Abstract
The SAMMY code system is mainly used in nuclear data evaluations for incident neutrons in the resolved resonance region (RRR), however, built-in capabilities also allow the code to describe the resonance structure produced by other incident particles, including charged particles. (α,n) data provide fundamental information that underpins nuclear modeling and simulation software, such as ORIGEN and SOURCES4C, used for the analysis of neutron emission and definition of source emission processes. The goal of this work is to carry out evaluations of charged-particle-induced reaction cross sections in the RRR. The SAMMY code was recently used in this regard to generate a Reich-Moore parameterization of the available 17,18 O(α,n) experimental cross sections in order to estimate the uncertainty in the neutron generation rates for uranium oxide fuel types. This paper provides a brief description of the SAMMY evaluation procedure for the treatment of 17,18 O(α,n) reaction cross sections. The results are used to generate neutron source rates for a plutonium oxide matrix.
Highlights
Evaluated nuclear data describing charged-particle interactions are essential for a large range of applications
The SAMMY code [5] was recently used to generate a Reich-Moore parameterization of the 17,18O(α,n) available experimental cross sections in order to estimate the uncertainty in the neutron generation rates for uranium oxide fuel types [6]
The objective of this paper is to present a brief description of the SAMMY method applied to compute (α,n) cross sections in the resolved resonance region (RRR) applied to isotopes essentially important to application as 17,18O
Summary
Evaluated nuclear data describing charged-particle interactions are essential for a large range of applications. Α-particle induced reactions on light nuclei are of particular importance to calculate neutron emission via (α,n) process as in neutron-based verification measurements of unirradiated uranium and plutonium nuclear materials in a non-metal matrix. In irradiated nuclear fuels that achieve a moderate to high burnup, the neutron source is typically dominated by 242,244Cm spontaneous fission. In low-burnup fuels such as those encountered in weapons material production, the (α,n) neutron processes can represent a large component of the total neutron source and are important to neutron measurement techniques. The ultimate need is to estimate the accuracy of applied quantities, e.g. neutron source rates, due to the uncertainties in basic data
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