What is Neutron flux: Definition and 27 Discussions
The neutron flux, φ, is a scalar quantity used in nuclear physics and nuclear reactor physics. It is the total length travelled by all free neutrons per unit time and volume. Equivalently, it can be defined as the number of neutrons travelling through a small sphere of radius
R
{\displaystyle R}
in a time interval, divided by
π
R
2
{\displaystyle \pi R^{2}}
(the cross section of the sphere) and by the time interval. The usual unit is cm−2s−1 (neutrons per centimeter squared per second).
The neutron fluence is defined as the neutron flux integrated over a certain time period, so its usual unit is cm−2 (neutrons per centimeter squared).
Hi,
I am sorry if my question seems a bit basic but I find it confusing to understand the differences between the angular neutron flux and the neutron current vector.
I read the definitions from multiple textbooks (Lamarsh, Stacey, Duderstadt) but my idea is that: despite the fact that the...
Why is it that when integrating over all angles, integration is over the solid angle omega composed of theta and phi but the vector angle within flux, phi( r, E, omega), is resolved into the cartesian coordinates by cos(theta)sin(phi), sin(theta)sin(phi) and cos(phi) which is essentially the dot...
Hi,
The figure given below was taken from "Nuclear Systems" by TODREAS and KAZIMI. it shows the effect of neutron reflector on the thermal neutron flux radial distribution.
Is this correct? Because it does not show the extrapolated distance. It seems to me that the reflector, somehow...
Homework Statement
"The 2200 m/s flux in an ordinary water reactor is 1.5*10^13 neutrons/cm^2*s. At what rate are the thermal neutrons absorbed by the water?"
Homework Equations
(unsure)
The Attempt at a Solution
I know that absorption of a thermal neutron (a neutron in thermal equilibrium)...
Hello everyone
I am trying to obtain the integral neutron flux based on the results obtained with MCNP (neutron spectrum calculation) for each energy bin (51 neutron energy bins). I have seen in many papers the calculation of the differential neutron flux multiplying the neutron flux results of...
Hi,
Below are the neutron flux spectra of a nuclear reactor. In the first spectrum, y-axis is differential flux and in the second spectrum, y-axis is flux (created by multiplying differential flux by energy in MeV). As far as I have seen so far, differential flux is used commonly. I am just...
how to convert MNCP F4 tally to neutron flux (n/cm2s). and how to calculate average neutron produced per fission from given nuclear fuel mixture (eg. mixture of 80.8% enriched U-235 and U-238 as fuel material).
In PWR, fast neutron produced from fission in fuel has been moderated into thermal neutron by the a series of collisiion with coolant,i.e. H2O. So the multi-group neutron flux in coolant and fuel pin has much diffenrce, e.g. the relative higher fast neutron in fuel pin and relative higher...
Homework Statement
reactor consists of nested spheres (sorry, this is my first time posting):
Sphere 1 --> Lead target with radius of R0 (Target-region)
Sphere 2 --> Actinide Fuel surrounding lead target; has outer radius of R1 (A-region)
Sphere 3 --> Reflector surrounding the fuel with outer...
Hi everyone,
I am supposed to derive the neutron flux equation provided for region A of my reactor. Just wondering if anyone can help me out since I stuck on the derivation for [1/r - 1/R2]; S/4πD aspect is very similar to a solving the constant for a point source spherical reactor
Here is my...
I remember reading that in addition to the ITER, that they're also building a facility to study the intense neutron flux that is expected from the reactor in order to study how to better protect materials from the flux. Anyone remember the name of the facility? I'm having trouble finding it.
I have a liquid scintillator, the electronics system, and a beam of neutrons from a reactor. How do I calculate the neutron flux? Cold anybody give me a formula or equation? Not this equation: phi = n X v, where phi - flux, n = number of neutrons per cm3, v = velocity of the neutrons, because I...
Homework Statement
A beam of thermal neutrons (10^12 neutrons per cm^2 second) strikes a 1cm thick water target normal to its surface. The target is a round disk with diameter 20cm. Find the exposure rate (R/second) 100cm beyond the water target (the middle of the disk) from only the...
I was reading that the ITER will take advantage of the energetic neutron flux by using them and lithium 6 to breed tritium for more fuel, and to use multiple heat exchangers to grab their energy. And just to make sure that I'm reading it right, the main problem with the neutrons will be their...
when we calculate the neutron flux in finite medium using sn method for steady state neutron transport equation, it gives us some numbers up to 1. I am sure its not the real flux, can someone explain how we can calculte the real flux using sn method.
I have to figure out how to prove that the neutron flux for a point source is given by ø=\frac{S}{4πr^2}.
I can get this type of solution, but I have an e^(-r/L) in the numerator. I'm assuming I'm missing some theory somewhere as apparently this is the solution for a point source in an...
Hello People
I need help with the following assignment:
It states:
Consider an ideal moderator with zero absorption cross section, Ʃa = 0, and a diffusion coefficient, D, which has a spherical shape with an extrapolated radius, R. If neutron sources emitting S neutrons/cm3sec are distributed...
Assumptions
1) a=absorption
2) f=fission
3) ∅=neutron flux
4) time independent
5) group 1, fast neutrons
6) group 2, thermal neutrons
7) All fission neutron are boring in fast group
8) All neutrons created by thermal group, thus vƩf2 exists vƩf2 does not
9) Down scattering occurs but up...
Hello people, I think the problem what I have is well known in experimental physics. Hope, somebody can helps me, thanks in advance. Here is:
There is a source of neutrons with a flux of λ neutrons per minute, which you do not know and want to estimate. You open the detector for one minute...
Deaar all
good morning
I am very interested to the flux in a slab of extrapolated thickness a, containing distributed sources of neutron. A I have an example in which the source is given as s(x)=S(x+a/2) where S is a constant and x distance from the center of the slab.
You mentioned in one...
What is measured in a fast reactor for power calculation : fast neutron flux or overall ( fast + thermal) neutron flux ?
My doubt is :
The fission chambers used for measuring neutron flux undergo chemical reactions due to which type of neutron : fast or thermal ?
Just over half-way down this page:-
http://www.embedded.com/columns/technicalinsights/220301380?cid=RSSfeed_embedded_news
it says:-
"These sunspot counts show variation on a cycle of roughly 11 years. By overlaying Oulu neutron measurements with sunspot count data since 1964, we see that...
well, that's the heading of a project i am doing...i need some help on neutron detectors...
how they are dectected and what is the probable graph of count rate vs, pulse height that i might get for a nuclar fission reaction of U-235...
can anyone help?
In the text I use for class, the examples and derivations for functions showing the neutron flux at some point, are all about sources within infinite mediums. Now I have a probelm where I must show that neutron flux, for a point source within a finite sphere, is found by the following equation...
Hello. I'm a little unsure of how to proceed on this problem... Here it is:
A 3-Gev proton flux is monitored by measuring 24Na activity induced in 25 microm (6.85 microg/cm2) aluminum foil via 27Al(p, 3pn)24Na reaction (for 3 Gev protons, cross section = 9.1 mb). Exactly 15 hr after the end...