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Warhead Activities (Life Extension Programs, Modifications, and Alterations)

NNSA’s warhead modernization activities ensure that the U.S. nuclear weapons stockpile continues to meet Department of Defense (DoD) requirements while enhancing safety and security. NNSA’s modernization efforts address aging, unavailability of replacement parts, and integration with DoD’s modernized nuclear weapons delivery systems without providing new military capabilities or the need for explosive nuclear testing.

Source record↗

The Next Generation NC3 Enterprise: Opportunities and Challenges

The 2018 Department of Defense Nuclear Posture Review emphasized the importance of modernizing the existing nuclear command, control, and communications (NC3) architecture. Heightened awareness of US reliance upon technology developed in the 1970s and emerging challenges posed by frontier technologies spurred Secretary of Defense Mattis to task USSTRATCOM with identifying the technologies, processes, and institutions needed to build a next-generation NC3 system. General Hyten, USAF Commander, has noted that this will require an “overhaul of our existing processes from requirements definition to systems engineering and integration, acquisition, and budgeting. A framework is required for modern processes to enable delivery of a flexible, continuously evolving, threat-driven set of capabilities.” This brief, prepared in response to Gen. Hyten’s call for memoranda to address the “Next Generation NC3 Enterprise,” outlines three distinct approaches to NC3 modernization, each with a system lifecycle of approximately 50 years (2030-2080): upgrading the existing system, a hybrid approach to integrating new technologies, and creating an entirely new NC3 architecture. It is organized as follows: First, it outlines the integral role of NC3 in strategic operations. Second, it examines the current status of NC3 tools and the contemporary challenges that future systems must address. Third, it provides a series of paths forward for USSTRATCOM’s consideration. Fourth and finally, it recommends a “hybrid” approach to maintain the strengths of the current NC3 architecture, increase the resilience of the system, and leverage the advantages of new sensing, ISR, and communication technology.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

An Experimental Investigation of Fuel Staging In A Model Gas Turbine Combustor

Modern gas turbine engines are typically operated in a lean-premixed configuration in order to meet emissions requirements. Lean operation makes gas turbines more susceptible to combustion instability, which can cause catastrophic hardware failure if left unchecked. The spatially uneven distribution of fuel in a gas turbine combustor, known as fuel staging, is an effective way of suppressing instabilities. This dissertation examines the effect of both steady-state and transient fuel-staging in a model gas turbine combustor. First, the steady-state effect of fuel staging on nominally unstable operating conditions is examined. At 200 degrees Celsius and a bulk flow velocity of 26 m/s, the combustor undergoes self-excited instabilities when the equivalence ratio is between φ = 0.70 and φ = 0.74 when all nozzles are fueled equally. These instabilities are suppressed when the center nozzle equivalence ratio is increased to φ Staging = 0.85 and the outer nozzle equivalence ratio maintained at φ = 0.70 or when the global equivalence ratio is maintained at φ = 0.70 by increasing the center nozzle equivalence ratio to φ Staging = 0.74 and decreasing the outer nozzle equivalence ratio to φ Staging = 0.69. The staged p' RMS amplitudes are found to decrease with increasing equivalence ratio and the damping rates found to increase with increasing equivalence ratio. Next, axisymmetric fuel staging transients are conducted valve opening/closing timescales of 1, 16, 58, 4,000, and 10,000 ms, and equivalence ratios of φ staging = 0.75, 0.80, 0.85. It should be noted that these valve opening/closing timescales, referred to as τ Transient , differ from the actual fuel convection timescale that are summarized in Table 5.2. It is found that the combustor pressure fluctuation at the end of the transient strongly depends on staging equivalence ratio but not on τ Transient . Increasing the staging equivalence ratio decreased the staged- stable combustor pressure amplitude, similarly to the steady-state results. The staged-stable damping rate trends are insensitive to to τ Transient and increased with increasing staging equivalence ratio. The timescale of the transition, τ, is found to be sensitive to both τ Transient and the staging equivalence ratio. For a Modern gas turbine engines are typically operated in a lean-premixed configuration in order to meet emissions requirements. Lean operation makes gas turbines more susceptible to combustion instability, which can cause catastrophic hardware failure if left unchecked. The spatially uneven distribution of fuel in a gas turbine combustor, known as fuel staging, is an effective way of suppressing instabilities. This dissertation examines the effect of both steady-state and transient fuel-staging in a model gas turbine combustor. First, the steady-state effect of fuel staging on nominally unstable operating conditions is examined. At 200 degrees Celsius and a bulk flow velocity of 26 m/s, the combustor undergoes self-excited instabilities when the equivalence ratio is between φ = 0.70 and φ = 0.74 when all nozzles are fueled equally. These instabilities are suppressed when the center nozzle equivalence ratio is increased to φ Staging = 0.85 and the outer nozzle equivalence ratio maintained at φ = 0.70 or when the global equivalence ratio is maintained at φ = 0.70 by increasing the center nozzle equivalence ratio to φ Staging = 0.74 and decreasing the outer nozzle equivalence ratio to φ Staging = 0.69. The staged p' RMS amplitudes are found to decrease with increasing equivalence ratio and the damping rates found to increase with increasing equivalence ratio. Next, axisymmetric fuel staging transients are conducted valve opening/closing timescales of 1, 16, 58, 4,000, and 10,000 ms, and equivalence ratios of φ staging = 0.75, 0.80, 0.85. It should be noted that these valve opening/closing timescales, referred to as τ Transient , differ from the actual fuel convection timescale that are summarized in Table 5.2. It is found that the combustor pressure fluctuation at the end of the transient strongly depends on staging equivalence ratio but not on τ Transient . Increasing the staging equivalence ratio decreased the staged-stable combustor pressure amplitude, similarly to the steady-state results. The staged-stable damping rate trends are insensitive to to τ Transient and increased with increasing staging equivalence ratio. The timescale of the transition, τ, is found to be sensitive to both τ Transient and the staging equivalence ratio. For a Modern gas turbine engines are typically operated in a lean-premixed configuration in order to meet emissions requirements. Lean operation makes gas turbines more susceptible to combustion instability, which can cause catastrophic hardware failure if left unchecked. The spatially uneven distribution of fuel in a gas turbine combustor, known as fuel staging, is an effective way of suppressing instabilities. This dissertation examines the effect of both steady-state and transient fuel-staging in a model gas turbine combustor. First, the steady-state effect of fuel staging on nominally unstable operating conditions is examined. At 200 degrees Celsius and a bulk flow velocity of 26 m/s, the combustor undergoes self-excited instabilities when the equivalence ratio is between φ = 0.70 and φ = 0.74 when all nozzles are fueled equally. These instabilities are suppressed when the center nozzle equivalence ratio is increased to φ Staging = 0.85 and the outer nozzle equivalence ratio maintained at φ = 0.70 or when the global equivalence ratio is maintained at φ = 0.70 by increasing the center nozzle equivalence ratio to φ Staging = 0.74 and decreasing the outer nozzle equivalence ratio to φ Staging = 0.69. The staged p' RMS amplitudes are found to decrease with increasing equivalence ratio and the damping rates found to increase with increasing equivalence ratio. Next, axisymmetric fuel staging transients are conducted valve opening/closing timescales of 1, 16, 58, 4,000, and 10,000 ms, and equivalence ratios of φ staging = 0.75, 0.80, 0.85. It should be noted that these valve opening/closing timescales, referred to as τ Transient , differ from the actual fuel convection timescale that are summarized in Table 5.2. It is found that the combustor pressure fluctuation at the end of the transient strongly depends on staging equivalence ratio but not on τ Transient . Increasing the staging equivalence ratio decreased the staged- stable combustor pressure amplitude, similarly to the steady-state results. The staged-stable damping rate trends are insensitive to to τT Transient and increased with increasing staging equivalence ratio. The timescale of the transition, τ, is found to be sensitive to both τ Transient and the staging equivalence ratio. For a given τ Transient , the decay timescale decreased with increasing staging equivalence ratio. For 1, 16, and 58 ms transients, τ does not depend on τ Transient , though it does strongly depend on direction, where instability onset cases have longer and more variable τ than instability decay cases. The same dependence of τ on direction is seen for the 4,000 and 10,000 ms transients. However, unlike the short duration transients, τ does depend on τ Transient as would be expected. Phase space reconstruction shows the instability is a harmonic limit cycle with a single frequency. The instantaneous phase images shows the center region of the combustor goes out-of-phase first followed by the rest of the combustor. Finally, asymmetric fuel staging tests are conducted. It is found that non-axisymmetric staging can be as effective as axisymmetric staging, although small hardware variations between nozzles can affect the the minimum staging equivalence ratio that consistently suppresses the instability. These changes in the bifurcation equivalence ratio can make staging configurations appear less effective; however, both axisymmetric and non-axisymmetric staging configurations exhibit similar staged-stable p' RMS , damping rates, and instability transition times for equivalence ratios greater than the bifurcation equivalence ratio. Instantaneous phase images show similar instantaneous phase relationship structures in the staged nozzle (center or outer), suggesting that the instability suppression mechanism is similar in each case. This result is surprising given that the time-averaged flame shapes are very different between axisymmetric and non-axisymmetric staging configurations. The similar effectiveness of axisymmetric and non-axisymmetric staging in this geometry is attributed to the fact that the instability mode is a nominally axisymmetric plane wave. It is hypothesized that non-axisymmetric staging configurations will likely be more effective than axisymmetric staging configurations when the instability mode shape is asymmetric (such as a circumferential wave).

Culler, Wyatt↗

Verification and Performance Impact of the New Parallel MCNP6.3 Particle Track Output Capability for Subcritical Multiplication Simulations

The MCNP6® code, version 6.3, has several new features that are intended to ultimately replace legacy features that are now marked for deprecation. One of these features is the new particle track output (PTRAC) format and capability, where the legacy PTRAC capability still exists alongside the modern PTRAC capability in MCNP6.3. While the MCNP6.3 code has been extensively verified and validated for many applications, the PTRAC feature is not exercised in any of the typical verification and validation (V&V) applications studied during the course of a typical MCNP code release. The primary goal of this paper is to verify that the legacy and modern PTRAC feature produces equivalent results for subcritical multiplication benchmarks previously studied. In the process of verifying that the simulated benchmark results are equivalent, the computational performance is compared between the legacy and modern PTRAC uses. In addition to verification of the update, which is important to the community as a whole, this effort also supports advances in the simulation of recent subcritical neutron noise measurements that require higher computational effort per second of real-time measurement than that of systems typically measured.

97 MATHEMATICS AND COMPUTING↗

Ensuring operator reliability in digital control rooms - 102

The U.S. fleet of nuclear power plants is aging, and efforts are being made to modernize instrumentation and control systems including integration of digital human-machine interfaces (HMIs) into the control room. Automation and control technologies have advanced since original systems were installed in the plants. This presents unique opportunities and challenges in ensuring the replacements are a benefit to the operators and the overall safe and reliable operation of the plant. We discuss trends we have observed from assisting utilities in their modernization efforts. Automation and evolving control rooms come with their own host of problems that existing human factors and engineering methods fail to capture. Traditional approaches are analytical approaches to ensuring reliability. Here we discuss some of the shortfalls of these traditional approaches as plants become increasingly more complex. We suggest that ensuring reliability for modernized and next generation control room will require expanding the scope of what was traditionally considered human factors as well as exploring novel approaches and methodologies for dealing with complexity and rapidly advancing technology. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Development of programmable logic in a rapid, automated, and repeatable fashion for safety-related applications - 385

Digital safety platform considerations for power plant modernization or new build efforts involve a burdensome task of implementing a large number of software-based functions onto a designated platform to match the plant's architecture. This problem is not unique to any particular safety-related platform solution, but is prevalent across the landscape of current Qualified and Licensed Safety-related Instrumentation and Control (I and C) Solutions. As an example when implementing a RG 1.97 compliant Post Accident Monitoring System (PAMS) solution with the Lockheed Martin Nuclear Protection and Control (NuPAC) safety platform, 72 unique software loads are needed to configure the Qualified Display System (QDS). Large scale development and customization efforts to upgrade existing plant functions or systems require the licensee to assume the cost and technical risk necessary to undergo a digital based upgrade. Often the cost, schedule, and technical risk outweighed the benefit of modernization efforts due to the unknown hurdles and past projects that overran cost and schedule projections. Lockheed Martin's Engineering Production Suite (EPS) minimizes the licensee risk and expense associated with modernization efforts through a developer's toolkit for rapid, automated, and repeatable development of Programmable Logic used to implement plant functions and architectures for use on the Lockheed Martin NuPAC platform. This paper describes a use case where by EPS can create system level representations, simulate system functions, partition logic designs to accommodate hardware constraints through logic partitioning, automate the assembly of Application Specific Programmable Logic (ASPL), generate system test cases, program the NuPAC hardware, and perform testing throughout the design workflow. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Flux in a 1D-Reactor Cell by Response Matrix

Legacy computational reactor physics software tools and protocols currently used in support of the Advanced Test Reactor (ATR) core fuel management and safety assurance, and to some extent, experiment management, are inconsistent with the state of modern nuclear engineering practice. Thus, they are difficult, if not impossible, to verify and validate (V and V) according to modern standards. Furthermore, the legacy staff knowledge required for effective application of their application and protocols from the 1960's and 1970's is rapidly being lost due to staff turnover and retirements. In late 2009, the Idaho National Laboratory (INL) initiated a focused effort, the ATR Core Modeling Update Project, to address this situation through the introduction of modern high-fidelity transport theory based computational software and models to replace the legacy methods, with appropriate (V and V) protocols according to applicable national standards. The new computational tools include: - The 2-dimensional HELIOS deterministic transport code for integrated ATR core fuel and experiment management and Core Safety Analysis Package (CSAP) physics data generation. The MCNP5 3-dimensional stochastic simulation (Monte Carlo) code for verification and validation applications and for specialized ATR core analysis along with its longstanding traditional role in experiment support - The 2-dimensional deterministic NEWT transport code and the 3-dimensional KENO6 stochastic simulation code, with their SCALE-based support infrastructure for use in V and V applications and as supporting modules for sensitivity-uncertainty analysis via the TSUNAMI system. The ATTILA 3-dimensional discrete-ordinates transport code for specialized experiment support and possibly additional applications. Regardless of application, every neutron transport code, new or legacy, requires verification. The aim of this presentation is to provide a neutron transport benchmark for a simple 1D-reactor cell in plane geometry. The benchmark finds use as a standard of comparison for the codes listed above intended for the ATR upgrade. The method developed here is in contrast to the finite Fourier transform solution found in Ref. 1, which is limited to isotropic scattering and particularly difficult to numerically implement unless the total cross section is constant over the cell.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NJOY21: Making NJOY Suitable for the 21. Century - 14409

NJOY has a long history of nuclear data processing. NJOY began as a successor to MINX (Multigroup Interpretation of Nuclear X-Sections), a Los Alamos code and was first publicly released in the summer of 1977. The initial version included modules such as RECONR for resonance reconstruction on a union energy grid, BROADR for Doppler broadening plus a new Multigroup averaging program, GROUPR and an associated output module, DTR. Soon thereafter the HEATR, ERRORR, and THERMR modules were added to NJOY. By the mid 1980's, NJOY had a worldwide user community. The current version of NJOY2012 was released in December 2012. NJOY2012 builds on the success of more than 45 years of nuclear data processing at Los Alamos National Laboratory. NJOY2012 was rewritten using Fortran 90/95, utilizing among other things the concepts of public and private variables, eliminating common blocks in lieu of Fortran modules and taking advantage of allocatable arrays. A new version of NJOY is being developed at Los Alamos National Laboratory; NJOY21: NJOY for the 21. Century. NJOY21 uses modern programming standards of quality, reliability, and maintainability. It is being designed from the ground up to utilize modern methods of interacting with computer codes by providing an interactive and scripting (e.g., Python) interface. This will allow scientists the ability to more easily work with the results of the various steps in the processing of nuclear data. NJOY21 will continue the legacy of NJOY by providing a trusted and verified processing code that is fast, flexible, easy to use and customize and capable of handling both modern and established nuclear data formats. This paper introduces NJOY21 by first relating the history of NJOY; the majority of the paper will discuss NJOY21. NJOY21 will take advantage of the many decades of nuclear data processing experience that is contained in NJOY2012 and previous versions of NJOY and extend the capabilities and convenience of processing nuclear data. We will discuss how NJOY21 will remain backwards compatible with NJOY2012 - i.e., how current NJOY2012 input decks will continue to work with NJOY21. We will also discuss the verification and validation of the NJOY21 code base. Finally we will discuss new features in NJOY21 that do not exist in NJOY2012. These features include changes in how a scientist can interact with NJOY to process and verify and validate nuclear data; and the ability to read and write new nuclear data formats. (authors)

42 ENGINEERING↗

Scheduling and Performance of Asynchronous Tasks in Fortran 2018 with FEATS

Most parallel scientific programs contain compiler directives (pragmas) such as those from OpenMP (Hermanns in Parallel programming in Fortran 95 using openMP, 2002. School of Aeronautical Engineering, Universidad Politécnica de Madrid, España, 2011), explicit calls to runtime library procedures such as those implementing the Message Passing Interface (MPI) (in A message-passing interface standard version 4.0, 2021. https://www.mpi-forum.org/docs/mpi-4.0/mpi40-report.pdf), or compiler-specific language extensions such as those provided by CUDA (Ruetsch and Fatica in CUDA Fortran for scientists and engineers: best practices for efficient CUDA Fortran programming, Elsevier, 2013). By contrast, the recent Fortran standards empower developers to express parallel algorithms without directly referencing lower-level parallel programming models (Numrich in Parallel programming with co-arrays, CRC Press, 2018, and Curcic in Modern Fortran: building efficient parallel applications, Manning Publications, 2020). Fortran’s parallel features place the language within the Partitioned Global Address Space (PGAS) class of programming models. When writing programs that exploit data parallelism, application developers often find it straightforward to develop custom parallel algorithms. Problems involving complex, heterogeneous, staged calculations, however, pose much greater challenges. Such applications require careful coordination of tasks in a manner that respects dependencies prescribed by a directed acyclic graph. When rolling one’s own solution proves difficult, extending a customizable framework becomes attractive. Further, the paper presents the design, implementation, and use of the Framework for Extensible Asynchronous Task Scheduling (FEATS), which we believe to be the first task scheduling tool written in modern Fortran. We describe the benefits and compromises associated with choosing Fortran as the implementation language, and we propose ways in which future Fortran standards can best support the use case in this paper.

97 MATHEMATICS AND COMPUTING↗

The evolution of the Human Systems and Simulation Laboratory in nuclear power research

The events at Three Mile Island in the United States brought about fundamental changes in the ways that simulation would be used in nuclear operations. The need for research simulators was identified to scientifically study human-centered risk and make recommendations for process control system designs. This paper documents the human factors research conducted at the Human Systems and Simulation Laboratory (HSSL) since its inception in 2010 at Idaho National Laboratory. The facility’s primary purposes are to provide support to utilities for system upgrades and to validate modernized control room concepts. In the last decade, however, as nuclear industry needs have evolved, so too have the purposes of the HSSL. Thus, beyond control room modernization, human factors researchers have evaluated the security of nuclear infrastructure from cyber adversaries and evaluated human-in-the-loop simulations for joint operations with an integrated hydrogen generation plant. Lastly, our review presents research using human reliability analysis techniques with data collected from HSSL-based studies and concludes with potential future directions for the HSSL, including severe accident management and advanced control room technologies.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Migration of Older to New Digital Control Systems in Nuclear Power Plant Main Control Rooms

The United States (U.S.) Department of Energy (DOE) Office of Nuclear Energy (NE) has the primary mission to advance nuclear power by resolving socio-technical issues through research and development (R&D). One DOE-NE activity supporting this mission is the Light Water Reactor Sustainability (LWRS) program. LWRS has the overall objective to sustain the operation of existing commercial nuclear power plants (NPPs) through conducting R&D across multiple “pathways,” or R&D focus areas. The Advanced Instrumentation, Information, and Control (II&C) Systems Technologies pathway conducts targeted R&D to address aging and reliability concerns with the legacy instrumentation and control (I&C) and related information systems in operating U.S. NPPs. This work involves (1) ensuring that legacy analog II&C systems are not life-limiting issues for the LWR fleet, and (2) implementing digital II&C technology in a manner that enables broad innovation and business improvement in the NPP operating model. Under the LWRS Advanced II&C pathway, Human Factors experts at Idaho National Laboratory (INL) have been conducting R&D in support of NPP main control room (MCR) modernization activities. Work in prior years has focused on migrating analog I&C systems to new digital I&C systems (). In fiscal year 2016 (FY16), one new focus area for this research is migrating older digital I&C systems to new and advanced digital I&C systems. This report summarizes a plan for conducting a digital-to-digital migration of a legacy digital I&C system to a new digital I&C system in support of control room modernization activities.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Impacts of Using Distributed Energy Resources to Reduce Peak Loads in Vermont

To help the United States develop a modern electricity grid that provides reliable power from multiple resources as well as resiliency under extreme conditions, the U.S. Department of Energy (DOE) is leading the Grid Modernization Initiative (GMI) to help shape the future of the nation's grid. Under the GMI, DOE funded the Vermont Regional Initiative project to provide the technical support and analysis to utilities that need to mitigate possible impacts of increasing renewable generation required by statewide goals. Advanced control of distributed energy resources (DER) can both support higher penetrations of renewable energy by balancing controllable loads to wind and photovoltaic (PV) solar generation and reduce peak demand by shedding noncritical loads. This work focuses on the latter. This document reports on an experiment that evaluated and quantified the potential benefits and impacts of reducing the peak load through demand response (DR) using centrally controllable electric water heaters (EWHs) and batteries on two Green Mountain Power (GMP) feeders. The experiment simulated various hypothetical scenarios that varied the number of controllable EWHs, the amount of distributed PV systems, and the number of distributed residential batteries. The control schemes were designed with several objectives. For the first objective, the primary simulations focused on reducing the load during the independent system operator (ISO) peak when capacity charges were the primary concern. The second objective was to mitigate DR rebound to avoid new peak loads and high ramp rates. The final objective was to minimize customers' discomfort, which is defined by the lack of hot water when it is needed. We performed the simulations using the National Renewable Energy Laboratory's (NREL's) Integrated Energy System Model (IESM) because it can simulate both electric power distribution feeder and appliance end use performance and it includes the ability to simulate multiple control strategies.

24 POWER TRANSMISSION AND DISTRIBUTION↗

Connecting LWRS Human Factor Engineering R&D to NUREG-0711 Elements and Modification Activities in Nuclear Power Plants

This Report documents the connection between DOE INL LWRS HFE R&D activities and the technical content of NUREG-0711 Results Summary Reports. Since the current and most recent LWRS HFE activities are associated with modernization and modification of nuclear power electric generating stations, this Report will limit its scope to these topics. NUREG-0711 Results Summary Reports content is shared with US nuclear generating stations by INL HFE staff through HFE Technical Reports to identify potential HEDs with planned upgrades. With such information, utilities have confidence that their upgraded HMIs, operating procedures, conduct of operations, and training with the upgraded systems will ensure their continued safe control of the nuclear plant. The utilities use the HFE Technical Reports as basis documents in their licensing processes used to modify the plant (LAR or 10CFR50.59). For plant modifications, LWRS HFE R&D activities commence as early as an agreement is reached between the utility and the LWRS Program staff. The multiple interfaces usually result in the two organizations meeting at INL, at the utility’s site, and at the vendor’s facility. Since there generally is a shortage of utility based HFE experience, INL HFE staff supports the HFE related utility modernization & modification (obsolescence) activities whenever requested.

10 CFR 50.59↗

Report for 2.2.1 Task 1: Support of Operator in the Loop Studies to be Conducted in the HSSL

Idaho National Laboratory (INL) has contracted IFE to support human factors research to understand how to perform control room modernizations in a manner that maintains or improves human and overall system performance. The US Nuclear industry and in particular several utilities are currently moving forward with a Main Control Room (MCR) modernization projects or future projects to replace current analogue systems with digital. The focus of the current project was to look at different interface platforms and different ways to present information to the interface end user, the plant operator. IFE’s focus was to develop digital soft controls HSI for the Generic Pressurized Water Reactor (gPWR) and a Larger Screen Display (LSD) to support the Control Room Operators. The IFE Advanced Control Room is a combination of the digital soft controls HSI and the LSD that provides a unique opportunity for testing in the HSSL at Idaho National Laboratory.

99 GENERAL AND MISCELLANEOUS↗

Development and evaluation of the conceptual design for a liquid radiological waste system in an advanced hybrid control room

As the current fleet of nuclear power plants are maintaining outdated or obsolete equipment it is common practice to replace worn-out equipment on an as-needed basis resulting in a series of like-for-like replacements of components on the control boards addressing an immediate need to replace equipment. Such upgrades rarely represent an encompassing or systematic vision for control room modernization and do not add the additional benefit of enhanced support for operators in the control room, which tend to overlook the opportunity to reduce operations and maintenance costs through enhanced efficiency. The purpose of the research is to provide guidance on how to realize these opportunities by designing control room human-system interfaces with advanced capabilities in mind. Further, the research seeks to ensure that control room modernizations are undertaken with a sound understanding of the impacts to human operators and are designed based on state-of-the art human factors principles. This report describes the conceptual design of the liquid radiological waste system and several research activities conducted to inform the design philosophy. The report also provides recommended updates to the original design philosophy (published by the researchers in 2018) based on the findings.

99 GENERAL AND MISCELLANEOUS↗

Evaluation of Fire Station 23-425, Mercury, Area 23, Nevada National Security Site, Nye County, Nevada

The U.S. Department of Energy, National Nuclear Security Administration Nevada Field Office (NNSA/NFO) plans to modernize the Mercury town site at the Nevada National Security Site (NNSS) in Nye County, Nevada, to meet current and future National Weapons Science, Global and Homeland Security Programs, and Environmental Management mission requirements. These plans—including anticipated demolitions, consolidations, and rebuilding at Mercury—are considered an undertaking subject to review under Title 54 U.S.C. § 306108, commonly known as Section 106 of the National Historic Preservation Act, Title 54 U.S.C. § 300101, et seq., and its implementing regulations, 36 C.F.R. Part 800. Desert Research Institute (DRI) completed an architectural survey of Mercury. This effort resulted in the identification, recordation, and evaluation of the Mercury Historic District, including the identification of its contributing components. This district was recommended eligible for listing to the National Register of Historic Places (NRHP or National Register) under Criteria A and C but was unevaluated under Criteria B and D at that time. The Mercury Historic District contains 153 individual landscapes, buildings, and structures. Of these resources, all except one are recommended as contributing elements of the historic district during its period of significance corresponding to nuclear testing from 1951 through 1992. During most of the Cold War, the land now called the NNSS was named the Nevada Test Site (NTS). This report concerns the proposed demolition of the 1966 Fire Station 23-425 in Mercury as a phase of the larger modernization program mentioned above. Further evaluation of this property indicates that it does not meet any of the Secretary of the Interior’s Significance Criteria for eligibility for the NRHP at either the national or local levels. However, it does remain an important contributing element of Mercury Historic District. This is part of a series of reports.

54 ENVIRONMENTAL SCIENCES↗

Evaluation of 1950s-Era Architectural Resources in Blocks 10, 11, and 17, Mercury, Area 23, Nevada National Security Site, Nye County, Nevada

The U.S. Department of Energy, National Nuclear Security Administration Nevada Field Office (NNSA/NFO) plans to modernize the Mercury town site at the Nevada National Security Site (NNSS) in Nye County, Nevada, to meet current and future National Weapons Science, Global and Homeland Security Programs, and Environmental Management mission requirements. These plans, including anticipated demolitions, consolidations, and rebuilding at Mercury, are considered an undertaking subject to review under Title 54 U.S.C. § 306108, commonly known as Section 106 of the National Historic Preservation Act, Title 54 U.S.C. § 300101, et seq., and its implementing regulations, 36 C.F.R. Part 800. The Desert Research Institute (DRI) completed an architectural survey of Mercury. This effort resulted in the identification, recording, and evaluation of the Mercury Historic District (Nevada State Historic Preservation Office [SHPO] Resource No. D230), including the identification of its contributing components (Reno et al. 2018). This district was recommended eligible to the National Register of Historic Places (NRHP or National Register) under Criteria A and C but was unevaluated under Criteria B and D at that time. The Mercury Historic District contains 154 individual landscapes, buildings, and structures. Of these resources, DRI recommends all except one as contributing elements of the historic district during its period of significance corresponding to nuclear testing from 1951 through 1992. During most of the Cold War, the land now called the NNSS was formerly the Nevada Test Site (NTS). This report is concerned with evaluating resources dating to the 1950s in Blocks 10, 11, and 17 in Mercury prior to the proposed demolition of 22 resources in Blocks 10 and 11 as a phase of the larger modernization program mentioned above. Although the Mercury Historic District as a whole is eligible to the National Register, further evaluation of these properties as detailed in this report indicates that none are individually eligible under any of the Secretary of Interior’s Significance Criteria, either at the national or local level. This is one report in the report series "Evaluation, Finding of Effect, and Mitigation Documentation for 1950s-Era Architectural Resources, Mercury, Area 23, Nevada National Security Site, Nye County, Nevada".

54 ENVIRONMENTAL SCIENCES↗

Report series: Evaluation, Finding of Effect, and Mitigation Documentation for the 1960s Dormitories and 1950s Quonset Hut Foundations, Mercury, Area 23, Nevada National Security Site, Nye County, Nevada [3 Reports Consolidated]

The U.S. Department of Energy, National Nuclear Security Administration Nevada Field Office (NNSA/NFO) plans to modernize the town of Mercury at the Nevada National Security Site (NNSS) in Nye County, Nevada, to meet current and future National Weapons Science, Global and Homeland Security Programs, and Environmental Management mission requirements. These plans—including anticipated demolitions, consolidations, and rebuilding at Mercury—are considered an undertaking subject to review under Title 54 U.S.C. § 306108, commonly known as Section 106 of the National Historic Preservation Act, Title 54 U.S.C. § 300101, et seq., and its implementing regulations, 36 C.F.R. Part 800. Desert Research Institute (DRI) completed an architectural survey of Mercury in 2017. This effort resulted in the identification, recording, and evaluation of the Mercury Historic District, as well as the identification of its contributing components. This district was recommended eligible to the National Register of Historic Places (NRHP or National Register) under Criteria A and C, but was unevaluated under Criteria B and D at that time. The Mercury Historic District contains 153 individual landscapes, buildings, and structures. Of these resources, DRI recommends all except one as contributing elements of the historic district during its period of significance corresponding to nuclear testing from 1951 through 1992. During most of the Cold War, the land now called the NNSS was referred to as the Nevada Test Site (NTS). This report evaluates resources dating to the 1950s and 1960s between Jangle and Knothole Streets and south of Ranger Avenue in Mercury prior to the proposed demolition of all resources in that area between Sandstone and Greenhouse Avenues as a phase of the larger modernization program mentioned above. Although the Mercury Historic District as a whole is eligible to the National Register, further evaluation of these properties, as detailed in this report, indicates that none are individually eligible under any of the Secretary of Interior’s Significance Criteria, either at the national or local level.

54 ENVIRONMENTAL SCIENCES↗