Calculating burnup/consumption rates in thermal/fast reactor

[Matthew92]

Homework Statement

I am currently taking an introductory course in nuclear engineering. We are using Lamarsh's Introduction to Nuclear Engineering, 3rd Ed.

My question revolves around calculating fuel burnup and consumption rates for thermal reactors and fast reactors. In particular, problem 9 from chapter 4:

If a nuclear reactor is assumed to have a recoverable energy per fission of 200 MeV, calculate the fuel burnup and consumption rates in g/MWd for:
(a) thermal reactors fueled with 233U or 239Pu; and (b) fast reactors fueled with 239Pu.
[Note: In part (b), take the capture-to-fission ratio to be 0.065]

2. The attempt at a solution

For part a, I was able to find thermal data for fissile nuclides (Table 3.4) and was able to come up with this solution:

I however can not find any relevant data to perform part b. I am a bit confused as to how to approach this last portion. Any help and guidance would be truly appreciated.

 

[SteamKing]

Homework Statement

I am currently taking an introductory course in nuclear engineering. We are using Lamarsh's Introduction to Nuclear Engineering, 3rd Ed.

My question revolves around calculating fuel burnup and consumption rates for thermal reactors and fast reactors. In particular, problem 9 from chapter 4:

If a nuclear reactor is assumed to have a recoverable energy per fission of 200 MeV, calculate the fuel burnup and consumption rates in g/MWd for:
(a) thermal reactors fueled with 233U or 239Pu; and (b) fast reactors fueled with 239Pu.
[Note: In part (b), take the capture-to-fission ratio to be 0.065]

2. The attempt at a solution

For part a, I was able to find thermal data for fissile nuclides (Table 3.4) and was able to come up with this solution:

I however can not find any relevant data to perform part b. I am a bit confused as to how to approach this last portion. Any help and guidance would be truly appreciated.

Which relevant data specifically are you looking for? You know how much energy each fission yields (approx. 200 MeV / fission), the molar mass of Pu-239, and the ratio of captures to fissions of 0.065.

This article also has a table of thermal neutron cross sections for various fissile isotopes:

http://en.wikipedia.org/wiki/Breeder_reactor

 

[Matthew92]

Which relevant data specifically are you looking for? You know how much energy each fission yields (approx. 200 MeV / fission), the molar mass of Pu-239, and the ratio of captures to fissions of 0.065.

This article also has a table of thermal neutron cross sections for various fissile isotopes:

http://en.wikipedia.org/wiki/Breeder_reactor

Thank you for your response. After taking what you said into consideration, I came up with this work for part b:

I am however unsure as to what value would be input into the "absorption x-section" value. For part a, I was able to use a table in my textbook that had all of these values for thermal reactors, but a similar table does not exist for fast reactors.

I really do appreciate the continued help!