40-MW(E) PROTOTYPE HIGH-TEMPERATURE GAS-COOLED REACTOR RESEARCH AND DEVELOPMENT PROGRAM. Quarterly Progress Report, January 1, 1961-March 31, 1961

Technical Report · Tue Oct 31 00:00:00 EST 1961

OSTI ID:4834272

Summary research and development studies directed toward the construction of a 40-Mw(e) prototype power plant employing a high-temperature, gas-cooled, graphitemoderated reactor (HTGR) are described. These include: the development of pyrolytic coating of carbide particles to a point of scaleup for production purposes; the development of nondestructive testing methods for graphite tubes; the assembly of a 4-ft-long element, together with a backup element, for in-pile testing; and the evaluation of lowpermeability graphite samples after an irradiation to approximately 10/sup 21/ nvt to 1400 deg C. Other accomplishments include the first preliminary successful results on the production of U - Th carbides, on a laboratory scale, and production of 4-ft lengths of graphite tubes. The most striking irradiation effects on impermeable graphites which were evaluated concern the relatively large degree of contraction observed on several samples. Installation of the in-pile loop at the General Electric Test Reactor is essentially complete. The main loop and the fission- product trapping and sampling systems have been leak-checked. Two fully rated circulators have been installed in the loop and have been run for approximately 5 hr each. Work on the development of codes for nuclear calculations continued with the completion of a one-dimensional burnup program designated FEVER and the final checkout and use of the two-dimensional code DDB. The half-scale now model was assembled, instrumented, and checked out. Heat-transfer, flow, and pressure data have been taken, and reduction and analysis are under way. The total pressure drop of the prototype HTGR has been calculated based on flow-model results. Studies on the pressure vessel and internals have included methods for emergency cooling of the vessel, analyses of neutron streaming out of the concentric nozzles, and attemperation of the outlet gas following a loss-of- pressure accident. The design of the rockerarm reflector is continuing with studies on the linkage mechanism and the vertical seals. The soundness of the principle of holding the core together by the gas pressure on the outer surface of the reflector blocks has been established in tests with the half-scale flow model. Results obtained from tests on structural materials in helium contaminated with H/sub 2/ and CO indicate that oxidation rather than carburization appears to be the major corrosion process influencing the materials studied. The design of the prototype control rod and drive was completed, and fabrication of the components for the rod and drive system is essentially complete. A flow model of the concentric pipe was designed and built to obtain heat-transfer and pressure-drop data. A simulated fuel compact containing coated fuel particles was irradiated to an exposure equivalent to about three years in the HTGR. The release characteristics of the irradiated compact showed evidence that a majority of the particles were still intact. Experiments were initiated to study the steady-state release of the short-lived isotopes of krypton and xenon. Studies on a coolant monitoring system and a failed-fuel-element location system were completed. An operational analysis effort was initiated to determine in detail the operational aspects of the plant during normal and abnormal operating conditions. Intensive studies on safety and hazards were made: of core oxidation resulting from the admittance of air following a rupture of the main coolant system; the transient differential pressures across internal components and the pressure increase in the plant container during blowdown following rupture; and the adiabatic pressure peak in the containment shell caused by ruptures in either the coolant system or the steam generator, or both. Design studies included emergency isolation of the primary loops, emergency cooling of the pressure vessel, and emergency shutdown systems. (auth)

• Research Organization: General Atomic Div.# General Dynamics Corp., San Diego, Calif
• DOE Contract Number: AT(04-3)-314
• NSA Number: NSA-15-032977
• OSTI ID: 4834272
• Report Number(s): GA-2204
• Resource Relation: Other Information: Orig. Receipt Date: 31-DEC-61
• Country of Publication: United States
• Language: English

Related Subjects

REACTOR TECHNOLOGY
AIR
BURNUP
CARBON MONOXIDE
COATING
COMPUTERS
CONFIGURATION
CONTAMINATION
CONTROL ELEMENTS
COOLING
CORROSION
DEFORMATION
DETECTION
EQUATIONS
FAILURES
FISSION PRODUCTS
FUEL ELEMENTS
GAS COOLANT
GAS FLOW
GRAPHITE MODERATOR
HEAT TRANSFER
HELIUM
HIGH TEMPERATURE
HTGR
IN PILE LOOPS
INSTRUMENTS
IRRADIATION
KRYPTON
MATERIALS TESTING
MECHANICAL STRUCTURES
MIXING
MOCKUP
MONITORING
MOTORS
NEUTRON BEAMS
NUCLEAR REACTIONS
OPERATION
OXIDATION
PARTICLES
POROSITY
POWER PLANTS
PRESSURE
PRESSURE VESSELS
PROGRAMMING
PYROLYSIS
REACTOR SAFETY
REACTORS
REFLECTORS
SAFETY
SAMPLING
SEALS
SHIELDING
SHUTDOWN
TESTING