CE60013 - Nuclear Chemical Engineering Flashcards

1
Q

List uranium ppm in v.high grade, high grade, low grade, and v.low grade:

A

Very high: 200,000 ppm

High: 20,000 ppm

Low: 1,000 ppm

Very low: 100 ppm

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2
Q

What are the main steps of the nuclear fuel cycle?

A

Ore mining and milling

Refining

Enrichment

Fabrication

Reactor

Transport and storage

Reprocessing

Recycling

Waste treatment

Waste disposal

(~ in order)

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3
Q

Give examples of uranium mineral types:

A

Oxides (uraninite, pitchblende)

Complex oxides (Davidite)

Silicates (uranophane)

Vanadates (Carnotite)

Phosphates (Autunite)

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4
Q

Describe uranium supply:

A

High and very-high grade uranium mineral ore resources (uraninite and pitchblende) are exhausted.

Uranium mineral sources now being worked are:
- by-product uranium recovery in gold extraction
- by product recovery from phosphate rock in the fertilizer industry
- Low and very-low grade ore bodies, e.g. Rossing mine, Namibia. Average ores processes in US: 800-1500 ppm U (2015 data).
- Former Soviet Union and United States bilateral agreement on nuclear weapons

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5
Q

What are the types of uranium mine?

A

Underground
Open pit
In-situ leaching (ISL)

Open-Pit Mines:
Open-pit mining involves removing large quantities of overburden (soil, rock, and other materials) to expose the uranium ore.
Usage: This method is suitable when the uranium deposits are close to the surface and spread over a large area.

Underground Mines:
Underground mining involves tunnelling into the Earth to reach the uranium deposits. This can be done using various methods, such as shafts or adits.
Usage: Used when uranium deposits are deeper or when environmental or geological conditions make open-pit mining less practical.

In Situ Recovery (ISR) or In Situ Leach (ISL):
ISR involves the injection of a leaching solution (usually containing oxygen and sometimes chemicals) directly into the ore deposit, allowing the uranium to be dissolved and pumped to the surface.
Usage: This method is used when the uranium deposits are in a porous or permeable rock formation, and it can be more environmentally friendly compared to traditional mining methods.

Heap Leaching:
Similar to ISR, heap leaching involves piling up crushed ore on a leach pad and applying a leaching solution to dissolve the uranium. The solution is then collected and processed to extract the uranium.
Usage: Suitable for low-grade ore or when economic considerations favor a less intensive mining process.

Placer Mining:
Placer mining involves extracting uranium from loose, unconsolidated sediments such as riverbeds or beach sands.
Usage: Rarely used for uranium, as deposits tend to be more commonly found in hard rock rather than in placer deposits.

By-Product Recovery:
Some uranium is recovered as a by-product of other mining operations, such as copper or phosphate mining, where uranium is present in small quantities.
Usage: This method leverages the extraction of uranium as a secondary product in conjunction with the primary mining activity.

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6
Q

List the main steps involved in uranium ore processing:

A

Mining ore

Crushing and grinding

Pre-concentration

Leaching

Liquid-solids separation

a. Tailings

b. Ion exchange

c. Solvent extraction

(Then from b and/or c…)

Precipitation and filtration

Drying

Final product

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7
Q

What does traditional beneficiation of mineral ore bodies involve?

A
  • crushing and grinding.
  • roasting of mineral to oxidise uranium
  • mineral upgrading using gravity and air
  • flotation, although they have limited applicability
  • thickening of ore pulps, solid/liquid separation
  • leaching of mineral slurry, which can be done in-situ.

(Beneficiation: the treatment of raw material (such as iron ore) to improve physical or chemical properties especially in preparation for smelting.)

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8
Q

How is uranium is extracted from the mineral ore body?

A

Via chemical extraction

Acid or base leaching of uranium:
- H2SO4 or Na2CO3/NaHCO3 mixture are the preferred reagents.
- Uranium forms soluble complexes with both sulphate and carbonate ions in solution
- Competing inorganic sulphates and carbonates are mostly insoluble

In-situ leaching is used when the uranium deposits are located in porous or permeable rock formations.
The leaching solution (acid or base) is injected into the ore body through wells drilled into the deposit.

The solution dissolves the uranium from the ore, and the uranium-bearing solution is then pumped to the surface for further processing.

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9
Q

For uranium solvent (acid) leaching, why is oxidation required before adding the solvent?

A

Tetravalent U (in UO2) needs to be oxidised to hexavalent U (UO3) to make it soluble (UO3 is soluble).

MnO2 or NaClO3 are often used to oxidise uranium.

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10
Q

What is gangue in nuclear engineering?

A

The commercially valueless material in which ore is found / the rest of the ore that is not uranium.

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11
Q

How do properties of the gangue affect the solvent used for leaching?

A

If gangue is silica (insoluble in acid), acid leaching with H2SO4 is employed (cheaper and faster dissolution than alkaline processes).

If gangue is limestone (it consumes acid), Na2CO3/NaHCO3 is preferred as leaching agent.

You want to dissolve the uranium only, and not the rest of the rock.

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12
Q

What are the necessary conditions for in-situ leaching (ISL)?

A

Ore deposit located in water saturated zone

Aquifer trapped between two impermeable layers

Deposit must have adequate permeability

The deposit must be easily oxidised

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13
Q

Describe in-situ leaching (ISL):

A
  1. Well Injection:
    Wells are drilled into the uranium-bearing ore zone. These wells may be cased to prevent the leaching solution from spreading to unwanted areas.
  2. Leaching Solution Injection:
    A leaching solution, often containing a dilute acid (such as sulfuric acid) or an alkaline solution, is injected into the ore zone through the wells.
    The leaching solution interacts with the uranium ore, causing the uranium to dissolve into the solution.
  3. Uranium Dissolution:
    As the leaching solution percolates through the ore, it dissolves the uranium from the rock matrix.
  4. Solution Recovery:
    The uranium-bearing solution, now referred to as the “pregnant” solution, is pumped back to the surface through recovery wells.
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14
Q

Why is uranium solvent extraction needed?

A

To separate it from the other, unwanted elements, which should stay in the aqueous phase whilst uranium is preferentially separated into the organic phase.

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15
Q

How does ion exchange work?

A

Ion exchange separation is a separation technique based on the reversible exchange of ions between a solid resin or polymer matrix and a solution.

The ion exchange resin, typically in the form of beads or a column, is designed with specific groups that attract and bind uranium ions preferentially.

As the solution passes through the resin, uranium ions replace other ions in the resin, leading to the retention of uranium while other ions are released.

Subsequent elution with a different solution, often an acid, reverses the process, releasing the bound uranium ions.

This enables the separation and concentration of uranium

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16
Q

Why is it better for uranium ion exchange columns to be in sulphate form?

A
  • Uranium in the hexavalent state (U(VI)) is more stable in sulfate solutions compared to other anions. Sulfate provides a stable environment for maintaining uranium in the desired oxidation state during the ion exchange process.
  • The sulfate environment helps prevent redox reactions that might occur with other anions, ensuring the preservation of uranium in its hexavalent state, which is typically the form targeted for recovery.
  • The leaching solutions used to extract uranium from ores or concentrates are often in sulfate form. Using a sulfate form in the ion exchange column ensures compatibility and consistency with the overall uranium recovery process.
  • Eluting or recovering uranium from the ion exchange resin is more straightforward when using sulfate solutions.
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17
Q

Why is uranium refining required?

A

To obtain pure uranium hexafluroide.

UF6 has a bp of 56C (whilst oxides are typically solids). It is easier to convert into a gas and then undergo enrichment.

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18
Q

Why does the uranium process require the production of uranium hexafluoride (UH6) (via reaction for UO2 and HNO3)?

A

UF6 has a bp of 56C (whilst oxides are typically solids).

The next step of the process is enrichment which requires gaseous feed.

It is easier/less costly and energy intensive to convert it into a gas.

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19
Q

List the main steps involved in uranium refining:

A

Dissolution (of uranium ore concentrate and nitric acid)

Purification by solvent extraction

Conversion to UO3

Reduction

Hydrofluorination (anhydrous HF and pure UO2 reacted)

a. Metallothermic reduction

b. Fluorination (obtaining pure UF6)

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20
Q

What is uranium enrichment?

A

The process of increasing the concentration of the isotope uranium-235 (U-235) in a sample of uranium, usually in the form of uranium hexafluoride (UF6).

Fissile isotope U-235 is the one that can sustain a nuclear chain reaction. Natural uranium consists mostly of U-238, which is not as effective for sustaining nuclear reactions.

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21
Q

What techniques could be used for uranium enrichment?

A

Distillation

Chemical exchange

Diffusion

Centrifuge

Aerodynamics

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22
Q

What does SWU stand for?

A

Separative Work Units (measured in kg)

SWU is used to describe the capacity of an enrichment plant

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23
Q

How isthe Separative Work Unit (SWU) calculated?

A

SWU = WV(x.w) + PV(x.p) - F*V(x.f)

Where V(x) is the value function,
V(x) = (1-2x)*ln((1-x) / x)

F = Feed (kg)
P = Product( kg)
W = Waste (kg)
xf = Feed composition
xp = Product composition
xw = Waste composition

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24
Q

What is yellow cake?

A

Yellowcake, also known as urania, is a concentrated form of uranium ore.

The mined uranium ore is crushed and chemically treated to separate the uranium. The result is ‘yellow cake’, a yellow powder of uranium oxide (U3O8). In yellow cake the uranium concentration is raised to more than 80%.

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25
Q

Why is solvent extraction needed in uranium processing?

A

Want to recover uranium of high purity and remove unwanted elements.

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26
Q

How is uranium recovered and concentrated?

A

Liquid-liquid extraction

Ion exchange
- conventional recovery from - clarified liquors
- resin-in-pulp processes

Combined ion exchange / liquid-liquid extraction

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27
Q

Describe a mixer-settler system for solvent extraction:

A

Solvent (usually solute free, e.g. TBP in kerosene) and a feed (aqueous solution containing uranium) are fed into a mixer.

Impeller produces a well mixed system, increasing contact between the feed and solvent.

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28
Q

What are the 2 main performance indicators of uranium mixer-settler systems for solvent extraction:

A

Uranium purity in extract phase

High recovery yield

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29
Q

What do the following represent in equilibrium stage theory?

L
H
y
x
i

A

L - light phase flowrate
H - heavy phase flowrate
y - conc of solute in light phase
x - conc of solute in heavy phase
i - stage number

Light phase is the kerosene / solvent
Heavy phase could be water (immiscible with kerosene)

Concentrations xi and yi (light and heavy phase leaving the stage in opposite directions) would be in equilibrium.

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30
Q

What is the mass balance for the solute in a singe stage, i?

A

MB:
Lᵢyᵢ₊₁ + Hᵢxᵢ-₁ = Lᵢyᵢ + Hᵢxᵢ

Operating line: [y = mx]
yᵢ - yᵢ₊₁ = (Hᵢ/Lᵢ)*(xᵢ-₁ - xᵢ)

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31
Q

What is the extraction factor, Ei?

A

Ei is the ratio of the solute leaving in the extract / the sulte leaving in raffinate

Ei = Lᵢyᵢ / Hᵢxᵢ

= (Lᵢ / Hᵢ)*mi

Where mi is the equilibrium slope.

Ei = slope of eq line / slope of operating line

We want a high Ei and more uranium in the light phase. This can be improved by increasing the amount of light phase, L, or having more favourable thermodynamic conditions to promote mi.

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32
Q

How is single stage extraction loss found?

A

Loss = xn / x0 = 1 / (1+E1)

This is the amount of solute (uranium) remaining in the heavy phase which was not able to be extracted into the light phase.

We want this as close to 0 as possible.

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33
Q

How is uranium recovery calculated?

A

Recovery = 1 - Loss

= Lyᵢ / Hxᵢ-₁ = E1/(1+E1)

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34
Q

How is overall solute loss for cross-current uranium extraction calculated?

A

= xn / x0 = 1/(1+E)^n

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35
Q

How is overall solute loss for counter-current uranium extraction calculated?

A

= xn / x0 = (E-1)/(E^(n+1)-1)

This is known as the Kremser-Souders-Brown Equation, where E is the extraction factor.

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36
Q

Regarding the Kremser-Souders-Brown Equation for counter-current solute loss,

xn / x0 = (E-1)/(E^(n+1)-1)

What do variations in E lead to? (i.e. what happens when E > 1 and E < 1)

A
  1. For E > 1, as n increases, Xn / X0 decreases,
    i.e. with a large extraction factor, as the number of stages is increased, the amount of solute extracted with each stage decreases.
  2. For E = 1, Xn / X0 = 1 / (1 + n)
    With increasing stages, Xn/X0 is decreasing in the above fashion.
  3. For E < 1 and n is large, E^(n+1) tends to 0.
    This leads to Xn / X0 = 1 - E
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37
Q

How does acid concentration effect uranium separation?

A

Separation is improved by increasing acid conc’ in the aqueous phase

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38
Q

When does a pinch condition occur?

A

When the line, H/L is greater than m.

Pinch scenarios suggest an infinite amount of stages would be needed.

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39
Q

What is the decontamination factor, f(AB)?

A

f(AB) = ((y.A1 / y.B1) / (x.A0 / x.B0)

We want this to be large.

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40
Q

What are the 3 main scenarios to beware of for pinch conditions?

A

a) At the feed point

b) at the effluent

c) at the intermediate point

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41
Q

What is the principle of ion exchange?

A

The ion exchange process involves a chemical reaction in which an ion in the solution is retained by a solid, which in turn releases a different ion into the solution.

Ion exchange is stoichiometric.
Electroneutrality is always preserved.

Ion Exchange Resins, in particular a cross-linked polystyrene / polydivinilbenzene (PS-PDVB) co-polymer, are the most relevant to nuclear applications.

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42
Q

Describe the ion exchange resin cation structure (in lecture):

A

Na+ cations are linked to SO3- groups by an ionic bond and can be exchanged with other cations.

SO3- anions are fixed to the resin by chemical covalent bonds

Polystyrene and divinylbenzene provide an organic structure insoluble in water

Water fills the pores in the resin structure and allows diffusion of the cations, which is necessary for the exchange process.

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43
Q

Describe the fundamental properties of ion exchange resins:

A

Ion exchange resins are porous

Ionic species diffuse within the porous structure

The greater the DVB (Divinilbenzene) content of the resins, the greater the degree of crosslinking

Resins with low crosslinking have larger pores and diffusion of ions is faster

If the crosslinking is too low, then moisture content will become excessive and resin will be softer and difficult to use

Cross-linkage of around 8% is commonly used

The styrene/divinylbenzene matrix is hydrophobic and does not absorb water

The resin absorbs water strongly and swells after introduction of ion exchange functional groups

Water absorption is due to hydration of the fixed functional groups and counter ions

Absorption of water results in swelling and hence stretching of the polymer chains, this is balanced by the elastic forces of the polymer

Highly crosslinked resins cannot swell as the polymer chains are restrained

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44
Q

How do sulphonic resins behave?

A

Sulphonic acid is strongly acidic and can dissociate not only in alkaline but also in acidic solutions over the entire pH range.

The sulphonic form can exchange positive ions such as Na+ and Ca2+:

R-SO3H + NaCl -> R-SO3Na + HCl (1)
2R-SO3H + CaCl2 -> (R-SO3)2Ca + 2 HCl (2)
R-SO3H + NaOH -> R-SO3Na + H2O (3)

Reactions 1 and 2 are reversible, regeneration is not 100% in a batch system hence a column system is needed.

Reaction 3 is not an equilibrium reaction, hence ion exchange may be performed efficiently even in a batch system.

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45
Q

How do carboxyl resins behave?

A

Weakly acidic carboxyl groups do not dissociate in acid solution and there is no ion exchange ability

No salt splitting ability in neutral solutions e.g. NaCl or Na2SO4

Selectivity similar to strongly acidic cation exchange resins.
- Although not with respect to H+ ions, which are taken up preferentially to any other univalent ion. Hence ease of regeneration with acid.

Can only be used in a limited pH (basic) range

More economical to regenerate
- Ease of loss of captured ions even a flow of water may be sufficient to hydrolyse them and elute ions in the water

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46
Q

How do quaternary ammonium resins behave?

A

Exchange anions like Cl- and SO4 2-

Resins which have a quaternary ammonium group (triple bond N+) as the exchanging group

  • Groups dissociate in the same way as strong alkalis NaOH or KOH and exhibit strong acidity
  • The quaternary ammonium exchange group dissociates not only in acidic but even in alkaline solutions
  • Resins can absorb mineral acids and split neutral salts (see eqs 1 and 2)
  • Ion exchange properties over the entire pH range.

R-NOH + NaCl -> R-NCl + NaOH (1)
R-NOH + HCl -> R-NCl + H2O (2)

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47
Q

How do tertiary ammonium resins behave?

A

Functional group does not dissociate in alkaline solutions and there is no ion exchange

Ion exchange does take place with mineral acids such as HCl and H2SO4 or with salts such as NH4Cl

Weak acids are generally difficult to capture

Some weakly basic ion exchange resins do exchange with H2CO3 but not with silicic acid

Easy to regenerate, not only with NaOH but also by Na2CO3 or NH3

Captured ions are easily eluted

1) R-N: + HCl -> [R-N:H]+ Cl-
2) [R-N:H]+ Cl- + HNO3 -> [R-N:H]+ [NO3]- + HCl

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48
Q

What is the organic and aqueous phase in U-solvent extraction?

A

The aqueous phase (Heavy phase) is water-based and can be an acidic, basic, neutral, or a saturated salt solution.

The organic phase (Light phase) is an organic solvent, usually diethyl ether or dichloromethane, which has minimal solubility in water.

For instance, ethanol would be a poor extraction solvent because it forms a solution with water.

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49
Q

How would you modify the system if the concentration of A in the extract needs
to be increased keeping the same initial and final concentrations of A in the aqueous phase?

A

Increase the H/L ratio (use less organic solvent) while adding more stages.

So, by increasing the H/L ratio and adding more stages:

  • More of compound A is absorbed into the organic phase due to increased contact between the phases.
  • The concentration of compound A in the organic phase increases, while the concentration of compound A in the aqueous phase decreases.
  • This results in an overall increase in the concentration of compound A in the extract phase while maintaining the same initial and final concentrations of A in the aqueous phase.
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50
Q

How to simply calculate extraction factor, Ei?

A

E = m * (L/H)

Which can then be used to calculate loss via the KSB equation,
Loss = Xn/X0 = (E-1)/(E^(n+1)-1)

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51
Q

Describe how a solvent extraction system fed at an intermediate stage operates and explain why this configuration is advantageous:

A

Intermediate feeding enables having a scrubbing section, where traces of the solute with less affinity for the extract are removed from the extract.

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52
Q

What is the distribution coefficient?

A

The distribution coefficient, m, is the ratio of the organic (L) to aqueous (H) phases.

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53
Q

What are the aspects of ion exchange resins and their functions?

A

Polystyrene substrate - provides an insoluble structure

Divinilbenzene (DVB) crosslink - maintains structure stiffness and rigidity

Ions

As water enters, the polytysrene substrate swells. The DVB cross links are needed to maintain structure and reduce swelling.

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54
Q

What is equilibrium of ion exchange on resin dependent on?

A

In equilibrium, both the ion exchanger and the solution contain both the competing and the counter ions X and Y
The exchange is normally reversible.

The concentration ratio of the two competing counter-ion species is usually different from that in the solution.
As a rule, the ion exchanger selects one species in preference to the other.

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55
Q

What is the rate determining step in ion exchange kinetics?

A

In ion exchange the rate determining step is the interdiffusion of counter-ions.

In the bulk solution, any concentration differences are constantly levelled out by agitation or turbulence (both involving convection).

Agitation does not affect the interior of the beads nor a liquid film which adheres to the bead surfaces.

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56
Q

What factors affect the rate of ion exchange?

A

Rate increases with inverse of the square of the resin bead radius

Rate decreases as resin crosslinking increases

Rate increases as temperature increases, about 4-8 per cent per degree C

Agitation of resin and solution has no effect on rate of exchange

The rate of diffusion within the bead is dependent on the concentration gradient

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57
Q

How do we determine whether ion exchange kinetics are film or particle diffusion controlled?

A

(XD’δ)/(CDro)*(5+2a[AB])

If above &laquo_space;1, particle diffusion controlled

If&raquo_space; 1, film diffusion controlled

Where:
X = concentration of fixed ionic groups;
C = concentration of solution;
𝐷’ = interdiffusion coefficient in the ion exchanger;
D = interdiffusion coefficient in the film;
ro = bead radius;
δ = film thickness;
a[AB] = separation factor

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58
Q

What are the different ways of radioactive decay?

A

Alpha - an α particle is the nucleus of a He atom
Beta - a β particle is an electron
Gamma - γ is high-energy electromagnetic radiation

Positron - a positron has positive charge equal to the electron
Electron capture (or K-capture) - when an atomic nucleus absorbs an inner-shell electron (usually from the K shell) and transforms a proton into a neutron while emitting a neutrino and a characteristic X-ray photon
Spontaneous fission - immediately forming fission products
Neutron emission
Proton emission

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59
Q

How is radioactive decay and half life calculated?

A

𝑁(𝑡)=𝑁0*𝑒^(−λ𝑡)

Where N is the number of atoms, t is time, and λ is a decay constant.

Considering that the half life is when N = 0.5N0,

𝑡(1/2)=0.693/λ

60
Q

What is external vs internal exposure?

A

External Exposure: Radiation incident on the body from the outside

Internal Exposure: Radiation from nuclides present in the body

61
Q

How is exposure calculated?

A

Exposure = ∆q/∆m
= the number of charges produced near the body per unit of air mass

Radiation can cause several types of cancer.
The effect of radiation depends not only on the dose but the linear energy transfer (LET).
LET is the amount of energy transferred to the material per unit of distance: higher for α particles and neutrons than β or γ rays.

62
Q

Write down the ion exchange reaction that takes place between a strongly basic amine in its chloride form and the predominant uranium species in an aqueous solution with high sulphate concentration:

A

4[R-N(CH3)3Cl] + [UO2(SO4)3]4- → [R-N(CH3)3)]4.UO2(SO4)3 + 4Cl-

Due to the high concentration of sulphate uranium forms uranyl trisulphate.

Or for a more general resin,
4[R4-NCl] + [UO2(SO4)3]4- → [R4-N]4.UO2(SO4)3 + 4Cl-

63
Q

What materials are typically used as moderators in the uranium fission process?

A

C (graphite), used in gas-cooled reactors, such as Magnox (UK) and the Russian RBMK.

D as D2O (heavy water), used in Canadian CANDU reactors.

H as H2O (light water), used in BWR and PWR

As neutrons are more effectively slowed-down by lighter atoms and collisions with 238U leads to absorption, a moderator is used in thermal reactors.

64
Q

How is nuclear power produced?

A

Nuclear power generation is based on the energy released in fission.

U235 absorbs a neutron then undergoes fission to form fission products and more neutrons.

Most of the energy is released as kinetic energy of the fission products.
As the fragments collide with other atoms and molecules, this is dissipated as heat.
Neutrons and radiation also interact with matter, leading to thermal vibrations.

[In order of highest to lowest energy], the energy comes from:
Fission products
Neutrinos
Gamma radiation
Beta radiation
Fission neutrons

65
Q

What’s the fuel in uranium fission?

A

The fuel consists of a fissile substance in a metal cladding containment.The fissile substance is typically 235U but 239Pu (used in MOX with U) and 232Th can also be used (breeding reactor).

The chemical form of U is UO2 or metallic U.

Different levels of enrichment are needed depending on the reactor type.

Cladding contains the fuel and avoids fission products mixing with the coolant.

66
Q

How is uranium fuel fabricated?

A

UF6 is reduced to UO2: Several processes have been used.

Reduction of UF6:
UF6 + H2 <-> UF4 + 2 HF

Hydrolysis of UF4:
UF4 + 2 H2O <-> UO2 + 4 HF

Alternative processes are:
1. Hydrolysis of UF6:
UF6 + 2H2O <-> UO2F2 + 4HF
Reduction to UO2:
UO2F2 + H2 <-> UO2 + 2 HF

  1. Hydrolysis of UF6:
    UF6 + 2H2O <-> UO2F2 + 4HF
    Reduction to UO2:
    UO2F2 + 6 NH4OH <-> (NH4)2U2O7 + 4 NH4F 2H2O
    (NH4)2U2O7 is then reduced to UO2 with H2
67
Q

List methods to produce pure U metal:

A

Electrolysis of fused salts
Reduction of UO2
Reduction of UF4
Reduction of UCl4

Reducing agents can be H, Ca, Na, Mg

68
Q

How is UO2 treated to form fuel?

A

UO2 undergoes:
1. Powder preparation
2. Pelletisation
3. Fuel rod loading
4. Fuel bundle assembling

Powder preparation: homogeneisation; blending with U3O8 and compaction

Pelletisation: pressed into pellets followed by sintering (H2 @ 1750 °C) to increase density. Pellets are ~ 1cm depending on the reactor.

The fuel bundle is assembled.
An assembly has control rod as well as fuel rods.

69
Q

What happens once UO2 pellets are made?

A

Pellets are loaded in a rod and cladded.

Rod cladding materials are typically Zircaloy (Zr and Sn alloy with smaller amounts of Nb, Fe, Cr and Ni) or a Mg alloy (in Magnox reactors).

70
Q

What are issues with Zircaloy (Zr and Sn alloy) rod cladding?

A

Zr has a low cross-section for thermal neutrons and is resistant to corrosion when passivated, but Hf needs to be removed (used in control rods) .

Hf absorbs neutrons and is often used in control rods, so we don’t want these in the fuel rods.

Also, it may suffer from H embrittlement

71
Q

Calculate work done by a system and efficiency:

A

W = Qh - Qc

η = W / Qh
η = 1 - Qc/Qh

71
Q

What is the thermodynamic limit to the efficiency of thermal power generation?

A

That if the Carnot cycle

η = 1 - Tc/Th

71
Q

What are the main components of a nuclear reactor?

A

Fuel
Control rods
Moderator
Coolant
Reflectors
Radiation Containment

72
Q

Why is U235 used?

A

U235 is the only fissile nuclide naturally found on Earth.

239Pu can be generated by conversion of 238U.
238U is fissionable but not fissile, i.e., it can only undergo fission at high neutron energies.
U233 is not naturally found on Earth.

73
Q

What is breeding in nuclear fission?

A

The net production of fissile species.

74
Q

What are control rods and/or poisons in nuclear reactors?

A

Materials with good neutron-absorption properties, like Hf or Cd.
Response times are of the order of several seconds.

Prompt neutrons have lifetimes of 0.1 to 1 ms. They represent 99-99.5% of the neutron population. The rest are delayed neutrons, which are key to control.

Boron can also be used in aqueous solution as a poison to decrease neutron numbers.

75
Q

What is k in nuclear chemistry?

A

A ratio of the number of neutrons from reaction 1 producing fission in reaction 2 : the number of fissions in step 1

k < 1 - subcritical reactor with decreasing power
k > 1 - supercritical reactor with increasing power
k = 1 - critical reactor with steady power

76
Q

What are moderators in nuclear reactors used for?

A

Slows down neutrons. Light atoms like H, D and C are effective.
Heavy atoms tend to cause deflection but not major kinetic energy losses.

77
Q

What are reflectors in nuclear reactors?

A

Materials that send neutrons back to the reaction zone.

78
Q

Why is there a containment around nuclear reactors?

A

To act as a barrier to radiation

79
Q

What’s cladding in nuclear reactors?

A

Typically an alloy (like Zircaloy) that contains the fuel and transfers heat to the coolant by conduction.

80
Q

What’s coolant in nuclear reactors?
What are the main coolant choices?

A

Cooling fluid with high conduction properties to take heat away quickly.

Heat conduction to the coolant takes place by convection. Rate of heat transfer depends on a heat transfer coefficient, h.

Coolants:
Water
Heavy water (D2O)
CO2
Liquid sodium
He

81
Q

Describe the Magnox reactor:

A

The core structure is given by graphite bricks (moderator) with holes in which the fuel (metallic U) is inserted in Mg alloy cans.

CO2 raises steam in the secondary circuit.

Typical efficiencies are around 31% and the volumetric power density and fuel rating (power generation per tonne of fuel) are low.

Metallic U maximum temperature is 650 °C (a change in crystal structure leads to swelling).

Metallic U thermal conductivity is very high.

Magnox cladding maximum temperature is 450 °C.

The coolant is CO2 at 20 bars. It circulates through the holes and leaves the core at 400 C.

As high pressure CO2 is a poor heat transfer fluid (low heat transfer coefficient), the fuel cladding has fins to increase heat transfer area.

82
Q

Describe the Advanced Gas Reactor (AGR) [nuclear]:

A

Builds on the Magnox design to improve on efficiency and low volumetric power density.

The coolant is still CO2 but at 40 bars. It leaves the core at 650 °C. Superheated steam is produced at 540 °C and 160 bar.

Higher temperatures lead to higher efficiencies.

Typical efficiencies are around 40% and the volumetric power density increased by 3x and fuel rating by 4x.

A change in fuel is required as higher temperatures are used: UO2 is used instead of metallic U. UO2 is only limited by its melting point (2800 °C).
The cladding is made of stainless steel (m.p. 1400 °C). Its maximum temperature is limited to 750 °C by oxidation reactions (notably with CO2).
This needs U enrichment due to absorption.

Boudouard reaction can lead to consumption of the moderator: C + CO2 <-> 2 CO
Therefore CO and CH4 are introduced, but they may lead to carbon deposition on fuel rods.
Inlet CO2 (cold) is fed through the moderator structure, cooling it.

83
Q

Describe the High Temperature Gas Reactor (HTGR) [nuclear]:

A

Graphite oxidation is sorted by using He, a noble gas, instead of CO2 as coolant.

Helium is chemically inert and does not react with neutrons, so it does not contribute to radioactivity (which may still arise from contaminants).

Argon has the same chemical properties and is cheaper and more abundant but it reacts with neutrons to from 41Ar (radioactive).

Overall efficiencies about 50% can be reached with a Gas Turbine (Brayton) Cycle.

Typical He coolant outlet temperatures can be of up to 950 °C. He at such high T can also be used directly in a gas turbine.

Alternatively, superheated steam produced at 540 °C and 160 bar is used.

There are several different designs with four
potential fuel cycles:
1. Low-enriched uranium. Based on UO2.
2. Mixed Oxide (mixture of UO2 and PuO2 oxides). Pu is obtained from reprocessing. This cycle reduces the need for uranium and 235U enrichment.
3. Pu only.
4. 232Th – 235U mixtures.

In initial designs, the loaded fuel was highly enriched (up to 93%) and mixed with fertile 232Th (5-10 Th to 235U atomic ratio).

High enrichment leads to more compact reactors but more recent designs use lower enrichments (in line with non-proliferation agreements).

The HTGR is not a breeding reactor (it does not generate a net amount of fissile material) as it is not possible to base a breeder reactor on 235U fission at thermal neutron energies).
However, the HT Fast Breeding Reactor is based on a similar concept.

84
Q

What are the advantages of ThO2 over UO2? [used in high temperature gas reactors for nuclear power]

A

Very high melting point (ThO2 3300 °C vs 2700 °C; Th 1750 °C vs 1130 °C for metal U).
Higher strength.
Higher Thermal conductivity.
Greater chemical stability (ThO2 does not oxidise).
Better irradiation behaviour leading to higher burnups possible.
Lower production of actinides

Th232 is converted to U233 in the reactor.

85
Q

In what 2 types of fuel cycles can Thorium, Th232, be used?

A

Closed cycle:
233U is recycled and made into fresh fuel. This reduces uranium consumption. The “THOREX” reprocessing route has been developed. The main challenge is to handle 233Pa, which requires long cooling times.

Open cycle:
Once-through process operating at high burnups (fuel sent to disposal or interim storage). This is the cycle current used.

86
Q

Describe the fuel used in High Temperature Gas Reactor (HTGR) [nuclear]

Describe all layers:

A

Fuel can be:
- UC (uranium carbide),
- UOC (uranium oxycarbide, preferred for high burn-up).
- UO2.
- 232Th can be added to the fuel, also as carbide (ThC).

  1. The fuel is in the centre of 650-850 μm particles.
  2. Outside this core is the low density porous carbon buffer. This attenuates fission products and provides void space to accommodate gases and kernel expansion.
  3. Next is the inner pyrolytic carbon which protects inner layers from chemical attack (Chlorine byproducts from SiC deposition) and provides surface for SiC deposition.
  4. Then a silicon carbide layer which is a high density layer that provides structural support. Impermeable barrier to gases and fission products.
  5. Then its the outer pyrolytic carbon which protects SiC layer from mechanical damage during fuel compaction to fabricate fuel elements. Acts as a barrier should internal layers fail.
87
Q

Name 3 main gas-cooled nuclear reactors:

A

Magnox
AGR (advanced gas reactor)
HTGR (high temperature gas reactor)

88
Q

Describe the Pressurised Water Reactor (PWR) [nuclear]:

A

The most widespread reactor ~66%.

The reactor is made of carbon steel with SS cladding.

High pressure liquid water (155 bar) with temperature in the range 290-330 °C is employed.

There is no boiling in the reactor. The coolant is used to produce steam in a secondary circuit.

Water cooled.

Efficiency ~32-33%.

Low capital cost.

High fuel rating: high heat generation even during decay.

The fuel is UO2 pellets in a Zircalloy cladding.

Pellets are loaded in 3.5 to 4 meter-long rods made of Zircaloy, which form an assembly.

As in other reactors, the flow of neutrons (and therefore criticality) is controlled by inserting and removing control rods.

The arrangement of assemblies in the reactor is such that those with higher enrichment are loaded at the periphery and are moved towards the centre as they become depleted.

89
Q

Describe the Boiling Water Reactor (BWR) [nuclear]:

A

Steam is generated directly in the reactor core from water at 70 bars at 270-290C.

About 10% of the water is converted to steam. Eliminates the need for the steam generator.

Similar to PWR the fuel is also UO2 pellets in a Zircalloy cladding.

The coolant also goes through the steam turbine and condenser.

Drawbacks:
They must be operated within the radioactive boundaries.
can lead to corrosion products in the reactor.
can also give higher (but controlled) radiation doses to operators.

90
Q

Describe the RBMK Reactor [nuclear]:

A

RBMK: Abbreviation (Russian) for High-Power Channel-Type Reactor.

Developed in the Soviet Union.

Boiling water reactor that uses graphite as a moderator.

It does not have a pressure vessel. Each fuel element is located in a pressure tube.

This allows the reactor to be refuelled online.

Steam is generated directly in the reactor core from water at 70 bar at 260-290C.

Vapour quality: 20% max. 14% average.

91
Q

What is used as the coolant in nuclear reactors?

A

Water (being turned to steam) is often used in the control rods. This may be boiling or pressurised (depending on reactor type)

He-N2 is used to cool the graphite structure.

92
Q

What caused the Chernobyl accident?

A
  1. The (RMBK) reactor was operating at full power: 3 GWth. Power was slowly decreased over one day. It was planned to carry out the experiment at 700-1000 MW.
  2. In preparation for the experiment, the emergency core cooling system was disconnected (low steam drum levels were expected).
  3. Nearer the time of the experiment, the local automatic system controlling 12 rods was disengaged. An error in the set point of the overall automatic regulation was made.
  4. There was a power drop to ~30 MW and it proved difficult to control power manually as an increase in 135Xe produced a drop in neutron population. (Xe absorbs neutrons)
  5. Power could only be stabilized at 200 MW (instead of 700-1000 MW), but it was decided to go ahead.
  6. The reactor power was lower than intended and with all (4 back up plus 4 experimental) pumps working, water flow was high, and steam generation was low despite being very close to boiling.

7 (a) Therefore, neutron absorption was high. Control rods were further withdrawn.
7 (b) Steam pressure dropped. Manual attempts to keep steam pressure and drum level above the tripping (automatic shutdown) point were unsuccessful. Trip point settings were overridden by operators to avoid shutdown.

  1. Water was let manually in the steam drum, reaching the desired water level in 30 s, but feeding continued. Cold water moved to the core.
    This led to a further reduction in steam generation and voidage. At this point control rods were fully withdrawn.
  2. Sharp manual decrease in the water flowrate by the operator led to an increase in water temperature and automatic control rods started to lower. 6-8 rods were in the core (less than the minimum 15 required). Normally, reactor should have been shutdown but the test continued.
  3. Experiment started: steam feed to the turbine stopped. Protection was overrun as the reactor would trip when both turbines were tripped. During the experiment, operation of the reactor was not required but it was not shutdown so that a second test could be carried out should the initial failed.
  4. The turbine and the pumps started to run down. Water flow decreased causing an increase in water inlet temperature and steam generation.
  5. Less than 40 s after the start of the experiment, reactor power increased and this could not be compensated by lowering automatic control rods.
  6. Power excursion. Manual shutdown attempted but rods could not be fully inserted (there would have been a 10 s delay before power could be controlled).
  7. Increased steam in the core led to prompt criticality. Reactor power reached 530 MW after 3 s. A second excursion in power was produced from voidage caused by rupture of tubes.
  8. Two explosions were heard, likely one from fuel-coolant interaction and the other from CO and H2 (from gasification of graphite with steam) combustion.
93
Q

Describe the CANDU reactor:

A

Stands for Canadian Deuterium-Uranium Reactor

Uses heavy water as moderator (it absorbs fewer neutrons than light water) at 1 bar and < 80 °C.

Heavy water is also the coolant (90 bars) and passes to a steam generator as in PWR.

The fuel is also UO2 pellets in a Zircaloy cladding. These are inside pressurised tubes (90 bars) with Zircaloy walls, as in the RBMK.

No U enrichment is needed but D2O (heavy water enrichment) is necessary.

Fast control achieved through Cd control rods and adding a “poison” to the D2O moderator in the calandria (dissolved substance that absorbs neutrons, such as a boron salt)

Volumetric power densities are 10% of PWR (large moderator volume) and 4x the AGR.
High capital cost due to large amounts of D2O.

94
Q

Describe properties of Fast Reactors (nuclear):

A

The coolant is a liquid metal (Na is the prevalent one due to good heat transfer properties).

Fast neutrons used, not thermal neutrons

Fuel in mixed PuO2 (20-25%) and UO2 with an austenitic alloy steel cladding.

The reactor can operate at low pressure.

The system has three loops: Sodium is primary and secondary coolant. The secondary coolant is fed to a steam generator.

The large Na reservoir provides a safe heat sink in case of circulation failure.

95
Q

What are the potential hazards of fast (nuclear) reacts?

A

Na (which is used as the coolant) reacts violently with water

Na is a very weak neutron absorber, but can produce radioactive 24Na (half-life: 15 h)

96
Q

What’s the Oklo reactor?

A

A natural nuclear fission reactor is a uranium deposit where self-sustaining nuclear chain reactions occur.
Oklo is the only location where this phenomenon is known to have occurred.

Natural 235U fraction: >3% (down from 25% at Earth formation).
Underground water was moderator and coolant.
Power level was of the order of 100 kW.
6 ton 235U consumption over some hundred thousand years.
Pu was produced but has decayed.

97
Q

What does the fuel being removed from a reactor contain?

A
  1. A fraction of original fissile material (e.g. 235U).
  2. New fissile material bred in the reactor (239Pu).
  3. Non-fissile material (238U).
  4. Fission products
  5. Actinides and other radioactive decay products
98
Q

Why is nuclear fuel removed from the reactor before total burn-up is achieved?

A

Physical and mechanical changes that occur cause profound changes in the properties of the fuel.

Reactivity of the fuel decreases with time due to the growth of fission product poisons like Xe and I.

Burn-up is 60-75% for thermal reactors (3-5 years) and about 25% for fast reactors (12-18 months).

99
Q

What are the different spent fuel strategies?

A
  1. Once-through cycle: Store the fuel as recovered without any separation of useful materials.
    USA, Switzerland and Sweden intend to store unprocessed fuel element assemblies - after about 40 years the spent fuel element assemblies are ready for encapsulation and permanent disposal underground.
  2. Reprocessing: Separate the fissile 235U and 239Pu and fissile-generating materials 238U and store the fission products.
    UK, French, German, Russian and Japanese nuclear industry have adopted a reprocessing strategy involving chemical separation of U, Pu and fission products.
100
Q

How is spent fuel stored?

A

In a cooling pond.

Water absorbs the heat while radioactivity decreases.
Criticality must be taking into account when designing the pool.

101
Q

What are the general principles of the PUREX process?

A

Should handle the required throughput

Should achieve the required flowsheet performance, i.e. acceptable losses of U and Pu to waste streams

Should avoid any risk of nuclear criticality

Should meet any process limitations, e.g. limit residence times to restrict degradation of the solvent (TBP)

102
Q

What’s PUREX?

A

Plutonium uranium reduction extraction

103
Q

What are the objectives of spent fuel reprocessing?

A

Recover Pu in pure form with >99.9% efficiency

Reduce U content of Pu to <1%

Reduce fission product activity in Pu by a factor of ~108

Recover depleted U in pure form decontaminated from fission products

Arrange for the fission products to be in a form suitable for permanent storage

Emission of fission products to the atmosphere and effluents is CONTROLLED and SAFE

104
Q

Outline the basic stages of the PUREX process?

A
  1. Irradiated fuel
  2. Cladding is opened (off-gases to gas treatment)
  3. Fuel dissolution in HNO3 (off-gases to gas treatment)
  4. Primary decontamination (removal of 99% fission products)
  5. U-Pu partitioning
    6a. UO2(NO3)2 purification and conversion to pure UF6
    6b. Pu(NO3)3 purification and conversion to pure PuO2

UF6 used in enrichment
PuO2 used in (fast) reactors

105
Q

What happens in the fuel dissolution stage of the PUREX process?

A

The fuel is dissolved in HNO3 (aqueous phase) while cladding does not dissolve.

For U, the following reactions take place:

3 UO2 + 8 HNO3 -> 3 UO2 (NO3)2 + 2 NO + 4 H2O
[low acid concentration]

UO2 + 4 HNO3 -> UO2 (NO3)2 + 2 NO2 + 2 H2O
[high acid concentration]
Note that U(IV) is oxidised to U(VI)

For Pu, the following reactions take place:

PuO2 2+ + N2O4 + 2 H+ -> Pu4+ + HNO3

4 Pu3+ + N2O4 + 4 H+ -> 4 Pu4+ + 2 NO + 2 H2O

106
Q

How is criticality avoided in fuel dissolution of the PUREX process?

A
  1. controlling the geometry in the dissolver
  2. controlling the concentration of fissile material
  3. adding a neutron absorber, such as B in the concrete or Gd as a nitrate in solution
107
Q

What is criticality?

A

The condition where a self-sustaining nuclear chain reaction is maintained.

In other words, it is the state where the rate of neutron production from fission reactions within the reactor core is balanced with the rate of neutron absorption by reactor components, including fuel, coolant, and control materials.

108
Q

What happens in U-Pu partitioning

A

Pu(IV) is reduced to Pu(III) and moves into the aqueous phase, leaving U in the organic phase.

Potential reductants include ferrous ammonium sulphate Fe(NH4)SO4 and hydroxylamine (NH2OH).

The key point for why Pu is reduced but not U is that the redox potential (for Fe(NH4)SO4) lies between that of U and Pu.
Pu is originally in the form Pu(NO3)4 [Pu 4+] with high m.
(Lower distribution coefficient, m, means that the material is more likely to move to the aqueous phase.).
It’s then reduced, forming Pu(NO3)3, which has a lower m and will move out of organic phase and into aqueous phase.

Pu(NO3)3 is removed into the aqueous phase together with 1% of the feed U some Np and fission products.

109
Q

What happens in Pu and U purification?

A

Pu(III) is oxidised to Pu(IV) again and extracted with TBP/Kerosene, followed by reduction and scrubbing into the aqueous solution.

U is extracted back to the aqueous phase in dilute HNO3 and then extracted with TBP/Kerosene

110
Q

What happens following reprocessing?

A

U can be sent to an appropriate stream in the enrichment process or directly to fuel fabrication.

Pu is sent to fast reactor fuel fabrication.

The fission product waste is further treated for long term disposal.

111
Q

What is the unit meq?

A

1meq = 1 mmol of charge

112
Q

Show the thermal de-nitration of uranyl nitrate:

A

Thermal de-nitration of uranyl nitrate:
UO2(NO3)2 -> UO3 + (NO2 + NO + O2)

At 190C and atmospheric pressure:
1. Uranyl nitrate evaporates to molten form
2. UO3 forms a solid pyrolysis product
3. Oxides of nitrogen are generated as off-gas

113
Q

What are the 3 key reactions in uranium refining?

A
  1. Reduction of UO3 to UO2: UO3 + H2 -> UO2 + H2O

The thermochemistry of the reaction is thought to pass through an intermediate stage:
3UO3 -> U3O8 + ½O2
U3O8 + 2H2 -> 3UO2 + 2H2O
Reduction is fast at about 700 oC

  1. Hydrofluorination of UO2 to UF4: UO2 + 4HF <-> UF4 + 2H2O
    The reaction is reversible and exothermic,
    dHf = -44 kcal/mole
    The rate of conversion of UO2 to UF4 increases up to about 600C. Above 600C the rate of conversion decreases with increasing temperature.
  2. Fluorination to form UF6: UF4 + F2 <-> UF6
    The reaction reaches a flame temperature of ~1600 oC with reactor walls cooled to ~500C. UF6 is recovered by condensation at -10 oC (as a solid).
114
Q

What are the main ways for U235 enrichment?

A
  • Gaseous Diffusion: uranium hexafluoride gas passed through membranes, allowing the lighter U235 isotopes to diffuse more readily than the heavier uranium-238 isotopes.
  • Gas Centrifuge: Uranium hexafluoride gas is spun at high speeds in gas centrifuges, causing the heavier uranium-238 isotopes to move towards the outer edge while the lighter uranium-235 isotopes collect closer to the centre.

Laser Isotope Separation: This method uses laser technology to selectively ionize uranium-235 atoms, allowing them to be separated from uranium-238.

Aerodynamic separation - A specially designed nozzle enables isotopes to be separated.

Note:
- For 235U/238U mixture centrifugation requires much fewer stages than a diffusion plant for U [10-20 compared to >1000 to reach ~4%]
- The centrifuge throughput is much less than a diffusion stage so large number of parallel centrifuges required

115
Q

List the types of gas-cooled nuclear reactors:

A

Magnox (U fuel, CO2 coolant)
AGR (advanced gas reactor) (UO2 fuel, CO2 coolant)
HTGR (high temp gas reactor) (UC, UO2, UCO fuel and He coolant)

All use graphite moderators

116
Q

List water cooled nuclear reactors:

A

BWR (Boiling Water Reactors)
PWR (Pressurised Water Reactors)
RBMK (Boiling Water Reactors - The Soviet-designed RBMK (reaktor bolshoy moshchnosty kanalny reactor)

117
Q

What off-gases are released when nuclear reactor cladding is opened?

A

He: used in some cases to increase heat transfer in the fuel
Kr: 85Kr is the only radioactive noble gas present in large quantities.
Xe: typically over 80 vol.% of the gas but less than 0.01-0.1% of the radioactivity in the off-gas.
14CO2
3H2 : may be oxidised to 3H2O

Additionally, during dissolution the following gases are released:
NO and NO2 (known as NOx) are recovered by absorption and converted back into HNO3.

118
Q

What are the nuclear waste classifications?

A

Very Low Level Waste VLLW: up to 10^5 Bq/kg

Low Level Waste LLW: up to 10^6 Bq/kg α radiation and 10^8 Bq/kg β/γ radiation

Intermediate Level Waste ILW: up to 10^11 Bq/kg β/γ radiation.
comprises a wide range of materials ranging from spent
fuel cladding, spent ion exchange resins, insoluble fission products,
aqueous raffinates, scrubber liquors, effluent treatment plant sludges and miscellaneous solid waste.

High Level Waste HLW: typically 10^11 Bq/kg α radiation or 1013 Bq/kg β/γ radiation
This material is produced in the first liquid-liquid extraction cycle
of processing. Storage, vitrification and disposal strategies are highly developed in the UK. It contains spent fuel and fission products.

VLLW and LLW materials are suitable for discharge to the environment in liquid
or gaseous form and solid wastes suitable for shallow land burial.
Most of this material is generated in hospitals, industry and in the
nuclear fuel cycle. It comprises paper, rags, tools, clothing, filters, etc.
To reduce its volume, it is often compacted or incinerated before
disposal.

119
Q

Describe LLW, ILW, and HLWs:

A

Low Level Waste:
This material is suitable for discharge to the environment in liquid
or gaseous form and solid wastes suitable for shallow land burial.
Most of this material is generated in hospitals, industry and in the
nuclear fuel cycle. It comprises paper, rags, tools, clothing, filters, etc.
To reduce its volume, it is often compacted or incinerated before
disposal.

ILW:
This comprises a wide range of materials ranging from spent
fuel cladding, spent ion exchange resins, insoluble fission products,
aqueous raffinates, scrubber liquors, effluent treatment plant sludges and miscellaneous solid waste.

HLW:
Typically 1011 Bq/kg α radiation or 1013 Bq/kg β/γ radiation). This material is produced in the first liquid-liquid extraction cycle of processing. Storage, vitrification and disposal strategies are highly developed in the UK. It contains spent fuel and fission products.

120
Q

List typical radionuclide gaseous wastes:

A

85Kr (t1/2: 10 y): difficult to recover as cryogenic technologies are hindered by solidification of Xe.

129I (t1/2: 17m y): The only isotope to survive initial cooling. Adsorption on Ag/zeolites is feasible.

3H (t1/2: 12.5 y): typically in water (as HTO), which can be bounded to solids, such as silica gel and concrete.

14C (t1/2: 5730 y): Low levels, so not necessarily considered a waste.

121
Q

Where are low level nuclear wastes produced?

A

Uranium mining

Nuclear plant decommissioning

122
Q

Where are high level nuclear wastes produced?

A

Fission products

Actinides

123
Q

What are actinides?

A

A chemical family / elements with similar masses to uranium, plutonium etc.

They are formed in the reactor from species decay.

They are not fission products - these are heavy products that remain long term.

124
Q

How long does uranium waste treatment take?

A

Cooling pond (from reactor): 0.5-3 yrs

Interim liquid/solid storage: 30 yrs

Engineered Surface Storage: ~ 100 yr

Repository: 1000s of yrs
Geological disposal allows time for nuclides to decay to the toxicity levels of uranium ore before they can reach the biosphere again

125
Q

What’s transmutation?

A

The conversion of one element into others.

The aim is to shorten the radioactivity half live by converting Pu and the minor actinides into species that decay faster.

These processes need to be carried out in sub-critical reactors to operate safely. Neutrons are generated in-situ by bombardment of a Pb target.

126
Q

What are the aims of nuclear High Level Waste Solidification?

A

Immobilisation: safer to handle, less corrosive.

Resistance to damage and degradation over geological timescales.

Three main alternatives for reprocessing wastes:
Vitrification - immobilisation in glass structure
Ceramics - immobilisation in crystalline structure
Cementation - more for ILW, materials embedded in cement

Spent fuel can be encapsulated.

127
Q

Describe Vitrification:

A

Vitrification: immobilisation in a glass matrix, normally Borosilicate, phosphate or Lanthanum Borosilicate.

Glass and a solution of fission products are fed together to a melting furnace.

The large atoms will be surrounded by the large 3D glass crystal structure, keeping the species immobilised.

128
Q

Advantages of nuclear waste vitrification:

A

Incorporates nuclides in the lattice.

High resistance to irradiation due to lattice flexibility (broken bonds recombine in picoseconds).

It incorporates He produced by α decay.

Little swelling or recrystallisation leading to stable properties.

Insoluble in water (1 nm per day @ 20 °C; 1 μm per day @ 100 °C)

129
Q

What do deep geological repositories rely on?

A

The matrix (glass)

Engineered barrier (clay layer)

Geological barrier (surrounding rock)

130
Q

Factors to be considered when deciding a deep geological repository location:

A

Stability: No faults or volcanic activity. Limited consequences of erosion and glaciations (plains).

Hydrogeology: will ultimately convey nuclides away from the repository and potentially into the biosphere.
Low permeability rocks are favourable:
- Aquitard: 1 mm per year
- Aquifer: 10 mm per year

Chemistry: Good adsorption capacity of the host rock.
Adsorption will delay the flow of actinides with water with retardation factors in clay of the order of 104 to 6x105.
Solubility also depends on oxidation state: actinides as +3 and +4 are largely insoluble (immobile) while +5 and +6 are soluble (mobile).

Thermal and mechanical properties: Feasibility of excavation; high thermal conductivity as in the repository heat is transferred by conduction.

131
Q

Describe features of possible Earth types around deep geological repositories:

A

Salt:
Free from circulating groundwater
Stable despite deforming under stress (creep)
[salt domes in northern Germany are 100m years old, about the age of the Atlantic Ocean].
Good thermal properties

Granite:
low porosity
high thermal conductivity
Main drawback: existence of large scale fractures (higher permeability)

Clay:
Impermeable
High chemical stability
Poor mechanical and thermal properties

132
Q

Describe the PUREX process using text, diagrams and chemical equations:

A
  • PUREX is a process for recovering Pu in pure form with > 99.9% efficiency
  • It should reduce the U content of Pu to <1%
  • It should reduce fission activity in Pu by a factor of 108
  • It should recover depleted U in pure form, decontaminated from fission products
  • PUREX helps to arrange for fission products to be in a form suitable for permanent storage
  • Emission of fission products and effluents to the atmosphere are controlled and safe
  • Risks of nuclear criticality must be avoided
  • The fuel is dissolved in HNO3 (aq phase) – the cladding does not dissolve

For U…
Low acid conc: 3UO2 + 8HNO3 → 3UO2 (NO3)2 + 2NO + 4H2O

High acid conc: UO2 + 4HNO3 → UO2 (NO3)2 + 2NO + 2H2O

Pu is converted to Pu(NO3)4 by the addition of N2O4:
PuO2^(2+) + N2O4 + 2H+ → Pu^(4+) + HNO3

4Pu^(3+) + N2O4 + 4H+ → 4Pu^(4+) + 2NO + 2H2

  • During dissolution, criticality is avoided by i) controlling the dissolver geometry, ii) controlling the fissile material concentration, and iii) adding a neutron absorber, such as B in the concrete, or Gd in solution as a nitrate.
  • In the scrubbing stages, Pu(IV) is reduced to Pu(III) using reductants such as Fe(NH4)SO4 (ferrous ammonium sulphate) or hydroxylamine (NH2OH)
133
Q

Describe the complete process for extraction of U, from an orebody through to production of yellowcake.

A

Ore is mined, then beneficiation of minerals ore includes:
* crushing and grinding
* roasting of mineral to oxidise uranium
* mineral upgrading using air flotation, etc
* thickening of ore pulps, solid/liquid separation
* leaching of mineral slurry

Leaching may be done using acid or base leaching, and may be in-situ, in heaps or in
agitated vessels.

Uranium is then recovered from the leachate using ion exchange, liquid-liquid extraction, or a combination of both.

The solution is precipitated with NaOH or NH4OH to form yellowcake concentrate.

134
Q

Discuss the interactions of radiation with water and the important considerations of this in nuclear systems:

A

Water undergoes radiolysis according to a wide range of primary and secondary reactions, producing many short-lived, highly reactive species.

This is very important since water is used as a moderator, coolant, and heat transfer fluid, and the interactions of the active species with other materials must be carefully managed, especially to avoid materials degradation leading to equipment failure.

135
Q

Discuss the interactions of radiation with carbon dioxide and the important considerations of this in nuclear systems:

A

CO2 undergoes radiolysis and forms reactive species. These are particularly of relevance in the core of AGR systems, where they can react with the graphite moderator, leading to corrosion and cracking.

Where CO and CH4 are added to control the corrosion, the deposition of carbon becomes an additional concern.

136
Q

Discuss the interactions of radiation with hydrocarbons and the important considerations of this in nuclear systems:

A

Radiation chemistry depends on the organic molecule:

  • Chlorinated hydrocarbons tend to lose Cl
  • Aromatic compounds are more radiation stable
  • Hydrocarbons give H2, unsaturated compounds, and dimers which can polymerise in unfavourable ways
  • In the presence of O2, peroxy compounds form
137
Q

As part of the uranium refining process pure UO3 is converted to a fluoride.

Show the reactions involved and explain the rationale behind obtaining this final product.

A

UO3 + H2 -> UO2 + H2O

UO2 + 4 HF -> UF4 + 2 H2O

UF4 + F2 -> UF6

Uranium hexafluoride (UF6) has low boiling point and can be easily converted to the gas form needed for enrichment.

138
Q

Explain the significance of 232Th as a potential nuclear fuel source:

A

By absorption of a neutron, 232Th is converted to 233Th, which then decays to 233U, a fissile nuclide:

Th(233) + n(1) → Th(233) → Pa(233) → U(233)

(Also, it is more abundant than U and produces waste with shorter half lives than U).

139
Q

What are the main contributing radioactive groups after short (<100 years) and long (>100 years) times of reprocessing?

A

Short term radioactivity is largely caused by fission products.

Long term radioactivity is caused by actinides and their daughters.

140
Q
A
141
Q

What occurs in the leaching stage, and what leaching agents may be used?

A

Leaching liberates the uranium from the solid ore. Usually leached with sulfuric acid (to
dissolve the uranium) and an oxidant (to convert U(IV) to U(VI) which is more soluble.

  • H2SO4 or Na2CO3/NaHCO3 mixture are the preferred reagents
  • uranium forms soluble complexes with both sulphate and carbonate ions in solution
  • competing inorganic sulphates and carbonates are mostly insoluble

Acid leaching process:
UO2 + ½O2 → UO3
UO3 + 2H+ → UO22+ + H2O
UO22+ + SO42- → UO2SO4
UO2SO4 + SO42- → [UO2(SO4)2]2-
[UO2(SO4)2]2- + SO42- → [UO2(SO4)3]4-

142
Q

What occurs in the precipitation and filtration stage, and what precipitating agents may be
used?

A

Precipitation is to convert the dissolved and purified uranium into a solid, and filtration is to
remove the solid from the solution.

  • Uranium eluates and/or raffinates are preciptated with NaOH or NH4OH to form “yellow-cake” concentrate, Na2U2O7, (NH4)2U2O7
143
Q

Where can the following types of waste be found?
VLLW
LLW
ILW
HLW

A

VLLW - Materials from hospitals, processing plants with extremely low levels of activity

LLW - Soft and hard waste, Packaging materials

ILW - Reprocessing wastes, power station wastes, decommissioning and historic waste, insoluble fission products, scrubber liquors, ion exchange resins, corroded magnox sludge, effluent treatment, floc, ion exchangers, filters, zeolites, combustible wastes, solid scrap

HLW - Spent fuel elements, materials from first LL extraction cycle

144
Q

What is involved in uranium refining?

A

After the uranium ore is extracted from an open pit or underground mine, it is refined into uranium concentrate at a uranium mill. The ore is crushed, pulverized, and ground into a fine powder.

Chemicals (sulphuric acid and sodium chlorate) are added to the fine powder, which causes a reaction that separates the uranium from the other minerals. Groundwater from solution mining operations is circulated through a resin bed to extract and concentrate the uranium.

145
Q

How is separation factor a[AB] calculated?

A

a (ᴬb) = (CA’ * CB) / (CB’ * CA)

Where:
CA’ - A resin conc
CA - A solution conc
CB’ - B resin conc
CB - B solution conc