Larry L. Humphries

Sandia Heat Flux Gauge Thermal Response and Uncertainty Models

The San&a Heat Flux Gauge (HFG) was developed as a rugged, cost-effective technique for performing steady state heat flux measurements in the pool fire environment. The technique involves reducing the time-temperature history of a thin metal plate to an incident heat flux via a dynamic thermal model, even though the gauge is intended for use at steady state. In this report, the construction of the gauge is reviewed. The thermal model that describes the dynamic response of the gauge to the f~e environment is then advanced and it is shown how the heat flux is determined from the temperature readings. This response model is based on first principles, with no empirically adjusted constants. A validation experiment is presented where the gauge was exposed to a step input of radiant heat flux. Comparison of the incident flux, determined from the thermal response model, with the known flux input shows that the gauge exhibits an noticeable time lag. The uncertainty of the measurement is analyzed, and an uncertainty model is put forth using the data obtained from “the experiment. The uncertainty model contains contributions from seventeen separate sources loosely categorized as being either from uncontrolled variability, missing physics, or simplifying assumptions. As part of the missing physics, an empirical constant is found that compensates for the gauge time lag. Because this compensation is incorporated into the uncertainty model instead of the response model, this information can be used to advantage in analyzing pool fire data by causing large uncertainties in non-steady state situations. A short general discussion on the uncertainty of the instrument is presented along with some suggested design changes that would facilitate the determination and reduction of the measurement uncertainty.

MELCOR/CONTAIN LMR Implementation Report - FY16 Progress.

This report describes the progress of the CONTAIN-LMR sodium physics and chemistry models to be implemented in MELCOR 2.1. In the past three years, the implementation included the addition of sodium equations of state and sodium properties from two different sources. The first source is based on the previous work done by Idaho National Laboratory by modifying MELCOR to include liquid lithium equation of state as a working fluid to model the nuclear fusion safety research. The second source uses properties generated for the SIMMER code. The implemented modeling has been tested and results are reported in this document. In addition, the CONTAIN-LMR code was derived from an early version of the CONTAIN code and many physical models that were developed since this early version of CONTAIN are not available in this early code version. Therefore, CONTAIN 2 has been updated with the sodium models in CONTAIN-LMR as CONTAIN2-LMR, which may be used to provide code-to-code comparison with CONTAIN-LMR and MELCOR when the sodium chemistry models from CONTAIN-LMR have been completed. Both the spray fire and pool fire chemistry routines from CONTAIN-LMR have been integrated into MELCOR 2.1 and debugging and testing are in progress. Because MELCOR only models the equation of state for liquid and gas phases of the coolant, a modeling gap still exists when dealing with experiments or accident conditions that take place when the ambient temperature is below the freezing point of sodium. An alternative method is under investigation to overcome this gap. We are no longer working on the separate branch from the main branch of MELCOR 2.1 since the major modeling of MELCOR 2.1 has been completed. At the current stage, the newly implemented sodium chemistry models will be a part of the main MELCOR release version (MELCOR 2.2). This report will discuss the accomplishments and issues relating to the implementation. Also, we will report on the planned completion of all remaining tasks in the upcoming FY2017, including the atmospheric chemistry model and sodium-concrete interaction model implementation.

Test Design, Results, and Archived Database for the OECD Lower Head Failure Program Integral Experiments

In the event of a severe core meltdown accident, core material can relocate to the lower head of a pressurized water reactor (PWR) vessel resulting in significant thermal and pressure loads to the vessel. The potential for failure of the pressure vessel makes possible the release of core material to the containment. The objective of this experimental/analytical program is to characterize the mode, timing, and size of lower head failure (LHF) under severe accident conditions. The OECD Lower Head Failure (OLHF) project investigates lower head failure for conditions of low reactor coolant system (RCS) pressure (2–5 MPa) and prototypic through-wall temperature differential (ΔTW >200K). Low RCS pressure is motivated by the desire to use the data to develop models for assessing accident management strategies involving reactor pressure vessel (RPV) depressurization. Pressure transient is useful in assessing the effect of water injection as part of accident management strategy. Prototypic through-wall temperature differential, ΔTW , is of importance because of the need to provide data where stress redistribution in the vessel wall occurs (as a result of decreasing material strength with temperature). Test design and results for the four OLHF integral tests are reported and summarized in this paper. A short description of the test conduct and heating history is followed by a description of the vessel failure site, the vessel deformation, temperature profiles, stress state, and rupture dynamics for each test. Key observations and conclusions are summarized for each test. The ∼1/5 scale tests are extensively instrumented to provide temperature, pressure, and displacement data. The vessel surfaces are mapped both before and after the test to provide measurements of pre-test thickness, post-test thickness, and cumulative vessel deformation. Data has been assessed and qualified in data reports for each test. The data has been preserved in MSEXCEL™ spreadsheets with macro utilities to facilitate access and analysis of the data. As a result, there exists a well-archived, well-qualified database for model development and validation.Copyright

Integration of CONTAIN Liquid Metal Models Into the MELCOR Code

A sodium coolant accident analysis code is necessary to provide regulators with a means of performing confirmatory analyses for future sodium reactor licensing submissions. MELCOR and CONTAIN, which are currently employed by the U.S. Nuclear Regulatory Commission (NRC) for light water reactor (LWR) licensing, have been traditionally used for level 2 and level 3 probabilistic analyses as well as containment design basis accident analysis. To meet future regulatory needs, new models will be added to the MELCOR code for simulation of Liquid Metal Reactor (LMR) designs. Existing models developed for separate effects codes will be integrated into the MELCOR architecture. This work integrates those CONTAIN code capabilities that feasibly fit within the MELCOR code architecture.Implementation of such models for sodium reactor simulation into an actively maintained, full-featured, integrated severe accident code fills a significant gap in capability for providing the necessary analysis tools for regulatory licensing. Current work scope will focus on the following implementation goals:• Phase 1: Implement sodium Equations of State (EOS) as a working fluid for a MELCOR calculation from:○ The fusion safety database○ The SIMMER-III Code○ The SAS4a Code• Phase 2: Examine and test changes to the CONTAIN-LMR Implemented by Japan Atomic Energy Agency, specifically:○ Aerosol Condensation○ Implementation of the capability for simultaneous sodium and water condensation modeling• Phase 3: Implementation and Validation of CONTAIN physics models:○ Sodium Spray Fires (including new test data)○ Sodium Pool Modeling○ Sodium Pool Fires• Phase 4: Implementation and Validation of CONTAIN chemistry models:○ Debris Bed/Concrete Cavity Interactions○ Sodium Pool Chemistry○ Atmospheric ChemistryAn option for changing the EOS for the MELCOR working fluid from water to liquid metal and the heat transfer from water/steam to liquid metal has been implemented into MELCOR. The property models implemented include an analytic EOS model developed for the SIMMER-III code and the fusion safety works done at Idaho National Laboratory (INL). This paper provides a summary of the status of the code development work. A description of the current models implemented together with user requirements and test calculations will be presented.© 2014 ASME