Results 181 to 190 of about 16,330 (259)

Urinary phthalate metabolites associated with increased prevalence of gallstone disease in U.S. adults: data from the NHANES study. [PDF]

BMC Public Health
Jia F, Chang Y, Li Y, Li F, Chen X, Liu X, Li W, Cui J.   +7 more
europepmc   +1 more source

A comparison and evaluation of statistical methods for mediation analysis with mixtures of environmental exposures. [PDF]

BMC Med Res Methodol
McGrath S, Wang Y, Lin YT, Meeker JD, Park SK, Warren JL, Mukherjee B.   +6 more
europepmc   +1 more source

Unveiling the radiation shielding efficacy of diorite, granodiorite, tonalite, and granite: experimental and simulation study. [PDF]

Sci Rep
Elsafi M, El-Nahal MA, Alawy MK, Nabil IM.   +3 more
europepmc   +1 more source

Development of PMMA based polymer composite incorporating WO₃ for gamma radiation shielding using synthesis and Monte Carlo simulation. [PDF]

Sci Rep
Mokhtari Dorostkar M, Abdi Saray A.
europepmc   +1 more source

The association between environmental endocrine-disrupting chemicals and allergic disorders in children: A comprehensive systematic review and meta-analysis. [PDF]

J Res Med Sci
Tavousi N, Yousefvand A, Maghami M, Kelishadi R, Amini N.   +4 more
europepmc   +1 more source

Association between phthalates exposure and urine flow rate. [PDF]

Sci Rep
Zhang S, Tang H, Zhou M.
europepmc   +1 more source

High-Temperature MCNP Cross Sections

, 1987
openaire   +2 more sources

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MCNP Output Data Analysis with ROOT (MODAR)

Computer Physics Communications, 2010
MCNP Output Data Analysis with ROOT (MODAR) is a tool based on CERN's ROOT software. MODAR has been designed to handle time–energy data issued by MCNP simulations of neutron inspection devices using the associated particle technique. MODAR exploits ROOT's Graphical User Interface and functionalities to visualize and process MCNP simulation results in a
C Carasco
exaly   +4 more sources

A new methodology to estimate stochastic uncertainty of MCNP-predicted isotope concentrations in nuclear fuel burnup simulations

Annals of Nuclear Energy, 2021
The Monte Carlo N-particle radiation transport code (MCNP) is used widely in nuclear fuel burnup simulation (FBS). In the FBS, MCNP predicts various neutron reaction rates and the associated stochastic relative error (SRE).
Sunil S Chirayath
exaly   +2 more sources

Evaluation of photon shielding properties for new refractory tantalum-rich sulfides Ta9(XS3)2 alloys: A study with the MCNP-5

Annals of Nuclear Energy, 2023
M Kh Hamad
exaly   +2 more sources