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All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

On-Site Inspection RadioIsotopic Spectroscopy (Osiris) System-Integration and Field-Testing Report

We have designed a system for the acquisition and analysis of high-resolution gamma-ray spectra during on-site inspections under the Comprehensive Nuclear-Test-Ban Treaty (CTBT). The On-Site Inspection RadioIsotopic Spectroscopy—OSIRIS —gamma-ray spectrometer is a mechanically-cooled, battery-powered ORTEC IDM-200V, chosen to avoid the need for liquid nitrogen during on-site inspections. The spectrometer is controlled and the spectral data analyzed by a Panasonic Toughbook notebook computer. Spectral data analysis is similar to that performed by traditional gamma-ray spectrometers, except the OSIRIS gamma-ray spectral data are filtered to display only radioisotopic information relevant to CTBT on-site inspections, for example, the fission-product isotope Zr-95 and its daughter isotope Nb-95. The OSIRIS software assists the user during operation with software checklists for startup and shutdown, and menu-selected display windows during energy calibration and data acquisition. OSIRIS can be operated in either sample-counting mode or the in-situ mode. In this report, the extensive system-integration tests of the OSIRIS system carried out in 2015-2017 are described, including tests at Idaho National Laboratory, at the Nevada National Security Site (formerly, the Nevada Test Site), and at Pacific Northwest National Laboratory in Richland, Washington. We also note recent software enhancements and the steps to our successful commercialization of OSIRIS.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2018

BOP-301 and BOP-302R: Test Definitions and Analyses

Following the conclusion of the International Atomic Energy Agency (IAEA) coordinated research project (CRP) on analysis of two EBR-II tests, Argonne has continued its EBR-II analysis with additional tests from the Shutdown Heat Removal Test (SHRT) series. The tests selected for this follow-up analysis are BOP-301 and BOP-302R, two unprotected loss of heat of sink tests in which the intermediate sodium pump tripped without scramming the control rods or tripping the primary pumps. The IAEA CRP focused on the SHRT-17 and SHRT-45R tests, the most severe protected and unprotected loss of flow tests performed during the SHRT series. With the primary pumps tripping, flow rates decreased to levels where flowmeter uncertainty was very high. Accurately capturing the progressions of the two tests despite this uncertainty was one of the key modeling and simulation challenges during the CRP. BOP-301 and BOP-302R were selected because the primary sodium pumps did not trip for those tests. Primary flow rate measurements are known with greater certainty. Power and temperature discrepancies can be explored without concern that flow predictions match test measurements but not the actual flow rates from the tests. These two tests were performed during the same test window as SHRT-45R and had the same core load configuration. Therefore, core models created for analysis of SHRT-45R were easily adapted for analysis of BOP-301 and BOP-302R. Another reason for analyzing BOP-301 and BOP-302R is to analyze tests in which the core inlet temperature changed significantly, which did not occur during the loss of flow tests. The transient progression of the unprotected SHRT-45R test was driven by a decreasing core flow rate while the core inlet temperature remained relatively constant. For the BOP tests, the transient progression was driven by an increasing core inlet temperature while the core flow rate was stable. This report defines the BOP-301 and BOP-302R tests and provides a selection of measured test data. Initial predictions from SAS4A/SASSYS-1 simulations of these tests are then presented. Following these simulations, additional analysis was performed with the Dakota uncertainty quantification and optimization toolkit to determine which parameters have the greater impact on the transient results and the optimal values of the most important parameters.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Benchmark Development for TREAT Minimum Critical Mass Core Loading to Support Validation of TREAT Operations

Benchmark experiment data is necessary to validate modeling and simulation activities to support Transient Reactor Experiment Test (TREAT) Facility restart and operations. Key measurements of interest include criticality, control rod worth, excess reactivity, and shutdown margin for varying core loading sizes. Benchmark evaluations are being developed according the guidelines provided in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook) that can also support advanced modeling and simulation activities, new experiment design for advanced reactors and accident tolerant fuel, and potential low-enriched uranium conversion of the TREAT reactor. This summary discusses development activities supporting a benchmark evaluation of the TREAT minimum critical mass core loading, which will be submitted for peer-review and publication in the IRPhEP Handbook with the identifier TREAT-FUND-RESR-001. Detailed models of TREAT were developed and evaluated to prepare benchmark models of the minimum critical mass configuration. Calculations were performed using MCNP6.1 with ENDF/B-VII.1 nuclear data libraries. Discussion of results.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Elementary Particle Physics

The UT Dallas High Energy Physics group collaborates on the ATLAS experiment which observes the world’s high energy proton-proton collisions. The UTD group utilized ATLAS 8 TeV Run 1 data to search for evidence of a “dark photon”, a gauge boson outside the “Standard Model of Particle Physics” that theorist hypothesize could mediate a force felt only by the dark matter particles that dominate out universe. The observable signature is a pair of prompt lepton jets. The results of this search were published in 2013. More recently, we have improved the sensitivity of the analysis using machine learning techniques, and we are extending the search to higher and lower mass dark photons using the ATLAS 13 TeV data set. Detection of a new fifth force would represent an important advance beyond the 4 known forces that influence baryonic matter. The LHC Long Shutdown 1 from 2013-2014 was a period of intense work on refurbishment and upgrades for the ATLAS Pixel community and the UTD group. The service quarter panels of the original Pixel detector were replaced, and the Insertable B Layer (IBL) detector was integrated. During IBL integration, UTD participated on the Stave Task Force that solved corrosion problems with IBL wire bonds. We developed a relative humidity sensor that was used during integration and transport to the beamline when detector was unpowered. The sensors were repurposed to monitor relative humidity inside the ID endplates until the built-in sensors were recommissioned. The UTD group worked on the connection of the Pixel and IBL at Point 1 and we were in charge of electric service testing. The IBL wire bond corrosion was the instigation for developing a new conformal coating method using an atomizing spray and Cellpack polyurethane. It is being considered for use in the ITk tracker upgrade. We have demonstrated its tolerance both of a 9 MGy ionizing dose and thermal expansion. Two UTD students are completing authorship qualification projects. One is helping understand variations in impact parameter resolution across different running periods. The second is developing a data-driven method to monitor the fraction of fake tracks in dense jets.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2018

Energy Infrastructure Dependency Curves Simulation Tool User’s Guide

Foundational studies of infrastructure interdependencies have attempted to make sense of the complex relationships among critical infrastructure. Beyond understanding the complexity of these interconnections, owner/operators and decision makers have a continuing need for a situational awareness capability tool to move beyond a conceptual understanding toward a more quantified, visual analytic method to enable more detailed real-world analysis and prioritization.The purpose and development of the Energy Infrastructure Dependency Curves Simulation Tool is to provide a means to display and analyze the complexities of infrastructure dependencies, in a way that is dynamic and can be easily manipulated. The tool focuses on the connections between two infrastructure assets, a natural gas-processing plant that relies on electric power, and an electric power-generating plant that depends on natural gas to produce electricity. The tool allows the user to enter information characterizing each asset and the dependence relationships, including details about core operations, backup function, shutdown, and resumption of service. Output is provided as a visualization of the performance of each asset over time, displaying impact, stages of operation, and resumption to full service. Input can be modified to enable the user to see how different scenarios or decisions would change the outcome. The intent of the tool is to provide the user with the ability to make informed decisions regarding operation of individual asset.This report presents an overview of the concepts used for developing the Energy Infrastructure Dependency Curves Simulation Tool and a detailed user guide that provides instructions on how to use the tool. Finally, three example scenarios are presented to illustrate dependency curve outputs.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2016

Specification of Fenix MPI Fault Tolerance library (V.0.9)

Fenix is a software library compatible with the Message Passing Interface (MPI) to support fault recovery without application shutdown. This specification is derived from a current implementation of Fenix that employs the User Level Fault Mitigation (ULFM) MPI fault tolerance module proposal. We only present the C library interface for Fenix; the Fortran interface will be added once the C version is complete.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2018

MP1 Intermediate Characterization Report Summary

Great effort worldwide has been invested in replacing high enriched uranium (HEU) fuel with low enriched uranium (LEU) fuel in nuclear plants. This effort aims in safeguarding nuclear material and promoting the proliferation resistance of current reactors. The United States Department of Energy (DOE), National Nuclear Security Administration's (NNSA) Materials Management and Minimization Reactor Conversion program is actively working to convert civilian research and test reactors from the use of HEU to LEU fuel. To date, DOE NNSA has converted or verified the shutdown of various domestic reactors and also reactors worldwide. The task to develop and test a new LEU fuel while maintaining current performances has been assigned to the High-Performance Research Reactor (USHPRR) program. U alloyed with Mo achieves the goal of high uranium density necessary for the conversion of the high performance research reactors, and has also demonstrated good performance in pile [1]. The alloying process of U with Mo stabilizes the body-centered cubic (bcc) gamma phase (?-U) at room temperature. Without this alloying property enhancement, this phase is only stable at temperatures only above 770 °C. Retaining this isotropic gamma phase within the U-Mo alloy provides swelling and oxidation resistance and desired mechanical properties [2]. Generally 10% w. has been considered sufficient to achieve stabilization. However, local phenomena within the alloy can affect the local Mo content and lead to gamma phase decomposition. Because of this, it has been suggested that controlling gamma phase decomposition is most critical during processing [2-3]. The work detailed in this document is part of the HPPR program and focuses on the characterization of fresh fuel (a.k.a. as fabricated, before reactor testing) in collaboration with other laboratories including PNNL (Pacific Northwest National Laboratory), LANL (Los Alamos National Laboratory) and BWXT (Babcock and Wilcox Technologies). This testing is part of a 4 way validation for which similar samples from master plates were taken and distributed to each laboratory to independently analyze utilizing a standardized set of measurement and analysis procedures. When complete this work will supply a “toolbox” of information for alloying and fabrication parameters with its direct effects on microstructure quantified. Various parameters have been analyzed in this characterization effort (including grain size, quality of the coating layer, composition of the samples, carbide content etc.), based on the direction of the working group in ref.[4]. The final aim of this work is to increase the understanding of the impact of processing conditions on the final fuel microstructure, using fabrication processes that meet “commercial viability” requirements [4]. The strength of this campaign rests on the extent of the samples studied and the variation of the manufacture parameters which will permit to achieve the required understanding. Thus permitting to use these experiments to down-select test samples used to evaluate irradiation performance [4] to be tested in various locations of the Idaho National Laboratory (INL) Advanced Test Reactor (ATR).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

A User Guide to PARET/ANL

PARET was originally created in 1969 at what is now Idaho National Laboratory (INL), to analyze reactivity insertion events in research and test reactor cores cooled by light or heavy water, with fuel composed of either plates or pins. The use of PARET is also appropriate for fuel assemblies with curved fuel plates when their radii of curvatures are large with respect to the fuel plate thickness. The PARET/ANL version of the code has been developed at Argonne National Laboratory (ANL) under the sponsorship of the U.S. Department of Energy/NNSA since the inception of the Reactor Conversion Program. Since Reduced Enrichment for Research and Test Reactors (RERTR) began in 1978, PARET/ANL has been benchmarked to experimental data including SPERT testing, and used to determine the expected transient behavior of a large number of reactors regardless of enrichment both inside and outside the Reactor Conversion Program. This document provides the pertinent information for the use of PARET/ANL Version 7.6. PARET/ANL models the various fueled regions of a reactor core as channels. Each of these channels consists of a single flat fuel plate/pin (including cladding and, optionally, a gap) with water coolant on each side. In slab geometry the coolant channels for a given fuel plate are of identical dimensions (mirror symmetry), but they can be of different thickness in each channel. There can be many channels, but each channel is independent and coupled only through reactivity feedback effects to the whole core. The time-dependent differential equations that represent the system are replaced by an equivalent set of finite-difference equations in space and time, which are integrated numerically. PARET/ANL uses fundamentally the same numerical scheme as RELAP5 for the time-integration of the point-kinetics equations. The one-dimensional thermal-hydraulic model includes temperature-dependent thermal properties of the solid materials, such as heat capacity and thermal conductivity, as well as the transient heat production and heat transfer from the fuel meat to the coolant. Temperature- and pressure-dependent thermal properties of the coolant such as enthalpy, density, thermal conductivity, and viscosity are also used in determining parameters such as friction factors and heat transfer coefficients. The code first determines the steady-state solution for the initial state. Then the solution of the transient is obtained by integration in time and space. Multiple heat transfer, DNB and flow instability correlations are available. The code was originally developed to model reactors cooled by an open loop, which was adequate for rapid transients in pool-type cores. An external loop model appropriate for Miniature Neutron Source Reactors (MNSR’s) was also added to PARET/ANL to model natural circulation within the vessel, heat transfer from the vessel to pool and heat loss by evaporation from the pool. PARET/ANL also contains models for decay heat after shutdown, control rod reactivity versus time or position, time-dependent pump flow, and loss-of-flow event with flow reversal as well as logic for trips on period, power, and flow. Feedback reactivity effects from coolant density changes and temperature changes are represented by tables. Feedback reactivity from fuel heat-up (Doppler Effect) is represented by a four-term polynomial in powers of fuel temperature. Photo-neutrons produced in beryllium or in heavy water may be included in the point-kinetics equations by using additional delayed neutron groups.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

IWTU TI-102 – Part 2 Sample Analyses

Samples from the Integrated Waste Treatment Unit (IWTU) were collected for analysis after TI-102 – Part 2 shutdown. The scope of analyses performed by Battelle Energy Alliance (BEA) included: • Scanning electron micrography (SEM) and energy dispersive spectroscopy (EDS) analysis for particle morphology, particle size, and elemental composition estimates • Loss on ignition (LOI) for determining the fraction that is coke and not bauxite • Proximate and ultimate analysis • Particle-size analysis using sieve trays • Bulk and particle density • Elemental analysis • Solubility • X-ray diffraction for mineralogy analysis • Thermal gravimetric analysis (TGA). This report provides the results of those analyses.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2019

University of Tennessee, Knoxville (UTK) contribution to: Deciphering the role of mixed-material deposition and temperature on lithium-coated PFCs in NSTX-U high-performance plasmas: Collaborative UIUC & UTK Proposal (Final Report)

This report describes the final product of model development to decipher the role of mixed material deposition and temperature on lithium-coated PFCs in NSTX-U. The computational modeling approach used to address these specific questions involved a pragmatic multiscale materials modeling framework, with the use of reaction-diffusion based continuum simulations, through modifications to the SDTrimSP code. This pragmatic modeling was supplemented with a few, select parametric ab-initio and molecular dynamics simulations, to investigate the effect of lithium and mixed lithium-carbon-refractory metal chemistry and surface morphology and chemistry on erosion yields and deuterium uptake, including the role of temperature and boronization. The main progress of the project was to develop the major framework and structure of the model, although validated application of the model was limited as a result of the unplanned shutdown and outage of NSTX-U.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

Neutronic Design and Analysis of the Holos-Quad Concept

HolosGen, LLC is proposing a highly innovative micro-reactor concept targeting both civilian and military applications. It consists of an advanced helium-cooled gas reactor using a turbojet-type turbine and compressor to achieve a highly condensed reactor that fits into a commercial ISO container. The Holos-Quad concept being considered for civilian applications will generate 22MWt, using four Subcritical Power Modules (SPMs) that fit into one 40-foot ISO container. It is a high-temperature gas-cooled reactor concept using TRISO fuel distributed in graphite hexagonal blocks, cooled with helium gas at 7MPa and a high outlet temperature. In FY2019, HolosGen, LLC received an ARPA-E MEITNER grant to team up with the Argonne National Laboratory (ANL) design team to demonstrate the feasibility of the Holos-Quad concept in terms of neutronics. In particular, the evaluation study of the Holos design features and alternative options for reactivity control was considered. Through these tasks, the ANL design team has been working on demonstrating the feasibility of the Holos-Quad concept in terms of neutronics to help advance the core design of the Holos-Quad. A rigorous design approach was employed for the Holos-Quad in order to ensure that the input space is fully investigated, and that the best solutions are considered. The first step of this approach consisted of properly defining the Holos-Quad design problem by identifying the competitive design objectives: its economic performance is maximized with longer lifetime and its transportability to remote locations is maximized with reduced weight. The design problem also required to identify the operational constraints (in terms of thermal hydraulics, reactivity feedback and shutdown margins) and the input parameters that could be varied. In the second step, a sensitivity analysis was performed to enable preliminary investigation of the input space to identify the correlations among input and output variables. It allowed, for instance, to find input parameters that have relatively poor impact on the core performance and that could be removed from the optimization problem to accelerate its convergence. This sensitivity analysis showed that the design problem defined is highly constrained since only 117 solutions out of 1,000 simulated cores met all the design constraints. The design optimization was performed in the third step employing a genetic algorithm to effectively explore this highly constrained input space and find global optimal solutions. More than 2,500 cores were fully evaluated to converge to a Pareto Frontier of optimum solutions providing best compromise between the life-time target and the total core weight, as shown in Figure 1. The main benefit of this approach was to reduce the human effort and to be able to automatically re-run the optimization problem when changing the design problem. The multi-objective optimization identified various core solutions that would be optimum solutions for different types of applications.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Evaluation of High Rotor Pole Switched Reluctance Motors to Control Condenser Fans in a Commercial Refrigeration System

A high rotor pole switched reluctance motor (HRSRM) designed by Software Motor Corporation was evaluated for use as a condenser fan motor due to its reported high motor efficiency over a broad range of speeds and added advanced control potential. The project was split into two phases; the first of which was to determine the direct energy savings due to the increased motor efficiency, the second was to leverage the additional benefits of the energy efficiency measure (EEM) to create additional energy savings. This report only discusses results from the first phase of the project. Nine 1.5hp HRSRMs were installed parallel to the same number of legacy induction motors of a supermarket commercial refrigeration system located in Lakeside, Colorado. At this site, the 1.5 hp legacy motors had historically adjusted condenser fan speed from 200-870 rpm based on outdoor air temperature and compressor discharge pressure. This type of control is known as variable head pressure control (VHPC). These nine new motors were given a signal from the legacy variable frequency drive (VFD) to match the motor speed of the legacy motors. Determination of the energy savings of the EEM was performed over a two-month period, at which time a correlation between legacy and EEM power consumption was developed based on motor speed. Annual data that had been collected from 2016-2017 (before the two-month test period) was then leveraged to predict the annual energy savings of the EEM for a complete retrofit using both VHPC and constant fan speed control. Table 1 displays the quantitative results. The EEM saved 27% of total energy consumption of the condenser fan motors for constant fan speed control, while the energy savings increased to 32% when using VHPC. This was due to the improved motor efficiency of the HRSR motor at low speeds. While the constant speed fan control kept the motors running at 870 rpm, VHPC adjusted fan speed between 200-870rpm based on the outdoor air temperature and the compressor discharge pressure. These savings are assuming a direct replacement with no additional controls. Retrofitting a constant speed fan control system with HRSRMs and switching to VHPC would provide significant energy savings looking at the condenser alone. Additional considerations of the EEM were noted during the project as well. The HRSRM system made it possible to identify and remotely shutdown an individual motor when a piece of foam locked the fan and thereby the motor. With the traditional system, a locked rotor could have caused motor damage unless a breaker or fuse was activated. Furthermore, due to vibration complaints from the installation technicians, the ramp-up time of the HRSRM had to be increased to a minute. After implementing this change no complaints for excessive vibration were recorded.

15 GEOTHERMAL ENERGY↗
DOE OSTI2019

RCCS Testing with the Water-based NSTF: Year-1 Single-Phase Results

Under support from the Department of Energy (DOE) and the Office of Advanced Reactor Technologies (ART), a large scale test facility has been constructed to generate validation data for passive decay heat removal systems of advanced reactors. Located at Argonne National Laboratory, the Natural convection Shutdown heat removal Test Facility (NSTF) reflects key features of a ½ scale, water-based, Reactor Cavity Cooling System (RCCS). This report serves as a summary of the single-phase testing results obtained during the programs first year of matrix testing. An introduction to the high-level program objectives is first presented, followed by a brief summary of facility design, instrumentation, and control systems. Details of the NQA-1 driven test method are included, which govern the programs quality assurance requirements and provide a background for the testing strategy and procedures. Checkout activities, including verification of gross system mass and energy balances, heat loss quantification, and isothermal facility characterization, were performed early in the first year and results presented in this report. These tests have identified relevant facility characteristics such as K-factors, frictional pressure losses, parasitic heat loss rates, and transient heating response. Results are then presented from four accepted matrix tests where the facility established a steady-state, single-phase mode of flow operation. All matrix tests were performed with an identical top-level procedure: first initiating a heater power ramp to target levels, and then holding power levels constant until the facilities structural members and network piping reached thermal equilibrium. Upon obtaining stable thermal and hydraulic conditions, active cooling was initiated off of the bulk storage tank to maintain constant riser inlet temperatures of 30°C. This steady-state period was then held for a minimum required duration, at which point the heaters were ramped down and test concluded. The two baseline tests completed reflect conditions prototypic to the full scale plant at both normal operation, and design basis Depressurized Loss of Forced Cooling (DLOFC) conditions. These parameters are defined by full RCCS decay heat loads of 1.4 MWt and 2.1 MWt, which translate to NSTF scale thermal power levels of 34.4 kWt and 51.6 kWt, respectively. Two additional tests were also conducted, which repeated baseline cases but included modification by single parameter variations in decay heat load, or heater power, at 700 kWt, and 2.8 MWt. Finally, a scoping test was performed to demonstrate the facilities ability to operate in a two-phase mode of operation. The facility successfully operated for a 3-hour period at this mode, reaching peak inventory temperatures of 105°C and condensing approximately 50 gallons of boil-off inventory. The testing parameters for this demonstration were based off the design basis DLOFC condition, and will be the reference for new baseline cases as the program moves forward with two-phase matrix testing.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Methods to Automate Assessing the Outage Risk and Technical Specification Compliance for Planned Work and Current Plant Conditions. Light Water Reactor Sustainability Program report

This research effort is a part of the Light Water Reactor Sustainability (LWRS) Program, which is a research and development (R&D) program sponsored by the Department of Energy. The LWRS Program is performed in close collaboration with industry R&D programs that provide the technical foundations for licensing and managing the long term, safe, and economical operation of current nuclear power plants (NPPs). The LWRS Program serves to help the United States (U.S.) nuclear industry adopt new technologies and engineering solutions that facilitate the continued safe operation of these NPPs and the extension of their current operating licenses. The Outage Risk Management Improvement Pilot Project seeks to improve the management of nuclear power plant outages. This is done through the development of tools to assist in evaluating pending activities against requirements to detect undesired interactions. Significant efforts are expended to manage the nuclear risk of an outage. The utilities conduct pre outage risk assessments, based on a very detailed review of the outage schedule, to identify where combinations of outage work and out of service equipment would result in degraded conditions with respect to nuclear safety or regulatory compliance. Probabilistic risk assessment studies are conducted to quantify incremental core damage frequency as a result of the outage activities and system unavailability. These studies are usually presented to site and fleet management, the site plant operational review committee, and the NPP’s independent Nuclear Safety Review Board for concurrence that the outage is planned safely and that reasonable measures have been taken to reduce the added risk of conducting the outage. During the outage, the plant configuration is monitored continuously to ensure that it conforms to the approved safety plan. Deviations must be assessed and approved by management committees and, in some cases, the plant operational review committee. In virtually all outage meetings and job briefings, the current nuclear safety status of the plant is communicated, including information on the specific equipment necessary to meet the requirements of the nuclear safety plan. In addition, the operations and outage organizations implement several layers of physical and administrative barriers to prevent unintended interaction with the systems and equipment credited for nuclear safety. In spite of all these efforts, nuclear safety challenges still occur too frequently during outages. While some of these are due to failures of equipment credited for safety, the majority occur because of human error. These typically involve some form of interaction between work activities and plant configuration changes. Some of them are very subtle and are extremely challenging to detect in advance. Nevertheless, they are not acceptable and represent clear opportunities to improve nuclear safety during outages. This particular report will provide a demonstration of the capabilities in the development of a predictive software platform for improving configuration control and accessing the plant risks associated with outage work protocols and illustrate how the software assists in strategies to minimize these interactions. In addition to maintaining nuclear safety, there are numerous administrative requirements that must be complied with during an outage. This report will discuss the use of logic modelling to monitor planned work against technical specification requirements for given plant conditions and to integrate these requirements with shutdown risk models, clearance order requirements and other plant condition requirements.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Simulation Center for Runaway Electron Avoidance and Mitigation (SCREAM) - NIMROD DIII-D Shattered Pellet Injection Disruption Mitigation and RE Modeling. Final report

Final technical report for UCSD and subcontractor contributions to DE-SC0016420, Simulation Center for Runaway Electron Avoidance and Mitigation (SCREAM). Project description. MHD modeling includeing runaway electron drift orbit calculations will be carried out for DII-D disruption mitigation scenarios. The effects of pitch angle scattering on the energy distribution of runaway electron population will be evaluated using one or more collision models integrated / coupled with the runaway orbit following model in the nimrod code. An extension of the test particle model will be developed to predict runaway electron generation considering both primary and secondary sources of runaway electron production. Simulations simultaneously modelling the effects of generation, confinement, and moderation-space losses during a DII-D rapid shutdown (mitigated disruption) will be completed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Specification of Fenix MPI Fault Tolerance library (V. 1.0)

This document provides a specification of Fenix, a software library compatible with the Message Passing Interface (MPI) to support fault recovery without application shutdown. The library consists of two modules. The first, termed process recovery, restores an application to a consistent state after it has suffered a loss of one or more MPI processes (ranks). The second specifies functions the user can invoke to store application data in Fenix managed redundant storage, and to retrieve it from that storage after process recovery.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2019

Hydraulic Testing of the Lost Lake Aquifer Near Recovery Well RWM 8

Groundwater beneath the M-Area HWMF is contaminated with chlorinated ethenes including trichloroethylene (TCE) and tetrachloroethylene (PCE). SRS operates a network of recovery wells designed to hydraulically contain and capture the high concentration VOC plume in the Lost Lake Aquifer Zone (LLAZ) (Figure 1). The recovery wells are connected to the M-1 Air Stripper and the system is permitted by the South Carolina Department of Environmental Control (SCDHEC) to operate at a total flow of 610 gpm. RWM 8, which is one of ten operational recovery wells, has historically operated at a flow rate of approximately 40 gpm. RWM 8 was installed with four separate 10-ft screens that spanned the LLAZ with a continuous gravel pack spaced as shown in Figure 2. The three uppermost screens are located within the LLAZ whereas the deepest screen is located within the middle sand of the Crouch Branch Confining Unit (CBCU). Area Closure Projects (ACP) installed a permanent packer in May 2019 to isolate the lowest screen from the pump with the goal being to target the zone of highest concentration in the LLAZ. ACP expressed concern that a substantial portion of the flow from RWM 8 comes from the lower screen. With this screen isolated from service, flow from the well may be below the desired operational envelope for the well. The reconfiguration of RWM 8 presented an opportunity to re-evaluate the performance characteristics of the well. Following placement of the packer and relocation of the pump, the well was redeveloped. To assess the effects of eliminating the lower screen, a step-drawdown test was planned. The purpose of this test was to establish new baseline well performance properties including specific capacity, well efficiency, and well loss coefficients. Data from the step-drawdown testing was also used to estimate a maximum pumping rate for RWM 8 under the new configuration. The setup for the step-drawdown test facilitated a subsequent constant rate aquifer pumping test aimed at determining aquifer hydraulic properties. Properties estimated for the LLAZ included transmissivity and storativity. Following shutdown of RWM 8, recovery data were also collected and analyzed to estimate aquifer properties. This report discusses the hydrologic tests conducted following the reconfiguration of RWM 8. The information provided in this report may serve as input to subsequent updates to the groundwater flow and contaminant transport model for A/M Area.

42 ENGINEERING↗