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DOE OSTI2016

Nuflood, Version 1.x

Nuflood Version 1.x is a surface-water hydrodynamic package designed for the simulation of overland flow of fluids. It consists of various routines to address a wide range of applications (e.g., rainfall-runoff, tsunami, storm surge) and real time, interactive visualization tools. Nuflood has been designed for general-purpose computers and workstations containing multi-core processors and/or graphics processing units. The software is easy to use and extensible, constructed in mind for instructors, students, and practicing engineers. Nuflood is intended to assist the water resource community in planning against water-related natural disasters.

Tasseff, Byron↗

DOE OSTI2013

The Future of Energy from Nuclear Fission

Nuclear energy is an important part of our current global energy system, and contributes to supplying the significant demand for electricity for many nations around the world. There are 433 commercial nuclear power reactors operating in 30 countries with an installed capacity of 367 GWe as of October 2011 (IAEA PRIS, 2011). Nuclear electricity generation totaled 2630 TWh in 2010 representing 14% the world’s electricity generation. The top five countries of total installed nuclear capacity are the US, France, Japan, Russia and South Korea at 102, 63, 45, 24, and 21 GWe, respectively (WNA, 2012a). The nuclear capacity of these five countries represents more than half, 68%, of the total global nuclear capacity. The role of nuclear power in the global energy system today has been motivated by several factors including the growing demand for electric power, the regional availability of fossil resources and energy security concerns, and the relative competitiveness of nuclear power as a source of base-load electricity. There is additional motivation for the use of nuclear power because it does not produce greenhouse gas (GHG) emissions or local air pollutants during its operation and contributes to low levels of emissions throughout the lifecycle of the nuclear energy system (Beerten, J. et. al., 2009). Energy from nuclear fission primarily in the form of electric power and potentially as a source of industrial heat could play a greater role for meeting the long-term growing demand for energy worldwide while addressing the concern for climate change from rising GHG emissions. However, the nature of nuclear fission as a tremendously compact and dense form of energy production with associated high concentrations of radioactive materials has particular and unique challenges as well as benefits. These challenges include not only the safety and cost of nuclear reactors, but proliferation concerns, safeguard and storage of nuclear materials associated with nuclear fuel cycles. In March of 2011, an unprecedented earthquake of 9 magnitude and ensuing tsunami off the east coast of Japan caused a severe nuclear accident in Fukushima, Japan (Prime Minister of Japan and His Cabinet, 2011). The severity of the nuclear accident in Japan has brought about a reinvestigation of nuclear energy policy and deployment activities for many nations around the world, most notably in Japan and Germany (BBC, 2011; Reuter, 2011). The response to the accident has been mixed and its full impact may not be realized for many years to come. The nuclear accident in Fukushima, Japan has not directly affected the significant on-going nuclear deployment activities in many countries. China, Russia, India, and South Korea, as well as others, are continuing with their deployment plans. As of October 2011, China had the most reactors under construction at 27, while Russia, India, and South Korea had 11, 6, and 5 reactors under construction, respectively (IAEA PRIS, 2011). Ten other nations have one or two reactors currently under construction. Many more reactors are planned for future deployment in China, Russia, and India, as well as in the US. Based on the World Nuclear Association’s data, the realization of China’s deployment plan implies that China will surpass the US in total nuclear capacity some time in the future.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Metrics for the technical performance evaluation of light water reactor accident-tolerant fuel

The safe, reliable, and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the nuclear industry. Continual improvement of technology, including advanced materials and nuclear fuels, remains central to the industry’s success. Enhancing the accident tolerance of light water reactors (LWRs) became a topic of serious discussion following the 2011 Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex. The overall goal for the development of accident-tolerant fuel (ATF) for LWRs is to identify alternative fuel system technologies to further enhance the safety, competitiveness, and economics of commercial nuclear power. Designed for use in the current fleet of commercial LWRs or in reactor concepts with design certifications (GEN-III+), fuels with enhanced accident tolerance would endure loss of active cooling in the reactor core for a considerably longer period of time than the current fuel system while maintaining or improving performance during normal operations. The complex multiphysics behavior of LWR nuclear fuel in the integrated reactor system makes defining specific material or design improvements difficult; as such, establishing desirable performance attributes is critical in guiding the design and development of fuels and cladding with enhanced accident tolerance. Research and development of ATF in the United States is conducted under the U.S. Department of Energy (DOE) Fuel Cycle Research and Development Advanced Fuels Campaign. The DOE is sponsoring multiple teams to develop ATF concepts within multiple national laboratories, universities, and the nuclear industry. Concepts under investigation offer both evolutionary and revolutionary changes to the current nuclear fuel system. This study summarizes the technical evaluation methodology proposed in the United States to aid in the optimization and prioritization of candidate ATF designs.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2022

Impact of Increased Latent Generations on Sensitivity Calculations with SCALE

Analyses of cross section sensitivity data from systems with fissile material allow analysts to associate an importance for each material, nuclide, reaction, and neutron energy by simulating real world criticality scenarios. Although criticality safety validation efforts can be guided by the cross-section sensitivity and uncertainty data generated for a particular system, these calculations can often be computationally expensive and sometimes cumbersome without proper guidance. The TSUNAMI suite within the SCALE code package has several methods for generating sensitivity data, including multigroup and continuous energy (CE) capabilities. The release of SCALE 6.3 has three different CE methods for generating cross section sensitivity data: (1) the Iterated Fission Probability (IFP) method with the KENO Monte Carlo transport solver, (2) the IFP method with the Shift Monte Carlo transport solver, and (3) the Contributon-Linked eigenvalue sensitivity/Uncertainty estimation via Tracklength importance CHaracterization (CLUTCH) method with the KENO Monte Carlo transport solver. Although the CLUTCH method has additional parameters for generating sensitivity data files relative to the IFP method, all three methods use latent generations, which are the generations between an event (i.e., fission) and the assessment of importance based on the asymptotic population of progeny neutrons. Increasing the number of latent generations in a calculation leads to increased discrimination of the sensitivity coefficients but at the cost of the increased uncertainty associated with those generated values. Analysts must balance the accuracy of the sensitivity calculations and its uncertainty with the associated computational cost involved in generating the values. This paper discusses the impact of adjusting the latent generation parameter for a range of sensitivity values and how these changes compare with the direct perturbation values obtained from a change of ±0.5% Δ k in both benchmark and safety application models. Two benchmarks from the International Handbook of Evaluated Criticality Safety Benchmark Experiments and the MPC-32 dual purpose canister for spent nuclear fuel are used for analysis.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2023

Preliminary Design of Critical Experiments Involving Commercially Available B 4 C Neutron Absorber Plates with Low-Enriched UO 2 Fuel

The International Criticality Safety Benchmark Evaluation Project (ICSBEP) is an initiative to provide high-quality benchmark data in a standardized format for criticality safety analysts to use to validate calculational tools and nuclear data. In the last released 2022 version, the ICSBEP handbook contains over 5,000 critical experiment descriptions, results, and associated models. This paper describes the critical configurations proposed and calculations performed as part of the conceptual design phase of Integral Experiment Request (IER) 554 (IER-554). The goal of IER-554 is to add to the benchmark experiments available to the criticality safety community by designing critical experiments that can analyze how adding commercially available Boralcan plates to an assembly of low-enriched UO2 fuel rods affects the effective neutron multiplication factor (k eff ). Boralcan, a neutron absorber product developed by Rio Tinto, is made of B4C in an Al 1100 matrix and is commonly used for criticality suppression in fuel storage pools. In the current version of the ICSBEP handbook, only a few dated critical experiments involve B4C materials, and none involve Boralcan in a thin plate shape. The experiments designed as part of IER-554 are intended to be performed at the Sandia Pulsed Reactor Facility/Critical Experiments (SPRF/CX) apparatus at Sandia National Laboratories. SPRF/CX is a well-characterized assembly considered trustworthy by the benchmarking community because of the numerous high-quality evaluations with very low experimental uncertainties included in the ICSBEP handbook. Figure 1 illustrates the assembly for a particular configuration from LCT-078, one of the published benchmarks in the ICSBEP handbook. The experiments will be moderated and reflected by light water at ambient atmospheric pressure. Before the plates are inserted in the critical assembly, they will be characterized with x-ray computed tomography (XCT) to identify the sizes and distribution of the B 4 C powder particles inside the plates. The preliminary design calculations were performed with SCALE 6.3.0 using the KENO V.a sequence for the criticality calculations, the CE-TSUNAMI-3D sequence for the sensitivity and uncertainty studies, and the ENDF/B-VII.1 Continuous Energy cross section library. All the calculation results presented have 10 pcm statistical uncertainties. A set of critical experiments with Boralcan plates could increase the confidence of the criticality safety community in its modeling methods when using this type of neutron absorber material. Experiments would also help validate k eff calculations, and the industry could use these validations to change the B loading credit limits from the US Nuclear Regulatory Commission standard review plan for dry cask storage of spent nuclear fuel. This paper summarizes only part of the analysis documented in the preliminary design report.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2011

Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2011

Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance and, to some extent, experiment management are obsolete, inconsistent with the state of modern nuclear engineering practice, and are becoming increasingly difficult to properly verify and validate (V&V). Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In 2009 the Idaho National Laboratory (INL) initiated a focused effort to address this situation through the introduction of modern high-fidelity computational software and protocols, with appropriate V&V, within the next 3-4 years via the ATR Core Modeling and Simulation and V&V Update (or 'Core Modeling Update') Project. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF). The ATR Core Modeling Update Project, targeted for full implementation in phase with the anticipated ATR Core Internals Changeout (CIC) in the 2014 time frame, began during the last quarter of Fiscal Year 2009, and has just completed its first full year. Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (SCALE, KENO-6, HELIOS, NEWT, and ATTILA) have been installed at the INL under various permanent sitewide license agreements and corresponding baseline models of the ATR and ATRC are now operational, demonstrating the basic feasibility of these code packages for their intended purpose. Furthermore, a capability for rigorous sensitivity analysis and uncertainty quantification based on the TSUNAMI system is being implemented and initial computational results have been obtained. This capability will have many applications in 2011 and beyond as a tool for understanding the margins of uncertainty in the new models as well as for validation experiment design and interpretation. Finally we note that although full implementation of the new computational models and protocols will extend over a period 3-4 years as noted above, interim applications in the much nearer term have already been demonstrated. In particular, these demonstrations included an analysis that was useful for understanding the cause of some issues in December 2009 that were triggered by a larger than acceptable discrepancy between the measured excess core reactivity and a calculated value that was based on the legacy computational methods. As the Modeling Update project proceeds we anticipate further such interim, informal, applications in parallel with formal qualification of the system under the applicable INL Quality Assurance procedures and standards.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2012

Phase II Characterization Survey of the USNS Bridge (T-AOE 10), Military Sealift Fleet Support Command, Naval Station, Norfolk, Virginia (Final Report)

In March 2011, the USNS Bridge was deployed off northeastern Honshu, Japan with the carrier USS Ronald Reagan to assist with relief efforts after the 2011 Tōhoku earthquake and tsunami. During that time, the Bridge was exposed to air-borne radioactive materials leaking from the damaged Fukushima I Nuclear Power Plant. The proximity of the Bridge to the air-borne impacted area resulted in the contamination of the ship’s air-handling systems and the associated components, as well as potential contamination of other ship surfaces due to either direct intake/deposition or inadvertent spread from crew/operational activities. Preliminary surveys in the weeks after the event confirmed low-level contamination within the heating, ventilation, and air conditioning (HVAC) ductwork and systems, and engine and other auxiliary air intake systems. Some partial decontamination was performed at that time. In response to the airborne contamination event, Military Sealift Fleet Support Command (MSFSC) contracted Oak Ridge Associated Universities (ORAU), under provisions of the Oak Ridge Institute for Science and Education (ORISE) contract, to assess the radiological condition of the Bridge. Phase I identified contamination within the CPS filters, ventilation systems, miscellaneous equipment, and other suspect locations that could not accessed at that time (ORAU 2011b). Because the Bridge was underway during the characterization, all the potentially impacted systems/spaces could not be investigated. As a result, MSFSC contracted with ORAU to perform Phase II of the characterization, specifically to survey systems/spaces previously inaccessible. During Phase II of the characterization, the ship was in port to perform routine maintenance operations, allowing access to the previously inaccessible systems/spaces.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗

DOE OSTI2013

DOE Robotic and Remote Systems Assistance to the Government of Japan

At the request of the Government of Japan, DOE did a complex wide survey of available remotely operated and robotic systems to assist in the initial assessment of the damage to the Fukushima Daiichi reactors following an earthquake and subsequent tsunami. As a result several radiation hardened cameras and a Talon robot were identified as systems that could immediately assist in the effort and were subsequently sent to Japan. These systems were transferred to the Government of Japan and used to map radiation levels surrounding the damaged facilities. This report describes the equipment, its use, data collected, and lessons learned from the experience.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2012

Direct Drive Wave Energy Buoy – 33 rd Scale Tank Testing

Columbia Power Technologies (ColPwr) and Oregon State University (OSU) jointly conducted a series of tests in the Tsunami Wave Basin (TWB) at the O.H. Hinsdale Wave Research Laboratory (HWRL). These tests were run between November 2010 and February 2011. Models at 33rd scale representing Columbia Power’s Manta series Wave Energy Converter (WEC) were moored in configurations of one, three and five WEC arrays, with both regular waves and irregular seas generated. The primary research interest of ColPwr is the characterization of WEC response. The WEC response will be investigated with respect to power performance, range of motion and generator torque/speed statistics. The experimental results will be used to validate a numerical model. The primary research interests of OSU include an investigation into the effects of the WEC arrays on the near- and far-field wave propagation. This report focuses on the characterization of the response of a single WEC in isolation. To facilitate understanding of the commercial scale WEC, results will be presented as full scale equivalents.

16 TIDAL AND WAVE POWER↗

DOE OSTI2013

A Brief Review of Past INL Work Assessing Radionuclide Content in TMI-2 Melted Fuel Debris: The Use of 144 Ce as a Surrogate for Pu Accountancy

This report serves as a literature review of prior work performed at Idaho National Laboratory, and its predecessor organizations Idaho National Engineering Laboratory (INEL) and Idaho National Engineering and Environmental Laboratory (INEEL), studying radionuclide partitioning within the melted fuel debris of the reactor of the Three Mile Island 2 (TMI-2) nuclear power plant. The purpose of this review is to document prior published work that provides supporting evidence of the utility of using 144 Ce as a surrogate for plutonium within melted fuel debris. When the TMI-2 accident occurred no quantitative nondestructive analysis (NDA) techniques existed that could assay plutonium in the unconventional wastes from the reactor. However, unpublished work performed at INL by D. W. Akers in the late 1980s through the 1990s demonstrated that passive gamma-ray spectrometry of 144 Ce could potentially be used to develop a semi-quantitative correlation for estimating plutonium content in these materials. The fate and transport of radioisotopes in fuel from different regions of the core, including uranium, fission products, and actinides, appear to be well characterized based on the maximum temperature reached by fuel in different parts of the core and the melting point, boiling point, and volatility of those radioisotopes. Also, the chemical interactions between fuel, fuel cladding, control elements, and core structural components appears to have played a large role in determining when and how fuel relocation occurred in the core; perhaps the most important of these reaction appears to be related to the formation of mixed-material alloys, eutectics, in the fuel cladding. Because of its high melting point, low volatility, and similar chemical behavior to plutonium, the element cerium appears to have behaved similarly to plutonium during the evolution of the TMI-2 accident. Anecdotal evidence extrapolated from open-source literature strengthens this logical feasibility for using cerium, which is rather easy to analyze using passive nondestructive analysis gamma-ray spectrometry, as a surrogate for plutonium in the final analysis of TMI-2 melted fuel debris. The generation of this report is motivated by the need to perform nuclear material accountancy measurements on the melted fuel debris that will be excavated from the damaged nuclear reactors at the Fukushima Daiichi nuclear power plant in Japan, which were destroyed by the Tohoku earthquake and tsunami on March 11, 2011. Lessons may be taken from prior U.S. work related to the study of the TMI-2 core debris to support the development of new assay methods for use at Fukushima Daiichi. While significant differences exist between the two reactor systems (pressurized water reactor (TMI-2) versus boiling water reactor (FD), fresh water post-accident cooing (TMI-2) versus salt water (FD), maintained containment (TMI-2) versus loss of containment (FD)) there remain sufficient similarities to motivate these comparisons.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2011

Summary of International Waste Management Programs (LLNL Input to SNL L3 MS: System-Wide Integration and Site Selection Concepts for Future Disposition Options for HLW)

The Used Fuel Disposition Campaign (UFDC) within the Department of Energy’s Office of Nuclear Energy (DOE-NE) Fuel Cycle Technology (FCT) program has been tasked with investigating the disposal of the nation’s spent nuclear fuel (SNF) and high-level nuclear waste (HLW) for a range of potential waste forms and geologic environments. This Lessons Learned task is part of a multi-laboratory effort, with this LLNL report providing input to a Level 3 SNL milestone (System-Wide Integration and Site Selection Concepts for Future Disposition Options for HLW). The work package number is: FTLL11UF0328; the work package title is: Technical Bases / Lessons Learned; the milestone number is: M41UF032802; and the milestone title is: “LLNL Input to SNL L3 MS: System-Wide Integration and Site Selection Concepts for Future Disposition Options for HLW”. The system-wide integration effort will integrate all aspects of waste management and disposal, integrating the waste generators, interim storage, transportation, and ultimate disposal at a repository site. The review of international experience in these areas is required to support future studies that address all of these components in an integrated manner. Note that this report is a snapshot of nuclear power infrastructure and international waste management programs that is current as of August 2011, with one notable exception. No attempt has been made to discuss the currently evolving world-wide response to the tragic consequences of the earthquake and tsunami that devastated Japan on March 11, 2011, leaving more than 15,000 people dead and more than 8,000 people missing, and severely damaging the Fukushima Daiichi nuclear power complex. Continuing efforts in FY 2012 will update the data, and summarize it in an Excel spreadsheet for easy comparison and assist in the knowledge management of the study cases.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

Advanced Fuels Campaign: Light Water Reactor Accident Tolerant Fuel Performance Metrics

The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the United States’ nuclear industry. As a result, continual improvement of technology, including advanced materials and nuclear fuels, remains central to industry’s success. Decades of research combined with continual operation have produced steady advancements in technology and yielded an extensive base of data, experience, and knowledge on light water reactor (LWR) fuel performance under both normal and accident conditions. In 2011, following the Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex, enhancing the accident tolerance of LWRs became a topic of serious discussion. As a result of direction from the U.S. Congress, the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) initiated an Accident Tolerant Fuel (ATF) Development program. The complex multiphysics behavior of LWR nuclear fuel makes defining specific material or design improvements difficult; as such, establishing qualitative attributes is critical to guide the design and development of fuels and cladding with enhanced accident tolerance. This report summarizes a common set of technical evaluation metrics to aid in the optimization and down selection of candidate designs. As used herein, “metrics” describe a set of technical bases by which multiple concepts can be fairly evaluated against a common baseline and against one another. Furthermore, this report describes a proposed technical evaluation methodology that can be applied to assess the ability of each concept to meet performance and safety goals relative to the current UO 2 – zirconium alloy system and relative to one another. The resultant ranked evaluation can then inform concept down-selection, such that the most promising accident tolerant fuel design option(s) can continue to be developed for lead test rod or lead test assembly insertion into a commercial reactor within the desired timeframe (by 2022).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

Fukushima Daiichi—A Case Study for BWR Instrumentation and Control Systems Performance During a Severe Accident, Rev 1

This report provides an assessment of the performance of various instrumentation and control (I&C) and monitoring systems, with a focus on the first few days of the Fukushima Daiichi accident starting on March 11, 2011. Similar to what occurred at the Three Mile Island Unit 2 (TMI-2) accident in 1979, operator responses were challenged when information on key parameters was lost, erroneous, or misleading. While numerous improvements in instrumentation capabilities have been made since the TMI-2 event, the accident at Fukushima Daiichi suggests the potential for further improvement. The objective of this research is to review the performance of the Fukushima Daiichi Units 1–3 instrumentation systems that could have helped ensure that plant operators and emergency responders had relevant, accurate, and timely knowledge of plant conditions as they managed the accident response. As was the case with TMI-2, forensic analyses of plant instrumentation failures will be a long-term effort. As an introduction to this long-term research, this report focuses on the information available in the control room during the accident and illustrates the challenges to operators in their understanding of plant conditions during the progression of the accident. I&C systems monitor the success or failure of achievement of light water reactor safety principles in which the protection of the reactor fuel, reactor coolant system, and reactor containment is accomplished. This report (1) provides an overview of the safety principles; (2) notes parameters associated with the boiling water reactor (BWR)/3 and BWR/4 protection, actuation, and monitoring systems as they relate to the achievement of these principles; (3) presents data from past research on predicted instrumentation performance during severe accidents, and (4) assesses performance of the key I&C and monitoring systems associated with BWR/3 and BWR/4 reactors like those at Fukushima Daiichi. Lessons learned or potential improvements associated with the performance of these systems in the context of severe accidents are delineated. As background information, an overview of key BWR/3- and BWR/4-Mark I reactor designs (BWR/3 and BWR/4 with a Mark I containment) reactor protection, engineered safety features actuation, and accident monitoring systems is provided. A brief description of plant safety systems used to prevent or mitigate reactor accidents is also provided. This background information provides context for the brief discussion of previous research that has been conducted on BWR severe accidents and for the discussion of the accident sequences, which led to core damage at Fukushima Daiichi Units 1–3. The progression of the accident sequences at Fukushima Daiichi Units 1–3 occurred very much as predicted in prior severe accident research conducted in the United States and internationally, given the complete station blackout that occurred when the tsunami generated by one of the most powerful earthquakes ever recorded rolled onshore.

Accident Tolerant Equipment↗

DOE OSTI2014

Advanced Small Modular Reactors: Prototype Development Capabilities of 3D Spatial Interactions and Failures During Scenario Simulation

Computers have been used for 3D modeling and simulation, but only recently have computational resources been able to give realistic results in a reasonable time frame for large complex models. This report addressed the methods, techniques, and resources used to develop a prototype for using 3D modeling and simulation engine to improve risk analysis and evaluate reactor structures and components for a given scenario. The simulations done for this evaluation were focused on external events, specifically tsunami floods, for a hypothetical nuclear power facility on a coastline.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2014

3D Modeling Engine Representation Summary Report

Computers have been used for 3D modeling and simulation, but only recently have computational resources been able to give realistic results in a reasonable time frame for large complex models. This summary report addressed the methods, techniques, and resources used to develop a 3D modeling engine to represent risk analysis simulation for advanced small modular reactor structures and components. The simulations done for this evaluation were focused on external events, specifically tsunami floods, for a hypothetical nuclear power facility on a coastline.

3D Modeling↗

DOE OSTI2014

Demonstration of NonLinear Seismic Soil Structure Interaction and Applicability to New System Fragility Seismic Curves

Risk calculations should focus on providing best estimate results, and associated insights, for evaluation and decision-making. Specifically, seismic probabilistic risk assessments (SPRAs) are intended to provide best estimates of the various combinations of structural and equipment failures that can lead to a seismic induced core damage event. However, in general this approach has been conservative, and potentially masks other important events (for instance, it was not the seismic motions that caused the Fukushima core melt events, but the tsunami ingress into the facility). SPRAs are performed by convolving the seismic hazard (the frequency of certain magnitude events) with the seismic fragility (the conditional probability of failure of a structure, system, or component given the occurrence of earthquake ground motion). In this calculation, there are three main pieces to seismic risk quantification, 1) seismic hazard and nuclear power plants (NPPs) response to the hazard, fragility or capacity of structures, systems and components (SSC), and systems analysis. Figure 1 provides a high level overview of the risk quantification process. The focus of this research is on understanding and removing conservatism (when possible) in the quantification of seismic risk at NPPs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2015

Lotic Water Hydrodynamic Model

Water-related natural disasters, for example, floods and droughts, are among the most frequent and costly natural hazards, both socially and economically. Many of these floods are a result of excess rainfall collecting in streams and rivers, and subsequently overtopping banks and flowing overland into urban environments. Floods can cause physical damage to critical infrastructure and present health risks through the spread of waterborne diseases. Los Alamos National Laboratory (LANL) has developed Lotic, a state-of-the-art surface water hydrodynamic model, to simulate propagation of flood waves originating from a variety of events. Lotic is a two-dimensional (2D) flood model that has been used primarily for simulations in which overland water flows are characterized by movement in two dimensions, such as flood waves expected from rainfall-runoff events, storm surge, and tsunamis. In 2013, LANL developers enhanced Lotic through several development efforts. These developments included enhancements to the 2D simulation engine, including numerical formulation, computational efficiency developments, and visualization. Stakeholders can use simulation results to estimate infrastructure damage and cascading consequences within other sets of infrastructure, as well as to inform the development of flood mitigation strategies.

16 TIDAL AND WAVE POWER↗

DOE OSTI2014

Advanced Seismic Probabilistic Risk Assessment Demonstration Project Plan

Idaho National Laboratories (INL) has an ongoing research and development (R&D) project to remove excess conservatism from seismic probabilistic risk assessments (SPRA) calculations. These risk calculations should focus on providing best estimate results, and associated insights, for evaluation and decision-making. This report presents a plan for improving our current traditional SPRA process using a seismic event recorded at a nuclear power plant site, with known outcomes, to improve the decision making process. SPRAs are intended to provide best estimates of the various combinations of structural and equipment failures that can lead to a seismic induced core damage event. However, in general this approach has been conservative, and potentially masks other important events (for instance, it was not the seismic motions that caused the Fukushima core melt events, but the tsunami ingress into the facility).

58 GEOSCIENCES↗