Nuclear Engineering ETDs

Publication Date

Spring 5-29-2019

Abstract

Neutronic and CFD-thermal hydraulic analyses are performed of the Very-Small, Long-LIfe, and Modular (VSLLIM) nuclear reactor. This reactor was developed at the University of New Mexico’s Institute for Space and Nuclear Power Studies (UNM-ISNPS) to generate 1.0 – 10 MWth for extended periods without refueling. It offers passive operation and safety features and redundant control and would be fabricated, assembled and sealed in the factory. During nominal operation and after shutdown, the VSLLIM is cooled by natural circulation of in-vessel liquid sodium, with the aid of an in-vessel chimney and annular helically-coiled tubes Na-Na heat exchanger (HEX) in the downcomer. In case of a malfunction of the Na-Na HEX, the reactor shuts down, and the decay heat generated in the core is removed by natural circulation of the in-vessel liquid sodium with the aid of redundant passive means. These include variable-conductance liquid-metal heat pipes that are embedded in the upper part of the reactor primary vessel wall, and natural circulation of ambient air along the outer surface of the reactor guard vessel wall.

The VSLLIM reactor core is loaded with hexagonal assemblies of 19 UN fuel rods clad in HT-9 steel and with scalloped BeO walls, clad also in HT-9 steel. In addition to helping achieve an almost uniform flow distribution in the fuel assemblies, the scalloped BeO walls, together with the BeO axial and radial reflectors, increase the hot-clean reactivity for achieving long full-power operation life, at a relatively low UN fuel enrichment. During nominal operation, the inlet and exit coolant temperatures in the reactor core are maintained at 610 K, and < 820 K to minimize embrittlement and corrosion of the HT-9 steel cladding, core structure, and reactor vessel by liquid sodium.

This research conducted neutronic and thermal-hydraulics analyses to investigate and quantify the passive operation and safety features of the Very-Small, Long-LIfe, and Modular (VSLLIM). The research tasks carried out include:

(a) Performing neutronic analyses of the VSLLIM to investigate the effects of several design and material choices on the cold and hot-clean reactivity, for achieving long operation life without refueling. Also calculate the cold-clean reactivity shutdown margins of the emergency shutdown system (ESS) and reactor control (RC), and the beginning of life (BOL) hot-clean reactivity. Investigated are the effect of UN fuel enrichment, and the material of the axial and radial reflectors. These analyses calculated the neutron energy spectrums and the radial and axial fission power distributions. These are in addition to determining the temperature reactivity feedback effects due to the UN fuel, sodium coolant, HT-9 steel cladding and core structure, BeO in the driver core and axial radial reflectors, and the Doppler broadening of neutron cross-sections. To estimate the full-power operation lives of the VSLLIM reactor at different thermal powers, fuel depletion calculations are carried out at hot-clean operation condition.

(b) Performing thermal-hydraulic and computational fluid dynamics (CFD) analyses during nominal reactor operation and after shutdown. These analyses estimated the friction number for the liquid sodium flow in the core hexagonal UN fuel assemblies, with scalloped wall as a function of the flow Reynolds number. The flow and temperature distributions in the UN fuel assemblies are calculated, at different reactor thermal powers and inlet and exit core temperatures of 610 K and < 820 K, respectively.

(c) Performing CFD analyses to quantify the passive decay heat removal by natural circulation of ambient air along the outer surface of the reactor guard vessel. Both, after reactor shutdown and in the case of a malfunction of the in-vessel Na-Na HEX.

The performed neutronics and fissile depletion analyses of the VSLLIM confirmed that a UN fuel enrichment of 13.76% is sufficient for achieving high enough hot-clean excess reactivity for operating VSLLIM reactor without refueling for ~92 and 5.8 full power years (FPY) at 1.0 and 10 MWth, respectively. Results confirmed sufficient cold-clean reactivity shutdown margins using either the reactor control (RC) or the emergency shutdown system (ESS), independently. In addition to having two independent reactor shutdown systems, results show that the negative temperature reactivity feedback is capable of shutting down the reactor with a modest increase in the temperatures of the UN fuel and the in-vessel liquid sodium. Results also show that the neutron energy spectrum in the VSLLIM reactor core is hard, which reduces the inventory of minor actinides in UN fuel during reactor operation. Because of its low operating temperatures, < 812 K at 10 MWth and UN fuel low average power density (< 23.47 MWth/m3), the fuel in the VSLLIM core experiences practically no swelling and retains practically all fission gas release.

The performed CFD-thermal-hydraulics analyses investigated the effects of using metal fins along the outer surface of the guard vessel wall and changing the width of the cold air intake duct on the decay heat removal rate and the time after shutdown for cooling the in-vessel liquid sodium to 400 K. Results show that without metal fins, the heat removal rate of the decay heat is 244 kWth immediately after shutdown. However, within 8 minutes after shutdown, natural circulation of ambient air along the outer surface of the guard vessel removes more heat than is being generated by radioactive decay in the core. Consequently, the average in-vessel sodium temperature drops from 682 K to 400 K, within 120 hrs after shutdown.

Using metal fins along the outer surface of the guard vessel increases the total area for heat transfer and the decay heat removal rate by 13.5%, reducing the time for the average temperature of the in-vessel liquid sodium to decrease to 400K in ~100 hrs. Without metal fins, reducing the width of the cold air intake duct by 50% decreases the decay heat removal rate by 35%, increasing to ~ 346 hrs the cooling time of the in-vessel sodium to 400 K. These results demonstrate that the decay heat removal by natural circulation of ambient air from the outer surface of the reactor guard vessel wall, with and without metal fins, is quite effective. Additionally, the results show that the temperature of the in-vessel sodium after shutdown remain safely, ~470 K below its boiling point (~1156 K at 0.1 MPa).

The performed CFD analyses investigated the friction factor for laminar, transition, and turbulent flows in hexagonal bundles of bare tube and also investigated with flat and scalloped walls. The results for the bundles with flat walls are in good agreement with previously reported experimental data by others. The CFD results and the reported experimental data for bundles with flat walls and various numbers of tubes, (7, 19, 37, and 61), in a triangular lattice with 1.07 < P/D < 2.416, are used to develop a continuous correlation of the friction factor as a function of P/D and Rein. The developed correlation, for P/D up to 3.0, and a wide range of tubes in the bundles (N = 7 – 331), spans all three flow regions (102 < Rein < 106) and is in excellent agreement with the compiled numerical and experimental database. The results also validated the applicability of the developed friction factor correlation to the VSLLIM reactor hexagonal bundles with scalloped walls.

The developed continuous friction factor correlation is within 5% - 8% of the CFD data generated for the scalloped walls bundles with 19 and 37 rods or tubes. Results also showed that scalloped walls reduce the bypass flow next to the wall, while increasing the flow in the interior subchannels in the bundles, thus improving heat transfer. Higher flow in interior subchannels enhances the thermal-hydraulics in the VSLLIM reactor core and reduces the maximum temperature at a given Rein.

Keywords

Small and very small modular reactors; Walk-away safe; UN Fuel; Natural circulation; portable; long-life; distributed grid; Passive operation; Walk-away safe; Auxiliary electric power; Thermoelectric conversion.

Sponsors

University of New Mexico’s Institute for Space and Nuclear Power Studies (UNM-ISNPS)

Document Type

Dissertation

Language

English

Degree Name

Nuclear Engineering

Level of Degree

Doctoral

Department Name

Nuclear Engineering

First Committee Member (Chair)

Distinguished and Regents’ Professor Mohamed S. El-Genk

Second Committee Member

Distinguished Professor Anil K. Prinja

Third Committee Member

Professor Fernando Garzon

Fourth Committee Member

Dr. Timothy M. Schriener

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