Heat transfer experiments simulating a failure of the inlet piping to a K reactor process tube

Technical Report · Fri Jan 20 04:00:00 EST 1961

DOI:https://doi.org/10.2172/10146505· OSTI ID:10146505

Waters, E D; Fitzsimmons, D E

Reported herein are the results of laboratory heat transfer experiments. These experiments were conducted to investigate fuel element temperatures which could result from coolant flow loss following a failure of the inlet piping to a process tube at a K reactor. The failure of the inlet coolant piping between the front header and the process tube on a reactor would cut off the flow of cooling water to the fuel elements but should immediately initiate a reactor scram by causing a low trip on the Panellit pressure monitor. However, the reactor power reduction would not be immediate nor absolute and would be dictated by the time required to insert the emergency control rods and by post-scram delayed fission and fission product decay heating. The only means of heat removal from the affected tube and fuel elements during the post-scram period would be by reverse flow of hot water from the rear header. The objective of the subject experiments was to determine what rear header pressures would be required to achieve adequate cooling of a K reactor fuel assembly during such a post-scram period. Such information is of value in updating of reactor hazards evaluation reports. Experimental studies were previously reported concerning failure of a front hydraulic fitting on a C reactor `Operational-charge-discharged` geometry. Preliminary experiments were also conducted for a standard K reactor coolant inlet geometry, but failure of the test section caused postponement of the program. The previous studies did indicate appreciable differences in the rear header pressures required for adequate post-scram cooling of the two assemblies and thus the C reactor data could not be applied to K reactor.

• Research Organization: General Electric Co., Richland, WA (United States). Hanford Atomic Products Operation
• Sponsoring Organization: USDOE, Washington, DC (United States)
• DOE Contract Number: AC06-76RL01830
• OSTI ID: 10146505
• Report Number(s): HW--68214; ON: DE94010872
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
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220900
CONTROL ELEMENTS
FAILURES
FUEL ELEMENTS
HANFORD PRODUCTION REACTORS
HEAT TRANSFER
HYDRAULICS
LOSS OF COOLANT
REACTOR CHANNELS
REACTOR SAFETY
RESEARCH
TEST
TRAINING
PRODUCTION
IRRADIATION
MATERIALS TESTING REACTORS
TESTING