Search Results - "Mai Nguyễn Trọng Nhân"

Propagation of radiation source uncertainties in spent fuel cask shielding calculations by Ebiwonjumi, Bamidele, Trong Mai, Nhan Nguyen, Lee, Hyun Chul, Lee, Deokjung

“…The propagation of radiation source uncertainties in spent nuclear fuel (SNF) cask shielding calculations is presented in this paper. The uncertainty…”
Get full text

EFFECTS OF COMPONENT AND THICKNESS OF METAL LAYER ON RESPONSE FUNCTIONS OF BONNER SPHERE EXTENDED SPECTROMETER USING MCNP CALCULATION by Mai Nguyễn Trọng Nhân, Trịnh Thị Tú Anh

“…The response functions of the Bonner Sphere Extended spectrometer were calculated using the MCNP program. For incident neutrons above 10MeV, Tungsten was an…”
Get full text

On-the-fly energy release per fission model in STREAM with explicit neutron and photon heating by Mai, Nhan Nguyen Trong, Lee, Woonghee, Kim, Kyeongwon, Ebiwonjumi, Bamidele, Kim, Wonkyeong, Lee, Deokjung

“…The on-the-fly energy release per fission (OTFK) model is implemented in STREAM to continuously update the Kappa values during the depletion calculation. The…”
Get full text

EFFECTS OF COMPONENT AND THICKNESS OF METAL LAYER ON RESPONSE FUNCTIONS OF BONNER SPHERE EXTENDED SPECTROMETER USING MCNP CALCULATION by Nhân, Mai Nguyễn Trọng, Anh, Trịnh Thị Tú

“…Hàm đáp ứng của phổ kế Bonner Sphere Extended (BSE) được tính toán dựa trên phần mềm mô phỏng MCNP. Trên 10MeV, Wolfram là vật liệu tốt nhất vì lớp lót Wolfram…”
Get full text

Modelling atomic relaxation and bremsstrahlung in the deterministic code STREAM by Mai, Nhan Nguyen Trong, Kim, Kyeongwon, Lee, Deokjung

“…STREAM, developed by the Computational Reactor Physics and Experiment laboratory (CORE) of the Ulsan National Institute of Science and Technology (UNIST), is a…”
Get full text

Extension of Monte Carlo code MCS to spent fuel cask shielding analysis by Mai, Nhan Nguyen Trong, Zhang, Peng, Lemaire, Matthieu, Ebiwonjumi, Bamidele, Kim, Wonkyeong, Lee, Hyunsuk, Lee, Deokjung

“…SUMMARY The shielding analysis of the spent fuel dry storage cask TN‐32 is carried out using the continuous‐energy Monte Carlo neutron‐ and photon‐transport…”
Get full text

Analysis of several VERA benchmark problems with the photon transport capability of STREAM by Mai, Nhan Nguyen Trong, Kim, Kyeongwon, Lemaire, Matthieu, Nguyen, Tung Dong Cao, Lee, Woonghee, Lee, Deokjung

“…STREAM - a lattice transport calculation code with method of characteristics for the purpose of light water reactor analysis - has been developed by the…”
Get full text

UNCERTAINTY ANALYSIS OF SUB-EXERCISES IN UAM-SFR BENCHMARK WITH THE MCS CODE by Jo, Yunki, Dos, Vutheam, Mai, Nhan Nguyen Trong, Lee, Hyunsuk, Lee, Deokjung

“…Uncertainty analysis in Modelling (UAM) for Design, Operation and Safety Analysis of Sodium-cooled Fast Reactors (SFRs) has been formed by OECD/NEA to assess…”
Get full text