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

Conference · Sun Jun 07 04:00:00 EDT 1987 · Transactions of the American Nuclear Society

OSTI ID:5460479

Smolen, G. R. ^[1]; Funabashi, H. ^[1]

---

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

Critical experiments have been conducted with organically moderated mixed-oxide (MOX) fuel pin assemblies at the Pacific Northwest Lab. Critical Mass Lab. These experiments are part of a joint exchange program between the US Dept. of Energy and the Power Reactor and Nuclear Fuel Development Corp. 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 organically moderated systems. Calculations presented in this paper indicated that the Scale code system and the 27-energy-group cross-section library accurately compute k_eff for organically 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)
• OSTI ID: 5460479
• Report Number(s): CONF-870601-
• Conference Information: Journal Name: Transactions of the American Nuclear Society Journal Volume: 54
• Country of Publication: United States
• Language: English

Related Subjects

050800* -- Nuclear Fuels-- Spent Fuels Reprocessing
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
42 ENGINEERING
420203 -- Engineering-- Handling Equipment & Procedures
97 MATHEMATICS AND COMPUTING
BATTELLE PACIFIC NORTHWEST LABORATORIES
BENCHMARKS
Benchmark Computer Codes
COMPARATIVE EVALUATIONS
COMPUTER CODES
CRITICALITY
CROSS SECTIONS
Cross-Section Libraries
EXTRACTION
FUEL ASSEMBLIES
FUEL ELEMENTS
FUEL PINS
FUEL REPROCESSING PLANTS
FUELS
HYDROGEN COMPOUNDS
MIXED OXIDE FUELS
MODERATORS
MULTIPLICATION FACTORS
Mixed-Oxide (MOX) Fuel Pin Assemblies
NATIONAL ORGANIZATIONS
NUCLEAR FACILITIES
Nuclear Criticality Safety Program (NCSP)
ORGANIC COMPOUNDS
OXYGEN COMPOUNDS
REACTIVITY
REACTOR COMPONENTS
Reactivity Difference Between Systems
SAFETY
SEPARATION PROCESSES
SOLID FUELS
SOLVENT EXTRACTION
US DOE
US ERDA
US ORGANIZATIONS
WATER