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Title: SPERT PROJECT QUARTERLY TECHNICAL REPORT, APRIL, MAY, JUNE, 1960

Abstract

< ; 9 3 : 8 7 rogram for kinetic testing of the ORNL Bulk Shielding Reactor BSR-II core was completed. The prompt neutron lifetime parameter 1 BETA /sub e//sub f//sub f/ for the BSR-II core was determined by anaylsis of data from approximately 40 super-prompt-critical power excursions. The dynamic measurements were compsred with those obtained from l/v-absorber, reactor noise, and pulsed neutron measurements. The results from the various techniques agreed within the assigned experimental errors. The previously tested "P" core was reinserted in Spert 1 for use as the driving source for the in-pile capsule studies of transient steam void formation. Prior to the initiation of the capsule program, a series of low-power, large-amplitude, reactivity oscillation measurements was performed to supplement an earlier series of low- and high-power transfer function measurements performed on the P-18/19 core. SPERT II. The intrinsic neutron source level was measured for three different light-water-moderated, highly enriched, uranium-aluminum plate-type cores in Spert II. The level was found to be approximately 3.6 x 10/sup -//sup 1/ neutrons per sec per g of U/sup 2//sup 3//sup 5/, in good agreement with calculated values attributed primarily to the Al( alpha ,n)P reaction where the alpha particles are suppliedmore » by the U/sup 2//sup 3//sup 4/ content of the fuel. Measurements were made in Spert II of the statistical fluctuation of the neutron flux at very low-power levels for two cores at various subcritical multiplications. The differential control rod worths, temperature coefficients and pressure coefficients were determined for a light-watermoderated core containing an excess reactivity of about at ambient temperature and atmospheric pressure. SPERT III. The first series of self-limiting power excursion tests was performed in the Spert III reactor to provide a base point for additional investigations of the effect of various system parameters on the kinetic behavior of the system and to provide a basis for comparison of the behavior of the Spert III core with that of the several cores previously tested in Spert 1. This fiducial test series included power excursions with initial asymptotic periods from 10 sec to 10 msec initiated from room temperature, at atmospheric pressure, and with no forced coolant flow. As expected, the performance of the Spert IIl core was quite similar to that of the other cores. An additional test series was performed to investigate the effect of system pressure on power excursions initiated from room temperature. For shortperiod excursions, as the pressure was raised from 0 to 2500 psig, the delay and eventual suppression of boiling as a shutdown mechanism caused approximately a two-fold increase in energy release and fuel-plate temperature rise with a resultant increase in the reactivity compensation arising from water and fuel plate expansion. The power burst was broadened slightly with only a small increase in the power maximum. ENGlNEERlNG. preliminary burnout heat flux calculations were made on the Spert III reactor core to help establish operational limits for the steady power experiments. As a first approximation, an IBM-650 computer program was coded to calculate the bunnout conditione in the high temperature (400 to 650 deg F) high pressure (2000 to 2500 psig) range. For simplicity, two-phase flow effects were neglected. However, a sample calculation showed burnout heat fiuxes to be in the bulk boiling region for some of the operational ranges investigated. Cut-ofpile hydraulic tests were conducted on a Spert IIl type "C" 1-S fuel assembly in a flow jig simulating the condition of a typical position in the Spert IIl reactor core. A pitot tube located in a position hydraulically similar to one used for the Spent IIl reactor engineering hydraulic tests was calibrated to establish the geometric correlation coefficient. This coefficient was found to be 1.073 plus or minus 0.005 which compares favorably with the value of 1.0 which was assumed for the initial evaluation of the engineering hydraulic test« less

Authors:
Publication Date:
Research Org.:
Phillips Petroleum Co. Atomic Energy Div., Idaho Falls, Idaho
Sponsoring Org.:
US Atomic Energy Commission (AEC)
OSTI Identifier:
4055499
Report Number(s):
IDO-16640
NSA Number:
NSA-15-020327
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; ABSORPTION; ERRORS; EXCURSIONS; LIFETIME; MEASURED VALUES; NEUTRONS; OSCILLATIONS; POWER; PROMPT NEUTRONS; REACTIVITY; REACTORS; SPERT- 1; STEAM; TRANSFER FUNCTIONS; TRANSIENTS; VARIATIONS; WATER COOLANT; WATER MODERATOR; ZONES

Citation Formats

Schroeder, F ed. SPERT PROJECT QUARTERLY TECHNICAL REPORT, APRIL, MAY, JUNE, 1960. United States: N. p., 1961. Web. doi:10.2172/4055499.
Schroeder, F ed. SPERT PROJECT QUARTERLY TECHNICAL REPORT, APRIL, MAY, JUNE, 1960. United States. doi:10.2172/4055499.
Schroeder, F ed. Sat . "SPERT PROJECT QUARTERLY TECHNICAL REPORT, APRIL, MAY, JUNE, 1960". United States. doi:10.2172/4055499. https://www.osti.gov/servlets/purl/4055499.
@article{osti_4055499,
title = {SPERT PROJECT QUARTERLY TECHNICAL REPORT, APRIL, MAY, JUNE, 1960},
author = {Schroeder, F ed},
abstractNote = {< ; 9 3 : 8 7 rogram for kinetic testing of the ORNL Bulk Shielding Reactor BSR-II core was completed. The prompt neutron lifetime parameter 1 BETA /sub e//sub f//sub f/ for the BSR-II core was determined by anaylsis of data from approximately 40 super-prompt-critical power excursions. The dynamic measurements were compsred with those obtained from l/v-absorber, reactor noise, and pulsed neutron measurements. The results from the various techniques agreed within the assigned experimental errors. The previously tested "P" core was reinserted in Spert 1 for use as the driving source for the in-pile capsule studies of transient steam void formation. Prior to the initiation of the capsule program, a series of low-power, large-amplitude, reactivity oscillation measurements was performed to supplement an earlier series of low- and high-power transfer function measurements performed on the P-18/19 core. SPERT II. The intrinsic neutron source level was measured for three different light-water-moderated, highly enriched, uranium-aluminum plate-type cores in Spert II. The level was found to be approximately 3.6 x 10/sup -//sup 1/ neutrons per sec per g of U/sup 2//sup 3//sup 5/, in good agreement with calculated values attributed primarily to the Al( alpha ,n)P reaction where the alpha particles are supplied by the U/sup 2//sup 3//sup 4/ content of the fuel. Measurements were made in Spert II of the statistical fluctuation of the neutron flux at very low-power levels for two cores at various subcritical multiplications. The differential control rod worths, temperature coefficients and pressure coefficients were determined for a light-watermoderated core containing an excess reactivity of about at ambient temperature and atmospheric pressure. SPERT III. The first series of self-limiting power excursion tests was performed in the Spert III reactor to provide a base point for additional investigations of the effect of various system parameters on the kinetic behavior of the system and to provide a basis for comparison of the behavior of the Spert III core with that of the several cores previously tested in Spert 1. This fiducial test series included power excursions with initial asymptotic periods from 10 sec to 10 msec initiated from room temperature, at atmospheric pressure, and with no forced coolant flow. As expected, the performance of the Spert IIl core was quite similar to that of the other cores. An additional test series was performed to investigate the effect of system pressure on power excursions initiated from room temperature. For shortperiod excursions, as the pressure was raised from 0 to 2500 psig, the delay and eventual suppression of boiling as a shutdown mechanism caused approximately a two-fold increase in energy release and fuel-plate temperature rise with a resultant increase in the reactivity compensation arising from water and fuel plate expansion. The power burst was broadened slightly with only a small increase in the power maximum. ENGlNEERlNG. preliminary burnout heat flux calculations were made on the Spert III reactor core to help establish operational limits for the steady power experiments. As a first approximation, an IBM-650 computer program was coded to calculate the bunnout conditione in the high temperature (400 to 650 deg F) high pressure (2000 to 2500 psig) range. For simplicity, two-phase flow effects were neglected. However, a sample calculation showed burnout heat fiuxes to be in the bulk boiling region for some of the operational ranges investigated. Cut-ofpile hydraulic tests were conducted on a Spert IIl type "C" 1-S fuel assembly in a flow jig simulating the condition of a typical position in the Spert IIl reactor core. A pitot tube located in a position hydraulically similar to one used for the Spent IIl reactor engineering hydraulic tests was calibrated to establish the geometric correlation coefficient. This coefficient was found to be 1.073 plus or minus 0.005 which compares favorably with the value of 1.0 which was assumed for the initial evaluation of the engineering hydraulic test},
doi = {10.2172/4055499},
journal = {},
number = ,
volume = ,
place = {United States},
year = {1961},
month = {4}
}