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DOE OSTI2015

Stochastic Modeling of Radioactive Material Releases

Nonreactor nuclear facilities operated under the approval authority of the U.S. Department of Energy use unmitigated hazard evaluations to determine if potential radiological doses associated with design basis events challenge or exceed dose evaluation guidelines. Unmitigated design basis events that sufficiently challenge dose evaluation guidelines or exceed the guidelines for members of the public or workers, merit selection of safety structures, systems, or components or other controls to prevent or mitigate the hazard. Idaho State University, in collaboration with Idaho National Laboratory, has developed a portable and simple to use software application called SODA (Stochastic Objective Decision-Aide) that stochastically calculates the radiation dose associated with hypothetical radiological material release scenarios. Rather than producing a point estimate of the dose, SODA produces a dose distribution result to allow a deeper understanding of the dose potential. SODA allows users to select the distribution type and parameter values for all of the input variables used to perform the dose calculation. SODA then randomly samples each distribution input variable and calculates the overall resulting dose distribution. In cases where an input variable distribution is unknown, a traditional single point value can be used. SODA was developed using the MATLAB coding framework. The software application has a graphical user input. SODA can be installed on both Windows and Mac computers and does not require MATLAB to function. SODA provides improved risk understanding leading to better informed decision making associated with establishing nuclear facility material-at-risk limits and safety structure, system, or component selection. It is important to note that SODA does not replace or compete with codes such as MACCS or RSAC, rather it is viewed as an easy to use supplemental tool to help improve risk understanding and support better informed decisions. The work was funded through a grant from the DOE Nuclear Safety Research and Development Program.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2015

Embedded Sensors and Controls to Improve Component Performance and Reliability -- Bench-scale Testbed Design Report

Embedded instrumentation and control systems that can operate in extreme environments are challenging due to restrictions on sensors and materials. As a part of the Department of Energy's Nuclear Energy Enabling Technology cross-cutting technology development programs Advanced Sensors and Instrumentation topic, this report details the design of a bench-scale embedded instrumentation and control testbed. The design goal of the bench-scale testbed is to build a re-configurable system that can rapidly deploy and test advanced control algorithms in a hardware in the loop setup. The bench-scale testbed will be designed as a fluid pump analog that uses active magnetic bearings to support the shaft. The testbed represents an application that would improve the efficiency and performance of high temperature (700 C) pumps for liquid salt reactors that operate in an extreme environment and provide many engineering challenges that can be overcome with embedded instrumentation and control. This report will give details of the mechanical design, electromagnetic design, geometry optimization, power electronics design, and initial control system design.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

10 MW, L-Band Klystron for Accelerators

This program developed a 10 MW, pulsed, Annular Beam Klystron (ABK) for accelerator applications. This is an alternative RF source to multiple beam klystrons MBKs), which are more complex and considerably more expensive. The ABK uses a single, annular cathode and a single beam tunnel with fundamental mode cavities. The operating specifications (voltage, efficiency, power, bndwidth, duty, etc.) are the same as for comparable MBKs.

43 PARTICLE ACCELERATORS↗

DOE OSTI2016

Multifamily Housing Rehabilitation Process Improvements

Rea Ventures Group, LLC, (Rea Ventures) partnered with Southface Energy Institute (Southface) on the rehabilitation of 418 low-income rental multifamily apartments located at 14 different properties in Georgia (Climate Zones 2-4). These 22-year old, individually-metered units were arranged in rowhouse or townhouse style units. Rehabilitation plans were developed using a process prescribed by the US Department of Agriculture (USDA) Rural Development program, who partially funded the building upgrades. The USDA is responsible for building, upgrading, and subsidizing housing in rural areas nationwide. In 2012, over $100 million was allocated in grants and loans. Due to the unique financing mechanism as well as long-term ownership requirements, property owners are especially motivated to invest in upgrades that will increase durability and tenant retention. These buildings represent a large stock of rural affordable housing that have the potential for significant energy and cost savings for property owners and tenants. Southface analyzed the energy upgrade potential of one stereotypical property in the Rea Ventures portfolio. This study will provide insight into the most cost-effective, implementable energy efficiency and durability upgrades for this age multifamily housing, having an enormous impact not only on the portfolio of Rea Ventures but on the vast USDA and larger Federal portfolio. Additionally, Southface will identify gaps in the current capital needs assessment process, examine available audit and simulation tools and protocols, and evaluate additional auditor training or certification needs.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗

DOE OSTI2015

Advanced Burner Reactor with Breed-and-Burn Thorium Blankets for Improved Economics and Resource Utilization

This study assesses the feasibility of designing Seed and Blanket (S&B) Sodium-cooled Fast Reactor (SFR) to generate a significant fraction of the core power from radial thorium fueled blankets that operate on the Breed-and-Burn (B&B) mode without exceeding the radiation damage constraint of presently verified cladding materials. The S&B core is designed to maximize the fraction of neutrons that radially leak from the seed (or “driver”) into the subcritical blanket and reduce neutron loss via axial leakage. The blanket in the S&B core makes beneficial use of the leaking neutrons for improved economics and resource utilization. A specific objective of this study is to maximize the fraction of core power that can be generated by the blanket without violating the thermal hydraulic and material constraints. Since the blanket fuel requires no reprocessing along with remote fuel fabrication, a larger fraction of power from the blanket will result in a smaller fuel recycling capacity and lower fuel cycle cost per unit of electricity generated. A unique synergism is found between a low conversion ratio (CR) seed and a B&B blanket fueled by thorium. Among several benefits, this synergism enables the very low leakage S&B cores to have small positive coolant voiding reactivity coefficient and large enough negative Doppler coefficient even when using inert matrix fuel for the seed. The benefits of this synergism are maximized when using an annular seed surrounded by an inner and outer thorium blankets. Among the high-performance S&B cores designed to benefit from this unique synergism are: (1) the ultra-long cycle core that features a cycle length of ~7 years; (2) the high-transmutation rate core where the seed fuel features a TRU CR of 0.0. Its TRU transmutation rate is comparable to that of the reference Advanced Burner Reactor (ABR) with CR of 0.5 and the thorium blanket can generate close to 60% of the core power; but requires only one sixth of the reprocessing and fabrication capacity per unit of core power. Nevertheless, these high-performance cores were designed to set upper bounds on the S&B core performance by using larger height and pressure drop than those of typical SFR design. A study was subsequently undertaken to quantify the tradeoff between S&B core design variables and the core performance. This study concludes that a viable S&B core can be designed without significant deviation from SFR core design practices. For example, the S&B core with 120cm active height will be comparable in volume, HM mass and specific power with the S-PRISM core and could fit within the S-PRISM reactor vessel. 43% of this core power will be generated by the once-through thorium blanket; the required capacity for reprocessing and remote fuel fabrication per unit of electricity generated will be approximately one fifth of that for a comparable ABR. The sodium void worth of this 120cm tall S&B core is significantly less positive than that of the reference ABR and the Doppler coefficient is only slightly smaller even though the seed uses a fertile-free fuel. The seed in the high transmutation core requires inert matrix fuel (TRU-40Zr) that has been successfully irradiated by the Fuel Cycle Research & Development program. An additional sensitivity analysis was later conducted to remove the bias introduced by the discrepancy between radiation damage constraints -- 200 DPA applied for S&B cores and fast fluence of 4x1023 n(>0.1MeV)/cm2 applied for ABR core design. Although the performance characteristics of the S&B cores are sensitive to the radiation damage constraint applied, the S&B cores offer very significant performance improvements relative to the conventional ABR core design when using identical constraint.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2014

Development of an Automated Decision-Making Tool for Supervisory Control System

This technical report was generated as a product of the Supervisory Control for Multi-Modular Small Modular Reactor (SMR) Plants project within the Instrumentation, Control and Human-Machine Interface technology area under the Advanced Small Modular Reactor (AdvSMR) Research and Development Program of the US Department of Energy. The report documents the definition of strategies, functional elements, and the structural architecture of a supervisory control system for multi-modular AdvSMR plants. This research activity advances the state of the art by incorporating real-time, probabilistic-based decision-making into the supervisory control system architectural layers through the introduction of a tiered-plant system approach. The report provides background information on the state of the art of automated decision-making, including the description of existing methodologies. It then presents a description of a generalized decision-making framework, upon which the supervisory control decision-making algorithm is based. The probabilistic portion of automated decision-making is demonstrated through a simple hydraulic loop example.

42 ENGINEERING↗

DOE OSTI2017

A physics-based crystallographic modeling framework for describing the thermal creep behavior of Fe-Cr alloys

This Report addresses the Milestone M2MS-16LA0501032 of NEAMS Program (“Develop hardening model for FeCrAl cladding), with a deadline of 09/30/2016. Here we report a constitutive law for thermal creep of FeCrAl. This Report adds to and complements the one for Milestone M3MS-16LA0501034 (“Interface hardening models with MOOSE-BISON”), where we presented a hardening law for irradiated FeCrAl. The last component of our polycrystal-based constitutive behavior, namely, an irradiation creep model for FeCrAl, will be developed as part of the FY17 Milestones, and the three regimes will be coupled and interfaced with MOOSE-BISON.

36 MATERIALS SCIENCE↗

DOE OSTI2016

STREAM2016: Streaming Requirements, Experience, Applications and Middleware Workshop

The Department of Energy (DOE) Office of Science (SC) facilities including accelerators, light sources and neutron sources and sensors that study, the environment, and the atmosphere, are producing streaming data that needs to be analyzed for next-generation scientific discoveries. There has been an explosion of new research and technologies for stream analytics arising from the academic and private sectors. However, there has been no corresponding effort in either documenting the critical research opportunities or building a community that can create and foster productive collaborations. The two-part workshop series, STREAM: Streaming Requirements, Experience, Applications and Middleware Workshop (STREAM2015 and STREAM2016), were conducted to bring the community together and identify gaps and future efforts needed by both NSF and DOE. This report describes the discussions, outcomes and conclusions from STREAM2016: Streaming Requirements, Experience, Applications and Middleware Workshop, the second of these workshops held on March 22-23, 2016 in Tysons, VA. STREAM2016 focused on the Department of Energy (DOE) applications, computational and experimental facilities, as well software systems. Thus, the role of “streaming and steering” as a critical mode of connecting the experimental and computing facilities was pervasive through the workshop. Given the overlap in interests and challenges with industry, the workshop had significant presence from several innovative companies and major contributors. The requirements that drive the proposed research directions, identified in this report, show an important opportunity for building competitive research and development program around streaming data. These findings and recommendations are consistent with vision outlined in NRC Frontiers of Data and National Strategic Computing Initiative (NCSI) [1, 2]. The discussions from the workshop are captured as topic areas covered in this report's sections. The report discusses four research directions driven by current and future application requirements reflecting the areas identified as important by STREAM2016. These include (i) Algorithms, (ii) Programming Models, Languages and Runtime Systems (iii) Human-in-the-loop and Steering in Scientific Workflow and (iv) Facilities.

96 KNOWLEDGE MANAGEMENT AND PRESERVATION↗

DOE OSTI2017

Crystalline and Crystalline International Disposal Activities

This report presents the results of work conducted between September 2015 and July 2016 at Los Alamos National Laboratory in the crystalline disposal and crystalline international disposal work packages of the Used Fuel Disposition Campaign (UFDC) for DOE-NE’s Fuel Cycle Research and Development program. Los Alamos focused on two main activities during this period: Discrete fracture network (DFN) modeling to describe flow and radionuclide transport in complex fracture networks that are typical of crystalline rock environments, and a comprehensive interpretation of three different colloid-facilitated radionuclide transport experiments conducted in a fractured granodiorite at the Grimsel Test Site in Switzerland between 2002 and 2013. Chapter 1 presents the results of the DFN work and is divided into three main sections: (1) we show results of our recent study on the correlation between fracture size and fracture transmissivity (2) we present an analysis and visualization prototype using the concept of a flow topology graph for characterization of discrete fracture networks, and (3) we describe the Crystalline International work in support of the Swedish Task Force. Chapter 2 presents interpretation of the colloidfacilitated radionuclide transport experiments in the crystalline rock at the Grimsel Test Site.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Enhanced LAW Glass Correlation - Phase 1

About 50 million gallons of high-level mixed waste is currently stored in underground tanks at the United States Department of Energy’s (DOE’s) Hanford site in the State of Washington. The Hanford Tank Waste Treatment and Immobilization Plant (WTP) will provide DOE’s Office of River Protection (ORP) with a means of treating this waste by vitrification for subsequent disposal. The tank waste will be separated into low- and high-activity waste fractions, which will then be vitrified respectively into Immobilized Low Activity Waste (ILAW) and Immobilized High Level Waste (IHLW) products. The ILAW product will be disposed in an engineered facility on the Hanford site while the IHLW product is designed for acceptance into a national deep geological disposal facility for high-level nuclear waste. The ILAW and IHLW products must meet a variety of requirements with respect to protection of the environment before they can be accepted for disposal. Acceptable glass formulations for vitrification of Hanford low activity waste (LAW) must meet a variety of product quality, processability, and waste loading requirements. To this end, The Vitreous State Laboratory (VSL) at The Catholic University of America (CUA) developed and tested a number of glass formulations during Part A, Part B1 and Part B2 of the WTP development program. The testing resulted in the selection of target glass compositions for the processing of eight of the Phase I LAW tanks. The selected glass compositions were tested at the crucible scale to confirm their compliance with ILAW performance requirements. Duramelter 100 (DM100) and LAW Pilot Melter tests were then conducted to demonstrate the viability of these glass compositions for LAW vitrification at high processing rates.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

Diffusivities of Ag, Cs, Sr, and Kr in TRISO fuel particles and graphite

Tri-structural isotropic (TRISO) coated particles have been developed and studied since the late 1950s when the concept of coated particles was invented by Roy Huddle of the United Kingdom Atomic Energy Authority. Several decades of work by half a dozen countries on fission product transport in TRISO fuel through numerous irradiation and heating experiments have led to several recommendations of transport data and to the adoption of various sets of diffusion coefficients. In 1997, the International Atomic Energy Agency (IAEA) gathered all these historical results and issued a technical document (TECDOC-978 [IAEA]) that summarizes these sets of recommended diffusion coefficients. Table 1 shows the reference literature articles for the diffusivities that have historically been recommended by the American and German TRISO fuel development programs and that are summarized in the IAEA report (see section 7 for full references of these articles).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Non-Invasive Acoustic-Based Monitoring of Heavy Water and Uranium Process Solutions

The goals of the project are to leverage laboratory scientific strength in physical acoustics for critical international safeguards applications; create hardware demonstration capability for noninvasive, near real time, and low cost process monitor to capture future technology development programs; and measure physical property data to support method applicability.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2017

Succession planning for technical experts

This report describes a methodology for identifying, evaluating, and mitigating the loss of key technical skills at nuclear operations facilities. The methodology can be adapted for application within regulatory authorities and research and development organizations, and can be directly applied by international engagement partners of the Department of Energy’s National Nuclear Security Administration (NNSA). The resultant product will be of direct benefit to two types of NNSA missions: (1) domestic human capital development programs tasked to provide focused technical expertise to succeed an aging nuclear operations workforce, and (2) international safeguards programs charged with maintaining operational safeguards for developing/existing nuclear power program in nations where minimal available resources must be used effectively. This report considers succession planning and the critical skills necessary to meet an institution’s goals and mission. Closely tied to succession planning are knowledge management and mentorship. In considering succession planning, critical skill sets are identified and are greatly dependent on the subject matter expert in question. This report also provides examples of critical skills that are job specific.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

DOE OSTI2017

A Methodology for Succession Planning for Technical Experts

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2016

A History of Sandia’s Water Decision Modeling and Analysis Program

This document provides a brief narrative, and selected project descriptions, that represent Sandia’s history involving data, modeling, and analysis related to water, energy-water nexus, and energy-water-agriculture nexus within the context of climate change. Sandia National Laboratories has been engaged since the early-1990s with program development involving data, modeling, and analysis projects that address the interdependent issues, risks, and technology-based mitigations associated with increasing demands and stresses being placed on energy, water, and agricultural/food resources, and the related impacts on their security and sustainability in the face of both domestic and global population growth, expanding economic development, and climate change.

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2018

Neutron Residual Stress Mapping for Spent Nuclear Fuel Storage Canister Weldment

Many radioactive materials within the nuclear fuel cycle present a significant hazard. Such materials include spent nuclear fuel, high-level waste, legacy waste and other nuclear materials that have no current outlet than storage. These materials are often held in long-term storage as an interim stage within their lifecycle. Lifecycles can include reuse or disposal. Safety and security of spent nuclear fuel (SNF) interim storage installations are very important, due to a great concentration of fission products, actinides and activation products. Stress corrosion cracking of interim storage containers has been identified as a high priority data gap by the Department of Energy. However, little has been done with regards to the canister weld material properties and their impact on stress corrosion, until recently from Sandia National Laboratory canister mockup program. Furthermore, because no post-weld heat treatment was required, the associated high tensile residual stress within these canister welds can drive the initiation and growth of stress corrosion cracking (SCC) cracks. In this work, we carried out in-situ neutron residual stress mapping on a weld plate sectioned from Sandia mock-up canister wall. Significant residual stress profiles were identified from the received longitudinal weld sample. The details of the weldment residual stress program development at ORNL are presented in this progress report. The generated residual stress profile from neutron diffraction technique is similar to that of Sandia residual stress data obtained from the deep hole drilling approach.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2018

Sensitivity ZAM Modeling Study of Tank 10H VDS Waste Sample (U)

The objective of this work is to study the sensitivity of cesium loading expected on the Tank Closure Cesium Removal (TCCR) columns. Savannah River National Laboratory (SRNL) utilized ZAM, a computer program developed by the research group of Professor Rayford G. Anthony of Texas A&M University, for prediction of cesium loading on the Crystalline Silicotitanate (CST) resin during the processing of Tank 10H Variable Depth Sample (VDS) waste solution.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2018

FY18Q2 Quarterly Report: Radiation Enhanced Diffusion of Ag, Ag-Pd, Eu, and Sr in Neutron Irradiated PyC/SiC Diffusion Couples

Obtaining accurate diffusion kinetics in materials representative of those found in tristructural-isotopic (TRISO) coated particle fuel is needed to predict the diffusive release of fission products in reactor. Planar diffusion couples with representative pyrocarbon (PyC) and silicon carbide (SiC) layers are being produced using the same fluidized-bed chemical vapor deposition (FBCVD) technology used to produce TRISO particles from the first irradiation experiment of the Advanced Gas Reactor (AGR) Fuel Qualification and Development Program. The layer properties of the planar diffusion couples are tailored to meet the specified PyC density and microstructure of the SiC layer as defined by the AGR program. The influence of these variables on diffusion is also being explored by producing PyC and SiC variants.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗