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A 50-100 Kwe Gas-Cooled Reactor for Use on Mars

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A 50-100 Kwe Gas-Cooled Reactor for Use on Mars SANDIA REPORT SAND2006-2189 Unlimited Release Printed April 2006 A 50-100 kWe Gas-cooled Reactor For Use On Mars Curtis D. Peters Prepared by Sandia National Laboratories Albuquerque, New Mexico 87185 and Livermore, California 94550 Sandia is a multiprogram laboratory operated by Sandia Corporation, a Lockheed Martin Company, for the United States Department of Energy’s National Nuclear Security Administration under Contract DE-AC04-94AL85000. Approved for public release; further dissemination unlimited. Issued by Sandia National Laboratories, operated for the United States Department of Energy by Sandia Corporation. NOTICE: This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government, nor any agency thereof, nor any of their employees, nor any of their contractors, subcontractors, or their employees, make any warranty, express or implied, or assume any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represent that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise, does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government, any agency thereof, or any of their contractors or subcontractors. The views and opinions expressed herein do not necessarily state or reflect those of the United States Government, any agency thereof, or any of their contractors. Printed in the United States of America. This report has been reproduced directly from the best available copy. Available to DOE and DOE contractors from U.S. Department of Energy Office of Scientific and Technical Information P.O. Box 62 Oak Ridge, TN 37831 Telephone: (865) 576-8401 Facsimile: (865) 576-5728 E-Mail: [email protected] Online ordering: http://www.osti.gov/bridge Available to the public from U.S. Department of Commerce National Technical Information Service 5285 Port Royal Rd. Springfield, VA 22161 Telephone: (800) 553-6847 Facsimile: (703) 605-6900 E-Mail: [email protected] Online order: http://www.ntis.gov/help/ordermethods.asp?loc=7-4- 0#online ii SAND2006-2189 Unlimited Release Printed April 2006 A 50-100 kWe Gas-cooled Reactor For Use On Mars Curtis D. Peters (Year Round Intern) Advanced Nuclear Concepts Sandia National Laboratories P.O. Box 5800 Albuquerque, New Mexico 87185 Abstract In the space exploration field there is a general consensus that nuclear reactor powered systems will be extremely desirable for future missions to the outer solar system. Solar systems suffer from the decreasing intensity of solar radiation and relatively low power density. Radioisotope Thermoelectric Generators are limited to generating a few kilowatts electric (kWe). Chemical systems are short-lived due to prodigious fuel use. A well designed 50-100 kWe nuclear reactor power system would provide sufficient power for a variety of long term missions. This thesis will present basic work done on a 50-100 kWe reactor power system that has a reasonable lifespan and would function in an extraterrestrial environment. The system will use a Gas-Cooled Reactor that is directly coupled to a Closed Brayton Cycle (GCR-CBC) power system. Also included will be some variations on the primary design and their effects on the characteristics of the primary design. This thesis also presents a variety of neutronics related calculations, an examination of the reactor’s thermal characteristics, feasibility for use in an extraterrestrial environment, and the reactor’s safety characteristics in several accident scenarios. While there has been past work for space reactors, the challenges introduced by thin atmospheres like those on Mars have rarely been considered. iii ACKNOWLEGEMENTS The author wishes to thank Ronald J. Lipinski (SNL), Robert Busch (UNM), Steven A. Wright (SNL), Roger X. Lenard (SNL), and Kendall R. DePriest (SNL) for providing technical assistance, opportunity and guidance in the creation of this report. The author also wishes to thank the Sandia National Laboratories (SNL) Student Internship Program (SIP) for making this report possible. iv TABLE OF CONTENTS LIST OF FIGURES ...................................................................................................ix LIST OF TABLES.....................................................................................................x 1.0 Problem Description and Scope of Design ....................................................1 1.1 Problem Description ..........................................................................1 1.2 Design Scope .....................................................................................3 2.0 Background and Literature Review ...............................................................4 2.1 Gas-Cooled Reactor...........................................................................4 2.1.1 History of Gas-cooled Reactors.............................................5 2.1.2 How a Gas-cooled Reactor Works.........................................5 2.2 Brayton Cycle ....................................................................................6 2.2.1 Brayton Cycle History ...........................................................7 2.2.2 Brayton Cycle: How It Works ...............................................7 2.3 Fast Spectrum Reactor .......................................................................9 2.4 Fuel ....................................................................................................10 2.5 Pin Type Geometry............................................................................11 2.6 Space Environment ............................................................................11 3.0 Materials Used in the Design.........................................................................13 3.1 Uranium Nitride.................................................................................13 3.2 Helium/Xenon....................................................................................17 3.3 Niobium 1% Zirconium .....................................................................18 3.4 Hastelloy X ........................................................................................21 3.5 Beryllium Oxide.................................................................................23 v 3.6 Rhenium.............................................................................................25 4.0 Design Description.........................................................................................27 4.1 Design of Reactor ..............................................................................27 4.2 Gas Flow in Reactor...........................................................................34 4.3 Pressure Vessel Composition.............................................................35 4.4 Core Block Composition....................................................................35 4.5 Reactor Control Methods...................................................................36 5.0 Neutronic Investigation into Core Design Results.........................................38 5.1 Initial Calculations.............................................................................38 5.2 Reactor Multiplication Vs Reflector Position....................................39 5.3 High Temperature Feedback effects ..................................................42 5.4 Accident Scenarios.............................................................................45 5.4.1 Immersion in Water ...............................................................46 5.4.2 Immersion in Wet Sand with Water Flooding .......................46 5.4.3 Dry Sand Burial .....................................................................48 5.5 Alternative Cores ...............................................................................50 5.5.1 Category-III Reactor ..............................................................50 5.5.1.1 Alternative 1 Reactor .................................................52 5.5.1.2 Alternative 2 Reactor .................................................55 5.5.2 Boron Carbide Central Safety Rod ........................................57 5.6 Quality Control Runs .........................................................................62 6.0 Thermal Evaluation of Core and Remainder of System ................................66 6.1 Radiator..............................................................................................67 vi 7.0 Conclusions and Future Work .......................................................................72 7.1 Conclusions........................................................................................72 7.2 Future Work.......................................................................................73 8.0 References......................................................................................................70 vii LIST OF FIGURES Figure 1-1. Regions of applicability for different power systems ............................1 Figure 2-1. Closed Brayton Cycle ............................................................................8 Figure 3-1. UN Phase Diagram .................................................................................14 Figure 3-2. UN Mechanical Properties .....................................................................15
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