CE60013 - Nuclear Chemical Engineering Flashcards

Question

Why is solvent extraction needed in uranium processing?

Answer 1

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

Answer 2

Liquid-liquid extraction Ion exchange - conventional recovery from - clarified liquors - resin-in-pulp processes Combined ion exchange / liquid-liquid extraction

Answer 3

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.

Answer 4

Uranium purity in extract phase High recovery yield

Answer 5

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.

Answer 6

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

Answer 7

Ei is the ratio of the solute leaving in the extract / the sulte leaving in raffinate Ei = Lᵢ*yᵢ / Hᵢ*xᵢ = (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.

Answer 8

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.

Answer 9

Recovery = 1 - Loss = Lyᵢ / Hxᵢ-₁ = E1/(1+E1)

Answer 10

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

Answer 11

= xn / x0 = (E-1)/(E^(n+1)-1) This is known as the Kremser-Souders-Brown Equation, where E is the extraction factor.

Answer 12

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

Answer 13

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

Answer 14

When the line, H/L is greater than m. Pinch scenarios suggest an infinite amount of stages would be needed.

Answer 15

f(AB) = ((y.A1 / y.B1) / (x.A0 / x.B0) We want this to be large.

Answer 16

a) At the feed point b) at the effluent c) at the intermediate point

Answer 17

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.

Answer 18

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.

Answer 19

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

Answer 20

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.

Answer 21

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

Answer 22

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-N*OH + NaCl -> R-N*Cl + NaOH (1) R-N*OH + HCl -> R-N*Cl + H2O (2)

Answer 23

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

Answer 24

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.

Answer 25

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.

Answer 26

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)

Answer 27

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

Answer 28

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

Answer 29

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.

Answer 30

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.

Answer 31

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.

Answer 32

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

Answer 33

(X*D'*δ)/(C*D*ro)*(5+2a[AB]) If above << 1, particle diffusion controlled If >> 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

Answer 34

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

Answer 35

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

Answer 36

External Exposure: Radiation incident on the body from the outside Internal Exposure: Radiation from nuclides present in the body

Answer 37

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.

Answer 38

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-

Answer 39

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.

Answer 40

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

Answer 41

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.

Answer 42

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 2. 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

Answer 43

Electrolysis of fused salts Reduction of UO2 Reduction of UF4 Reduction of UCl4 Reducing agents can be H, Ca, Na, Mg

Answer 44

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.

Answer 45

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).

Answer 46

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

Answer 47

W = Qh - Qc η = W / Qh η = 1 - Qc/Qh

Answer 48

That if the Carnot cycle η = 1 - Tc/Th

Answer 49

Fuel Control rods Moderator Coolant Reflectors Radiation Containment

Answer 50

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.

Answer 51

The net production of fissile species.

Answer 52

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.

Answer 53

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

Answer 54

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

Answer 55

Materials that send neutrons back to the reaction zone.

Answer 56

To act as a barrier to radiation

Answer 57

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

Answer 58

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

Answer 59

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.

Answer 60

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.

Answer 61

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.

Answer 62

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.

Answer 63

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.

Answer 64

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.

Answer 65

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

Answer 66

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.

Answer 67

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.

Answer 68

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.

Answer 69

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.

Answer 70

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. 8. 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. 9. 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. 10. 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. 11. The turbine and the pumps started to run down. Water flow decreased causing an increase in water inlet temperature and steam generation. 12. Less than 40 s after the start of the experiment, reactor power increased and this could not be compensated by lowering automatic control rods. 13. 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). 14. 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. 15. Two explosions were heard, likely one from fuel-coolant interaction and the other from CO and H2 (from gasification of graphite with steam) combustion.

Answer 71

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.

Answer 72

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.

Answer 73

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)

Answer 74

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.

Answer 75

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

Answer 76

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).

Answer 77

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.

Answer 78

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

Answer 79

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)

Answer 80

Plutonium uranium reduction extraction

Answer 81

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

Answer 82

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

Answer 83

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

Answer 84

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

Answer 85

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.

Answer 86

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.

Answer 87

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

Answer 88

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.

Answer 89

1meq = 1 mmol of charge

Answer 90

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

Answer 91

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 2. 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. 3. 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).

Answer 92

* 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

Answer 93

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

Answer 94

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

Answer 95

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.

Answer 96

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.

Answer 97

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.

Answer 98

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.

Answer 99

Uranium mining Nuclear plant decommissioning

Answer 100

Fission products Actinides

Answer 101

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.

Answer 102

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

Answer 103

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.

Answer 104

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.

Answer 105

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.

Answer 106

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)

Answer 107

The matrix (glass) Engineered barrier (clay layer) Geological barrier (surrounding rock)

Answer 108

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.

Answer 109

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

Answer 110

- 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)

Answer 111

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.

Answer 112

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.

Answer 113

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.

Answer 114

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

Answer 115

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.

Answer 116

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).

Answer 117

Short term radioactivity is largely caused by fission products. Long term radioactivity is caused by actinides and their daughters.

Answer 118

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-

Answer 119

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

Answer 120

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

Answer 121

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.

Answer 122

a (ᴬb) = (CA' * CB) / (CB' * CA) Where: CA' - A resin conc CA - A solution conc CB' - B resin conc CB - B solution conc

CE60013 - Nuclear Chemical Engineering Flashcards

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