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Ross, Kyle

Publications and source records attributed to Ross, Kyle.

12 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

MELCOR 2.2 Benchmarks of Peach Bottom NUREG/CR 7155 Uncertainty Analysis

The U.S. Nuclear Regulatory Commission (NRC) performed a first-of-a-kind uncertainty analysis (UA) of the accident progression, radiological releases, and offsite consequences for the State-of- the-Art Reactor Consequence Analyses (SOARCA) of an unmitigated long-term station blackout (LTSBO) severe accident scenario at the Peach Bottom Atomic Power Station. The objective of the UA was to evaluate the robustness of the SOARCA deterministic "best estimate results and conclusions documented in NUREG-1935, and to develop insight into the overall sensitivity of the SOARCA results to uncertainty in key modeling inputs. The study was completed in 2015 and documented in NUREG/CR-7155. Since 2015, two other SOARCA UAs were completed for two pressurized water reactor (PWR) plants. The PWR UAs incrementally updated the approach and methodology, including using the latest release of the MELCOR 2.2 computer code. There were also advances made in the state-of-the-art modeling related to NRC efforts using the Peach Bottom model in NUREG-2206, which provided the technical basis for the containment protection and release reduction rulemaking for boiling water reactors with Mark I and Mark II containments. This report documents the input model changes from the NUREG/CR-7155 study and performs a small number of reference calculations to assess the changes of the new computer code and the model input updates. The objective of the work is to verify whether the updated Peach Bottom MELCOR model and updated version of MELCOR support the conclusions formed in the Peach Bottom SOARCA UA by performing these representative calculations.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

A New Method to Contain Molten Corium in Catastrophic Nuclear Reactor Accidents

The catastrophic nuclear reactor accident at Fukushima damaged public confidence in nuclear energy and a demand for new engineered safety features that could mitigate or prevent radiation releases to the environment in the future. We have developed a novel use of sacrificial material (SM) to prevent the molten corium from breaching containment during accidents as well as a validated, novel, high-fidelity modeling capability to design and optimize the proposed concept. Some new reactor designs employ a core catcher and a SM, such as ceramic or concrete, to slow the molten corium and avoid the breach of the containment. However, existing reactors cannot easily be modified to include these SMs but could be modified to allow injectable cooling materials (current designs are limited to water). The SM proposed in this Laboratory Development Research and Development (LDRD) project is based on granular carbonate minerals that can be used in existing light water reactor plants. This new SM will induce an endothermic reaction to quickly freeze the corium in place, with minimal hydrogen explosion and maximum radionuclide retention. Because corium spreading is a complex process strongly influenced by coupled chemical reactions (with underlying containment material and especially with the proposed SM), decay heat and phase change. No existing tool is available for modeling such a complex process. This LDRD project focused on two research areas: experiments to demonstrate the feasibility of the novel SM concept, and modeling activities to determine the potential applications of the concept to actual nuclear plants. We have demonstrated small-scale to large-scaled experiments using lead oxide (Pb0) as surrogate for molten corium, which showed that the reaction of the SM with molten Pb0 results in a fast solidification of the melt and the formation of open pore structures in the solidified Pb0 because of CO 2 released from the carbonate decomposition.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Modeling for Existing Nuclear Power Plant Security Regime

This document details the development of modeling and simulations for existing plant security regimes using identified target sets to link dynamic assessment methodologies by leveraging reactor system level modeling with force-on-force modeling and 3D visualization for developing table-top scenarios. This work leverages an existing hypothetical example used for international physical security training, the Lone Pine nuclear power plant facility for target sets and modeling.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Modeling for Existing Nuclear Power Plant Security Regime

This document details the development of modeling and simulations for existing plant security regimes using identified target sets to link dynamic assessment methodologies by leveraging reactor system level modeling with force-on-force modeling and 3D visualization for developing table-top scenarios. This work leverages an existing hypothetical example used for international physical security training, the Lone Pine nuclear power plant facility for target sets and modeling.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Terry Turbopump Analytical Modeling Efforts in Fiscal Year 2018. Progress Report

This document details the computational fluid dynamic and system-level modeling, including a mechanistic representation of a Terry turbopump. Until this recent effort, data and modeling results show that a Terry turbine, flowing air (or steam) at a certain rate, can develop the same power at two very different speeds, and has large implications with respect to understanding how a boiling water reactor's reactor core isolation cooling system or a pressurized water reactor turbine driven auxiliary feedwater system would respond to a loss of electrical power for Terry turbine speed governing. This work has provided insights in modeling uncertainties and provides confirmation for experimental efforts for the Terry turbopump expanded operating band being conducted at Texas A&M University.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗