Calculational Assessment of Critical Experiments with Mixed Oxide Fuel Pin Arrays Moderated by Organic Solution

Conference · Sun Jun 07 04:00:00 EDT 1987

OSTI ID:6589340

Smolen, G. R. ^[1]

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1. Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)

Critical experiments have been conducted with organic-moderated mixed oxide (MOX) fuel pin assemblies at the Pacific Northwest Laboratory (PNL) Critical Mass Laboratory (CML). These experiments are part of a joint exchange program between the United States Department of Energy (USDOE) and the Power Reactor and Nuclear Fuel Development Corporation (PNC) of Japan in the area of criticality data development. The purpose of these experiments is to benchmark computer codes and cross-section libraries and to assess the reactivity difference between systems moderated by water and those moderated by an organic solution. Past studies have indicated that some organic mixtures may be better moderators than water. This topic is of particular importance to the criticality safety of fuel processing plants where fissile material is dissolved in organic solutions during the solvent extraction process. In the past, it has been assumed that the codes and libraries benchmarked with water-moderated experiments were adequate when performing design and licensing studies of organic-moderated systems. Calculations presented in this paper indicated that the SCALE code system and the 27-energy-group cross-section accurately compute k-effectives for organic moderated MOX fuel-pin assemblies. Furthermore, the reactivity of an organic solution with a 32-vol-% TBP/68-vol-% NPH mixture in a heterogeneous configuration is the same, for practical purposes, as water.

• Research Organization: Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)
• Sponsoring Organization: USDOE National Nuclear Security Administration (NNSA), Nuclear Criticality Safety Program (NCSP)
• DOE Contract Number: AC05-84OR21400
• OSTI ID: 6589340
• Report Number(s): CONF-870601--1; ON: DE86014691
• Country of Publication: United States
• Language: English

Related Subjects

050800* -- Nuclear Fuels-- Spent Fuels Reprocessing
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
22 GENERAL STUDIES OF NUCLEAR REACTORS
CALCULATION METHODS
COMPARATIVE EVALUATIONS
COMPUTER CODES
CRITICALITY
Critical Mass Laboratory (CML)
ENERGY SOURCES
FUEL ELEMENTS
FUEL PINS
FUEL REPROCESSING PLANTS
FUELS
HYDROGEN COMPOUNDS
MATERIALS
MIXED OXIDE FUELS
MODERATORS
NUCLEAR FACILITIES
NUCLEAR FUELS
Nuclear Criticality Safety Program (NCSP)
ORGANIC MODERATORS
ORGANIC SOLVENTS
OXYGEN COMPOUNDS
Organic-Moderated Mixed Oxide (MOX)
Power Reactor and Nuclear Fuel Development Corporation (PNC)
REACTIVITY
REACTOR COMPONENTS
REACTOR MATERIALS
SCALE
SOLID FUELS
SOLVENTS
SPENT FUELS
WATER