Measurement of {sup 235}U content and flow of UF{sub 6} using delayed neutrons or gamma rays following induced fission

Abstract

SummaryFeasibility experiments conducted at Pacific Northwest National Laboratory demonstrate that either delayed neutrons or energetic gamma rays from short-lived fission products can be used to monitor the blending of UF6 gas streams. A 252Cf neutron source was used to induce 235U fission in a sample, and delayed neutrons and gamma rays were measured after the sample moved "downstream." The experiments used a U02 powder that was transported down the pipe to simulate the flowing u F 6 gas. Computer modeling and analytic calculation extended the test results to a flowing u F 6 gas system. Neutron or gamma-ray measurements made at two downstream positions can be used to indicate both the 235U content and UF6 flow rate. Both the neutron and gamma-ray techniques have the benefits of simplicity and long-term reliability, combined with adequate sensitivity for lowintrusion monitoring of the blending process. Alternatively, measuring the neutron emission rate from (a, n) reactions in the UF6 provides an approximate measure of the 235U content without using a neutron source to induce fission.The implementation for blend monitoring will place a 252Cf neutron source and moderator (polyethylene) adjacent to the pipes carrying the UF6 gas. Neutron detectors located downstream from the neutron source will measure the delayed neutrons emitted as the result of the induced fission. Measurements made at two different downstream locations will be used to establish the flow rate as well as measure 235U content. Additional detectors located upstream from the neutron source will provide a measure of the neutron background for subtraction during data analysis. At a minimum, the installation will include monitoring of the pipes carrying the highly enriched gas to the blending point.

Neutron experiments at Portsmouth for measuring flow and {sup 235}U content in UF{sub 6} gas

“…Computer simulations using the code MCNP (Briesmeister 1993) were performed after the experiments at Portsmouth were completed to gain insight on design parameters that could explain observed results or indicate improvements for future experiments. These computer simulations supplement those performed before the experiment and described in previous reports (Stromswold et al 1996a and1996b). The simulations did not attempt to reproduce the detailed geometry of the experimental setup.…”

Neutron methods for measuring {sup 235}U content in UF{sub 6} gas

“…One method monitors the delayed fission neutrons emitted downstream in the flowing gas following the induced fission. This method was described in a previous report (Stromswold et al 1996). A second method, described in this report, measures the attenuation of thermal neutrons crossing a pipe containing m 6 .…”

“…The MCNP calculations presented in a previous report (Stromswold et al 1996) give the number of expected induced fissions in the m 6 for the Russian blending conditions per source neutron emitted from a Z2Cf source. Adjusting those results to the gas pressure, temperature, and pipe diameter for Portsmouth gives an estimate of the induced fission rate for a source located adjacent to the pipe.…”