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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 181 records · Page 10
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

RIA Testing in TREAT

The Transient Reactor Test facility (TREAT) was constructed in the late 1950’s. The reactors flexible design and multi-mission nature provided thousands of transient irradiations for numerous reactor fuels and transient types before being placed in standby in 1994. The reactor resumed operations in 2017 to reclaim its crucial role in nuclear-heated safety research. A principal goal of restarting the reactor was to perform transient testing of Accident Tolerant Fuels (ATF), supporting their safe adoption by the commercial Light Water Reactor (LWR) fleet. In support of the Accident Tolerant Fuels mission, multiple capsules have been designed and are currently being designed to test fuel specimens under varying conditions. The Minimal Activation Retrievable Capsule Holder (MARCH) system was designed to support high throughput testing by using a modular approach where an outer containment structure will act as the primary safety containment. The MARCH system is designed to house various modules for testing different fuel systems in different environments. The near-term tests for the ATF concepts will use the Static Environment Rodlet Transient Test Apparatus (SERTTA), aka MARCH-SERTTA. MARCH-SERTTA is capable of testing a single ATF rodlet 10 cm in length in a static pool of water with initial conditions up to 200°C and 500 psi. The purpose of the MARCH-SERTTA device is to perform integral tests evaluating the Pellet Clad Mechanical Interaction (PCMI) performance of LWR fuel concepts in design basis Reactivity Initiated Accidents (RIA). This will be accomplished by testing a variety of ATF fuel concepts under varying initial thermal-hydraulic conditions and varying prompt energy pulses. The flexibility of TREAT and the MARCH-SERTTA capsule enables testing of a unique, non-fueled, rod specimen under similar conditions as the fueled specimens to perform separate-effect tests. This non-fueled rod will be used to study the thermal-hydraulic behavior (specifically determine the critical heat flux) during these fast prompt transients.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

The History of a Decision: A Standard Vibration Test Method for Qualification

As Mil-Std-810G and subsequent versions have included multiple degree of freedom vibration test methodologies, it is important to understand the history and factors that drove the original decision in Mil-Std-810 to focus on single degree of freedom (SDOF) vibration testing. By assessing the factors and thought process of early Mil-Std-810 vibration test methods, it enables one to better consider the use of multiple degree of freedom testing now that it is feasible with today’s technology and documented in Mil-Std-810. This paper delves into the details of the decision made in the 1960s for the SDOF vibration testing standards in Mil-Std-810 beyond the limitations of technology at the time. We also consider the implications for effective test planning today considering the advances in test capabilities and improvements in understanding of the operational environment.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Application of printed strain gauges in prototypical nuclear reactor conditions

This report discusses the current development and testing of additively manufactured temperature sensors and resistive/ capacitive strain gauges. This has potential to improve the sensor design and manufacturing techniques to meet the requirements of in-pile monitoring of mechanical properties and structural health of materials and specimens in current and advanced nuclear test reactors (i.e., in terms of environment conditions, sample geometry, and materials compatibility). The developmental additively manufactured strain gauges are exposed to separate effects testing (i.e., mechanical strain (up to 1000 µe), high temperature (up to 700 °C)) to determine environmental factors that affect the performance of the strain gauge. The robustness and integrity of a printed strain gauge layer exposed to a molten salt environment for up to 500 hours was evaluated, with the goal of assessing material compatibility of printed strain sensors for instrumented surveillance test articles in molten salt reactors. In addition, sensor qualification methodologies are further developed for determining the reliability and robustness at the interface of the additively manufactured strain gauge materials.

36 - MATERIALS SCIENCE↗
DOE OSTI2012

Experimental Studies of NGNP Reactor Cavity Cooling System With Water

This project will investigate the flow behavior that can occur in the reactor cavity cooling system (RCCS) with water coolant under the passive cooling-mode of operation. The team will conduct separate-effects tests and develop associated scaling analyses, and provide system-level phenomenological and computational models that describe key flow phenomena during RCCS operation, from forced to natural circulation, single-phase flow and two-phase flow and flashing. The project consists of the following tasks: Task 1. Conduct separate-effects, single-phase flow experiments and develop scaling analyses for comparison to system-level computational modeling for the RCCS standpipe design. A transition from forced to natural convection cooling occurs in the standpipe under accident conditions. These tests will measure global flow behavior and local flow velocities, as well as develop instrumentation for use in larger scale tests, thereby providing proper flow distribution among standpipes for decay heat removal. Task 2. Conduct separate-effects experiments for the RCCS standpipe design as two-phase flashing occurs and flow develops. As natural circulation cooling continues without an ultimate heat sink, water within the system will heat to temperatures approaching saturation , at which point two-phase flashing and flow will begin. The focus is to develop a phenomenological model from these tests that will describe the flashing and flow stability phenomena. In addition, one could determine the efficiency of phase separation in the RCCS storage tank as the two-phase flashing phenomena ensues and the storage tank vents the steam produced. Task 3. Develop a system-level computational model that will describe the overall RCCS behavior as it transitions from forced flow to natural circulation and eventual two-phase flow in the passive cooling-mode of operation. This modeling can then be used to test the phenomenological models developed as a function of scale.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

RELAP5-3D validation studies based on the High Temperature Test facility

In the spring and summer of 2019, experiments were conducted at the High Temperature Test Facility (HTTF) that form the basis of an upcoming high-temperature gas-cooled reactor (HTGR) thermal hydraulics (T/H) benchmark. HTTF is an integral effects test facility for HTGR T/H modeling validation. This paper presents RELAP5-3D models of two of those experiments: PG-27, a pressurized conduction cooldown (PCC); and PG-29, a depressurized conduction cooldown (DCC). These models used the RELAP5-3D model of HTTF originally developed by Paul Bayless as a starting point. The sensitivity analysis and uncertainty quantification code, RAVEN was used to perform calibration studies for the steady-state portion of PG-27. Here we developed four PG-27 calibrations based on steady-state conditions. These calibrations all used an effective thermal conductivity equal to 36 % of the measured thermal conductivity, but they differed with respect to the frictional pressure drops and radial conduction models. These models all captured the trends in steady-state temperature distributions and transient temperature behavior well. All four calibrations show room for improvement in predicting the transient temperature rise. The smallest error in temperature rise during the transient was a 21 % underprediction, and the largest was a 48 % underprediction. The errors in transient temperature rise are largely a result of a mismatch in power density between the RELAP5-3D model and the experiment due to the location of active heater rods along the boundary between heat structures in the model. The best of these calibrations was applied to PG-29 to model the DCC. Once again, temperatures during the transient were underpredicted but trends in temperature were captured. The RELAP5-3D model captured trends in the data but could not reproduce measured temperatures exactly. This result is not attributed to deficiencies in the experimental data or to RELAP5–3D itself. Rather, this result likely arises due to the some of the assumptions and decisions made when the RELAP5-3D model was first developed, prior to the execution of HTTF experiments. An agreement in prediction of temperature trends but challenges reproducing HTTF temperatures within measurement uncertainty is consistent with previous analyses of HTTF in the literature. Future RELAP5-3D validation activities centered around HTTF may be able to provide greater insight into the code’s capabilities for HTGR modeling with a more finely nodalized model.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Nuclear characterization of a general-purpose instrumentation and materials testing location in TREAT

The Transient Reactor Test facility (TREAT) was constructed in the late 1950s, provided thousands of transient irradiations before being placed in standby in 1994, and resumed operations in 2017 in order to reclaim its crucial role in nuclear-heated safety research. The latter half of TREAT’s historic operation was best known for integral-scale testing of fuel specimens under postulated reactor plant accident conditions, while TREAT’s earlier history included extensive simpler phenomena identification tests that elucidated fundamental behaviors and paved the way for these integral-scale tests. Advances in modern computational capabilities and a resurgence of interest in novel reactor technology have created an opportunity for emphasizing modernized science-based and separate effects test capabilities once again at TREAT. An innovative approach to this type of testing has been developed to leverage minor radioactivity built-in during brief TREAT irradiations in low activation hardware to facilitate handling for materials and instrumentation testing. This capability, termed the Minimal Activation Retrievable Capsule Holder (MARCH) irradiation vehicle system, will be used for inaugural fueled experiments in TREAT’s modern era as well as novel approaches to study materials undergoing neutron irradiation and instrumentation development and qualifications. This paper describes a comprehensive nuclear characterization, obtained via computational modeling, of the test position in the MARCH system. Though directly applicable to the MARCH system, the results also provide general quantification of the nuclear performance of the reactor and potential test materials, crucial for evaluating potential experiment design and response in TREAT. The neutron and photon flux environment was calculated via MCNP with ENDF/B-VII.1 nuclear data. Wire heating rate and DPA calculations were also performed and evaluated using historic TREAT test data. Furthermore, these calculations were performed to provide steady-state baseline reference values typical in both half- and full-slotted TREAT core configurations, enabling design scoping analysis prior to development of more specific design testing needs and transient testing experimentation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

A Cultural Resources Inventory and Historical Evaluation of the Smoky Atmospheric Nuclear Test, Areas 8, 9, and 10, Nevada National Security Site, Nye County, Nevada

This report presents the results of a National Historic Preservation Act Section 106 cultural resources inventory and historical evaluation of the 1957 Smoky atmospheric test location on the Nevada National Security Site (NNSS). The Desert Research Institute (DRI) was tasked to conduct a cultural resources study of the Smoky test area as a result of a proposed undertaking by the Department of Energy Environmental Management. This undertaking involves investigating Corrective Action Unit (CAU) 550 for potential contaminants of concern as delineated in a Corrective Action Investigation Plan. CAU 550 is an area that spatially overlaps portions of the Smoky test location. Smoky, T-2c, was a 44 kt atmospheric nuclear test detonated at 5:30 am on August 31, 1957, on top of a 213.4 m (700 ft) 200 ton tower (T-2c) in Area 8 of the NNSS. Smoky was a weapons related test of the Plumbbob series (number 19) and part of the Department of Defense Exercise Desert Rock VII and VIII. The cultural resources effort involved the development of a historic context based on archival documents and engineering records, the inventory of the cultural resources in the Smoky test area and an associated military trench location in Areas 9 and 10, and an evaluation of the National Register eligibility of the cultural resources. The inventory of the Smoky test area resulted in the identification of structures, features, and artifacts related to the physical development of the test location and the post-test remains. The Smoky test area was designated historic district D104 and coincides with a historic archaeological site recorded as 26NY14794 and the military trenches designed for troop observation, site 26NY14795. Sites 26NY14794 and 26NY14795 are spatially discrete with the trenches located 4.3 km (2.7 mi) southeast of the Smoky ground zero. As a result, historic district D104 is discontiguous and in total it covers 151.4 hectares (374 acres). The Smoky test location, recorded as historic district D104 and historic sites 26NY14794 and 26NY14795, is the best preserved post-shot atmospheric nuclear tower test at the NNSS and possibly in the world. It is of local, national, and international importance due to nuclear testing’s pivotal role in the Cold War between the United States and the former Soviet Union. The district and sites are linked to the historic theme of atmospheric nuclear testing. D104 retains aspects of the engineering plan and design for the Smoky tower, instrument stations used to measure test effects, German and French personnel shelters, and military trenches. A total of 33 structures contribute to the significance of D104. Artifacts and features provide significant post-test information. Historic district D104 (discontiguous) and historic site 26NY14794 (the Smoky test area) are eligible for listing on the NRHP under Criteria A, B, C, and D. The historic site 26NY14795 (the Smoky military trenches) is eligible for listing under Criteria A, C, and D. Several items have been identified for removal by the CAU 550 investigation. However, none of them is associated with the Smoky atmospheric test, but with later activities in the area. The military trenches are not part of CAU 550 and no actions are planned there. A proposed closure of the Smoky test area with restrictions will limit access and contribute to the preservation of the cultural resources. It is recommended that the Smoky historic district and sites be included in the NNSS cultural resources monitoring program.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2014

Reversible Bending Fatigue Testing on Zry-4 Surrogate Rods

Testing high-burnup spent nuclear fuel (SNF) presents many challenges in areas such as specimen preparation, specimen installation, mechanical loading, load control, measurements, data acquisition, and specimen disposal because these tasks are complicated by the radioactivity of the test specimens. Research and comparison studies conducted at Oak Ridge National Laboratory (ORNL) resulted in a new concept in 2010 for a U-frame testing setup on which to perform hot-cell reversible bending fatigue testing. Subsequently, the three-dimensional finite element analysis and the engineering design of components were completed. In 2013 the ORNL team finalized the upgrade of the U-frame testing setup and the integration of the U-frame setup into a Bose dual linear motor test bench to develop a cyclic integrated reversible-bending fatigue tester (CIRFT). A final check was conducted on the CIRFT test system in August 2013, and the CIRFT was installed in the hot cell in September 2013 to evaluate both the static and dynamic mechanical response of SNF rods under simulated loads. The fatigue responses of Zircaloy-4 (Zry-4) cladding and the role of pellet pellet and pellet clad interactions are critical to SNF vibration integrity, but such data are not available due to the unavailability of an effective testing system. While the deployment of the developed CIRFT test system in a hot cell will provide the opportunity to generate the data, the use of a surrogate rod has proven quite effective in identifying the underlying deformation mechanism of an SNF composite rod under an equivalent loading condition. This paper presents the experimental results of using surrogate rods under CIRFT reversible cyclic loading. Specifically, monotonic and cyclic bending tests were conducted on surrogate rods made of a Zry-4 tube and alumina pellet inserts, both with and without an epoxy bond.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

A methodology for selection of solid desiccants in energy recovery ventilators

Controlling indoor humidity levels is essential for maintaining acceptable indoor air quality in buildings. The use of energy recovery ventilators (ERVs) is an energy-efficient way to regulate indoor air humidity. Fixed-bed regenerators and rotary wheels are widely used ERVs because of their high sensible and latent effectiveness. These ERVs are made of desiccant-coated substrates, which enable them to transfer moisture between the supply and exhaust air streams. However, the moisture transfer ability of ERVs depends on the physiochemical and sorption properties of desiccants. Extensive, full-scale experiments are required to determine the best desiccant material for these systems. This paper presents a simplified method of selecting suitable desiccant materials for ERVs. The methodology involves important characterization methods, literature correlations for performance prediction, and cost-effective testing methods prior to full-scale testing, and full-scale test methods are discussed in detail. Furthermore, the performance of a few newly derived materials is evaluated and compared with that of conventional desiccants such as silica gel and molecular sieves. The highest latent effectiveness was obtained for composite of super absorbent polymer (SAP) with potassium formate (SAP-HCO2K-50 %), all-polymer porous solid desiccant (APPSD) and metal organic framework (MOF)–MIL–101 (Cr), followed by activated carbon fibre felt (ACFF) Silica sol-LiCl30, SAP, silica gel, MOF–303, and molecular sieve. Researchers and manufacturers would benefit from the proposed methodology and presented data in developing new desiccant materials for ERV applications.

Energy recovery↗
DOE OSTI2019

Progress Report on Modified Burst Testing and Alternative Test Methodologies

Performance of accident-tolerant fuel candidates must be assessed during the pellet-cladding mechanical interaction (PCMI) phase of a design-basis reactivity-initiated accident (RIA). Therefore, in this report, we describe Fiscal Year (FY) 2019 research activities on the development of separate-effects testing for ATF cladding candidate iron-chromium-aluminum (FeCrAl) tube specimens under loading conditions comparable to the PCMI. The first section of this report provides modified-burst test (MBT) results of unirradiated FeCrAl tubes at various pressurization-rates and an RIA-relevant temperature (275°C). The failure strain of FeCrAl was observed and calculated using a high-speed digital-image correlation (DIC) technique developed recently under the Department of Energy (DOE)’s Advanced Fuel Campaign (AFC). To overcome the complexities of performing MBT in a hot-cell environment on irradiated samples, a “plane-strain tension” (PST) test was proposed as an alternative methodology for PCMI testing. The relevance of the PST test as an alternative to MBT for irradiated specimens was explicated using the data available in the literature. The design of hot-cell compatible PST fixtures and a specific holder for PST hot-cell specimen machining were developed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Post-Irradiation Examination of Low Neutron Dose Metal Fuel Samples

The FCT irradiation series at the High Flux Isotope Reactor (HFIR) using the Hydraulic Tube Facility (HT) was initiated in FY-2012 with the goal of performing irradiation testing on advanced fuels and materials. Data from the FCT irradiation series was to be used to further the fundamental scientific understanding of phenomena involved in the evolution of metal fuels microstructure under irradiation based on separate effects testing of factors including elemental composition, temperature, and neutron fluence. This data would then be used to inform theoretical models for in-service performance of metallic fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Improved Verification and Validation Testing and Tools including Nuclear Criticality Safety Applications with the MCNP6.3® Code [Abstract]

A new Python-based framework has been developed to enable a more consistent layout with automatable setup, execution, and documentation of all verification and validation (V&V) test suites previously established for use with the MCNP code. In this paper, the new general framework for the V&V test suites is discussed, including information on all of the current capabilities and plans for future capabilities. For nuclear criticality safety applications, the existing V&V benchmark problems within the criticality, extended criticality, and analytic k-effective test suites have been ported into this new framework. In addition, the status and updates to the Rossi-alpha and subcritical multiplicationtest suites will be discussed. Some V&V results exercising new MCNP6.3 capabilities will be demonstrated.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2022

Pressure Drop in Seven-Pin Wire-Wrapped Rod Bundle for the Sodium Cartridge Loop in Versatile Test Reactor

This work studies the hydrodynamics of the seven pin wire-wrapped rod bundle in the sodium cartridge loop for the Versatile Test Reactor (VTR) through scaled water experiments and computational fluid dynamics (CFD) simulations. Here, the scaling analysis is first performed to demonstrate the hydrodynamic similarity between water and sodium flows at the same Reynolds number Re. A separate-effects test facility is designed and constructed based on the scaling analysis. Detailed experimental data on the pressure drop covering a wide range of Re values (1165 to 27 689) are obtained, which are used to evaluate existing correlations for friction factor and to benchmark CFD simulations. The experimentally determined friction factors f Exp agree well with the Upgraded Cheng and Todreas Detailed Correlation and Pacio-Chen-Todreas Detailed Model within ±7% but are significantly underpredicted by Rehme's correlation by 25%. Various CFD near-wall treatment methods are tested using ANSYS Fluent and evaluated by experimental data. It is found that when the recommended wall y + values are met, most of the near-wall treatment methods can give accurate friction factor predictions. The resolved near-wall method (y + ~ 1) with the Shear Stress Transport k - ω turbulence model and the scalable wall functions (y + > 12) with the realizable k - ε turbulence model can predict f Exp within ±5% The standard wall functions (y + > 30) and nonequilibrium wall functions (y + > 30) with the realizable k - ε model can predict f Exp within ±10%.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Analysis of Transient Groundwater Flow Near the Cheshire Test at Pahute Mesa, Nevada National Security Site

This report focuses on analysis and determination of the causes of transient water levels measured between 1964 and 2017 near the Cheshire underground nuclear test, which was detonated on February 14, 1976 in Area 20 (USDOE, 2015) of the Pahute Mesa (PM) testing areas of the Nevada National Security Site (NNSS) located in southern Nevada (Figure 1). Transient water levels near Cheshire indicate influences of a water-production well and yet-unexplained water level rise attributed to testing effects (Jackson and Fenelon, 2018; Jackson et al. 2021). Transient groundwater flow is relevant to the implementation of the hydrologic source term (HST), which prescribes the quantity and release rates of radionuclide source terms to be applied at individual test locations in assessment of regional-scale radionuclide transport (Carle and Tompson, 2021; Carle et al., 2021), as previously investigated for NNSS testing areas at PM (Pawloski et al., 2001), Frenchman Flat (Carle et al, 2007), Yucca Flat (Mcnab, 2009; Tompson, 2009), and Rainier Mesa (Carle, 2011; Tompson et al., 2011).

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2022

Analysis of Transient Groundwater Flow Near the Cheshire Test at Pahute Mesa, Nevada National Security Site

This report focuses on analysis and determination of the causes of transient water levels measured between 1964 and 2017 near the Cheshire underground nuclear test, which was detonated on February 14, 1976 in Area 20 (USDOE, 2015b) of the Pahute Mesa (PM) testing areas of the Nevada National Security Site (NNSS) located in southern Nevada (Figure 1). Transient water levels near Cheshire indicate influences of a water-production well and yet-unexplained water level rise attributed to testing effects (Jackson and Fenelon, 2018; Jackson et al. 2021). Transient groundwater flow is relevant to the implementation of the hydrologic source term (HST), which prescribes the quantity and release rates of radionuclide source terms to be applied at individual test locations in assessment of regional-scale radionuclide transport (Carle and Tompson, 2021; Carle et al., 2021), as previously investigated for NNSS testing areas at PM (Pawloski et al., 2001), Frenchman Flat (Carle et al, 2007), Yucca Flat (Carle et al., 2009; Mcnab, 2009; Tompson et al., 2009), and Rainier Mesa (Carle, 2011; Tompson et al., 2011).

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2023

FY23 Report on Water NSTF Testing at Two-Phase Conditions: Off-normal Scenarios

Under support from the Department of Energy (DOE) and the Office of Advanced Reactor Technologies (ART), a large-scale test facility has been constructed at Argonne National Laboratory to generate NQA-1 qualified validation data for passive decay heat removal systems in advanced reactors. The Natural convection Shutdown heat removal Test Facility (NSTF) reflects key features of a ½ scale, water-based, Reactor Cavity Cooling System (RCCS) and is intended to study the behavior, bound performance, and ultimately guide design decisions for passive decay heat removal systems for advanced reactors. In addition to the experimental activities detailed in this report, a supportive computational modeling effort is on-going which has been demonstrated to significantly strengthen the experimental program while also improving accuracy of the computer models. Together these create a mutually beneficial relationship integral to meeting the overall program objective of examining the heat removal performance of the RCCS concept. This report details the experimental activities and upgrades performed during the program’s fifth year of water-based operation. The theme of this year of testing was examining off-normal scenarios. In practice, that meant examining scenarios with blockages or other design basis conditions. From a maintenance and capability standpoint that meant assessing valving options for implementing blockages, assessing drain and refill capabilities, and ensuring in all cases that operators could safely, efficiently, and repeatably bring the facility to an off-normal condition and return it to the normal operating state. In one notable case, no solution existed, nor was a solution commercially available, that met the facility needs for implementing blockages within the two-phase region of the chimney. A valve and feedthrough were custom-designed for NSTF to operate submerged in saturated water while being safely actuated from outside the tank. The fifth year of testing included eight matrix tests consisting of 162 hours of active heating, 9,735 kWh of electrical heating, with seven tests classified as Accepted per NQA-1 and one as Trending. All matrix tests were performed at two-phase flow conditions. The power parametric series from previous years was extended to include another scenario with a decay load equivalent to 1.75 MWt, full scale, further resolving the system flow oscillation response as a function of input power. A depletion test was performed to extend previous work and examine the system’s approach to stagnation and geysering that were not previously observed. This test began at an initial inventory level of 50%, prototypic power of 2.1 MWt, and employed an accelerated drain of 0.75 gpm plus boiloff until stagnation conditions were achieved and multiple geysering events were recorded. Further exploring the geysering phenomenon was a separate effects test examining static boiling in the risers and chimney. This static boiling test was performed at 48 kWe, a higher power than testing in the previous year, and included enhancements to the drain system to remove inventory in a better-controlled and more reliable manner. Two tests were performed, at prototypic conditions of 1.4 MWt and 2.4 MWt, respectively, examining the system response to throttled conditions in the singlephase region of the piping. A ball valve immediately before he riser inlet header was increasingly throttled, progressing the system through a repeatable pattern of two-phase oscillations and stable regimes before ultimately inducing stagnation and geysering. Concluding the year, two tests were performed at baseline conditions of 70% initial inventory level and 2.1 MWt of prototypic decay heat to examine both repeatability and the response to throttling within the two-phase region, at the inventory tank inlet using the custom valve mentioned above. The first such test repeated transient operating conditions and revealed significant sensitivity t

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Increased abundance of a common scavenger affects allocation of carrion but not efficiency of carcass removal in the Fukushima Exclusion Zone

Abstract The 2011 nuclear accident in Fukushima, Japan caused the evacuation of > 100,000 people and prompted studies on environmental impacts of radiological contamination. However, few researchers have explored how the human evacuation has affected ecosystem processes. Despite contamination, one common scavenger (wild boar, Sus scrofa ) is 2–3× more abundant inside the Fukushima Exclusion Zone (FEZ). Shifts in abundance of some scavenger species can have cascading effects on ecosystems, so our objective was to investigate impacts of the evacuation and the resulting increase in wild boar on vertebrate scavenger communities. We deployed cameras at 300 carcasses in the FEZ and a nearby inhabited area, and quantified carcass fate, scavenger species, and detection/persistence times. We also tested effects of carcass size and habitat on scavenger community composition and efficiency by balancing trials across two carcass sizes and habitats in each zone. Overall scavenger richness and carcass removal rates (73%) were similar in the FEZ and inhabited area, but species-specific carcass removal rates and occurrence differed between zones. Wild boar removed substantially more carcasses inside the FEZ, with implications for nutrient and contaminant distribution. Our results suggest carcass size affects scavenging dynamics more than human activity or habitat, and abundance changes of common scavengers can influence carrion resource allocation.

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2022

Advanced Reactor Control and Operations (ARCO): A University Research Facility for Developing Optimized Digital Control Rooms

The Advanced Reactor Control and Operations (ARCO) facility was constructed in January 2018 to serve as a test bed for advanced reactor control rooms and operator support systems. Since then, it has supported human-machine interface user experience research, fault detection and mitigation technology development, control room concept of operations development, and remote operations research. ARCO serves as the control room for the Compact Integral Effects Test (CIET) facility, which replicates the primary-side flow paths and thermal-hydraulic behavior of a fluoride-salt-cooled high-temperature reactor (FHR) using simulant fluids and scaling principles. New reactor designs feature different operating conditions and scenarios than those in existing reactors. ARCO supports the research and development of digital tools for operator communications, intuitive real-time data analysis, online health monitoring and prognostics, and control room cybersecurity. By integrating these different technologies, ARCO acts as a prototypical control system to iteratively develop methods and tools of operation in advanced small modular nuclear reactors. This paper describes the features of and challenges to operating advanced small modular reactors underlying the design basis for ARCO and its operator support systems.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗