QUARTERLY PROGRESS REPORT FOR JANUARY, FEBRUARY, MARCH 1959
8 : 8 8 7 > < 8 9 < " reactor shutdown of heat transfer to the moderator, a series of experiments were performed in the SPERT I reactor using insulated Type A fuel assemblies. Three heat transfer conditions were investigated: (1) boiling temperatures not attained prior to reactor self- shutdown, (2) boiling temperatures existent throughout the tests, and (3) boiling temperatures attained prior to shutdown for an uninsulated core, but not for the insulated core. Comparison of the results of the insulated core experiments with previous Acore data revealed no difference in behavior. The need for detailed studies of spontaneous shutdowu mechanisms stimulated a series of capsule-type experiments designed to measure moderator expulsion and radlolytic gas formation during a reactor excursion. Temperature and void coefficient calculations for the SPERT I B-12/64 core have been made and are in essential agreement with available experimental results. A tabulation of the 20'C steptransient test data for the P-18/19 core has been completed. SPERT II. A calculation has been made to determine the excess reactivity of a proposed Al-H/sub 2/O, 8-plate modified B- type core at the operational conditions of 400 psig and 400 deg F. The results indicate that alternate plate loadings of 7 grams per plate and 17 grams per plate of U/sup 235/, respectively, are sufficient to attain the 2% excess reactivity required for kinetic experiments. SPERT II The initial cold critical experiments with the 27-assembly loading was repeated and additional cold critical experiments were performed with successively larger loadings through a 48-assembly core. A total of 11 different loadings was investigated. Hot critical experiments were commenced with the 48-assembly loading. Operating experience gained during the initial hot experiments and subsequent pressure operation has led to a number of component modifications. Theoretical. The way in which the compensated reactivily at the time of peak power versus reciprocal period depends on reactor properties bas been determined analytically. The shape of this curve is a function of delayed neutron distribution, the prompt neutron lifetime and the shape of the transient power burst. A minimum in the curve turns out to be a non-essential characteristic for some practical cases. The results of this analysis may be used to incorporate the effects of delayed neutrons in the predictions of peak power and energy in the long period region. A variable reactivity coefficient having a positive or zero initial value at power levels below the operating level of a reactor and becoming negative above the operating level will allow a much lower investment in flexible k than that required for a reactor having a strong negative coefficient. The effect of such a reactivity coefficient on selflimiting power bursts bas been investigated analytically for a simple case in terms of previously considered energy models and it appears that a pcsitive or zero initial reactivity coefficient can result in an over-all gain in reactor safety. System requirements in terms of nnsximum allowable energy have been determined. The analysis may be extended to other systems. Engineering. Corrosion and galling of threaded fasteners used in reactor internal structures under high temperature (650 deg F) demineralized water service has initiated an investigation to determine more suitable materials. Warping and blistering of P18/19 core fuel assembly plates has prompted a metallurgical investigation to determine the extent of damage and possible cause of the failure. Fuel and cladding temperatures and heat transfer rates for a 5 msec transient test were calculated. Center-line meat temperatures were shown to reach approximately 1500 deg F with peak heat fluxes of 1.1 x 10/sup 7/ Btu/ (hr)(ft/sup 2/). Construction of the facility is estimated at 80% completed. Reactor components are being modified, based on SPERT III experience, with design and fabrication estimated at 90% complete. Fuel assembly fab
- Research Organization:
- Phillips Petroleum Co. Atomic Energy Div., Idaho Falls, Idaho
- DOE Contract Number:
- AT(10-1)-205
- NSA Number:
- NSA-14-004876
- OSTI ID:
- 4214299
- Report Number(s):
- IDO-16539
- Resource Relation:
- Other Information: Orig. Receipt Date: 31-DEC-60
- Country of Publication:
- United States
- Language:
- English
Similar Records
SPERT PROJECT. Quarterly Technical Report for April, May, June 1959
SPERT PROJECT QUARTERLY TECHNICAL REPORT, JANUARY, FEBRUARY, MARCH 1960
Related Subjects
ALUMINUM- FUEL ELEMENTS- PLATES- PRESSURE- QUANTITY RATIO- - REACTIVITY- REACTOR CORE- REACTOR FUELING- REACTORS- SPERT-2- TEMPERATURE- URANIUM 235- WATER
BOILING- DECOMPOSITION- EXCURSIONS- GASES- HEAT TRANSFER- MODERATORS- RADIATION CHEMISTRY- RADIOLYSIS- REACTOR CORE- REACTORS- SHUTDOWN- SPERT-1- TABLES- TEMPERATURE- THERMAL INSULATION- TRANSIENTS- ZONES
CLEANING- CONTROL ELEMENTS- CORROSION- FAILURES- FUEL ELEMENTS- HEAT TRANSFER- HEAVY WATER- HIGH TEMPERATURE- INSTRUMENTS- LEAKS- LOSSES- MECHANICS- METALLURGY- PLANNING- PREPARATION- REACTORS- SPERT-3- TEMPERATURE- TRANSIENTS- VALVES- WATER