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Title: Thermal-Hydraulic Analyses of the Submersion-Subcritical Safe Space (S and 4) Reactor

Abstract

Detailed thermal-hydraulic analyses of the S and 4 reactor are performed to reduce the maximum fuel temperature of the Submersion-Subcritical Safe Space (S and 4) reactor to below 1300 K. The fuel pellet diameter is reduced from 1.315 cm to 1.25 cm, decreasing the thermal resistance of the pellets and each of the 1.54 cm diameter coolant channels in the reactor core are replaced with several 0.3 cm ID channels to increase the effective heat transfer area and to encourage mixing of the flowing helium-28% xenon coolant. The calculated maximum fuel temperature decreased from more than 1900 K to 1302 K and the relative pressure drop across the reactor core increased from 1.98% to 2.57% of the inlet pressure. Moving the concentric inlet and outlet pipes 1 cm towards the center of the reactor core encouraged more flow through the center region, further reducing the maximum fuel temperature by 14 degrees to 1288 K, with a negligible effect on the core pressure losses.

Authors:
 [1];  [2];  [3]
  1. Nuclear Engineering Department, University of Missouri-Rolla, 1870 Miner Circle, Rolla, MO 65409 (United States)
  2. Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM 87131 (United States)
  3. (United States)
Publication Date:
OSTI Identifier:
21054543
Resource Type:
Journal Article
Resource Relation:
Journal Name: AIP Conference Proceedings; Journal Volume: 880; Journal Issue: 1; Conference: International forum-STAIF 2007: 11. conference on thermophysics applications in microgravity; 24. symposium on space nuclear power and propulsion; 5. conference on human/robotic technology and the vision for space exploration; 5. symposium on space colonization; 4. symposium on new frontiers and future concepts, Albuquerque, NM (United States), 11-15 Feb 2007; Other Information: DOI: 10.1063/1.2437464; (c) 2007 American Institute of Physics; Country of input: International Atomic Energy Agency (IAEA)
Country of Publication:
United States
Language:
English
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS; COOLANTS; DESIGN; FISSION; FUEL PELLETS; HEAT TRANSFER; HELIUM; NUCLEAR FUELS; PIPES; POWER SYSTEMS; PRESSURE DROP; REACTOR CORES; SPACE; SPACE VEHICLES; SUBCRITICAL ASSEMBLIES; THERMAL HYDRAULICS; XENON

Citation Formats

King, Jeffrey C., El-Genk, Mohamed S., and Chemical and Nuclear Engineering Department, University of New Mexico, Albuquerque, NM 87131. Thermal-Hydraulic Analyses of the Submersion-Subcritical Safe Space (S and 4) Reactor. United States: N. p., 2007. Web. doi:10.1063/1.2437464.
King, Jeffrey C., El-Genk, Mohamed S., & Chemical and Nuclear Engineering Department, University of New Mexico, Albuquerque, NM 87131. Thermal-Hydraulic Analyses of the Submersion-Subcritical Safe Space (S and 4) Reactor. United States. doi:10.1063/1.2437464.
King, Jeffrey C., El-Genk, Mohamed S., and Chemical and Nuclear Engineering Department, University of New Mexico, Albuquerque, NM 87131. Tue . "Thermal-Hydraulic Analyses of the Submersion-Subcritical Safe Space (S and 4) Reactor". United States. doi:10.1063/1.2437464.
@article{osti_21054543,
title = {Thermal-Hydraulic Analyses of the Submersion-Subcritical Safe Space (S and 4) Reactor},
author = {King, Jeffrey C. and El-Genk, Mohamed S. and Chemical and Nuclear Engineering Department, University of New Mexico, Albuquerque, NM 87131},
abstractNote = {Detailed thermal-hydraulic analyses of the S and 4 reactor are performed to reduce the maximum fuel temperature of the Submersion-Subcritical Safe Space (S and 4) reactor to below 1300 K. The fuel pellet diameter is reduced from 1.315 cm to 1.25 cm, decreasing the thermal resistance of the pellets and each of the 1.54 cm diameter coolant channels in the reactor core are replaced with several 0.3 cm ID channels to increase the effective heat transfer area and to encourage mixing of the flowing helium-28% xenon coolant. The calculated maximum fuel temperature decreased from more than 1900 K to 1302 K and the relative pressure drop across the reactor core increased from 1.98% to 2.57% of the inlet pressure. Moving the concentric inlet and outlet pipes 1 cm towards the center of the reactor core encouraged more flow through the center region, further reducing the maximum fuel temperature by 14 degrees to 1288 K, with a negligible effect on the core pressure losses.},
doi = {10.1063/1.2437464},
journal = {AIP Conference Proceedings},
number = 1,
volume = 880,
place = {United States},
year = {Tue Jan 30 00:00:00 EST 2007},
month = {Tue Jan 30 00:00:00 EST 2007}
}
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