Reaction rate measurements for nuclear reactor analyses and critically data

Abstract

Neutron reaction rate measurements with solid state track recorders (SSTR) and radiometric (RM) neutron dosimeters have been conducted at the Pacific Northwest Laboratory (PNL) Critical Mass Laboratory (CML) as part of the U.S. Department of Energy (DOE) and the Power Reactor and Nuclear Fuel Development Corporation of Japan (PNC) Critically Data Development Program. These reaction rate measurements represent benchmark data that can rigorously test the adequacy of neutron transport calculations performed for nuclear reactor analyses as well as for critically safety assessments. In the 220 series of experiments, fast test reactor (FTR) plutonium fuel pins were assembled in a 0.761 cm square lattice array, which was immersed in an organic moderator. In-fuel and in-moderator dosimetry measurements were conducted near axial midplane. To obtain results at a single axial location, corrections were applied for the spatial variation (axial buckling) of the reaction rates. For the in-fuel measurements, dosimeters were placed between fuel pellets thereby creating gaps in the fuel pin column. Consequently, for these in-fuel measurements, the gap-perturbation effect was measured so that reaction rate data could be corrected to zero fuel gap, i.e. the reaction rates in the unperturbed fuel. Experimental uncertainties range from a low of 2–3 percent for U-238 and Th-232 fission rates to a high of 15–16 percent for U-238 capture rates. The uncertainty in the U-238 (n, Σ) moderator results is dominated by the uncertainty in the neutron self-shielding correction factor, which has been estimated to be approximately 15 percent. Uncertainties in U-235, Np-237 and Pu-239 fission rates range from approximately 4 to 7 percent. These latter uncertainties are larger than uncertainties normally achievable in SSTR neutron dosimetry and reflect the fact that the quality of the SSTR electrodeposits prepared for these isotopes was not completely satisfactory. Least-squares spectral analyses of these data were performed with the FERRET-SAND II computer code. These analyses confirm the general consistency of the experimental data and furnish absolute neutron fluxes with assigned uncertainties.