Radiation dose distribution of liquid fueled thorium molten salt reactor

Li, Chang-Yuan; Xia, Xiaobin; Cai, Jun; Zhang, Guoqing; Wang, Jianhua; Qian, Zhi-Cheng

doi:10.1007/s41365-021-00857-3

Cited by 15 publications

(3 citation statements)

References 28 publications

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“…Accurate estimation of the dose received by astronauts is an essential precondition for successful manned deep-space exploration missions. In addition, with the wide use of nuclear technology in material physics and medical physics [38][39][40][41][42][43][44][45][46], it is also important to be able to calculate the radiation dose received by a typical human in the corresponding radiation environment. Based on the prior simulation results, we calculated the radiation dose that would have been received by astronauts on the Mars surface.…”

Section: Radiation Dose Received By a Typical Human Bodymentioning

confidence: 99%

Studies of the radiation environment on the Mars surface using the Geant4 toolkit

et al. 2022

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The radiation environment on the surface of Mars is a potential threat for future manned exploration missions to this planet. In this study, a simple geometrical model was built for simulating the radiation environment on the Mars surface caused by galactic cosmic rays; the model was built and studied using the Geant4 toolkit. The simulation results were compared with the data reported by a radiation assessment detector (RAD). The simulated spectra of neutrons, photons, protons, $$\alpha$$ α particles, and particle groups $$Z=3$$ Z = 3 –5, $$Z=6$$ Z = 6 –8, $$Z=9$$ Z = 9 –13, and $$Z=14$$ Z = 14 –24 were in a reasonable agreement with the RAD data. However, for deuterons, tritons, and $$\,^{3}{\mathrm{He}}$$ 3 He , the simulations yielded much smaller values than for the corresponding RAD data. In addition, the particles’ spectra within the $$90^{\circ }$$ 90 ∘ zenith angle were also obtained. Based on these spectra, we calculated the radiation dose that would have been received by an average human body on Mars. The distribution of the dose throughout the human body was not uniform. The absorbed and equivalent doses for the brain were the highest among all of the organs, reaching 62.0 ± 1.7 mGy/y and 234.1 ± 8.0 mSv/y, respectively. The average absorbed and equivalent doses for the entire body were approximately 44 mGy/y and 153 mSv/y, respectively. Further analysis revealed that most of the radiation dose was owing to $$\alpha$$ α particles, protons, and heavy ions. We then studied the shielding effect of the Mars soil with respect to the radiation. The body dose decreased significantly with increasing soil depth. At the depth of 1.5 m, the effective dose for the entire body was 17.9 ± 2.4 mSv/y, lower than the dose limit for occupational exposure. At the depth of 3 m, the effective dose to the body was 2.7 ± 1.0 mSv/y, still higher than the accepted dose limit.

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Section: Radiation Dose Received By a Typical Human Bodymentioning

confidence: 99%

Studies of the radiation environment on the Mars surface using the Geant4 toolkit

et al. 2022

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“…Li et al, [7] modelled and calculated the neutron and gamma dose rate distributions around the liquid-fueled thorium molten salt reactor (TMSR-LF) after applying a 60 cm serpentine concrete layer as a radiation shield. Te analysis of the results showed that the maximum thermal neutron fux density outside the reactor biological shield was 1.89× 10 3 cm −2 s −1 , which was below the 1 × 10 5 cm −2 s −1 limit.…”

Section: Introductionmentioning

confidence: 99%

“…Also, serpentine loses its water of crystallization when heated to a very high temperature (above 480 °C) and as such serpentine shields could be subjected to high temperatures without a need for complex and expensive cooling systems for radiation shielding. So far, three deposits have been found in Ghana, at Anum, Peki-Dzake and Golokwati [7,8]. Tey have been investigated and found to be suitable for photon attenuation [9].…”

Section: Introductionmentioning

confidence: 99%

Effectiveness of Serpentine Concrete as Shielding Material for Neutron Source Facility Using Monte Carlo Code

Abrefah

Tuffour-Achampong

Amoah

2023

Science and Technology of Nuclear Installations

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In recent years, much attention has been dedicated to finding techniques to reduce exposure doses. This work examines the effectiveness of using serpentine concrete to shield a neutron source using a 241Am-Be neutron source facility at the National Nuclear Research Institute (NNRI) as a case study. The results obtained for both neutrons and gamma indicate that serpentine concrete provides better shielding as compared to ordinary concrete. At a distance of 100 cm from the Am-Be source, when shielded with serpentine concrete, it was found that personnel will receive an average gamma dose of 4.395.395 ± 0.122 μSv/h while a dose of 10.399 ± 0.083 μSv/h will be received for ordinary concrete shield. The average neutron dose equivalent at 100 cm, for ordinary concrete and serpentine concrete were 32.189 ± 0.277 and 9.276 ± 0.505, respectively. All dose equivalents obtained were also within internationally accepted limits.

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Experimental study on the penetration characteristics of leaking molten salt in the thermal insulation layer of aluminum silicate fiber

et al. 2021

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Radiation dose distribution of liquid fueled thorium molten salt reactor

Cited by 15 publications

References 28 publications

Studies of the radiation environment on the Mars surface using the Geant4 toolkit

Studies of the radiation environment on the Mars surface using the Geant4 toolkit

Effectiveness of Serpentine Concrete as Shielding Material for Neutron Source Facility Using Monte Carlo Code

Experimental study on the penetration characteristics of leaking molten salt in the thermal insulation layer of aluminum silicate fiber

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