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Neutron transport - WikiMili, The Best Wikipedia Reader

Neutron transport - WikiMili, The Best Wikipedia Reader

Neutron transport is the study of the motions and interactions of neutrons with materials. Nuclear scientists and engineers often need to know where neutrons are in an apparatus, what direction they are going, and how quickly they are moving. It is commonly used to determine the behavior of nuclear

Radioactive waste - Wikipedia

Radioactive waste - Wikipedia

Radioactive waste is a type of hazardous waste that contains radioactive material.Radioactive waste is a by-product of various nuclear technology processes. Industries generating radioactive waste include nuclear medicine, nuclear research, nuclear power, manufacturing, construction, coal and rare-earth mining and nuclear weapons reprocessing. Radioactive waste is regulated by government .

Price Increase to Take Place May 1, 2001

Price Increase to Take Place May 1, 2001

The University of Texas, Austin, Texas, contributed these continuous-energy neutron cross section libraries for temperatures 300 to 1365 K developed for use with MCNP. UTXS6 is based on ENDF/B-VI except for zinc and erbium for which the Russian (BROND-2) and the Chinese (CENDL-2) .

Neutron Flux Spectra - Nuclear Power

Neutron Flux Spectra - Nuclear Power

Thermal Reactors. Almost all of the current reactors which have been built to date use thermal neutrons to sustain the chain reaction.. These reactors contain neutron moderator that slows neutrons from fission until their kinetic energy is more or less in thermal equilibrium with the atoms (E < 1 eV) in the system.; Fast Neutron Reactors. Fast reactors contains no neutron moderator and use .

Energies | Free Full-Text | Exploring Stochastic . - MDPI

Energies | Free Full-Text | Exploring Stochastic . - MDPI

In the case of reactor dosimetry the quantities of calculation and validation interest are basically the neutron fluxes and fluence monitors' reaction rates. Details and some illustrative examples on the topic can be found in [60,61,62,63]. For the case of the problems discussed in this paper the main parameter of interest is the fast neutron .

Neutron production evaluation from a ADS target utilizing .

Neutron production evaluation from a ADS target utilizing .

The main purpose of the spallation target in an ADS is to provide the primary neutron flux for driving the fission process in the surrounding subcritical core. . Flux and Dose Rate Evaluation of ITER System Using MCNP5, Brazilian Journal of Physics 40, (2010 . H. Nifenecker et al., Basics of accelerator driven subcritical reactors, Nucl. Instr.

Optimum Utilization of Fission Power with Gas Core Reactors

Optimum Utilization of Fission Power with Gas Core Reactors

Any reuse of this item in excess of fair use or other copyright exemptions requires permission of the copyright holder. Embargo Date: . OPTIMUM UTILIZATION OF FISSION POWER WITH GAS CORE REACTORS By ROBERT NORRING A THESIS PRESENTED TO THE GRADUATE SCHOOL OF THE UNIVERSITY OF FLOR IDA IN PARTIAL FULFILLMENT OF THE REQUIREMENTS FOR THE .

UserGuide - LPSC

UserGuide - LPSC

Between 2 MCNP steps, reaction rates are keep ``constant'' ; a new reaction rate evaluation is performed at the next MCNP run for the next step. In fact, they are not really constant: due to renormalization of the flux to keep constant power, reaction rates used in Bateman equations are in fact, ασφ where α is the flux renormalization .

Neutron production evaluation from a ADS target utilizing .

Neutron production evaluation from a ADS target utilizing .

The main purpose of the spallation target in an ADS is to provide the primary neutron flux for driving the fission process in the surrounding subcritical core. . Flux and Dose Rate Evaluation of ITER System Using MCNP5, Brazilian Journal of Physics 40, (2010 . H. Nifenecker et al., Basics of accelerator driven subcritical reactors, Nucl. Instr.

SINBAD REACTOR, Shielding Benchmark Experiments.

SINBAD REACTOR, Shielding Benchmark Experiments.

- Fast and thermal neutron fluxes, dose rates of neutron and photon radiation at various altitudes from the reactor cover; - Spatial differential energy distributions of neutrons at various distances from the reactor axis (100, 200, 400, 500, 600, 800, 1000 m); - Spatial distribution of neutron radiation dose rate at 50 from 1500 m;

Compact D-D Neutron Source-Driven Subcritical Multiplier .

Compact D-D Neutron Source-Driven Subcritical Multiplier .

Abstract. This work assesses the feasibility of using a small, safe, and inexpensive keff 0.98 subcritical fission assembly [subcritical neutron multiplier (SCM)] to amplify the treatment neutron beam intensity attainable from a compact deuterium-deuterium (D-D) fusion neutron source delivering [approximately]1012 n/s.

PPT – Space Nuclear Power PowerPoint presentation | free .

PPT – Space Nuclear Power PowerPoint presentation | free .

Dose Next to a JIMO-Class Reactor Before Operation. Radiation transport calculations have shown that the radiation dose is below nominal background levels next to the SAFE-400 prior to operation. 12 hour hike up Colorado 14,000 ft peak 0.48 mR, Chest X-ray 40 mR, Average background 360 mR, Occupational dose limit 5 R. Dose rate data

Materials for Advanced Energy Systems and Fission & Fusion .

Materials for Advanced Energy Systems and Fission & Fusion .

Aug 02, 2002 · High Power Density Blanket Design Study for Fusion Reactors (J H Huang et al.) Study on Conceptual Design of Tritium Production Reactor Based on ST (K H He & J H Huang) and other papers; Readership: Researchers in the areas of systems & design, fission & fusion materials, and radiation damage.

Characterization of prompt neutron spectrum of the .

Characterization of prompt neutron spectrum of the .

Aug 15, 2017 · The foil set cover energies from 0.025 eV to 7.2 MeV for a broad spectrum analysis. After irradiation, the sample was cooled using water until the dose rate reached a level that allowed safe handling. For each material, both bare and cadmium covers were used to obtain the thermal and epithermal neutron flux.

How can I calculate the neutron dose by MCNP5?

How can I calculate the neutron dose by MCNP5?

By using the DE and DF cards, you can define this flux-to-dose function. You can find values for that in the MCNP manual or in radioprotection books and detector manuals. Cite

A Critical Review of the Recent Improvements in Minimizing .

A Critical Review of the Recent Improvements in Minimizing .

Equivalent Dose Rate could be calculated as follows: if we consider a gamma-ray intensity of photons/m 2 · s with an energy of MeV, the energy flux will be MeV/m 2 s. The rate of energy deposition per unit volume (in a small volume) will be MeV/m 3 s, with as the linear absorption coefficient ().

A NEW EXPERIMENTAL DESIGN AND METHOD FOR IMPROVED .

A NEW EXPERIMENTAL DESIGN AND METHOD FOR IMPROVED .

If k eff is bigger than 1.0, the reactor is supercritical, and the reactor power level is rising. Oppositely, k eff is less than 1.0, the reactor is subcritical, and the reactor power level is decre a sing [ 4 ]. The effective multiplication factor indicates the change of the reactor power level, dir ectly prop ortional to neutron population.

Full text of

Full text of "DTIC ADA452712: Investigation into the .

Full text of "DTIC ADA452712: Investigation into the Feasibility of Highly Enriched Uranium Detection by External Neutron Stimulation (Expanded Study)" See other formats

Reactor Dosimetry in the 21st Century - World Scientific

Reactor Dosimetry in the 21st Century - World Scientific

Aug 23, 2002 · This book presents the state of the art in reactor dosimetry as applied to nuclear power plants and to high performance research reactors, accelerator-driven systems and spallation sources. The reader will also find the latest advances in computer code development for .

A Critical Review of the Recent Improvements in Minimizing .

A Critical Review of the Recent Improvements in Minimizing .

Equivalent Dose Rate could be calculated as follows: if we consider a gamma-ray intensity of photons/m 2 · s with an energy of MeV, the energy flux will be MeV/m 2 s. The rate of energy deposition per unit volume (in a small volume) will be MeV/m 3 s, with as the linear absorption coefficient ().

Neutrons and Gamma-Ray Dose Calculations in Subcritical .

Neutrons and Gamma-Ray Dose Calculations in Subcritical .

atoms Article Neutrons and Gamma-Ray Dose Calculations in Subcritical Reactor Facility Using MCNP Ned Xoubi Nuclear Engineering Department, King .

Nuclear reactor - Wikipedia

Nuclear reactor - Wikipedia

A nuclear reactor, formerly known as an atomic pile, is a device used to initiate and control a self-sustained nuclear chain reaction.Nuclear reactors are used at nuclear power plants for electricity generation and in nuclear marine propulsion.Heat from nuclear fission is passed to a working fluid (water or gas), which in turn runs through steam turbines.

Neutron Diffusion Theory - Nuclear Power

Neutron Diffusion Theory - Nuclear Power

Neutron Diffusion Theory. In previous section we dealt with the multiplication system and we defined the infinite and finite multiplication factor.This section was about conditions for a stable, self-sustained fission chain reaction and how to maintain such conditions. This problem contains no information about the spatial distribution of neutrons, because it is a point geometry problem.

NEA - Abstract list

NEA - Abstract list

DOSE-SGTR, Iodine Release During Steam Generator Tube Rupture (SGTR) in PWR: ccc-0536: DOSFACTOR-DOE, Dose Rate Conversion Factors for Photon and Electron Exposure: ccc-0276: DOT-3.5, 2-D Neutron Transport, Gamma Transport Program DOT with New Space-Scaling: ccc-0320: DOT-4.2, 2-D Neutron Transport, Gamma Transport with Space Dependent Mesh and .

Shielding of Neutron Radiation

Shielding of Neutron Radiation

Slow down neutrons (the same principle as the neutron moderation).First point can be fulfilled only by material containing light atoms (e.g. hydrogen atoms), such as water, polyethylene, and concrete.The nucleus of a hydrogen nucleus contains only a proton. Since a proton and a neutron have almost identical masses, a neutron scattering on a hydrogen nucleus can give up a great amount of its .

Neutron transport - Wikipedia

Neutron transport - Wikipedia

Neutron transport (also known as neutronics) is the study of the motions and interactions of neutrons with materials. Nuclear scientists and engineers often need to know where neutrons are in an apparatus, what direction they are going, and how quickly they are moving. It is commonly used to determine the behavior of nuclear reactor cores and experimental or industrial neutron beams.

O.Petit, E.Dumonteil, F.X.Hugot, Y.K.Lee, A.Mazzolo, C .

O.Petit, E.Dumonteil, F.X.Hugot, Y.K.Lee, A.Mazzolo, C .

Research reactors . Nuclear Wastes . Transmutation, Hybrid System . Thermonuclear Fusion . Particle Accelerators . Medicine . Industrial plants, . 2. TRIPOLI-4 : Application fields Neutron and gamma dose rate calculations inside PWR reactors OSIRIS : conception of experiments in OSIRIS research reactor Dampierre PWR Keff calculation and ex .

Crystalline Structure, Synthesis, Properties and .

Crystalline Structure, Synthesis, Properties and .

Thus, for example, in the synthesis of PHT long fibers using Li 2 O-K 2 MoO 4 as flux, the raw materials K 2 CO 3, Li 2 CO 3, K 2 MoO 4 and TiO 2 reagent grade were used. Flux and raw materials were mixed and melted into a platinum crucible at 1200 °C for 4 h and then rapidly cooled at room temperature.

Neutron transport - Wikipedia

Neutron transport - Wikipedia

Neutron transport (also known as neutronics) is the study of the motions and interactions of neutrons with materials. Nuclear scientists and engineers often need to know where neutrons are in an apparatus, what direction they are going, and how quickly they are moving. It is commonly used to determine the behavior of nuclear reactor cores and experimental or industrial neutron beams.

Neutrons and Gamma-Ray Dose Calculations in Subcritical .

Neutrons and Gamma-Ray Dose Calculations in Subcritical .

A three-dimensional (3D) Monte Carlo model was developed to calculate the dose rate from neutrons and gamma, using the ANSI/ANS-6.1.1 and the ICRP-74 flux-to-dose conversion factors. Estimation for the dose was conducted across 39 areas located throughout the reactor .