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DOE OSTI2013

Report on Project Action Sheet PP05 task 3 between the U.S. Department of Energy and the Republic of Korea Ministry of Education, Science, and Technology (MEST)

This report documents the results of Task 3 of Project Action Sheet PP05 between the United States Department of Energy (DOE) and the Republic of Korea (ROK) Ministry of Education, Science, and Technology (MEST) for Support with Review of an ROK Risk Evaluation Process. This task was to have Sandia National Laboratories collaborate with the Korea Institute of Nuclear Nonproliferation and Control (KINAC) on several activities concerning how to determine the Probability of Neutralization, P N , and the Probability of System Effectiveness, P E , to include: providing descriptions on how combat simulations are used to determine P N and P E ; comparisons of the strengths and weaknesses of two neutralization models (the Neutralization.xls spreadsheet model versus the Brief Adversary Threat-Loss Estimator (BATLE) software); and demonstrating how computer simulations can be used to determine P N . Note that the computer simulation used for the demonstration was the Scenario Toolkit And Generation Environment (STAGE) simulation, which is a stand-alone synthetic tactical simulation sold by Presagis Canada Incorporated. The demonstration is provided in a separate Audio Video Interleave (.AVI) file.

Snell, Mark Kamerer↗
DOE OSTI2013

Medical Radioisotopes [Slides]

Medical radioisotopes are an important part of the physician’s diagnostic and therapeutic toolkit; - LANL is a major producer of Sr-82 and Ge-68 for clinical use; and - Ongoing R&D aimed at establishing new isotopes to improve human health.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2013

Monte Carlo Simulation Tool Installation and Operation Guide

This document provides information on software and procedures for Monte Carlo simulations based on the Geant4 toolkit, the ROOT data analysis software and the CRY cosmic ray library. These tools have been chosen for its application to shield design and activation studies as part of the simulation task for the Majorana Collaboration. This document includes instructions for installation, operation and modification of the simulation code in a high cyber-security computing environment, such as the Pacific Northwest National Laboratory network. It is intended as a living document, and will be periodically updated. It is a starting point for information collection by an experimenter, and is not the definitive source. Users should consult with one of the authors for guidance on how to find the most current information for their needs.

CRY↗
DOE OSTI2013

RELAP-7: Demonstrating the integration of two-phase flow components for an ideal BWR loop

The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at the Idaho National Laboratory. RELAP-7 will become the main reactor systems toolkit for the Risk-Informed Safety Margin Characterization Pathway of the Light Water Reactor Sustainability (LWRS) Program and the next generation tool in the RELAP reactor safety/systems analysis application series (i.e., the replacement for RELAP5). The code is being developed based on Idaho National Laboratory’s modern scientific software development framework – MOOSE (the Multi-Physics Object-Oriented Simulation Environment). During Fiscal-Year 2013, a number of physical components have been developed to support the demonstration calculations presented in this report. The case selected for demonstration of an idealized BWR loop in RELAP-7 includes the major components for the primary system of a BWR. The BWR loop has been successfully simulated to steady state with the RELAP-7 code. The next major stage of development is to demonstrate two-phase flow modeling capability through a simplified boiling water reactor station blackout analysis, which will be reported in the subsequent demonstration simulation report.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

RELAP-7 Simulation Resolving an SBO Scenario on a Simplified Geometry of a BWR

The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at the Idaho National Laboratory (INL). RELAP-7 will become the main reactor systems toolkit for the Risk-Informed Safety Margin Characterization Pathway of the Light Water Reactor Sustainability Program and the next generation tool in the RELAP reactor safety/systems analysis application series (i.e., the replacement for RELAP5). The code is being developed based on Idaho National Laboratory’s modern scientific software development framework – MOOSE (the Multi-Physics Object-Oriented Simulation Environment). During Fiscal Year 2013, a number of physical components with two-phase flow capability have been developed to support the simplified boiling water reactor (BWR) station blackout analyses. The case selected for demonstration calculation is built from the specifications documented in an Organization for Economic Cooperation and Development benchmark problem for BWR turbine trip analysis. The reference design for the benchmark problem was from the Peach Bottom-2 nuclear station, which is a General Electric BWR-4 design. The demonstration case includes the major components for the primary system of a BWR, as well as the safety system components for reactor core isolation cooling (RCIC) and the wet well of a BWR containment. The case was initially run to steady-state with RELAP-7. The station blackout transient simulations were subsequently initiated by using the INL-developed RAVEN code. Two scenarios for the station blackout simulations have been considered. Scenario I represents an extreme station blackout accident with no safety injection functioning. The reactor core could experience dry out fairly quickly with this scenario. Scenario II represents a more probable station blackout accident progression with the RCIC system functioning. In this scenario, the RCIC system is fully coupled with the reactor primary system and the safety injection to provide makeup cooling water to the reactor core from the suppression pool is dynamically simulated. With the RCIC system functioning, the core dry out is significantly postponed when compared to the results from Scenario I. This fully coupled RCIC system simulation capability represents the first-of-a-kind simulation capability. Sensitivity studies have also been carried out to study the effect of RCIC control strategy with varying core makeup cooling water mass flow rates. The next stage of development will be to demonstrate more refined BWR station blackout analyses with more realistic geometries, to be reported in the next demonstration simulation report.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Light Water Reactor Sustainability Program FY13 Status Update for EPRI - RISMC Collaboration

The purpose of the Risk Informed Safety Margin Characterization (RISMC) Pathway research and development (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. Goals of the RISMC Pathway are twofold: (1) Develop and demonstrate a risk-assessment method coupled to safety margin quantification that can be used by NPP decision makers as part of their margin recovery strategies. (2) Create an advanced "RISMC toolkit" that enables more accurate representation of NPP safety margin. In order to carry out the R&D needed for the Pathway, the Idaho National Laboratory (INL) is collaborating with the Electric Power Research Institute (EPRI) in order to focus on applications of interest to the U.S. nuclear power industry. This report documents the collaboration activities performed between INL and EPRI during FY2013.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2014

A pervasive parallel framework for visualization: final report for FWP 10-014707

We are on the threshold of a transformative change in the basic architecture of highperformance computing. The use of accelerator processors, characterized by large core counts, shared but asymmetrical memory, and heavy thread loading, is quickly becoming the norm in high performance computing. These accelerators represent significant challenges in updating our existing base of software. An intrinsic problem with this transition is a fundamental programming shift from message passing processes to much more fine thread scheduling with memory sharing. Another problem is the lack of stability in accelerator implementation; processor and compiler technology is currently changing rapidly. This report documents the results of our three-year ASCR project to address these challenges. Our project includes the development of the Dax toolkit, which contains the beginnings of new algorithms for a new generation of computers and the underlying infrastructure to rapidly prototype and build further algorithms as necessary.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2014

NOVA making stuff: Season 2

Over the course of four weeks in fall 2013, 11.7 million Americans tuned in to PBS to follow host David Pogue as he led them in search of engineering and scientific breakthroughs poised to change our world. Levitating trains, quantum computers, robotic bees, and bomb-detecting plants—these were just a few of the cutting-edge innovations brought into the living rooms of families across the country in NOVA’s four-part series, Making Stuff: Faster, Wilder, Colder, and Safer. Each of the four one-hour programs gave viewers a behind-the-scenes look at novel technologies poised to change our world—showing them how basic research and scientific discovery can hold the keys to transforming how we live. Making Stuff Season 2 (MS2) combined true entertainment with educational value, creating a popular and engaging series that brought accessible science into the homes of millions. NOVA’s goal to engage the public with such technological innovation and basic research extended beyond the broadcast series, including a variety of online, educational, and promotional activities: original online science reporting, web-only short-form videos, a new online quiz-game, social media engagement and promotion, an educational outreach “toolkit” for science educators to create their own “makerspaces,” an online community of practice, a series of nationwide Innovation Cafés, educator professional development, a suite of teacher resources, an “Idealab,” participation in national conferences, and specialized station relation and marketing. A summative evaluation of the MS2 project indicates that overall, these activities helped make a significant impact on the viewers, users, and participants that NOVA reached. The final evaluation conducted by Concord Evaluation Group (CEG) confidently concluded that the broadcast, website, and outreach activities were successful at achieving the project’s intended impacts. CEG reported that the MS2 series and website content were successful in raising awareness and sparking interest in innovation, and increased public awareness that basic research leads to technological innovation; this interest was also sustained over a six month period. Efforts to create an online community of practice were also successful: the quality of collaboration increased, and community members felt supported while using Maker pedagogy. These findings provide clear evidence that large-scale science media projects like MS2 are an effective means of “moving the needle” on attitudes about and excitement for science. NOVA’s broadcast audience and ratings have always indicated that a large portion of the population is interested in and engages with educational science media on a weekly basis. Yet these evaluation results provide the empirical evidence that beyond being capable of attracting, maintaining, and growing a dedicated group of citizens interested in science, these shows—with their diverse content provided on a variety of media channels—are capable of sparking new interest in science, raising public awareness of the importance of science, and maintaining and growing that interest over time. In a country where approximately a quarter of the population doesn’t know the earth rotates around the sun,1 roughly half still don’t accept evolution,2 and about 20% don’t think climate change is happening,3 the importance of these findings cannot be overstated. The success of MS2 suggests that large-scale media projects dedicated to and linked by coverage of scientific “big ideas” are an effective means of shifting public opinion on—and improving understanding of—science. REFERENCES 1, 2 National Science Foundation, Science and Engineering Indicators (2014). Chapter 7: Science and Technology: Public Attitudes and Understanding. 3 Leiserowitz, A., Maibach, E., Roser-Renouf, C., Feinberg, G., & Rosenthal, S. (2014) Climate change in the American mind: April, 2014. Yale University and George Mason University. New Haven, CT: Yale Project on Climate Change Communication.

36 MATERIALS SCIENCE↗
DOE OSTI2014

Accident Tolerant Fuel Analysis

Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about light water reactor design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the Risk Informed Safety Margin Characterization (RISMC) Pathway research and development (R&D) is to support plant decisions for risk-informed margins management by improving economics and reliability, and sustaining safety, of current NPPs. Goals of the RISMC Pathway are twofold: (1) Develop and demonstrate a risk-assessment method coupled to safety margin quantification that can be used by NPP decision makers as part of their margin recovery strategies. (2) Create an advanced “RISMC toolkit” that enables more accurate representation of NPP safety margin. In order to carry out the R&D needed for the Pathway, the Idaho National Laboratory is performing a series of case studies that will explore methods- and tools-development issues, in addition to being of current interest in their own right. One such study is a comparative analysis of safety margins of plants using different fuel cladding types: specifically, a comparison between current-technology Zircaloy cladding and a notional “accident-tolerant” (e.g., SiC-based) cladding. The present report begins the process of applying capabilities that are still under development to the problem of assessing new fuel designs. The approach and lessons learned from this case study will be included in future Technical Basis Guides produced by the RISMC Pathway. These guides will be the mechanism for developing the specifications for RISMC tools and for defining how plant decision makers should propose and evaluate margin recovery strategies.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2014

HyRAM Testing Strategy and Quality Design Elements

Strategy document and tentative schedule for testing of HyRAM, a software toolkit that integrates data and methods relevant to assessing the safety of hydrogen fueling and storage infrastructure. Because proposed and existing features in HyRAM that support testing are important factors in this discussion, relevant design considerations of HyRAM are also discussed. However, this document does not cover all of HyRAM desig n, nor is the full HyRAM software development schedule included.

Reynolds, John Thomas↗
DOE OSTI2014

Final Technical Report -- Bridging the PSI Knowledge Gap: A Multiscale Approach

The Plasma Surface Interactions (PSI) Science Center formed by the grant undertook a multidisciplinary set of studies on the complex interface between the plasma and solid states of matter. The strategy of the center was to combine and integrate the experimental, diagnostic and modeling toolkits from multiple institutions towards specific PSI problems. In this way the Center could tackle integrated science issues which were not addressable by single institutions, as well as evolve the underlying science of the PSI in a more general way than just for fusion applications. The overall strategy proved very successful. The research result and highlights of the MIT portion of the Center are primarily described. A particular highlight is the study of tungsten nano-tendril growth in the presence of helium plasmas. The Center research provided valuable new insights to the mechanisms controlling the nano-tendrils by developing coupled modeling and in situ diagnostic methods which could be directly compared. For example, the role of helium accumulation in tungsten distortion in the surface was followed with unique in situ helium concentration diagnostics developed. These depth-profiled, time-resolved helium concentration measurements continue to challenge the numerical models of nano-tendrils. The Center team also combined its expertise on tungsten nano-tendrils to demonstrate for the first time the growth of the tendrils in a fusion environment on the Alcator C-Mod fusion experiment, thus having significant impact on the broader fusion research effort. A new form of isolated nano-tendril “columns” were identified which are now being used to understand the underlying mechanisms controlling the tendril growth. The Center also advanced PSI science on a broader front with a particular emphasis on developing a wide range of in situ PSI diagnostic tools at the DIONISOS facility at MIT. For example the strong suppression of sputtering by the certain combination of light-species plasmas and metals was experimentally studied with independent measurement methods across the Center. This surprising result challenges the universal use of the binary-collision approximation in sputtering predictions and continues to be the subject of study. In order to address this issue MIT developed a new in situ erosion measurement technique based on ion beam analysis which can be used at elevated material temperatures. This exciting new technique is now being used to study material erosion in high performance plasma thrusters for space exploration and is being adopted to fusion experimental devices. This is an indicator of the positive synergies that arise from such a Center, with the research having impact beyond the initial area of study. The Center also served successfully as an organizing force for communication to the science community. The MIT members of the Center provided many high-profile overview presentations at prestigious international conferences and national workshops. The research resulted in three student theses and 24 peer-reviewed publications. PSI research continues to be identified as a critical area for fusion energy.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

Verification and validation plan for the SFR system analysis module

This report documents the Verification and Validation (V&V) Plan for software verification and validation of the SFR System Analysis Module (SAM), developed at Argonne National Laboratory for sodium fast reactor whole-plant transient analysis. SAM is developed under the DOE NEAMS program and is part of the Reactor Product Line toolkit. The SAM code, the phenomena and computational models of interest, the software quality assurance, and the verification and validation requirements and plans are discussed in this report.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2014

RISMC Advanced Safety Analysis Project Plan – FY 2015 - FY 2019

In this report, a project plan is developed, focused on industry applications, using Risk-Informed Safety Margin Characterization (RISMC) tools and methods applied to realistic, relevant, and current interest issues to the operating nuclear fleet. RISMC focuses on modernization of nuclear power safety analysis (tools, methods and data); implementing state-of-the-art modeling techniques (which include, for example, enabling incorporation of more detailed physics as they become available); taking advantage of modern computing hardware; and combining probabilistic and mechanistic analyses to enable a risk informed safety analysis process. The modernized tools will maintain the current high level of safety in our nuclear power plant fleet, while providing an improved understanding of safety margins and the critical parameters that affect them. Thus, the set of tools will provide information to inform decisions on plant modifications, refurbishments, and surveillance programs, while improving economics. This set of tools will also benefit the design of new reactors, enhancing safety per unit cost of a nuclear plant. The proposed plan will focus on application of the RISMC toolkit, in particular, solving realistic problems of important current issues to the nuclear industry, in collaboration with plant owners and operators to demonstrate the usefulness of these tools in decision making.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Accident Tolerant Fuel Analysis

Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about light water reactor design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the Risk Informed Safety Margin Characterization (RISMC) Pathway research and development (R&D) is to support plant decisions for risk-informed margins management by improving economics and reliability, and sustaining safety, of current NPPs. Goals of the RISMC Pathway are twofold: (1) Develop and demonstrate a risk-assessment method coupled to safety margin quantification that can be used by NPP decision makers as part of their margin recovery strategies. (2) Create an advanced “RISMC toolkit” that enables more accurate representation of NPP safety margin. In order to carry out the R&D needed for the Pathway, the Idaho National Laboratory is performing a series of case studies that will explore methods- and tools-development issues, in addition to being of current interest in their own right. One such study is a comparative analysis of safety margins of plants using different fuel cladding types: specifically, a comparison between current-technology Zircaloy cladding and a notional “accident-tolerant” (e.g., SiC-based) cladding. The present report begins the process of applying capabilities that are still under development to the problem of assessing new fuel designs. The approach and lessons learned from this case study will be included in future Technical Basis Guides produced by the RISMC Pathway. These guides will be the mechanism for developing the specifications for RISMC tools and for defining how plant decision makers should propose and evaluate margin recovery strategies.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Enhanced Severe Transient Analysis for Prevention Technical Program Plan

This document outlines the development of a high fidelity, best estimate nuclear power plant severe transient simulation capability that will complement or enhance the integral system codes historically used for licensing and analysis of severe accidents. As with other tools in the Risk Informed Safety Margin Characterization (RISMC) Toolkit, the ultimate user of Enhanced Severe Transient Analysis and Prevention (ESTAP) capability is the plant decision-maker; the deliverable to that customer is a modern, simulation-based safety analysis capability, applicable to a much broader class of safety issues than is traditional Light Water Reactor (LWR) licensing analysis. Currently, the RISMC pathway’s major emphasis is placed on developing RELAP-7, a next-generation safety analysis code, and on showing how to use RELAP-7 to analyze margin from a modern point of view: that is, by characterizing margin in terms of the probabilistic spectra of the “loads” applied to systems, structures, and components (SSCs), and the “capacity” of those SSCs to resist those loads without failing. The first objective of the ESTAP task, and the focus of one task of this effort, is to augment RELAP-7 analyses with user-selected multi-dimensional, multi-phase models of specific plant components to simulate complex phenomena that may lead to, or exacerbate, severe transients and core damage. Such phenomena include: coolant crossflow between PWR assemblies during a severe reactivity transient, stratified single or two-phase coolant flow in primary coolant piping, inhomogeneous mixing of emergency coolant water or boric acid with hot primary coolant, and water hammer. These are well-documented phenomena associated with plant transients but that are generally not captured in system codes. They are, however, generally limited to specific components, structures, and operating conditions. The second ESTAP task is to similarly augment a severe (post-core damage) accident integral analyses code with high fidelity simulations that would allow investigation of multi-dimensional, multi-phase containment phenomena that are only treated approximately in established codes.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

PFLOTRAN User Manual: A Massively Parallel Reactive Flow and Transport Model for Describing Surface and Subsurface Processes

PFLOTRAN solves a system of generally nonlinear partial differential equations describing multi-phase, multicomponent and multiscale reactive flow and transport in porous materials. The code is designed to run on massively parallel computing architectures as well as workstations and laptops (e.g. Hammond et al., 2011). Parallelization is achieved through domain decomposition using the PETSc (Portable Extensible Toolkit for Scientific Computation) libraries for the parallelization framework (Balay et al., 1997). PFLOTRAN has been developed from the ground up for parallel scalability and has been run on up to 218 processor cores with problem sizes up to 2 billion degrees of freedom. Written in object oriented Fortran 90, the code requires the latest compilers compatible with Fortran 2003. At the time of this writing this requires gcc 4.7.x, Intel 12.1.x and PGC compilers. As a requirement of running problems with a large number of degrees of freedom, PFLOTRAN allows reading input data that is too large to fit into memory allotted to a single processor core. The current limitation to the problem size PFLOTRAN can handle is the limitation of the HDF5 file format used for parallel IO to 32 bit integers. Noting that 2 32 = 4; 294; 967; 296, this gives an estimate of the maximum problem size that can be currently run with PFLOTRAN. Hopefully this limitation will be remedied in the near future.

58 GEOSCIENCES↗