XYLENE POWER LTD.

FNR U-233 PRODUCTION

By Charles Rhodes, P.Eng., Ph.D.

INTRODUCTION:
The ability to sustainably produce the fissile isotope U-233 from Th-232 has huge implications for the sustainable energy supply and hence for the sustainability of human society on planet Earth. It is believed that thorium is about 4X more abundant than uranium. Hence in terms of providing long term energy supply for human civilization, a fast neutron reactor technology that can operate using Th-232 as its primary energy source has sustainability advantages over U-238.

This web page examines a novel process for making U-233.

A major impediment to use of thorim as a source of prime energy is the relatively low number of neutrons per fission yielded by U-233 in a slow neutron flux. It is believed that the number of neutrons per fission yielded by U-233 substantially improves in a fast neutron flux. The issue is how to utilize this knowledge in a practical nuclear reactor.

In March 2021 this author conceived what might be a workable idea. Essentially the idea involves adding novel fuel tubes to a fuel breeding fast neutron reactor that operates using the U-238 to Pu-239 fuel cycle. One of the important functions of these novel fuel tubes is to breed Pa-233 and then immediately shift the Pa-233 out of the fast neutron flux until it decays to form U-233, at which time it can re-enter the fast neutron flx for subsequent fissioning.

If the Pa-233 remains in the neutron flux it will become Pa-234 and the desired U-233 breeding will not occur.

The desired process is thought to be possible. The concept is to partially fill the fuel tube with FLiBe (LiF-BeF2) to which ThF4 is added. The FLiBe melts between 350 C and 400 C. When Th-232 absorbs a neutron it quicky becomes Pa-233. At an appropriate temperature the PaF5 will concentrate either at the top or bottom of the FLiBe column where it should remain out of the neutron flux for at least a month until it naturally decays to form U-233. At that point it will form UF6 which as a gas will concentrate in the tube plenum. Some of the UF6 will fission and some will remain for later recovery via fuel tube reprocessing.

The concept is to finely grind the ThF4 so that it exists as a powder immersed in molten FLiBe.

Note that it is essential that PaF5 be mobile in the aforemetioned mixture and that the PaF5 concentrate at either the top or bottom of the FLiBe column. It is essential that the ThF4 remain in the high neutron flx region. At this time we are uncertain as to whether the relative densities of FLiBe and ThF4 will permit that to be the case.

Then if the liquid sodium cooled reactor operates at 450 degrees C the ThF4 will be immersed in liquid FLiBe and the PaF5 will accumulate at either the top or bottom of the FLiBe column.

In initial investigation of this idea this author received assistance from Mr. John Rudesill and Dr. Darryl Siemer.

It is assumed that there is an existing FNR built as described elsewhere on this website. This FNR describes a core region with a high fast neutron flux. Assume that we insert into this FNR a 6 m long fuel tube about 2 / 3 full of finely ground ThF4 surrounded by molten FLiBe.

Assume that:
Li is mono-isotopic Li-7

Molten salt references Dock Thesis- Page 72 (sheet 78) and Molten Salts: Eutectic Data indicates that at a Th weight fraction of 20.4%, 500 degrees C thorium chloride is in solution in a low temperature melting point KCl - LiCl salt eutectic.

When exposed to fast neutrons at 450 degrees C it is believed that the Th-232 atoms will each absorb one neutron and convert to Pa-233 which will tend to accumulate at the top or bottom of the FLiBe/ThF4 column outside the neutron flux. After about a month the Pa-233 will convert to U-233 and the U-233 will form UF6. The UF6 will bubble up into the tube plenum. It might eventually fission providing more fast neutrons to support the process.

After some years of operation these tubes can be reprocessed to obtain the surplus U-233.

We need to examine this process to see if it makes sense for extending the world fissionable isotope supply. Some of the neutrons released by U-233 fission will react with the U-238 blanket yielding more Pu-239.

What fraction of the incident fast neutrons are lost by absorption by Li-7, Be, F, Mo? What fraction of neutrons are lost by absorption by Pa-233 to make undesired Pa-234?

Does operation in the fast neutron spectrum significantly increase the number of neutrons / fission released by U-233?

The mono-isotopic Li-7 and isotopically modified Mo will be expensive. Hence the process for recovering metallic U-233 must conserve both the Li-7 and Mo components.

In the following table the cross sections are fast neutron absorption cross sections in barns

NEUTRONICS:
Note that the following table is slow neutron data.

ISOTOPE		WEIGHT	CROSS-SECTION	PRODUCT
Th-232		7.37 b
Li-7		0.0454 b
Mo-92		0.019 b
Mo-94		0.015 b
U-233		574.7 b
U-238		2.68 b
Pa-233		200.6 b
F		0.0096 b
O		0.00019 b
Be		0.0076 b
Na		0.53 b
Bi		0.03 b
Pb		0.171 b

 

This web page last updated April 4, 2021