SECOND LCTL TEST OF HNPF MODERATOR CANS

Technical Report · Thu Apr 20 04:00:00 EST 1961

OSTI ID:4840853

Phillips, E C

The Hallam Power Reactor moderator can incorporating a 6-segment graphite log and a 1/9 hard Type 304 stainless steel sheath was tested in sodium at 1000 deg F and found to be suitable for use in the reactor. Three full-size moderator cans were fabricated; can 3 was essentially identical to a production HNPF can while cans 1 and 2 differed tn that they utilized annealed 304 stainless steel for the sheath. In addition, can 1 did not contain a zirconiumsponge getter and utilized a single unit log rather than a log of 6 separate segments. Cans 1 and 2, installed in the core tank of the LCTL, were slowly heated in a gas atmosphere. The strains which developed in the can sheaths, because of differential expansion between the graphite logs and the sheaths, were measured. Inconclusive strain data on can 2 were obtained and data for 6-segment cans were subsequently obtained on can 3. Strains developed in a single-log can were approximately 6 times higher than those developed in the 6-segment can. In addition to the friction-strain tests, the cans were thermally cycled in sodium in each phase of the program. These exposures demonstrated that the cans do not leak, the zirconium getter becomes effective above 850 deg F, and the 1/8 hard material develops less permanent deformation than annealed material. (M.C.G.)

• Research Organization: Atomics International. Div. of North American Aviation, Inc., Canoga Park, Calif.
• NSA Number: NSA-15-030261
• OSTI ID: 4840853
• Report Number(s): NAA-SR-Memo-6332
• Country of Publication: United States
• Language: English

Related Subjects

ANNEALING
CANNING
DEFORMATION
EXPANSION
FRICTION
GRAPHITE MODERATOR
HEAT TREATMENTS
HIGH TEMPERATURE
HNPF
LEAKS
LIQUID METAL COOLANT
POWER PLANTS
REACTOR TECHNOLOGY
REACTORS
SODIUM
STAINLESS STEELS
TESTING
THERMAL STRESSES
TRAPS
ZIRCONIUM