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Title: SPERT PROJECT QUARTERLY TECHNICAL REPORT, JANUARY, FEBRUARY, MARCH 1960

Abstract

SPERT 1. A series of experiments relating to the problem of reactor startup from very low initial power was performed with a cold, clean, stainless steel core in the Spert I facility. tne intrinsic neutron source level for th core was measured and found to be approximately 500 n/sec Time delays were observed between the step-wise injection of reactivity at power levels of abeut 10/sup -//sup 5/ watts and the attainment of a stable period. In some cases these delays, which are attributed in part to the statistical properties of neutron chains, were as long as 2 seconds. In one sourceless startup test, control rods were withdrawn at a rate of 20(/sec with period and level scram circuits operative. The period circuit scrammed the reactor at a trivial power level before the reactor period became shorter than 50 msec. SPERT II. Construction of the Spert II facility was completed. Initial criticality was achieved in light water with 24, type "B" fuel assemblies containing 24 plates each and 8, 17-plate, control rod assemblies. The total U/sup 2//sup 3//sup 5/ mass of this initial core was 4.69 kg. On the basis of differential rod worth dsta, the critical mass for this configurationmore » was estimsted to be 4.6 kg of U/ sup 2//sup 3//sup 5/. An operational core loading containing 6.03 kg of U/sup 2// sup 3//sup 5/ was found to have an available excess reactivity of about st a tempersture of 400 F and abeut 8 at ambient tempersture. ln anticipstion of future transient testing of a core with a central positive void coefficient, a core loading was investigsted in which the plate spacing was increased for the central fuel assemblies. The central void coefficient of approximately +2 x 10/ sup -//sup 2/ (/cm/sup 3/ was measured for this core. SPERT III. In order to obtain power calibration dsta for the Spert III neutron chambers and to provide information for future use in the analysis of transient data, the neutron flux distribution was measured st ambient temperature in the Spert III operational core by activation of cobalt wires. The average power level during the irradiation was calculated from the measured flux values and combined with the neutron chamber output dsta to yield approximste calibration factors for each chamber. The results indicate thst power levels from 5 w to 20 Gw can be measured with the present arrangement of four chambers. ENGINEERING. The hydraulic characteristics of the type "Dr" fuel assemblies were investigated. These loose-plate-type assemblies have been designed for use in the Spert IV reactor. The pressure drop as a function of flow was calculated and found to agree very well with the experimental data obtained in the ETR flow test loop. Plate flutter is not excessive with flow rates up to 610 gpm through the assembly and it is concluded that the type "D" assembly will be acceptable hydrodynamically for use in Spert IV. Pressureflow relationships were calculated as a function of water temperature for type "B" fuel assemblies with 8, 12, and 24 fuel plates. The calculations agree with available experimental data at 85 deg F for the 24- late assembly. (auth)« less

Authors:
Publication Date:
Research Org.:
Phillips Petroleum Co. Atomic Energy Div., Idaho Falls, Idaho
Sponsoring Org.:
US Atomic Energy Commission (AEC)
OSTI Identifier:
4091695
Report Number(s):
IDO-16617
NSA Number:
NSA-15-020461
DOE Contract Number:  
AT(10-1)-205
Resource Type:
Technical Report
Resource Relation:
Other Information: Orig. Receipt Date: 31-DEC-61
Country of Publication:
United States
Language:
English
Subject:
REACTOR TECHNOLOGY; COBALT; CONFIGURATION; CONTROL ELEMENTS; CONTROL SYSTEMS; COOLANT LOOPS; CRITICALITY; DISTRIBUTION; EQUATIONS; FUEL ELEMENTS; HYDRAULICS; LIQUID FLOW; MASS; MEASURED VALUES; NEUTRON DETECTION; NEUTRON FLUX; NEUTRON SOURCES; OPERATION; PLANNING; PLATES; PRESSURE; REACTIVITY; REACTOR CORE; REACTOR SAFETY; REACTORS; SHUTDOWN; SPERT-1; SPERT-2; SPERT-3; STAINLESS STEELS; STARTUP; STATISTICS; TRANSIENTS; URANIUM 235; VARIATIONS; WATER MODERATOR; WIRES; ZONES

Citation Formats

Wilson, T.R. ed. SPERT PROJECT QUARTERLY TECHNICAL REPORT, JANUARY, FEBRUARY, MARCH 1960. United States: N. p., 1961. Web. doi:10.2172/4091695.
Wilson, T.R. ed. SPERT PROJECT QUARTERLY TECHNICAL REPORT, JANUARY, FEBRUARY, MARCH 1960. United States. doi:10.2172/4091695.
Wilson, T.R. ed. Fri . "SPERT PROJECT QUARTERLY TECHNICAL REPORT, JANUARY, FEBRUARY, MARCH 1960". United States. doi:10.2172/4091695. https://www.osti.gov/servlets/purl/4091695.
@article{osti_4091695,
title = {SPERT PROJECT QUARTERLY TECHNICAL REPORT, JANUARY, FEBRUARY, MARCH 1960},
author = {Wilson, T.R. ed.},
abstractNote = {SPERT 1. A series of experiments relating to the problem of reactor startup from very low initial power was performed with a cold, clean, stainless steel core in the Spert I facility. tne intrinsic neutron source level for th core was measured and found to be approximately 500 n/sec Time delays were observed between the step-wise injection of reactivity at power levels of abeut 10/sup -//sup 5/ watts and the attainment of a stable period. In some cases these delays, which are attributed in part to the statistical properties of neutron chains, were as long as 2 seconds. In one sourceless startup test, control rods were withdrawn at a rate of 20(/sec with period and level scram circuits operative. The period circuit scrammed the reactor at a trivial power level before the reactor period became shorter than 50 msec. SPERT II. Construction of the Spert II facility was completed. Initial criticality was achieved in light water with 24, type "B" fuel assemblies containing 24 plates each and 8, 17-plate, control rod assemblies. The total U/sup 2//sup 3//sup 5/ mass of this initial core was 4.69 kg. On the basis of differential rod worth dsta, the critical mass for this configuration was estimsted to be 4.6 kg of U/ sup 2//sup 3//sup 5/. An operational core loading containing 6.03 kg of U/sup 2// sup 3//sup 5/ was found to have an available excess reactivity of about st a tempersture of 400 F and abeut 8 at ambient tempersture. ln anticipstion of future transient testing of a core with a central positive void coefficient, a core loading was investigsted in which the plate spacing was increased for the central fuel assemblies. The central void coefficient of approximately +2 x 10/ sup -//sup 2/ (/cm/sup 3/ was measured for this core. SPERT III. In order to obtain power calibration dsta for the Spert III neutron chambers and to provide information for future use in the analysis of transient data, the neutron flux distribution was measured st ambient temperature in the Spert III operational core by activation of cobalt wires. The average power level during the irradiation was calculated from the measured flux values and combined with the neutron chamber output dsta to yield approximste calibration factors for each chamber. The results indicate thst power levels from 5 w to 20 Gw can be measured with the present arrangement of four chambers. ENGINEERING. The hydraulic characteristics of the type "Dr" fuel assemblies were investigated. These loose-plate-type assemblies have been designed for use in the Spert IV reactor. The pressure drop as a function of flow was calculated and found to agree very well with the experimental data obtained in the ETR flow test loop. Plate flutter is not excessive with flow rates up to 610 gpm through the assembly and it is concluded that the type "D" assembly will be acceptable hydrodynamically for use in Spert IV. Pressureflow relationships were calculated as a function of water temperature for type "B" fuel assemblies with 8, 12, and 24 fuel plates. The calculations agree with available experimental data at 85 deg F for the 24- late assembly. (auth)},
doi = {10.2172/4091695},
journal = {},
number = ,
volume = ,
place = {United States},
year = {1961},
month = {3}
}