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Jain, Prashant

Publications and source records attributed to Jain, Prashant.

22 records · Page 2

Influence of dissolved salts and impurities on seawater on heat transfer and fluid flow through fuel channels and debris bed

This project aims to better understand this mechanism and its effect on heat transfer. Specifically, experimental research was conducted to investigate the long term coolability issues in scaled test setups. Advanced diagnostic and measurement techniques were investigated and incorporated into an annular test facility and a debris bed test facility. The diagnostic techniques investigated included: radioactively tagged salt solution, Frisch collar radiation detector array, neutron and x-ray imaging systems, and a Rayleigh backscatter-based fiber-optic temperature system. Models were checked against the in-situ experimental data and recommendations were made for modeling purposes.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Experimental investigation on the coolability of nuclear reactor debris beds using seawater

During the Fukushima nuclear reactor accident, seawater was injected into the reactor core to cool the decay heat from the heat generating porous debris bed. However, the impact of dissolved salts in water on the coolability of a debris bed is not well understood. This paper is one of the first works to investigate the cooling mechanisms for debris bed using seawater. An experimental system was built where the packed bed test section was volumetrically heated via a high frequency induction heater. The temperature along the axis of the test section was measured using a high resolution temperature sensor based on Optical Frequency Domain Reflectometry. Experimental results show that the dryout heat flux increases with an increase in concentration, with pure water having a dryout volumetric heat flux of 1.8 MW/m and salt water of concentration 7% having a dryout volumetric heat flux of 2.8 MW/m. However, this enhanced coolability using salt water is short term, as the crystallization fouling mechanism becomes significant over time and leads to gradual temperature excursion in the bed. The fouling mechanism eventually leads to a cross section of the debris bed to plug with the deposited salts, and this leads to a pressure build up in the lower regions of the debris bed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Flattening the Radial Temperature Profile across the Transformational Challenge Reactor Core

The Transformational Challenge Reactor (TCR) program is demonstrating an agile development approach to advanced nuclear reactor design, which has traditionally utilized a linear design process. In leveraging artificial intelligence, additive manufacturing, advanced materials, and cutting-edge modeling and simulation, the TCR program aims to minimize the high cost and lengthy deployment timelines now standard in the nuclear industry. Within a relatively short period of time, a robust and mature advanced gas-cooled reactor was iteratively designed under the TCR program, using these cutting-edge technologies. The TCR is a 3 MWt gas-cooled microreactor fueled with uranium nitride (UN) tristructural isotropic (TRISO) fuel particles. Though manufactured via traditional means, these UN TRISO particles are loaded into additively manufactured silicon carbide (SiC) cans [4]. Once loaded with TRISO particles, the SiC cans are densified using a chemical vapor infiltration process. The additively manufactured SiC enables significantly more freedom in the design of the fuel form than could ever be achieved using traditionally manufactured SiC. The helium coolant, pressurized to 5 MPa, enters the core at 300°C and nominally exits it at 500°C. Typically, the most thermally limiting components in any reactor design are the fuel assemblies in the core center. To provide a wide thermal margin in these central fuel assemblies, the flow may be biased toward the center of the core to more effectively cool these fuel assemblies with more power deposition and flatten the core’s radial temperature distribution. An analytical fluid model of the TCR core was developed to explore methods for biasing the flow away from the cooler outer fuel assemblies and towards the hotter inner ones. Higher-fidelity models developed in STAR-CCM+ 2020.3.1, a computational fluid dynamics code, were then utilized to verify the analytical model’s findings.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗