Design of particle bed reactors for the space nuclear thermal propulsion program

doi:10.1016/0149-1970(95)00080-4

Progress in Nuclear Energy

Volume 30, Issue 1, 1996, Pages 1-65

https://doi.org/10.1016/0149-1970(95)00080-4 Get rights and content

Abstract

This paper describes the design for the Particle Bed Reactor (PBR) that we considered for the Space Nuclear Thermal Propulsion (SNTP) Program. The methods of analysis and their validation are outlined first. Monte Carlo methods were used for the physics analysis, several new algorithms were developed for the fluid dynamics, heat transfer and transient analysis; and commercial codes were used for the stress analysis. We carried out a critical experiment, prototypic of the PBR to validate the reactor physics; blowdown experiments with beds of prototypic dimensions were undertaken to validate the power-extraction capabilities from particle beds. In addition, materials and mechanical design concepts for the fuel elements were experimentally validated.

Four PBR rocket reactor designs were investigated parametrically. They varied in power from 400 MW to 2000 MW, depending on the mission's goals. These designs all were characterized by a negative prompt coefficient, due to Doppler feedback, and a moderator feedback coefficient which varied from slightly positive to slightly negative. In all practical designs, we found that the coolant worth was positive, and the thrust/weight ratio was greater than 20.

References (53)

J.R. Powell
High Flux Research Reactors based on Particulate Fuel
Nuc. Inst. and Methods in Physics Research
(1986)
E. Achenbach
Heat Transfer and Pressure Drop of Pebble Beds up to High Reynolds Numbers
J.W. Adams
Performance of CVD and CVR Coated Carbon-Carbon in High Temperature Hydrogen
R.M. Ball
Neutron Lifetime in the CX
R.M. Ball
CX ‘Peek-A-Boo’ Reactivity Worth Around Central Stalk
R.M. Ball
SPND Detector Response in the CX
R.E. Barletta
Ceramic Coatings for PBR Components
R.E. Barletta
The Development of CVR Coatings for PBR fuels
J.A. Bernard
Formulation and Experimental Evaluation of Closed-Form Control Laws for the Rapid Maneuvering of Reactor Neutronic Power
Brownlee, Private Communication, D. Husser, Babcock and Wilcox, Lynchburg, VA....

Caldwell, C.S., Private Communication, Babcock and Wilcox, Lynchburg, VA....

M. Charmchi

Analysis of Gas Cooled Particle Bed Reactors-SIMBED

(1993)

G.J. DeSalvo

D. Dobranich

Heat Transfer and Thermal Stress Analysis of Multi-Layered Spherical Fuel Particles of a Particle Bed Space Nuclear Reactor

F.P. Durham

Nuclear Engine Definition Study Preliminary Report (Vol1) - Engine Description

Los Alamos Scientific Laboratory Informal Report

(1972)

D'Yakov, E.K., Private Communication to D.G. Schweitzer, Brookhaven National Laboratory, Upton, NY. (Jan....

S. Ergun

A.Y. Goldin

Development of Nuclear Rocket Engines in the USSR

Gomozov, Private Communication from D. Husser, Babcock and Wilcox, Lynchburg, VA....

L.P. Hatch

Fluidized Beds for Rocket Propulsion

Nucleonics

(1960)

F.L. Horn

Transient Thermal-Hydraulic Measurements on the Particle Bed Reactor Fuel Element

W. Katscher

Nuclear Technology

(1977)

D.R. Koenig

Experience Gained from the Space Nuclear Rocket Program (Rover)

H. Kusters

Burning of Actinides and Fission Products in Reactors and in Accelerator-Driven Spallation Sources

Trans. Am. Nucl. Soc.

(1991)

K.D. Lathrop

TWOTRAN-II: An Interfaced, Exportable Version of the TWOTRAN Code for Two-Dimensional Transport

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Design of particle bed reactors for the space nuclear thermal propulsion program

Abstract

Nuc. Inst. and Methods in Physics Research

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