What Is a Neutron Beam?

A neutron beam is a physical quantity that describes the neutron flux.

Neutron beam overview

Hospital neutron irradiator is the world's first miniature neutron source reactor built specifically for boron neutron capture therapy built in China. It includes a core and three neutron beam channels: thermal neutron beam channels, Thermal neutron beam channels and experimental neutron beam channels. The first two beam channels are used for boron neutron capture treatment in different conditions; the experimental neutron beam channels are used for on-line measurement of blood boron concentration. Before using the beam channel to conduct research and clinical application of boron neutron capture therapy, the neutron spectrum of the beam channel needs to be calculated and experimentally measured to provide theoretical and experimental data for subsequent cell experiments, animal experiments, and clinical research. This article will use MCNP to build a hospital neutron irradiator model to obtain the calculated energy spectrum.

Neutron beam method

1.MCNP program simulates hospital neutron irradiator

The center of the active area of the core is used as the origin, and the size and material parameters of each part are determined according to the design drawing. Irregular parts are treated with a uniform method. During calculation, the central control rod is located at the critical rod position at full power, about 4.3cm from the origin; the auxiliary control rod is raised outside the reactor, and the core normalized nuclear power is 30kW.

2.Measurement by gold foil activation method

For the absolute neutron flux density, two pieces of Au foil and Mn foil were selected. One of them was placed in a cadmium box and the other was placed in an aluminum box. Irradiation was performed at the center of the beam exit. Absolute flux density can be calculated with the absolute activity, where: D ₀ is the out-of-stack activity of gold detection foil; F _c is the attenuation factor of neutrons on cadmium in cadmium; R _Au is cadmium of gold detection foil Ratio; R _Mn is the cadmium ratio of the manganese detection foil; N _m Is the total number of nuclei of the gold detection foil; is the modified g factor of the cross section of the gold detection foil that deviates from 1 / v ; is the inverse 0 cross section of gold and neutrons with a velocity of 2200m / s; G _t is the self-screen factor of the gold detection foil itself.

3. Neutron energy spectrum measurement

The neutron spectra of 37, 172, and 642 groups at the exit were calculated using MCNP. Finally, the neutron spectra of 642 groups were selected as the input spectrum. Irradiate each metal activated foil under different flux conditions at the center position, and measure each foil with a high-purity germanium probe. Taking In as an example, the measured activity is 4.1 × 10B _q , and the stack saturation is obtained from the cooling time and irradiation time. The activity is 6.78 × 10B _q , and the number of nuclei is 3.63 × 10 from the mass calculation. The single-core saturation activity is 1.87 × 10B _q . When converted to full power, the single-core saturation activity is 3.73 × 10B _q . The MCNP calculation spectrum was used as the input spectrum of the SAND-method to resolve the spectrum. The MCNP calculation result was used as the initial spectrum, and iterated through the SAND-program three times.

Neutron beam discussion

1) At full power operation, the thermal neutron flux density at the center of the thermal neutron channel exit is _{2.038 × 10 cm} · _s , which meets the design requirements of _{> 1.0 × 10 cm} · _s , and can be used as BNCT Neutron source of technology.

2) From the results of horizontal and vertical neutron flux density, r <3cm, the thermal neutron flux density is (1.99 ± 0.15) × 10cm · s, the change is small; 3cm r 6cm, the thermal neutron flux density is (1.58 ± 0.74) × _{10 cm} · _s , which varies greatly. Therefore, the neutron flux density at the center r <3cm and the neutron flux> 6cm should be used in the treatment to reduce the neutron flux density rapidly, reducing unnecessary doses.

3) It can be seen from the energy spectrum results that after the graphite moderating layer, the neutron beam is sufficiently slowed. Thermal neutrons account for more than 95% of total neutrons, and fast neutrons account for about 1%. They can be used as BNCT. Neutron source. The larger error of the energy spectrum calculation value in the higher energy range is caused by the small fast neutron flux density. More types of detectors can be used to reduce the error in SAND-spectrum resolution. ^[3]