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Title: IDAHO DIVISION SUMMARY REPORT, JULY, AUGUST, SEPTEMBER 1960

Abstract

Eperimental Breeder Reactor I. The fully ribbed and rigid Mark III loading of EBR-I was found to be govenned by feedback processes which guarantee safe and stable operation under normal operating conditions and to give a large radial contribution to the power coefficient. Nonlinearities in the power coefficient were investigated and found to be no problem. If the stabilizing ribs are removed from the fuel rode, a strong positive effect appears which is associated with the inward bowing of fuel rods. The prompt positive coefficient obseved in Mark II is discussed from the standpoint of Mark III tests. A 800-Mwh irradiation run was made on a number of samples, and some bric cladding failures are reported. Data are given for the dimensional changes in EBR-I, Mark III fuel rods used for a total of 2,682 Mwh operating time; the fuel rods usually increased in diameter and decreased in length, and some bowing was obseved. The growth and temperature profiles of the fuel rode are compared, and the effects of radial restraint on the rod growth are discussed. The EBR-I, Mark FV core design is then discussed. The fuel rod will incorporate four plutonium-10 at.% aluminum fuel slugs with two depletedmore » uranium blanket slugs. Calculations were made on the critical mass of Mark IV, which is shown to be 28.3 kg of total plutonium. Zero-power Reactor III (ZPR-III). A mockup of EBR-II was studied in ZPR-III, and the wonth of the mockup control rods was evaluated with tantalum and B/sub 4/ C followers. An EBR-II B/sub 4/C oscillator rod experiment was made in which the excess reactivity was measured as a function of the angular position of the oscillator. The worths of sodium, aluminum, and stainless steel were mapped throughout the core and blanket. From substitution experiments, there appears to be some spectral differences at the center of the clean cores, depending on whether the core is filled with sodium or aluminum. A two-dimensional mapping of the worth of U/sup 235/ and U/sup 238/ was also performed, and their fission rates were determined. Since the EBR-II shields and thimble holes have been redesigned, new mockups were made and their counter responses studied. Transient Reactor Test Facil1ty. The equipment and procedures used to obtain constant power or flat top pulse burats for transient testing of fuel elements are described. The poison sections of the control rode were modified by mixing epoxy resin with graded boron carbide to prevent poison movement. The radiation effects of pulsed bursts on pressure transducers were studied, and it was found that the extraneous pressure signal following the instantaneous power, but not that following the integrated power, can be eliminated. Boiling Reactor Experiment V (Borax V). A general review is given of work done to date, and the design of the reactor and plant is discussed in detail. The core structure is discussed, particularly the spring which allows differential expansion between the core structure and the reactor vessel. A comparison of boiling fuel rods with different diameters is given. The superheater fuel assembly was redesigned with 4 instead of 5 fuel plates because the maximum surface temperature has been lowered from 1200 to 1100 deg F. The reactor control system is compared with those of previous BORAX reactors and EBWR. The construction of the control rods is discussed; the control rod drives which are to be used are those originally used on EBWR. The fuel handling system is discussed. The programs for testing superheater-fiuel assembly seals and developing in-core instrumentation are described. Argonne Fast Source Reactor (AFSR). The present status of AFSR is discussed. Data are presented for the neutron fluxes at various points in AFSR and for the AFSR dimensions. (D.L.C.)« less

Publication Date:
Research Org.:
Argonne National Lab., Ill.
Sponsoring Org.:
USDOE
OSTI Identifier:
4041835
Report Number(s):
ANL-6301
NSA Number:
NSA-15-015275
DOE Contract Number:  
W-31-109-ENG-38
Resource Type:
Technical Report
Resource Relation:
Other Information: Orig. Receipt Date: 31-DEC-61
Country of Publication:
United States
Language:
English
Subject:
REACTOR TECHNOLOGY; AFSR; DISTRIBUTION; FAST NEUTRONS; NEUTRON FLUX; NEUTRON SOURCES; REACTORS

Citation Formats

. IDAHO DIVISION SUMMARY REPORT, JULY, AUGUST, SEPTEMBER 1960. United States: N. p., 1961. Web. doi:10.2172/4041835.
. IDAHO DIVISION SUMMARY REPORT, JULY, AUGUST, SEPTEMBER 1960. United States. https://doi.org/10.2172/4041835
. Tue . "IDAHO DIVISION SUMMARY REPORT, JULY, AUGUST, SEPTEMBER 1960". United States. https://doi.org/10.2172/4041835. https://www.osti.gov/servlets/purl/4041835.
@article{osti_4041835,
title = {IDAHO DIVISION SUMMARY REPORT, JULY, AUGUST, SEPTEMBER 1960},
author = {},
abstractNote = {Eperimental Breeder Reactor I. The fully ribbed and rigid Mark III loading of EBR-I was found to be govenned by feedback processes which guarantee safe and stable operation under normal operating conditions and to give a large radial contribution to the power coefficient. Nonlinearities in the power coefficient were investigated and found to be no problem. If the stabilizing ribs are removed from the fuel rode, a strong positive effect appears which is associated with the inward bowing of fuel rods. The prompt positive coefficient obseved in Mark II is discussed from the standpoint of Mark III tests. A 800-Mwh irradiation run was made on a number of samples, and some bric cladding failures are reported. Data are given for the dimensional changes in EBR-I, Mark III fuel rods used for a total of 2,682 Mwh operating time; the fuel rods usually increased in diameter and decreased in length, and some bowing was obseved. The growth and temperature profiles of the fuel rode are compared, and the effects of radial restraint on the rod growth are discussed. The EBR-I, Mark FV core design is then discussed. The fuel rod will incorporate four plutonium-10 at.% aluminum fuel slugs with two depleted uranium blanket slugs. Calculations were made on the critical mass of Mark IV, which is shown to be 28.3 kg of total plutonium. Zero-power Reactor III (ZPR-III). A mockup of EBR-II was studied in ZPR-III, and the wonth of the mockup control rods was evaluated with tantalum and B/sub 4/ C followers. An EBR-II B/sub 4/C oscillator rod experiment was made in which the excess reactivity was measured as a function of the angular position of the oscillator. The worths of sodium, aluminum, and stainless steel were mapped throughout the core and blanket. From substitution experiments, there appears to be some spectral differences at the center of the clean cores, depending on whether the core is filled with sodium or aluminum. A two-dimensional mapping of the worth of U/sup 235/ and U/sup 238/ was also performed, and their fission rates were determined. Since the EBR-II shields and thimble holes have been redesigned, new mockups were made and their counter responses studied. Transient Reactor Test Facil1ty. The equipment and procedures used to obtain constant power or flat top pulse burats for transient testing of fuel elements are described. The poison sections of the control rode were modified by mixing epoxy resin with graded boron carbide to prevent poison movement. The radiation effects of pulsed bursts on pressure transducers were studied, and it was found that the extraneous pressure signal following the instantaneous power, but not that following the integrated power, can be eliminated. Boiling Reactor Experiment V (Borax V). A general review is given of work done to date, and the design of the reactor and plant is discussed in detail. The core structure is discussed, particularly the spring which allows differential expansion between the core structure and the reactor vessel. A comparison of boiling fuel rods with different diameters is given. The superheater fuel assembly was redesigned with 4 instead of 5 fuel plates because the maximum surface temperature has been lowered from 1200 to 1100 deg F. The reactor control system is compared with those of previous BORAX reactors and EBWR. The construction of the control rods is discussed; the control rod drives which are to be used are those originally used on EBWR. The fuel handling system is discussed. The programs for testing superheater-fiuel assembly seals and developing in-core instrumentation are described. Argonne Fast Source Reactor (AFSR). The present status of AFSR is discussed. Data are presented for the neutron fluxes at various points in AFSR and for the AFSR dimensions. (D.L.C.)},
doi = {10.2172/4041835},
url = {https://www.osti.gov/biblio/4041835}, journal = {},
number = ,
volume = ,
place = {United States},
year = {1961},
month = {10}
}