Corrosion-Resistant Alloys for Nuclear Reactors

[candice_84]

What is the most corrosion resistant alloy other than Zirconium and SS that could be used in reactors but it is not economical?

 

[Astronuc]

Is one referring to a cladding/structural alloy for use in an aqueous environment, such as an LWR? It's hard to be the various Zr-alloys, e.g., Zircaloys, Zr-Sn-Nb-Fe, Zr-Nb-O-Fe, and various derivatives.

There are specialty stainless stainless (austenitic and martensitic) and duplex steels (ferritic-martensitic) that have been developed for liquid metal fast reactors and supercritical water reactors.

Gas reactors use SiC and pyrolytic C.

And in fact, there is a program to look at SiC for LWR applications.

 

[candice_84]

Thanks for your reply, Do SiC or pyroltic C have low neutron absorption? and also do they crack after 10 or 20 years?

 

[Astronuc]

Thanks for your reply, Do SiC or pyroltic C have low neutron absorption? and also do they crack after 10 or 20 years?

It's possible that they will crack eventually.

Is one asking about fuel structural material or reactor vessel structural material. There is a difference in residence time and fluence levels.

In LWRs, fuel is typically resident for 2 or 3 18-month or 24-month cycles. Design lifetime is however up to 8 years. Reactor vessel structural materials must be resident for the life of the plant which was originally 40 years, but now has been extended to 60 years. However, the fluence/dose rate is about an order of magnitude less than that of the fuel.

Naval nuclear fuel is specially designed for lifetimes greater than those achieved in commercial LWRs.

 

[QuantumPion]

What is the most corrosion resistant alloy other than Zirconium and SS that could be used in reactors but it is not economical?

Stronger and better alloys exist, as far as structural integrity in a high-temperature/pressure environment is concerned (e.g. inconel). However I don't think there are any other known that also have good neutronic properties.

 

[candice_84]

Is one asking about fuel structural material or reactor vessel structural material.

My question is about structural material. For example in Molten Salt Reactor the the temperature of fuel or coolant is very high and I assume salt is corrosive, what can be used that could stay there for 60 years and not let the coolant leak into the moderator? I think it would be a disaster if coolant or fuel leak into moderator since it is carrying fission product. Also what kind of fuel do they use in Naval reactors?

 

[QuantumPion]

My question is about structural material. For example in Molten Salt Reactor the the temperature of fuel or coolant is very high and I assume salt is corrosive, what can be used that could stay there for 60 years and not let the coolant leak into the moderator? I think it would be a disaster if coolant or fuel leak into moderator since it is carrying fission product. Also what kind of fuel do they use in Naval reactors?

In a commercial power reactor, the pressure vessel itself is carbon steel with a stainless steel liner. The fuel rods are zirconium alloy (which is not very strong but is transparent to neutrons). The structural supports for the fuel assemblies are made of Inconel, which is probably what you are looking for. It is very resilient to corrosion, radiation, and high temperature environments.

 

[candice_84]

The structural supports for the fuel assemblies are made of Inconel, which is probably what you are looking for.

According to this website the melting point of Inconel, http://www.engineeringtoolbox.com/melting-temperature-metals-d_860.html the melting point of Inconel is around 1400 C. While the fuel center line temperature of UO2, which is used in PWR is 1400 C. I think Inconel cannot work for Molten Salt Reactor. I am in my 2nd year university, so don't take my opinion as granted, i might be wrong. :)

 

[QuantumPion]

According to this website the melting point of Inconel, http://www.engineeringtoolbox.com/melting-temperature-metals-d_860.html the melting point of Inconel is around 1400 C. While the fuel center line temperature of UO2, which is used in PWR is 1400 C. I think Inconel cannot work for Molten Salt Reactor. I am in my 2nd year university, so don't take my opinion as granted, i might be wrong. :)

First of all, the centerline temperature for UO2 is much higher than the surface temperature due to its low thermal conductivity.

Secondly, the fuel cladding is not a structural support, it only has to withstand the internal pressure of the fuel or external pressure of the coolant. If your original question is thus: "is there a better material than zirconium alloy that could be used for fuel rod cladding" the answer is, for a LWR, not really. Zirconium has decent thermal properties, is transparent to neutrons, and has pretty good resistance to the harsh environment of the reactor core.

Now if you are designing a fast reactor, there may be a better alloy to use since you aren't worried as much with the cladding absorbing neutrons. I believe stainless steel has been used in earlier LFR's. I'm not an expert on metallurgy or fast reactor design though so I don't know what benefit there would be to using newer superalloys for this purpose.

 

[Astronuc]

According to this website the melting point of Inconel, http://www.engineeringtoolbox.com/melting-temperature-metals-d_860.html the melting point of Inconel is around 1400 C. While the fuel center line temperature of UO2, which is used in PWR is 1400°C. I think Inconel cannot work for Molten Salt Reactor. I am in my 2nd year university, so don't take my opinion as granted, i might be wrong. :)

FYI - http://www.gen-4.org/Technology/systems/msr.htm

It's challenging at 650°C, but it has low pressure which is beneficial from the stress on the PV.

Corrosion will be challenge, primarily from fission products in addition to the actinides.

Ferritic steels with 12% Cr are being considered for the SFR, since they possesses better strength at high temperatures than austenitic steels. Ferritic steels would have corrosion/IASSC/IGSSC problems in LWRs.


http://217.33.105.254/Energy/EnergyFinal/Abram paper - Electricity section.pdf <save target as>
http://www.foresight.gov.uk/Energy/GenerationIVnuclearpower.pdf


With respect to LWRs, Zr-alloys are commonly used for most of the fuel structure, e.g., fuel rod cladding, guide tube and spacer grids in the flux regions. Top and bottom grids are usually Inconel-718, but could be Zr-alloy or bimetallic (Zr-alloy strip with Inconel springs).

I can't comment on the materials and conditions in the naval reactors.


I have some additional resources on GenIV materials buried in my library. I'll see if I can dig them up.

 

[candice_84]

Thanks for the paper, in the case of MSR, Why don't they mix uranium in water, so it would be less corrosive.

 

[Astronuc]

Thanks for the paper, in the case of MSR, Why don't they mix uranium in water, so it would be less corrosive.

Water is rather corrosive, especially when heated to operating temperatures of nuclear reactor. Uranium eventually hydrolyzed to UO3 which is soluble in water. In addition, fission products would also form compounds with water and each other.

In a reactor environment, radiolysis of water is a factor. That changes the chemical nature and more aggressive species are produced.

High temperature water is generally deleterious to many metals, and more so in a radiation environment.

 

[Hologram0110]

A simple example, although not of structural materials, would be the use of heavy-water instead of light-water as a coolant and/or moderator. This offers better neutron economy but significantly higher costs(and physically larger reactors).

 

[Astronuc]

Reactor grade zirconium alloys have less than 0.010% by weight (100 ppm) of Hf, which is down from the natural value of ~2% Hf in ziron ores (ores with higher Hf concentrations have been found). Production values can be less than 100 ppm.

The total EBC (equivalent boron content) of cladding and structural material is also limited. This is more of a concern for LWR materials, rather than epithermal or fast reactor materials.

Hf is also commerically valuable.

 

[Hologram0110]

Reactor grade zirconium alloys have less than 0.010% by weight (100 ppm) of Hf, which is down from the natural value of ~2% Hf in ziron ores (ores with higher Hf concentrations have been found). Production values can be less than 100 ppm.

The total EBC (equivalent boron content) of cladding and structural material is also limited. This is more of a concern for LWR materials, rather than epithermal or fast reactor materials.

Hf is also commerically valuable.

You're absolutely right Astronuc. I mixed-up the content of commercial grade and reactor grades. For the commercial grade they do not remove nearly as much hafnium to save on cost.

 

[candice_84]

Uranium eventually hydrolyzed to UO3 which is soluble in water. .

If I understand you correctly, This means if uranium is dissolved in water, at high temperature it breaks from water? (making to different compounds?

 

[Astronuc]

If I understand you correctly, This means if uranium is dissolved in water, at high temperature it breaks from water? (making to different compounds?

Yes - it's one of the problems with degraded fuel rods which have failed. U-metal would oxidize to an oxide, which is one of the reasons that LWR fuel is UO₂. But under reactor operating conditions, UO₂ reacts with water/steam to oxidize first to U₃O₈, then to U₄O₉, and then finally to UO₃, which in water/steam and a radiation environment forms a soluble hydroxide, e.g, UO₂(OH)₂. Irradiated fuel is more complicated because of the presence of volatile and soluble fission products.

In reactor coolant, tramp uranium, and transuranics such as Np-239 and Pu-isotopes are a problem.

 

[candice_84]

Yes - it's one of the problems with degraded fuel rods which have failed. U-metal would oxidize to an oxide, which is one of the reasons that LWR fuel is UO₂. But under reactor operating conditions, UO₂ reacts with water/steam to oxidize first to U₃O₈, then to U₄O₉, and then finally to UO₃, which in water/steam and a radiation environment forms a soluble hydroxide, e.g, UO₂(OH)₂. Irradiated fuel is more complicated because of the presence of volatile and soluble fission products.

In reactor coolant, tramp uranium, and transuranics such as Np-239 and Pu-isotopes are a problem.

In reactor operation the three golden rules are "control, cool and contain". The fuel clad is one of the defense in depth barriers (Contain fission product). But according to your explanation, in the case of fuel failure, Uranium can escape through the ceramic into the gap of fuel cladding. Then escape from the clad to the water becoming oxidized (I am assuming there has to be a crack in the zirconium clad otherwise UO2 won't contact water to become oxidized). But this is a serious issue because fission product can also follow the same path as the UO2 and get into the coolant.

Also since you have mentioned about it is complicated to explain Irradiated fuel, Is there a difference between Irradiated fuel and degrading fuel? or you mean, irradiation of fuel by-product that escaped into the coolant. I mean neutron bombardment of any substance or compound which becomes activated and therefore emitting radiation (the creation of tritium for instance)

 

[minerva]

My question is about structural material. For example in Molten Salt Reactor the the temperature of fuel or coolant is very high and I assume salt is corrosive, what can be used that could stay there for 60 years and not let the coolant leak into the moderator? I think it would be a disaster if coolant or fuel leak into moderator since it is carrying fission product. Also what kind of fuel do they use in Naval reactors?

Early MSR work, such as the ORNL Molten-Salt Reactor Experiment, used Hastelloy N, which is a Ni-based superalloy with good chemical performance. Similar alloys are still promoted by advocates of liquid-fluoride thorium reactors today.

You're absolutely right Astronuc. I mixed-up the content of commercial grade and reactor grades. For the commercial grade they do not remove nearly as much hafnium to save on cost.

Hf is very chemically similar to Zr, which means that most commercial Zr contains a small amount of Hf, since it is found together with Zr in nature and most chemical processing won't separate it.

For most non-nuclear applications this doesn't matter, but since Hf has a high neutron cross-section but Zr has a low neutron capture cross section, it has a large effect when it is used for LWR fuel cladding.

When the first LWRs were developed, using Zr-alloys for fuel cladding, a special solvent extraction process was invented at ORNL specifically to allow nuclear-grade Zr with a very low Hf content to be produced.

 

[Astronuc]

In reactor operation the three golden rules are "control, cool and contain". The fuel clad is one of the defense in depth barriers (Contain fission product). But according to your explanation, in the case of fuel failure, Uranium can escape through the ceramic into the gap of fuel cladding. Then escape from the clad to the water becoming oxidized (I am assuming there has to be a crack in the zirconium clad otherwise UO2 won't contact water to become oxidized). But this is a serious issue because fission product can also follow the same path as the UO2 and get into the coolant.

Also since you have mentioned about it is complicated to explain Irradiated fuel, Is there a difference between Irradiated fuel and degrading fuel? or you mean, irradiation of fuel by-product that escaped into the coolant. I mean neutron bombardment of any substance or compound which becomes activated and therefore emitting radiation (the creation of tritium for instance)

For nuclear fuel, there are four functional requirements, three of which one mentioned:

1. Generate thermal energy in a controlled manner
2. Retain fission products (the hermeticity requirement) - which one mentioned as contain
3. Controlability - which infers dimensional stability such that control elements (rods/blades) may be inserted in order to rapidly shutdown the reaction
4. Coolability - which infers dimensional stability such that the fuel rods/elements are coolable under normal operating conditions, anticipated operational occurrences (AOOs) and postulated accidents (PAs).

3 and 4 are related to 2. Basically the fuel is supposed to retain fission product radionuclides as the first barrier between radioactive elements and the environment.

Failures do occur in nuclear fuel. The principal causes are debris fretting (small pieces of metal, e.g., wire or shavings from maintenance or ex-core systems), grid-to-rod fretting (GTRF, related to flow induced vibration), anomalous corrosion (a complex phenomenon related to duty and to some extent water chemistry/crud formation), PCI/PCMI, and manufacturing defects.

When the cladding is breached, volatile fission products, e.g., Cs, I, Rb, and noble gases Xe, Kr are released. Breaches in the cladding (Zr-alloys in LWRs) can lead to secondary hydriding , which may result in cracked or ruptured hydride blisters, circumferential (guillotine) fractures, or axial splits. The more open the primary and secondary breaches, the more likely the UO₂ ceramic will oxidize and release more fission products. UO₂ ceramic oxidizes preferentially along the grain boundaries, so as the fuel is oxidized, fuel particles (grains) will be released.

Fuel under normal conditions is obviously irradiated, and the product of the fission process is a variety of fission products. In addition, structural materials become activated, which is why it is desirable to minimize the use of Ni-based alloys (e.g., Inconels) and Stainless steels in the high flux regions of the core, and instead use Zr-alloys (e.g., Zircaloy, Zr-Nb-based alloys), which have relatively low thermal neutron absorption cross-sections, and hence relatively low activation.

Another concern for LWRs is the deposition of corrosion products from the primary system, e.g., PWR steam generator tubing, and the various Inconel and Stainless Steel structures, including pipes, vessel liner, core internals, etc. The industry has spent several decades learning to optimize the primary system water chemistry, and we're still tweaking the chemistry. Each LWR represents a unique thermochemical environment.

 

[Astronuc]

Hf is very chemically similar to Zr, which means that most commercial Zr contains a small amount of Hf, since it is found together with Zr in nature and most chemical processing won't separate it.

For most non-nuclear applications this doesn't matter, but since Hf has a high neutron cross-section but Zr has a low neutron capture cross section, it has a large effect when it is used for LWR fuel cladding.

When the first LWRs were developed, using Zr-alloys for fuel cladding, a special solvent extraction process was invented at ORNL specifically to allow nuclear-grade Zr with a very low Hf content to be produced.

The Kroll and Van Arkel processes are typically used in the prepration of purified Zr sponge. The hafnium is separated by solvent extraction.

See stage 6 in http://www.wahchang.com/pages/products/data/pdf/ZrProductionFlowChart.pdf

 

[candice_84]

Is it possible to make PWR fuel rods from SiC in 17 x 17 fuel assembly? Does it cost more than Zirconium?

 

[Astronuc]

Is it possible to make PWR fuel rods from SiC in 17 x 17 fuel assembly? Does it cost more than Zirconium?

There is ongoing research in this area. The motivation is to have a material that does not corrode under PWR coolant conditions. There are however challenges, e.g., how to seal the ends.

I'm not sure what the cost of SiC cladding would be compared to a Zr-alloy.