oxide fuel. The starting inventory of a 1 GWe reactor is about 100 t of UO2. It generates about 200 kg of plutonium per GWe×year.
Channel-type reactors (RBMK). RBMK reactors using graphite moderator and light-water coolant run on uranium oxide fuel of low enrichment (1.8–2 % 235U). The first core of a 1 GWe reactor contains about 150–180 t of UO 2. It produces approximately 250 kg of plutonium per GWe×year.
The proliferation hazard presented by RBMKs lies in the possibility of channel-by-channel refueling without reactor shutdown.
4. Heavy water CANDU reactors
CANDU reactors are fuelled with natural uranium. The first core of a 1 GWe reactor takes approximately 100 t of UO2. Every year, it will generate 350 kg/GWe.
CANDU reactors operate under conditions of continuous refueling carried out without reactor shutdown. This can pose certain difficulties for inspecting the way primary fuel is used and secondary fuel accumulated.
5. Fast breeder reactors
Fast reactors run on uranium oxide fuel of medium enrichment (15–25 % 235U). The inventory of a 1 GWe reactor is 10–15 t, wh ich means that the
core contains 2–3 t of 235U. It generates plutonium at a rate of about 1500 kg/(GWe×year) in an open fuel cycle. If the fuel cycle is closed, about 80 % of plutonium will be recycled, and its net output will approach 250 kg/(GWe)×year.
Natural NM mining and primary processing
High chemical activity of uranium and thorium explains why they occur in nature only in chemical compounds. All in all, about 200 uranium and thorium minerals were discovered.
The known rich uranium deposits are relatively scarce. The total uranium reserves in the Earth crust are estimated at 1014 t, with another ~ 4×109 t found in the water of seas and oceans (3.3 mg/m3 on the average).
Ore varies in uranium content, appearing as:
·very rich ore: more than 1 % of U;
·rich ore: 0.5–1 % of U;
·medium-grade ore: 0.25–0.5 % of U;
·run-of-mine ore: 0.09–0.25 % of U;
·lean ore: less than 0.09 % of U.
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Generally, mined ore contains about 0.1 % of uranium – these are lean and run-of-mine ores.
Uranium reserves are estimated by two cost criteria as:
∙cheap uranium, costing under $ 80 per 1 kg U3O8; and
∙expensive uranium, costing over $ 80 per 1 kg U3O8.
The threshold value of $ 80 per 1 kg U3O8 was adopted at the time when it marked the border of NPP competitiveness against coal-fired plants.
More than 80 % of proven and probable uranium reserves are found in 8 countries of America (USA, Canada, Brazil), Australia and Africa (SAR, Niger, Angola, Namibia). Two countries, namely, Canada and Australia, produce over 50 % of globally mined uranium. The CIS accounts for about 15 % of the world’s resources. The annual uranium output is not large enough to meet the requirements of global nuclear power. The deficit is offset by using reserves, waste processing, and recycling.
There are four methods of uranium production, i.e.:
∙underground mining;
∙quarrying;
∙in-situ leaching;
∙recovery from seawater.
The stage of ore mining is followed by its hydrometallurgical processing. Uranium can be extracted owing to high solubility of uranium oxides in acid and alkaline solutions.
Hydrometallurgical processing results in dry concentrate of uranium oxides (mainly, U3O8). The concentrate will contain all the uranium found before in the ore, as well as impurities (with 95–9 6 % accounted for by uranium oxides and 4–5 % by impurities). The impuri ties, which have to be removed from the concentrate, include powerful neutron absorbers: B, Cd, Hf and rare-earth elements (Eu, Gd, Sm).
The next stage is uranium concentrate purification using a refining process.
The most common method is refining by extraction, with tributylphosphate (TBP) serving as an extracting agent. The TBP density (0.973 g/cm3) is slightly lower than that of water. To reduce its viscosity, TBP is dissolved in neutral organic compounds. An important property of TBP is its capability of selectively extracting uranium compounds. Uranyl nitrate UO2(NO3)2 is extracted by this agent from a mixture 104 times more efficiently than are impurities.
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Isotopic enrichment of uranium
The chief material of nuclear power is enriched uranium, which contains more fissionable 235U than its natural variety does.
Uranium enrichment may be conceptually sketched as follows. The feed material is natural uranium, with its quantity in the formula below represented by F and enrichment in 235U by XF. As a result, two materials arise: 1) uranium enriched in 235U (product), its quantity denoted by P and
enrichment in 235U, by XP; and 2) uranium depleted of 235U (waste) found in an amount of W and enriched in 235U as XW.
Then, the NM balance is described by two equations:
·F = P + W for the uranium mass balance and
·XF × F = XP × P + XW × W for the 235U isotope balance.
These two equations in three unknowns (F, P, W) can be rearranged to have two equations in two unknowns (F/P, W/P):
F/P = 1 + W/P;
XF × F/P = XP + XW × W/P.
By solving this system, it is possible to find:
a) the rate of natural uranium consumption per unit of product
F/P = (XP – X W)/(XF – X W);
b) the rate of waste generation per unit of product
W/P = (XP – X F)/(XF – X W);
c) the product-to-feed material ratio q:
F = P + W = q × F + (1 – q) × F; q = P/F = (XF – X W)/(XP – X W).
We shall introduce the concept of relative concentration of isotope 235U in a binary mixture (235U,238U), appearing in the input as R = XF /(1 – XF); in
the product output as R¢ = XP /(1 – XP); and in the waste output as R¢¢ = XW /(1 – XW). Then:
the separation factor for the enriched fraction is:
a = R¢/R = [XP /(1 – XP)]/[XF /(1 – XF)]; 33
the separation factor for the depleted fraction is:
β = R/ R′′ = [XF /(1 – XF)]/[XW /(1 – XW)];
enrichment factor: ε′ = α – 1; depletion factor: ε′′ = β – 1;
total enrichment factor for one stage :
ε = ε′ + ε′′.
Table 2.3 presents some data descriptive of enrichment processes in terms of separation factors and power consumption.
Table 2.3
Separation factors and power consumption specific to enrichment processes
Enrichment |
Separation factor, |
Power |
process |
α |
consumption, |
|
|
kWh/SWU |
Electromagnetic separation |
No data |
4000 |
Gas diffusion |
1.0043 |
2300–2600 |
Gas centrifugation |
1,25 |
100–300 |
Separation nozzle |
1,025 |
3000–3500 |
Laser techniques |
3–15 |
10–50 |
Chemical techniques |
1.0025 |
400–700 |
Gas diffusion and gas centrifugation are the two NFC methods most extensively used on a commercial scale. Both of them rely on the difference in mass between molecules of a gaseous uranium compound (UF6).
Uranium hexafluoride UF6 is often used as a starting material in uranium enrichment. This compound has a number of attractive properties:
∙natural fluorine contains only one stable isotope, namely, 19F;
∙uranium hexafluoride can appear as a solid, liquid or gas at moderate temperatures and pressures;
∙uranium hexafluoride can pass from a solid to a gaseous state and back, skipping the liquid phase.
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One of the shortcomings of uranium hexafluoride is its high chemical activity. It reacts with air and steam, giving rise to uranium tetrafluoride UF4 (powder), which deposits in process circuits. That is why circuits and containers should be made leak-tight, should be desiccated and degreased. Ni, Al, Mg, Cu and their alloys as well as teflon show resistance when exposed to UF6.
Uranium oxide conversion to uranium hexafluoride. For isotopic enrichment, the concentrate of U3O8 has to be turned into UF6. This can be done in a two-step process. First, a reaction with fluorine gas yields uranyl fluoride UO2F2:
U3O8 + 3F2 → 3UO2F2 + O2 at 350–370 оС,
whereupon uranium hexafluoride is formed by a reaction of uranyl fluoride with fluorine at a lower temperature of 270 оС:
UO2F2 + 2F2 → UF6 + O2.
A single-stage process (flame method of direct fluorination) can only take place with excess fluorine and at much higher temperatures (900–1000 оС):
U3O8 + 9F2 → 3UF6 + 4O2 .
Production of uranium hexafluoride from uranium dioxide (UO2) is achieved by using another two-step process. First, UO2 reacts with hydrofluoric acid (HF) at 500–600 оС to form uranium tetrafluoride (UF4):
UO2 + 4HF → UF4 + 2H2O, whereupon UF4 reacts with fluorine at 400 оС:
UF4 +F2 = UF6.
1. Uranium enrichment by gas diffusion (GD)
The enrichment GD technology is based on different speeds of thermal motion of heavy and light molecules, and on the ease with which the latter go through thin porous partitions.
In a mixture of two gases with the same temperature, light and heavy molecules have the same average kinetic energy:
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