Machine Learning Proposal

Nuclear data is the linchpin underwriting several fundamental capabilities and mission needs at LANL. New techniques such as machine learning can be brought to bear to solve old problems such as multigroup cross-section accuracy. In neutron transport, generating multigroup cross sections is a complex and arcane task, but a crucial one, as accurate solutions require appropriate cross sections. There are two key challenges when generating multigroup cross sections: (1) choosing an accurate weight function, and (2) choosing appropriate energy boundaries. Often, energy boundaries are chosen using “expert judgment” that is not documented and is difficult to replicate. The long-standing Los Alamos 30-group structure has been in use since at least 1969 and is still in use today. Simplistic attempts over the years since to improve on the 30-group structure have been met with limited success. Machine learning algorithms would enable the selection of appropriate, problem-dependent group boundaries without an inordinate investment of scientist time. We will develop workflows and tools to enable these improved group boundary choices, which will reduce uncertainty and increase predictive capability of neutron-transport applications at LANL.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗