Photo: RBHT Test Facility. US NRC.
Modelling reactor core behaviour under accident conditions with delayed re-introduction of cooling water (commonly referred to as reflood) is a challenge for safety analysis computer codes. Reflood thermal-hydraulics (e.g. post-critical heat flux flow and heat transfer, entrainment, quench) remains a major contributor to code uncertainties in the simulation of many accident scenarios and must be more deeply understood to enhance nuclear safety. Moreover, as the nuclear industry evolves, there is a need for additional data for power up-rates and new designs.
Reflood heat transfer and rod bundle thermal-hydraulics have been studied since the 1960s when concerns about the effectiveness of the Emergency Core Cooling System (ECCS) of a nuclear plant were raised. Reflood thermal-hydraulics were extensively studied and significant improvements were made in modelling and simulation. With the recent needs to improve further plant operating conditions, efforts to develop more “mechanistic” models for reflood thermal-hydraulics have been undertaken; that is, models for physical processes would be based on the fundamental mechanisms that govern thermal-hydraulics rather than be based on empirical correlations that were often restricted to a specific range of applicability. While existing reflood experiments had demonstrated the effectiveness of the ECCS and provided information suitable for the development and licensing of evaluation models, a reflood facility with advanced and detailed measurement capabilities was considered an important next step.
Data from such a facility would be needed for evaluation of applicant licensing submittals, and for development and assessment of the NRC confirmatory analysis code, and they are of interest to industry, regulatory bodies, TSO and research organisations. As a result, the Rod Bundle Heat Transfer (RBHT) test facility was designed and constructed beginning in 1998. Research has continued since then, and a wide variety of tests have been conducted.
The RBHT facility was therefore designed to:
By providing these data and conditions, members' expectation was that data evaluation would be enhanced and more “mechanistic” models of reflood thermal-hydraulics would be produced.
The objective of the Rod Bundle Heat Transfer (RBHT) project was to conduct new experiments in the Rod Bundle Heat Transfer (RBHT) facility at Pennsylvania State University under the United States Nuclear Regulatory Commission (NRC) co-ordination. The high quality data and measurements generated (flow rates, temperature distributions, heat fluxes, droplet size distribution and velocity, spacer grid dry-out and rewet, carryover and quench front movement) served to evaluate system hydraulics and subchannel codes in the simulation of reflood tests in a full height rod bundle, prototypical of PWRs, for complex variable or oscillatory inlet flows which are more likely in hypothetical accident scenarios.
The project was conducted in two major sub-phases. In the first sub-phase, 11 “open” tests were conducted and their results distributed to the participants. These experiments were simulated by most participants using an analysis code of their choice. In the second sub-phase, five “semi-blind” tests were conducted. In these tests, thermal-hydraulic conditions similar to those in the “open” tests were imposed on the experiments and the test data recorded. However, only the as-measured initial and boundary conditions were provided to the participants for calculations. Simulations of these blind tests used an uncertainty methodology of the participant’s choice with the goal of capturing one or more of the several “figures of merit” that were defined based on measured quantities.
Overall, the project can be characterised as having two distinct products. One major product is the experimental data itself, which provided participants with new and unique reflood data. The second product is comparative code assessment, where simulations of the data suggest strengths and weaknesses of the analysis codes used by the participants.
The objective of Phase II is to conduct new experiments and evaluate and improve the accuracy of system and sub-channel thermal hydraulics codes in the simulation of reflood tests in a full height rod bundle for complex inlet flows. Like Phase I, Phase II is a combined experimental/analytical project. Experiments performed as part of Phase II will be simulated by participants with an analysis code of their choice. Phase I of this project produced experimental data that was used for code assessment and investigation of code uncertainty methods. The code assessment in Phase I identified some possible deficiencies in the thermal-hydraulic codes employed, and while spacer grids are known to have important effects on reflood behavior, Phase I did not focus on these effects. Phase II will build upon the results and findings of Phase I. Additional experimental data will be produced, and analytical methods refined based on those data.
Phase II will focus on thermal-hydraulic conditions that proved to be difficult for codes to simulate and on reflood phenomena including quench front entrainment and spacer grid effects. Numerous experimental tests have been conducted for steady, constant inlet conditions with flooding rates equal to and greater than 2.5 cm/sec. Relatively few experiments have examined very low flooding rates (less than 2.5 cm/sec) or variable inlet flows which are more likely in a hypothetical accident scenario. Data at low flooding rates and low inlet subcooling may be useful for some small modular reactors, and phenomena at these conditions are typical of those in long term cooling. The benchmark exercise in Phase II will produce and provide to participants a new set of reflood test data with low flooding rates, and variable, reflood inlet flows all of which are data that have never been obtained previously. Advanced instrumentation will be used to measure the real-time droplet field and the steam temperatures and velocities.
Rod Bundle Heat Transfer Thermal-Hydraulic Program, Stephen M. Bajorek, Fan-Bill Cheung, Nuclear Technology, 2019, 205:1-2, 307-327, DOI:10.1080/00295450.2018.1510697
Rod Bundle Heat Transfer Test Facility Description (NUREG/CR-6976), E.R. Rosal, T.F. Lin, I.S. McClellan, R.C. Brewer, 2010, https://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr6976/index.html
Measurements of droplet size and velocity distributions during reflood, Grant Robert Garrett Penn State University), Brian Lowery, Molly Hanson (Applied Research Laboratory), Douglas J. Miller, Turki Almudhhi, Fan-Bill Cheung (Penn State University), Stephen M. Bajorek, Kirk Tien, Chris L. Hoxie (NRC), 2022, 19th International topical meeting on NUclear REactor Thermal Hydraulics (NURETH-19)
RBHT: Belgium, Czechia, Finland, France, Germany, Italy, Japan, Korea, Spain, Sweden, Switzerland and United States.
RBHT-II: Belgium, Czechia, France, Germany, Italy, Korea, Spain, Sweden, Switzerland and United States.
RBHT: October 2019 - October 2022
RBHT-II: October 2023 - September 2026
RBHT: EUR 1.44 million
RBHT-II: EUR 1.52 million