Reactor Engineering Division (Quarterly Report July, August, September 1955: Section II)

Technical Report · Sun Jan 01 00:00:00 EST 1956

DOI:https://doi.org/10.2172/4351215· OSTI ID:4351215

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 shielding are tabulated. The mechanical arrangement of the thermal shielding is given. The pressure vessel containment structure is described in detail, 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 D₂O reactor is briefly discussed. Data given on boiling heat transfer studies include the ratio of steam-to-liquid velocity as a function of the variable inlet temperature, power, flow, and pressure. The analytical and experimental studies made of the 600 psi, unrestricted, single-channel 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°F degassed water. The corrosion behavior of Al-clad, natural U plates is compared to that of EBWR reference alloy plates. The corrosion of Al--Ni alloy in 500°F H₂O 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 H₂O 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.

• Research Organization: Argonne National Lab. (ANL), Argonne, IL (United States)
• Sponsoring Organization: USDOE National Nuclear Security Administration (NNSA), Nuclear Criticality Safety Program (NCSP); US Atomic Energy Commission (AEC)
• Contributing Organization: Reactor Engineering Division
• DOE Contract Number: W-31-109-ENG-38
• NSA Number: NSA-12-000977
• OSTI ID: 4351215
• Report Number(s): ANL-5511
• Resource Relation: Other Information: Decl. Jan. 17, 1957. Orig. Receipt Date: 31-DEC-58
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
42 ENGINEERING
73 NUCLEAR PHYSICS AND RADIATION PHYSICS
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
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
Nuclear Criticality Safety Program (NCSP)
Corrosion
Prototype Testing
Biological Shielding
Steam Plant
Forced Circulation
H2O