Simulation Design of Thermal Neutron Collimators for Neutron Capture Studies at The Dalat Research Reactor

Quyet, Pham Dang; Dien, Nguyen Nhi; Anh, Trịnh Thị Tú; Vũ, Cao Đông

doi:10.3923/ajsr.2020.214.218

Cited by 1 publication

(2 citation statements)

References 6 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…In this study, a conical collimator with 240.3 cm in length is proposed to increase neutron flux compared to a cylindrical collimator, presented in Ref. [5], which is used to simulate the neutron beam with the crystal filters of Al 2 O 3 and Bi. This collimator model is presented in Fig.…”

Section: Methodsmentioning

confidence: 99%

“…A thermal neutron beam of a nuclear reactor can be employed for neutron capture studies and applications such as prompt gamma neutron activation analysis (PGNAA), boron neutron capture therapy (BNCT) [1][2][3][4][5][6], and neutron radiography. To produce a thermal neutron beam from a nuclear reactor, single crystals of Si, Al 2 O 3 , or Bi are often used as neutron filters [5,6]. A high-performance thermal neutron beam for neutron radiography using the Al 2 O 3 sapphire crystal filter has been designed and developed at the PULSTAR reactor [7].…”

Section: Introductionmentioning

confidence: 99%

See 1 more Smart Citation

Design and simulation of a thermal neutron beam for neutron capture studies at the Dalat research reactor

Tu¹,

Dang

Ngoc³

et al. 2023

Comm. in Phys.

View full text Add to dashboard Cite

This paper presents the application of the MCNP5 code to conduct a modification design and simulation of a thermal neutron beam for neutron capture studies at the Dalat Nuclear Research Reactor (DNRR). The designed configuration of the horizontal neutron channel No.2 at the DNRR, which contains a conical collimator of 240.3 cm in length, and neutron filters of crystal Al2O3 and Bi with alternative thickness, were simulated. A pure thermal neutron beam can be obtained at the irradiation position when a composition of crystal filter of 20 cm Al2O3 and 6 cm Bi is employed. The thermal and epithermal neutron fluxes are 1.02´108 n/cm2/s (account for 97.92% of total neutron flux) and 0.22´107 n/cm2/s (account for 2.08% of total neutron flux), respectively.

show abstract

Section: Methodsmentioning

confidence: 99%