Comparing Geant4 hadronic models for the WENDI-II rem meter response function

Measurement of Ambient Dose Equivalent in Compact Proton Therapy using In-house Neutron Moderator-based Poly Allyl Diglycol Carbonate

Purpose

The high-energy proton produces the unwanted dose contribution from the secondary neutron. The main purpose of this study is to report the validation results of in-house neutron moderator based on poly allyl diglycol carbonate (CR-39) detector, Chulalongkorn University Neutron Moderator (CUMOD) through the ambient dose equivalent, H*(10) measurement.

Materials and Methods

The Particle and Heavy Ion Transport code System (PHITS) Monte Carlo code was used to simulate the neutron response function. The CUMOD was calibrated with 241AmBe source calibrator in the range of 100-1000 μSv. The variation of neutron fields was generated employing different proton treatment plans covering most of the clinical scenarios. The ambient dose equivalents, H*(10), evaluated employing CUMOD were compared to those obtained with WENDI-II dosimeter.

Results

The linear relationship between CUMOD and WENDI-II responses showed an R2 value close to 1. The H*(10) per Gy delivered dose was in the range of 22-105 μSv for a 10 cm × 10 cm field.

Conclusion

The in-house CUMOD neutron moderator can expand the neutron detection dose range of CR-39 detector for ambient dose equivalent. The advantage of CUMODs is its capability to evaluate H*(10) in various positions simultaneously.

Comparison of PHITS2.88 and MCNP6.1 for the characterization of a LUPIN-II neutron area monitor

The use of different nuclear data libraries and physics models can be a source of discrepancies in neutron transport simulation. Different Monte Carlo simulation toolkits can be used to characterize neutron monitors, these codes usually employ by default different nuclear data libraries and physics models. This work presents, for the first time, a comparison of MCNP and PHITS for the characterization of a LUPIN-II neutron rem-meter. The most significant discrepancies between the codes have been found around 100 MeV.

Determination of the uncertainties associated to the use of different nuclear data libraries in the analysis of extended-range rem-meters

In proton therapy centers stray neutron radiation of up to 230-250 MeV is yielded by (p, Xn) nuclear reactions. To monitor ambient dose in such facilities, extended-range rem-meters are needed. The aim of this project was to characterize the response of two extended-range rem-meters, WENDI-II and LUPIN-II, by Monte Carlo methods, for energies ranging from 10^-9 MeV up to 230 MeV. Different nuclear data libraries (ENDF/B-VII.1, TENDL2017, TENDL2019, JEFF-3.3) have been used, determining the uncertainties associated with the application of the libraries in the calculation of the response functions of both monitors. The differences found are very significant at energies around 150-200 MeV. This is an issue for predicting by Monte Carlo methods the response of these instruments at high energies. The results point to the necessity of testing experimentally the response of rem-meters at 150 MeV-200 MeV neutron beams.

Comparison of FANT results using the ENDF/B-VII.1, JEFF-3.3 and TENDL2017 nuclear data libraries

FANT (Fuente Ampliada de Neutrones Térmicos; in Spanish) is a thermal neutron irradiation facility with an extended and very uniform irradiation area, that has been developed by the Neutron Measurements Laboratory of the Energy Engineering Department at Universidad Politecnica de Madrid (LMN-UPM). This device is a parallelepiped box made of high-density polyethylene (HDPE), moderator material, that uses an A95241m/B49e neutron source of 111 GBq nominal activity for irradiating materials. The facility design was previously optimized, and the neutron spectra were estimated by extensive calculations with the MCNP6.1 code and carrying out experimental measurements (Bedogni et al., 2017). The facility takes advantage of the scattering reactions of neutrons with the HDPE surfaces of the chamber, where the moderation process is effective, achieving relevant thermal neutron fluence rates. The main goal of this work has been to simulate and analyse the FANT system by Monte Carlo methods using the MCNP6.1 code, employing 3 different nuclear data libraries: ENDF/B-VII.1, JEFF-3.3 and TENDL 2017. The transport of thermal neutrons in HDPE, E < 1eV, has been calculated in all the cases taking into account the thermal S (α,β) treatment. The results achieved in this work have been compared with those previously obtained in the former development of FANT, using the MCNP6.1 code with the ENDF/B-VII.1 nuclear data, and experimental measurements. These results have shown that the JEFF-3.3 nuclear data library is the nuclear data library that provides of the best matching between the MCNP computational results, and the experimental data collected at FANT. Hence, the JEFF-3.3 nuclear data library seems to be the most correct library to design and benchmark thermal neutron activation devices.