Experimental investigation of thermal stratification development in boiling water reactor suppression pools during reactor core isolation cooling system operation
Not provided.
Nuclear Science & Technology↗
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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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Effect of nonuniform isoflux heating on natural convection heat transfer in a prismatic modular reactor
Not provided.
Energy & Fuels↗
RNN-Based online anomaly detection in nuclear reactors for highly imbalanced datasets with uncertainty
Not provided.
Nuclear Science & Technology↗
Study on start-up of molten salt fast modular reactor from shutdown based on two-phase flow CFD analysis
Not provided.
Nuclear Science & Technology↗
Transient behavior of a molten salt fast reactor under two-phase flow conditions with helium bubbling
Not provided.
Nuclear Science & Technology↗
Helium-air mixing in simulated reactor cavities of high temperature gas reactors
Not provided.
Nuclear Science & Technology↗
Conservation of migration area by transport cross sections using Cumulative Migration Method in deterministic heterogeneous reactor transport analysis
Not provided.
Nuclear Science & Technology↗
Miniature Sensor Irradiation Tests Under Steady-State and Transient Conditions at MIT Research Reactor (MITR) and Transient Reactor Test Facility (TREAT)
Not provided.
Nuclear Science & Technology↗
Reactor Critical Experiments of the Rensselaer Polytechnic Institute Reactor Critical Facility with Noteworthy Non-fissile Stainless Steel Elements Sensitivity
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Dupont, Mathieu [ORNL] (ORCID:0000000183441714)↗
Reactor Physics Characteristics of High Flux Isotope Reactor Core Designs with Low-Enriched Uranium Fuels
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Betzler, Benjamin↗
High-Temperature Reactor Testing of Prototype Fission Chamber for Next-Generation Reactors
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Wilson, Brandon [ORNL] (ORCID:0000000320651647)↗
Data Testing with Zero Power Reactor (ZPR), Zero Power Physics Reactor (ZPPR), and New Thermal Epithermal eXperiments (TEX) Plutonium Benchmarks [Slides]
This presentation states that new TSLs in ENDF/B-VIII.0 will reduce the overprediction of the library for the most moderated cases. The differences between VII.1 and VIII.0 are driven by the 239 Pu changes and are consistently higher for all configurations. The VIII.0 239 Pu thermal cross sections were specifically adjusted lower to better predict plutonium thermal solutions criticality benchmarks, yet they overpredict the TEX thermal configurations, moderated by PE. Other, smaller differences have been seen for 240 Pu and 56 Fe. PMM-002 has shown utility in testing cross sections over a wide range of neutron energies.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗