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Title: REACTOR ENGINEERING DIVISION QUARTERLY REPORT FOR JULY, AUGUST, SEPTEMBER 1955. SECTION II

Abstract

: 2 9 9 3 9 8 9 8 8 < 9 9 < corrosion of Zircaloy-2 clad stainless steel thimbles is shown. Designs of the EBWR control rod and drive mechanism are given. The deceleration curve for the drive shaft at 600-psig operation is given. The prototype test facility for the control mechanisms is also shown. Calculated heat generation and thermal stresses in the pressure vessel wall for each type of thermal shieding are tabulated. The mechanical arrangement of the thermal shielding is given. The pressure vessel containment structure is described in detall, and calculated impact loading stresses are summarized. A redwood, steel, and celotex blast shield for the pressure vessel is also shown. Fast and thermal neutron flux calculations for the heavy water-reflected EBWR in the adjusted shield are summarized. The physical arrangement of this shielding is also described. An emergency boric acid injection system to reduce the reactor core activity below criticality was designed. The reactor water purification system is discussed. A schematic flow diagram of the air drying and fluid recovery system is given. Estimated flows in this system are also given, and the operation of the system is briefly discussed. A schematic plan view ismore » shown of the EBWR power plant and control areas. A simplified line diagram is also given of the EBWR electrical system. The schematic for the emergency airlock doors for the EBWR is shown. Supporting and Alternate Resign Research and Development. The design of a boiling D/sub 2/O reactor is briefly discussed. Data given on boiling heat transfer studies include the ratio of steam-toliquid velocity as a function of the variable inlet temperature, power, flow, and pressure. The analytical and experimental studies made of the 600 psi, unrestricted, singlechannel circulation loop (Armadilla) are described, and preliminary data are plotted. These plots show modified Martinelli-Nelson factors as a function of slip ratio and exit and local void fraction. Data from single channel forced and natural circulation boiling density tests at 150 psig are summarized. Slip ratios are also plotted as a function of inlet water velocity at 150 and 250 psig, indicating that there is no significant pressure effect on the slip ratio pattern in going from 150 to 250 psig. The corrosion of Zircaloy 2-clad, U--Zr fuel plates in 550 deg F degassed water. The corrosion behavior of Al-clad, nataral U plates is compared to that of EBWR reference alloy plates. The corrosion of Al--Ni alloy in 500 deg F H/sub 2/O is also shown. Corrosion data are also reported for austenitic stainless steel, and its weight changes are plotted. Corrosion and irradiation tests were continued on urania --thoria bodies. Several methods were tested for the detection and location of ruptured fuel elements in the core of a reactor. These included comparisons of fission product activities in various sampling devices and of gamma activities in H/sub 2/O samples. It was, however, concluded that neither of these methods would be useful. A proposed alternate control rod for the EBWR is described and discussed. Progress in the development of boron- stainless steel alloy is discussed. Boron--stainless steel corrosion tests are summarized. Results of experiments on the removal of boric acid using Amberlite IRA-400 resin are summarized. Experimental Breeder Reactor I. The conversion ratio was measured in the second U fuel loading of the EBR-I. U/sup 238/ capture distributions and U/sup 235/ fission distributions in the reactor are shown. The conversion ratio with the cup about 1/4 in. below its uppermost position was found to be 1.03 plus or minus 0.05. Activation cross section ratios at various locations in the outer blanket of the EBR were surveyed. Experimental and theoretical values are compared. Experimental Breeder Reactor II. The dissipation of simulated decay heat generated in the EBR-II core subassembly is shown. Water pressure drop tests were conducted on the complete EBR-II core subassembly, and the results of the tests are summarized. Calculations of hold« less

Publication Date:
Research Org.:
Argonne National Lab., Lemont, Ill.
OSTI Identifier:
4351215
Report Number(s):
ANL-5511
NSA Number:
NSA-12-000977
DOE Contract Number:  
W-31-109-ENG-38
Resource Type:
Technical Report
Resource Relation:
Other Information: Decl. Jan. 17, 1957. Orig. Receipt Date: 31-DEC-58
Country of Publication:
United States
Language:
English
Subject:
PHYSICS; ABSORPTION; ACCIDENTS; AIR; ALLOYS; ALUMINUM; ALUMINUM ALLOYS; AMBERLITE; AUSTENITE; BOILING; BORON; BORON OXIDES; BUILDINGS; CANNING; CIRCUITS; CONDENSERS; CONFIGURATION; CONTAMINATION; CONTROL ELEMENTS; CONVECTION; COOLANT LOOPS; CORROSION; CORROSION PROTECTION; CRACKS; CRITICALITY; DEGASSING; DENSITY; DIAGRAMS; DISTRIBUTION; EBR-1; EBWR; EFFICIENCY; ELECTRICITY; ENVIRONMENT; EQUATIONS; EVAPORATION; EXPLOSIONS; FAILURES; FAST NEUTRONS; FISSION PRODUCTS; FLUID FLOW; FUEL ELEMENTS; GAMMA RADIATION; GAS FLOW; HEAT TRANSFER; HEATING; HEAVY WATER; HEAVY WATER COOLANT; HIGH TEMPERATURE; HYDRAULICS; IMPACT SHOCK; IMPURITIES; INJECTION

Citation Formats

. REACTOR ENGINEERING DIVISION QUARTERLY REPORT FOR JULY, AUGUST, SEPTEMBER 1955. SECTION II. United States: N. p., 1956. Web. doi:10.2172/4351215.
. REACTOR ENGINEERING DIVISION QUARTERLY REPORT FOR JULY, AUGUST, SEPTEMBER 1955. SECTION II. United States. https://doi.org/10.2172/4351215
. Sun . "REACTOR ENGINEERING DIVISION QUARTERLY REPORT FOR JULY, AUGUST, SEPTEMBER 1955. SECTION II". United States. https://doi.org/10.2172/4351215. https://www.osti.gov/servlets/purl/4351215.
@article{osti_4351215,
title = {REACTOR ENGINEERING DIVISION QUARTERLY REPORT FOR JULY, AUGUST, SEPTEMBER 1955. SECTION II},
author = {},
abstractNote = {: 2 9 9 3 9 8 9 8 8 < 9 9 < corrosion of Zircaloy-2 clad stainless steel thimbles is shown. Designs of the EBWR control rod and drive mechanism are given. The deceleration curve for the drive shaft at 600-psig operation is given. The prototype test facility for the control mechanisms is also shown. Calculated heat generation and thermal stresses in the pressure vessel wall for each type of thermal shieding are tabulated. The mechanical arrangement of the thermal shielding is given. The pressure vessel containment structure is described in detall, and calculated impact loading stresses are summarized. A redwood, steel, and celotex blast shield for the pressure vessel is also shown. Fast and thermal neutron flux calculations for the heavy water-reflected EBWR in the adjusted shield are summarized. The physical arrangement of this shielding is also described. An emergency boric acid injection system to reduce the reactor core activity below criticality was designed. The reactor water purification system is discussed. A schematic flow diagram of the air drying and fluid recovery system is given. Estimated flows in this system are also given, and the operation of the system is briefly discussed. A schematic plan view is shown of the EBWR power plant and control areas. A simplified line diagram is also given of the EBWR electrical system. The schematic for the emergency airlock doors for the EBWR is shown. Supporting and Alternate Resign Research and Development. The design of a boiling D/sub 2/O reactor is briefly discussed. Data given on boiling heat transfer studies include the ratio of steam-toliquid velocity as a function of the variable inlet temperature, power, flow, and pressure. The analytical and experimental studies made of the 600 psi, unrestricted, singlechannel circulation loop (Armadilla) are described, and preliminary data are plotted. These plots show modified Martinelli-Nelson factors as a function of slip ratio and exit and local void fraction. Data from single channel forced and natural circulation boiling density tests at 150 psig are summarized. Slip ratios are also plotted as a function of inlet water velocity at 150 and 250 psig, indicating that there is no significant pressure effect on the slip ratio pattern in going from 150 to 250 psig. The corrosion of Zircaloy 2-clad, U--Zr fuel plates in 550 deg F degassed water. The corrosion behavior of Al-clad, nataral U plates is compared to that of EBWR reference alloy plates. The corrosion of Al--Ni alloy in 500 deg F H/sub 2/O is also shown. Corrosion data are also reported for austenitic stainless steel, and its weight changes are plotted. Corrosion and irradiation tests were continued on urania --thoria bodies. Several methods were tested for the detection and location of ruptured fuel elements in the core of a reactor. These included comparisons of fission product activities in various sampling devices and of gamma activities in H/sub 2/O samples. It was, however, concluded that neither of these methods would be useful. A proposed alternate control rod for the EBWR is described and discussed. Progress in the development of boron- stainless steel alloy is discussed. Boron--stainless steel corrosion tests are summarized. Results of experiments on the removal of boric acid using Amberlite IRA-400 resin are summarized. Experimental Breeder Reactor I. The conversion ratio was measured in the second U fuel loading of the EBR-I. U/sup 238/ capture distributions and U/sup 235/ fission distributions in the reactor are shown. The conversion ratio with the cup about 1/4 in. below its uppermost position was found to be 1.03 plus or minus 0.05. Activation cross section ratios at various locations in the outer blanket of the EBR were surveyed. Experimental and theoretical values are compared. Experimental Breeder Reactor II. The dissipation of simulated decay heat generated in the EBR-II core subassembly is shown. Water pressure drop tests were conducted on the complete EBR-II core subassembly, and the results of the tests are summarized. Calculations of hold},
doi = {10.2172/4351215},
url = {https://www.osti.gov/biblio/4351215}, journal = {},
number = ,
volume = ,
place = {United States},
year = {1956},
month = {1}
}