Sm-2 Reactor Core and Vessel Review Report for Period December 15, 1959 to March 18, 1960

Technical Report · Wed Mar 30 00:00:00 EST 1960

DOI:https://doi.org/10.2172/4161742· OSTI ID:4161742

Hoover, H. L. ^[1]

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1. Alco Products, Inc., Schenectady, NY (United States)

The SM-2 core lifetime was calculated and stuck rod criteria formulated. The control system could shutdown the core under the most adverse conditions with any single rod stuck out. The one-half-inch Eu₂O₃ flux suppressors adequately suppressed the power spike at the bottom of the core. The nuclear effects of increasing the dimensions of the fuel matrix were calculated. A new analysis of various reactivity measurements performed during the SM-2 critical experiments was made to test the validity of the calculational modes. Two-dimensional temperature distributions for the stainless steel vessel shell and flange and for the SM-2 control rod plates were plotted. The loss of flow experiment was completed on the SM-1 at Fort Belvoir, and the results agreed with analytical predictions. The SM-2 fuel element, control rod, core support structure, and vessel designs were completed. Three steel mills indicated they could supply low cobalt, low tantalum type 347 stainless steel. Corrosion testing of irradiated boron-stainless steel was reactivated. Corrosion and impact testing of nut end bolting materials was scheduled. Dysprosium oxide was considered as an alternate absorber material. Two modified SM-2 fuel assemblies were fabricated for testing in the Westinghouse Test Reactor loop, and two fuel elements were fabricated for insertion in the SM-1 core. Work on controlled addition of a burnable poison, such as ZrB₂, in fuel plates was continued. Procedures for welding fuel plates to side plates were completed. Automatic weld sequence timing was adopted for improving product quality. Pressure losses through the complete flow circuit in the reactor were recalculated and the pressure drop analysis brought up to date. Good flow distribution in the control rod fuel element was obtained in a single element rig. Methods of temperature regulation and corrosion control were established. Extended SM-2 critical experiments were initiated. Core support design modifications necessary for an accurate flow divider mockup were completed. Effects of flux suppression upon reactivity and neutron flux at the bottom of the core were measured. Gamma dose rates were measured over the core support plate and along the control rod basket guide. Prototype testing of the control rod drive mechanism was started. The blocked channel method of measuring fuel element internal temperature in SM-2 Core I instrumental fuel assembly was selected as the reference design.

• Research Organization: Alco Products, Inc., Schenectady, NY (United States)
• Sponsoring Organization: USDOE National Nuclear Security Administration (NNSA), Nuclear Criticality Safety Program (NCSP); US Atomic Energy Commission (AEC)
• DOE Contract Number: AT(30-3)-326
• NSA Number: NSA-14-017594
• OSTI ID: 4161742
• Report Number(s): APAE-Memo-250
• Resource Relation: Other Information: With this is bound: Battelle Memorial Inst., Columbus, Ohio. SM-2 CORE MATERIALS DEVELOPMENT PROGRAM. Mar. 11, 1960. Orig. Receipt Date: 31-DEC-60
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
73 NUCLEAR PHYSICS AND RADIATION PHYSICS
ABSORPTION
BORIDES
COBALT ALLOYS
CONTROL ELEMENTS
CONTROL SYSTEMS
CORROSION
CRITICALITY
DISTRIBUTION
DYSPROSIUM OXIDES
EQUATIONS
EUROPIUM OXIDES
FABRICATION
FUEL ELEMENTS
GAMMA RADIATION
INSTRUMENTS
LIFETIME
LIQUID FLOW
MACHINE PARTS
MEASURED VALUES
MECHANICAL STRUCTURES
NEUTRON FLUX
PLANNING
POISONING
POWER PLANTS
PRESSURE
PRESSURE VESSELS
RADIATION DOSES
REACTIVITY
REACTOR CORE
REACTORS
RODS
SM-2
STAINLESS STEELS
TANTALUM ALLOYS
TEMPERATURE
TESTING
WATER COOLANT
WELDING
ZIRCONIUM BORIDES
Nuclear Criticality Safety Program (NCSP)
CANDLE-2
IBM-650 Code
Struck Rod Criteria
T.P. 600
BM 1 Development
Test in WTR
Test in SM-1