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The MTR-Phoenix Fuel Experiment: Critical Test and Burnup Result

Technical Report ·
DOI:https://doi.org/10.2172/4703336· OSTI ID:4703336
 [1];  [1]
  1. Battelle Pacific Northwest Labs., Richland, WA (United States)
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The MTR-Phoenix Fuel Experiment was a burnup experiment in the Materials Testing Reactor (MTR) with a plutonium-aluminum alloy fueled core, following a comprehensive set of critical experiments. This experiment utilized Phoenix fuel, a mixture of plutonium isotopes with a relatively high content of 240Pu. The work described in this report includes the loading of the MTR to critical with the Phoenix fuel core; the zero power measurements of shim rod and regulating rod worths, power distributions throughout the core, kinetic parameters, and reactivity worths of loss of coolant and loss of fuel plates; and the variation of some of the above parameters as the burnup of the core proceeded. The burnup history of the MTR-Phoenix core is also presented. The results are compared, where possible, with results obtained in the PRCF-Phoenix experiment and with calculations.
Research Organization:
Battelle Pacific Northwest Labs., Richland, WA (United States)
Sponsoring Organization:
USDOE National Nuclear Security Administration (NNSA), Nuclear Criticality Safety Program (NCSP); US Atomic Energy Commission (AEC)
DOE Contract Number:
AT(45-1)-1830
NSA Number:
NSA-25-048298
OSTI ID:
4703336
Report Number(s):
BNWL-1593
Country of Publication:
United States
Language:
English

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Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
73 NUCLEAR PHYSICS AND RADIATION PHYSICS
BURNUP
Burnup
CRITICALITY
Criticality
Experiment
FUEL ELEMENTS
Kinetic Parameters
MTR
MTR-Phoenix Core
Materials Testing Reactor
N38510* --Research & Test Reactors & Critical Assemblies-- Kinetics & Dynamics
Nuclear Criticality Safety Program (NCSP)
PLUTONIUM ISOTOPES
PRCF
Phoenix Fuel
Phoenix Fuel Experiment
REACTIVITY WORTHS
REACTOR CORE
Shim Rod