For those who wonder, this seems to be the content or a video originated from the content of Kloosterman's course AP3631 Nuclear Reactor Kinetics' second chapter: Nuclear Reactor Kinetics at . Hopefully his YT channel will post more often this kind of content. I find it very helpful for nuclear workers who did not have the chance to have a real training and yet are still interested by the physics behind it. Although, this particular chapter seems to be close to most resources that you can find online and particularly Introductory Nuclear Reactor Dynamics, which this course is based on.
I'm a software developer who just happened to catch this video. Two questions 1. How can the fraction of delayed neutrons be calculated in the case of a MSR where part of the fuel will be continously removed from the core? 2. The value of 2.45 neutrons per fission tells me that this is for thermal spectrum reactors, will the equations be the same for fast spectrum reactors as well?
Hey Niklas Paulson, I am not a scholar, but I might be able to answer you. To both of your questions, the answer lies in which kind of reactor you use. In the case of MSR, the nuclear fuel are U238 or Th232 mostly. Then, one must find the nuclear data for those fuels and fortunately they are available, as well as the ones for U235 on IAEA's website if you look up "average number of neutrons emitted per fission" A-6 for nu_t and nu_d. Fractions of delayed neutrons is also present on Table A-7 of the same website. Have a great day.
For those who wonder, this seems to be the content or a video originated from the content of Kloosterman's course AP3631 Nuclear Reactor Kinetics' second chapter: Nuclear Reactor Kinetics at .
Hopefully his YT channel will post more often this kind of content. I find it very helpful for nuclear workers who did not have the chance to have a real training and yet are still interested by the physics behind it. Although, this particular chapter seems to be close to most resources that you can find online and particularly Introductory Nuclear Reactor Dynamics, which this course is based on.
I'm a software developer who just happened to catch this video.
Two questions
1. How can the fraction of delayed neutrons be calculated in the case of a MSR where part of the fuel will be continously removed from the core?
2. The value of 2.45 neutrons per fission tells me that this is for thermal spectrum reactors, will the equations be the same for fast spectrum reactors as well?
Hey Niklas Paulson,
I am not a scholar, but I might be able to answer you.
To both of your questions, the answer lies in which kind of reactor you use. In the case of MSR, the nuclear fuel are U238 or Th232 mostly.
Then, one must find the nuclear data for those fuels and fortunately they are available, as well as the ones for U235 on IAEA's website if you look up "average number of neutrons emitted per fission" A-6 for nu_t and nu_d.
Fractions of delayed neutrons is also present on Table A-7 of the same website.
Have a great day.
hello, do you have point kinetics equation's solution for second third delayed neutron groups? if have, could you please share it, thanks so much
Great video, but Dutch is not a real language. Its just a joke that the Dutch like to play on people.
I have hink lecture is in English.
I have a question: Why are the fluxes on the right side without 𝑟 and 𝑡 dependencies?