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All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Enhanced Component Performance Study: Motor Driven Pumps 1998–2022

This report presents an enhanced performance evaluation of motor-driven pumps (MDPs) at U.S. commercial nuclear power plants. The data used in this study are based on the operating experience failure reports from calendar year 1998 through 2022 as reported in the Institute of Nuclear Power Operations (INPO) Industry Reporting and Information System (IRIS). The MDP failure modes considered for standby systems are fail to start (FTS), fail to run (FTR) for one hour of operation (FTR=1H), FTR after one hour of operation (FTR>1H), and for normally running systems FTS and FTR. An eight-hour unreliability estimate is also calculated and trended. The component reliability estimates and the reliability data are trended for the most recent 10-year period while yearly estimates for reliability are provided for the entire study period. The following increasing trends were identified for MDPs for the most recent 10-year period: (1) Standby MDP frequency of start demands (demands per reactor year), (2) Standby MDP frequency of FTR=1H hours (hours per reactor year), (3) Standby MDP frequency of FTR>1H hours, and (4) Normally running MDP frequency of run hours. The following decreasing trends were identified for MDPs for the most recent 10-year period: (1) Standby MDP FTR=1H failure probability, (2) Normally running MDP FTR failure rate, (3) Standby MDP unavailability, (4) Standby MDP total unreliability (8-hour mission), and (5) Normally running MDP total unreliability (8-hour mission).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

U.S. Industry Opportunities for Advanced Nuclear Technology Development Phase II (Final Report)

This limited scope study aims at preserving the results of experimental programs for the safety of light water cooled and advanced reactors by locating and documenting, to the extent possible, where the experimental test information for these programs has been archived. The Light Water Reactor Data Preservation Activity Team initially identified seven (7) experimental programs, exceeding the required five (5) in the scope of work, which were determined to be “at risk” of potentially losing valuable data. The seven (7) experimental programs/subject areas identified are (descriptions of each experiment/experience are provided in Section 3.0): 1. FERMI-1 Reactor Accident 2. Fission Product Behavior During the In-Pile Severe Fuel Damage Test SFD I-4 3. Containment Iodine Computer Code Exercise Based on Radioiodine Test Facility (RTF) Experiment 4. Wide Range Piping Integrity Demonstration (WIND) Project 5. Iodine Chemical Research in Canada 6. High Temperature Fission Product Chemistry and Transport in Steam 7. Anything related to radioactive Methyl Iodide.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Versatile semi-batch apparatus for manometric measurement of gas–solid reaction rates

In this study, we describe the design, construction, operation, and performance of a simple and versatile semi-batch reactor that is especially useful for measurement of gas/solid reaction rates at pressures in the range of 1 mTorr to 1500 Torr. The reactor operates by repeatedly imposing small AC modulations of reactant gas pressure on top of a much larger DC pressure background. Based on the rate of pressure relaxation following each AC pulse, the reaction rate is determined. Our design is characterized by modular construction from off-the-shelf, ultra-high-vacuum-compatible components, which facilitate easy retrofitting and adaptation to a range of experimental conditions. Automated experiment control and data acquisition is accomplished via a custom National Instruments © LabView virtual instrument. Data analysis is automated using a custom series of Mathworks © Matlab scripts. We demonstrate reactor performance through measurements of hydrogenation kinetics for a composite H 2 getter material consisting of 1,4-bis(phenylethynyl)benzene mixed with a palladium/carbon catalyst.

47 OTHER INSTRUMENTATION↗
DOE OSTI2023

Cross-Section Comparison for Pu-238 Production in the Advanced Test Reactor at Idaho National Laboratory

Qualification of Advanced Test Reactor (ATR) positions for Pu-238 production has been ongoing at Idaho National Laboratory (INL). The ATR qualifications have stretched over multiple years during which new techniques have been developed and made available for ATR experiment neutronic analysis. As part of the transition to newer codes, new cross-section libraries have been evaluated for use in the Pu-238 production experiment analysis. A comparative study was done using the MCNP ORGIEN Activation Analysis (MOAA) tool between ENDF/B-VII.0 and ENDF/B-VIII.0 cross sections to capture the impact of the change in cross-sections on the analysis needed to qualify Pu-238 production targets. All comparisons were done assuming the ATR GEN-I targets were located in the south flux trap of the ATR. Finally, an overview of how this qualification and potential irradiation fits into Pu-238 is discussed.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2023

Enhanced Component Performance Study: Motor Operated Valves 1998–2022

This report presents an enhanced performance evaluation of motor-operated valves (MOVs) at U.S. commercial nuclear power plants. The data used in this study are based on the operating experience failure reports from calendar year 1998 through 2022 as reported in the Institute of Nuclear Power Operations (INPO) Industry Reporting and Information System (IRIS). The MOV failure modes considered are fail to open or close (FTOC), fail to operate or control (FTOP), and spurious operation (SO). The component reliability estimates and the reliability data are trended for the most recent 10-year period while yearly estimates for reliability are provided for the entire study period. The following increasing trend was identified for MOVs for the most recent 10-year period: • Low-demand MOV frequency of FTOC demands (demands per reactor year). The following decreasing trends were identified for MOVs for the most recent 10-year period: • Low-demand MOV FTOC failure probability • Low-demand MOV frequency of FTOC events (failures per reactor year).

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2020

Active Irradiation Testing of Temperature Sensing Capability of Clad Sapphire Optical Fibers with Type 2 Bragg Gratings using Optical Backscatter Reflectometry

This work investigated the temperature-sensing capabilities of Type II fiber Bragg gratings (FBGs) inscribed in sapphire optical fiber in the Ohio State University Research Reactor (OSURR). A Luna Innovations 4600 optical backscatter reflectometer (OBR) was used to interrogate FBGs in three sapphire optical fibers with an internal cladding that was developed by Ohio State. The sapphire fibers were fusion-spliced to silica lead fibers. In addition to the sapphire fibers, two silica fibers with the same Type II FBGs were included in the experiment as a reference. The fiber-based sensors were irradiated in the Central Irradiation Facility (CIF) of the OSURR for a total of 40 hours over 5 days at a reactor power of 450 kW, resulting in a total neutron fluence of approximately 2.5 × 10 18 n/cm 2 and a gamma dose of approximately 3.48 Grad. Two K-type thermocouples were included within the rig: one at the core centerline and the other 12 inches above the top of the core. The temperature was not actively controlled, but the thermocouples were used to provide a reference to which the fiber-based temperature measurements could be compared.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Equivalent Fundamental-Mode Source Simulations for Spherical Uranium and Plutonium Systems [Abstract]

The equivalent fundamental-mode source refers to a “source” which is identically distributed in space, energy, and angle as that of a fundamental-mode fission source distribution. It has previously been shown that the system fixed source multiplication ($M_{fs}$) relates to the effective multiplication factor ($\kappa_{eff}$) as $M_{fs} = \frac{g^*}{1-k_{eff}}$ The term $\mathcal{g}$* therefore relates the system source distribution to the system equivalent fundamental-mode source. As previously shown, the $\mathcal{g}$* term is approximately 1 for deeply subcritical systems and then diverges as the system multiplication increases. For a system with a point source located in the center of an assembly (such a sphere), $\mathcal{g}$* will be greater than 1; this happens because there is a higher probability of induced fission when all of the starter neutrons originate from the center of the sphere. For a uniform source, $\mathcal{g}$* is less than 1, but does not diverge from 1 as dramatically as seen by a point source. One reason that $\mathcal{g}$* is an important parameter is that it is used in measurement methods to infer reactor kinetics parameters ($β_{eff}$ being one example). This work will discuss multiple ways to simulate $\mathcal{g}$* using MCNP®6.2, which include use of traditional methods versus a single input file method that was recently published. A discussion related to methods to simulate $M_{fs}$ will be presented. Spherical systems of Highly Enriched Uranium (HEU) will be compared with previous works. In addition, results of plutonium including multiple reflector materials will be investigated. Simulated systems will include both parametric studies as well as configurations used in recent experiments performed at the National Criticality Experiments Research Center (NCERC). These studies will include subtleties associated with system geometry that have not been previously shown. Last, this work will investigate the relationship between $\mathcal{g}$* and reactor kinetics parameters (such as $β_{eff}$).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Final Design of Vacuum Pumping Systems for the Material Plasma Exposure eXperiment

The Material Plasma Exposure eXperiment (MPEX) vacuum pumping system is responsible for creating prototypic conditions in the plasma material interaction chamber that mimic those in a fusion reactor divertor region. Additionally, the vacuum system needs to minimize the pressure in the plasma heating region to improve the coupling of the electron cyclotron heating and ion cyclotron heating to the plasma, minimizing waste heat exhausted to high heat flux components. Here, the final design of the system has been sized to comply with the vacuum pump operating environment and to reduce the number of unique pumps required while meeting performance requirements. Bounding cases with and without the plasma present have been developed in the pumping analysis, and an initial calculation has been performed based on the plasma pumping identified in Proto-MPEX operation though this result will remain unverified until MPEX operation.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Phase I Summary of Management Activities - Advanced Construction Technology Initiative Project

As the end of the year is approached, it is essential to review the National Reactor Innovation Center (NRIC) management activities undertaken during the Advanced Construction Technology Initiative (ACTI) project. The ACTI project was launched to support a transformation in nuclear energy construction, management, and deployment costs, enabling nuclear energy to make important contributions to the energy systems of the future. This transformation will increase the confidence of investors, energy system planners, policymakers, and ultimately consumers in the capability of nuclear energy to meet future needs; thus, it represents a critical element of advanced nuclear energy system demonstration. Development and/or demonstration project(s) will consider regulatory requirements for commercial nuclear implementation and will incorporate strategies to develop regulator experience and review of the technology. GE Hitachi (GEH), a leading player in the nuclear industry that offers a range of products and services including nuclear reactor design, fuel supply, and decommissioning solutions, submitted a proposal for the ACTI project. GEH was awarded the contract for the first phase of the project.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

High-Fidelity CFD Modeling of Cryogenic Hydrogen Isotope Extrusion for Fusion Reactor Pellet Fueling

This study investigates the extrusion processes of deuterium and protium using ANSYS-Polyflow. The geometries and computational fluid dynamics (CFD) settings closely replicate the experimental setups and data acquired from the extruder experiments at Oak Ridge National Laboratory (ORNL) for validation purposes. We explore the impacts of (1) slip versus non-slip boundary conditions and (2) the use of constant, temperature-, and shear rate–dependent viscosities, concluding that the implementation of non-slip wall boundary conditions combined with shear rate–dependent viscosity produced more accurate predictions. The simulations achieved excellent agreement with the experimental data, with relative differences of only 5% for deuterium, and 3% to 6% for protium. This is the first time that experimental extrusion data at ORNL have been accurately predicted through high-fidelity CFD modeling. In conclusion, the advancements offer valuable insights and a foundational modeling tool for optimizing pellet injectors for ITER and other future reactor-scale devices.

ANSYS-Polyflow↗
DOE OSTI2025

Potential Supply Chain and Cost Reduction Strategies for Advanced Nuclear Reactors

This milestone represents a level of effort to provide recommendations for a supply chain plan and establish relationships between advanced reactor companies and large companies in the oil & gas and chemical industries. It represents a starting point to develop the plan with options, general equipment lists, and rough associated cost estimates of major equipment and the associated systems needed. This effort will provide an opportunity to leverage the experience that oil & gas and chemicals companies possess in large capital equipment procurement, modularization, acquisition and installation. Potential supply chain solutions will be discussed and preliminary options for worldwide manufacturers will be elucidated. The goal of this effort will be to leverage the years of accumulated experience in the oil & gas and chemical industries in the area of capital equipment acquisition to understand the potential supply chain options and to reduce potential acquisition costs by optimizing the supply chain options for advanced nuclear reactor modularization, manufacturing and acquisition.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Reactor Physics Simulations of the High Flux Isotope Reactor Permanent Beryllium Reflector Number 5 Design

This study evaluates reactor physics aspects of the proposed High Flux Isotope Reactor (HFIR) permanent beryllium reflector number 5 design. HFIR is a pressurized, light water–cooled, and beryllium reflected research reactor that operates at 85 MW for cycle lengths of approximately 24 to 26 days. A new permanent reflector design is highly desirable for increased versatility in irradiation experiments, to arrange the vertical experiment facilities (VXFs) to minimize their impact on neutron scattering if loaded with neutronabsorbing experiments, to enhance reflector thermal-structural-hydraulic performance, and to simplify the complex fabrication process. The current number 4 and concept number 5 reflectors have 22 and 28 VXFs, respectively, and are modeled and analyzed in this study with the MCNP5 and SCALE 6.1.3 nuclear simulation codes.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS2006

Experiences in managing the Prometheus Project

Congress authorized NASA?s Prometheus Project in February 2003, with the first Prometheus mission slated to explore the icy moons of Jupiter. The Project had two major objectives: (1) to develop a nuclear reactor that would provide unprecedented levels of power and show that it could be processed safely and operated reliably in space for long-duration, deep-space exploration and (2) to explore the three icy moons of Jupiter - Callisto, Ganymede, and Europa - and return science data that would meet the scientific goals as set forth in the Decadal Survey Report of the National Academy of Sciences. Early in Project planning, it was determined that the development of the Prometheus nuclear powered Spaceship would be complex and require the intellectual knowledge residing at numerous organizations across the country. In addition, because of the complex nature of the Project and the multiple partners, approaches beyond those successfully used to manage a typical JPL project would be needed. This paper1 will describe the key experiences in managing Prometheus that should prove useful for future projects of similar scope and magnitude

nuclear↗
DOE OSTI2024

Broad sterile neutrinos & the reactor/gallium tension

Significant evidence exists for the apparent disappearance of electron-type neutrinos in radioactive source experiments. Yet, interpreted within the standard ‘3+1 sterile neutrino scenario’, precision short-baseline measurements of electron antineutrinos from nuclear reactors strongly disagree with these results. Recently, it has been demonstrated that allowing for a finite wavepacket size for the reactor neutrinos can ameliorate such a tension, however the smallness of the required wavepackets is a subject of intense debate. In this work, we demonstrate that a ‘broad’ sterile neutrino may relax this tension in much the same way. Such a phenomenological possibility can arise in plausible hidden sector scenarios, such as a clockwork-style sector, for which we provide a concrete microscopic model.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2024

Mechanical degradation of duplex SiC-fiber reinforced SiC matrix composite tubes under a controlled high-temperature steam environment

This paper describes results from the mechanical evaluation of unirradiated SiC fiber–reinforced SiC matrix composite tubes under a controlled high-temperature steam environment. The experiments were aimed at identifying key material degradation behavior under environments relevant to loss-of-coolant accidents of light water reactors. Mechanical tests of the SiC composite tubes at 1000 °C under steam and inert environments were conducted using a unique test capability. Further, the material tested was a duplex tube with a thick monolithic SiC layer on the outer surface. The tubes were subjected to preloading at ~100 MPa in tension before exposure to high-temperature steam with up to 75% of the preload at a constant displacement. In the presence of matrix cracks, the steam exposure caused embrittlement of the SiC composite tubes and failure at a stress level below the pretest stress. The material degradation was explained by a fiber oxidation model, which can be applied to various SiC cladding concepts. The embrittlement could be a limiting factor for using SiC cladding subjected to loss-of-coolant accident conditions.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Density functional theory calculations of the mixing enthalpy of ternary uranium carbide compounds

The high melting point of uranium-zirconium carbides (U,Zr)C makes them an ideal fuel for nuclear thermal propulsion (NTP) reactors. Gaps remain in the current understanding of the U-Zr-C system due to the difficulty of conducting thermodynamic experiments at NTP operation conditions. Density functional theory calculations using the Hubbard U model (DFT+U) were performed using orbital matrix occupation (OMC) to obtain the mixing enthalpy for UC and ZrC for (U,Zr)C ternary compounds. Similarly, DFT+U calculations were also carried out for the (U,Nb)C and (U,Ta)C systems. In conclusion, the DFT results are envisioned to be used in thermodynamic assessments of the uranium carbide systems based on the CALPHAD approach to supplement the lack of experimental data for the mixing thermodynamics.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Investigating Spillover Energy as a Descriptor for Single-Atom Alloy Catalyst Design

The identification of thermodynamic descriptors of catalytic performance is essential for the rational design of heterogeneous catalysts. Here, we investigate how spillover energy, a descriptor quantifying whether intermediates are more stable at the dopant or host metal sites, can be used to design single-atom alloys (SAAs) for formic acid dehydrogenation. Further, using theoretical calculations, we identify NiCu as a SAA with favorable spillover energy and demonstrate that formate intermediates produced after the initial O–H activation are more stable at Ni sites where rate-determining C–H activation occurs. Surface science experiments demonstrated that NiCu(111) SAAs are more reactive than Cu(111) while they still follow the formate reaction pathway. However, reactor studies of silica-supported NiCu SAA nanoparticles showed only a modest improvement over Cu resulting from surface coverage effects. Overall, this study demonstrates the potential of engineering SAAs using spillover energy as a design parameter and highlights the importance of adsorbate–adsorbate interactions under steady-state operation.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2022

Current Progress of the Final Design of a Subcritical Assembly at the Oak Ridge National Laboratory

During the past decade, the US Department of Energy (DOE)/National Nuclear Security Administration (NNSA) Nuclear Criticality Safety Program (NCSP) has been coordinating nuclear criticality safety (NCS) courses to provide training and qualification to new staff, supervisors, managers, regulators, and other professionals involved with NCS. To address the growing need for more NCS training possibilities, a subcritical assembly was designed at Oak Ridge National Laboratory (ORNL). The Oak Ridge Subcritical Assembly (ORSA), operated at ORNL, could be used in a new course for fissile material handlers/operators or as a backup training facility for currently existing NCSP hands-on NCS training. This paper describes some of the design steps necessary for obtaining a final ORSA design, beginning with a successful feasibility study performed in 2020. Numerous parameters within each main element of the assembly were considered, such as the reflector used, the type of fuel coating or plating used, and the number of fuel plates used. The design choices related to each parametric study described in this paper are detailed and justified at the end of each subsection. An overview of the current final design characteristics of ORSA and the experiments’ possibilities is also provided. ORSA will be built with fuel plates from legacy Aerojet General Nucleonics (AGN)-201M research reactors currently stored at the Y-12 National Security Complex (Y-12). The cylindrical fuel plates comprise UO 2 particles embedded in polyethylene and have a nominal enrichment of 19.5 ± 0.5 wt % 235 U. The basis for the final design was the feasibility study from 2020. In the feasibility study, ORSA achieved k eff = 0.95, corresponding to a multiplication factor of 20, with approximately 620 g of 235 U (four 4 cm plates, three 2 cm plates, and two 1 cm plates) and an 8.5 cm thick graphite reflector radially and on top of the assembly. The bottom of the assembly has a graphite thickness of 5 cm, and the chosen graphite density was 2.3 g/cm 3 . All design calculations were performed using SCALE 6.2.4 KENO-VI and with the ENDF/B VII.1 Continuous Energy cross-section library.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

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