Talk:Burnup

Reducing the amount of nuclear waste

How does high burnup reduce waste? Pu-239 is burned, if you consider that waste; on the other hand, heavier actinides build up that can be even more troublesome. Only a few radioactive fission products like Sm-151 have such a high neutron cross section that much of their inventory is destroyed, and they're not the important ones. --JWB 23:35, 2 June 2007 (UTC)

Bizarre as it may seem (and I think it is), the most popular fuel cycle right now does consider Pu-239 as waste... the plan is to permanently bury spent LWR fuel elements whole at Yucca Mountain, Olkiluoto, and Forsmark, and I guess more to come.

So while the main driver to the (very successful) quest for longer burnup LWR fuel is greater availability owing to longer periods between shutdowns, the reduction in the number of fuel assemblies to be buried is also a plus, and a secondary driver. You're quite right, the simple answer is that it just reduces the number of fuel elements to be buried. They are also a little more radioactive in the short term, as most of the short-term activity is the intense gamma from fission product decay chains, but not enough to make them any more expensive to dispose of per fuel element, partly because the planned repositories are so over-engineered anyway.

A small comment: You have to shut down a stem power plant about every 18 months to take it apart and put it together again. If you have the time you check for damage before putting it together again. REASON: If the hot parts stay together to long they weld together and you will have great problem taking them apart when needed. If anything breaks down it can take three days or more to get the parts apart to correct a small problem against 3 hors if you have dismantled it repeatedly. (Three days was after 20 months hot running of a high pressure valve, personal experience, the other gang with low temperature valve more inaccessible spent 5 days during the same maintenance period.)Seniorsag (talk) 15:21, 8 June 2015 (UTC)

And now I understand this edit... I thought the original was simple, accurate and easily understood, and I still think it's accurate, but perhaps it was too simple, and it obviously wasn't understood. I don't think we're quite there yet. Hmmmm...

With any form of reprocessing, it gets a lot more complicated, and I wouldn't claim that higher burnup will necessarily reduce the waste in those fuel cycles. But I do claim it's one reason that people have gone to a lot of trouble to increase the burnup of LWR fuel. Andrewa 07:22, 8 June 2007 (UTC)

Yup, I would agree it is mostly the reduction in fuel elements and fueling periods. (and don't forget spent fuel pool capacity which is now running short at some sites) Feel free to edit my text to explain more clearly. (I have read about short-term heat being a limiting factor at Yucca Mountain though.) ::--JWB 23:09, 8 June 2007 (UTC)

Yes. But some (not all) of the stuff I've seen proposed for Yucca Mountain read like it was straight out of E. E. Smith, the engineering seemed to be of that genre. Part of the problem is that there's more money available than is really needed, and an even bigger part is that there are people who want it to cost as much as possible, for various reasons ranging from attempts to sabotage the economics of nuclear power to honeypot opportunism. I think the Swedish and Finnish proposals (which are both also simplified by being only for spent LWR fuel, while Yucca will take other sorts of HLW as well) are better ones to study. Certainly the heat is a consideration, but I'd be very surprised if the extra heat were significant enough to negate the benefits of high burnup, which is the way the article now reads.

The effect of higher burnup on spent fuel pool capacity is about neutral, perhaps even a slight advantage... that stuff needs to be stored under water for a little longer, but there's longer between fuel changes. The main reason for running out of spent fuel pool capacity is delays in getting Yucca etc organised. Most of the stuff now stored under water is already cool enough to be moved to dry storage, but there's uncertainty about how long this dry storage will be needed, and but for politics it wouldn't be needed at all. And again, those delaying the program don't take any responsibility for the extra costs they are causing, quite deliberately in some cases. Andrewa 00:54, 9 June 2007 (UTC)

I'll keep an eye out for cites on the thermal limitation stuff. I have seen a study of the feasibility of partitioning cesium to reduce medium-term heat load as a repository limitation,[1] so there are some nuclear scientists considering it, not just antinuclear activists.

I was mentioning fuel pool capacity as an advantage of higher burnup / fewer fuel elements. If the pool were limited by heat instead of volume, it would be burnup-neutral, but I doubt this is the case. --JWB 01:17, 9 June 2007 (UTC)

Quotes from: [2]

the statutory limit for the planned [Yucca Mt] geological repository, 63,000 Mt [metric tons] of civilian nuclear spent fuel, will be reached in 2015... DOE-NE is therefore developing a strategy to reprocess U.S. spent fuel that would first extract pure uranium from dissolved spent fuel and then process, “mixtures of plutonium and selected minor actinides for preparing proliferation-resistant fuels. . . . If implemented successfully, this treatment technology could significantly reduce the cost of the first repository and potentially eliminate the technical requirement for a second.”⁶

The accumulation of decay heat in the rock between the emplacement tunnels, in the period after which forced ventilation ceases, limits the capacity of an above-ground-water-level repository. With the removal of the longlived transuranics, the cumulative output of thermal energy by the radioactive waste in the first few thousand years would be significantly reduced. For Yucca Mountain, it has been estimated that removing the transuranics would make it possible to store the remaining radioactive waste from about five times as much spent fuel as could be stored there directly without reprocessing.⁷

DOE-NE is therefore also exploring the benefits of separating and storing on the surface the 30-year-halflife isotopes, ¹³⁷Cs and ⁹⁰Sr, which would otherwise dominate the heat output from the radioactive waste during the period immediately after repository closure. This would allow the storage of the residual radioactive waste from about 40 times as much spent fuel as could be stored in the proposed Yucca Mountain repository in the absence of reprocessing.

Most actinide transmutation proposals also include ⁹⁹Tc and ¹²⁹I, and most of the <30 year isotopes will be gone by the time the waste goes to the repository. ¹³⁵Cs will go with the ¹³⁷Cs. After excluding all that, what's left, ⁹³Zr? Looks like full reprocessing and non-repository disposal are returning through the back door! --JWB 16:15, 9 June 2007 (UTC)

Effects on total radioactivity at various periods in the future

Here are further thoughts on the effects of higher burnup on radioactivity from actinides (per unit energy or per number of fissions, not per fuel element) at various times after fuel use:

• ²⁴¹Pu (halflife 14 years) per unit of energy produced is a nonlinear function of burnup, first increasing, but not sure if it starts decreasing at some point. Higher burnup allows more time for neutron capture, but then ²⁴¹Pu itself absorbs neutrons with 5 times the cross section of ²⁴⁰Pu. ²⁴¹Am (halflife 432 years) is the decay product, and may dominate radiation production in that time range.
• ²³⁸Pu (86 years) increases as the 2.5 power of burnup. [3]
• ²⁴⁰Pu, ²⁴³Am, ²⁴⁵Cm, ²⁴⁶Cm all have halflives in the 5k-9k range and will be increased with burnup. ²⁴⁰Pu is abundant enough to dominate radiation production in this range.
• ²³⁹Pu (24 ky) will be burned up with higher burnup, but is also constantly produced from ²³⁸U. The net effect should be a slight decrease in ²³⁹Pu per unit energy, and possibly a slight decrease in net radiation in the range after most ²⁴⁰Pu has decayed.
• ²⁴²Pu (372 ky), and smaller amounts of ²⁴⁸Cm (340 ky), will be increased with burnup. ²⁴²Pu and the fission product ⁹⁹Tc will be the dominant radiation sources in this period. ⁹⁹Tc is not an alpha emitter, but may be more mobile.
• ²³⁷Np (2.14 my) is the decay product of ²⁴¹Am, neutron capture product of ²³⁶U, and (n,2n) product of ²³⁸U. It may dominate radiation in this period; competitors are the weak beta-emitting fission products ⁹³Zr and ¹³⁵Cs. Effect of higher burnup unclear, perhaps slight increase.
• ²³⁶U (23 my) will be decreased by lower reliance on ²³⁵U fission and increased by decay of increased ²⁴⁰Pu. Since the fission/capture ratio is about 6-7 for ²³⁵U, 4 for ²³⁹Pu, and ²⁴¹Pu does not produce any ²³⁶U, the net effect is a slight increase. Also, ²⁴⁷Cm (15.6 my) will increase with burnup.

--JWB 23:09, 8 June 2007 (UTC)

Hmmm. Getting close to WP:OR here I fear. Andrewa 00:42, 9 June 2007 (UTC)

Well, that's why I'm posting it here rather than this article or another article, so we can look out for citations confirming or denying some of this stuff. Most likely the dependence on burnup is not overwhelming, so that detailed an analysis would be more appropriate for an article on long-term waste storage in general. For this article, the main point is that waste radioactivity is close to proportional to total fuel fissioned / energy produced, regardless of the number of fuel elements the fuel was packaged in. --JWB 01:17, 9 June 2007 (UTC)

Disagree. This article is on burnup, not waste disposal. Much of the work done (in the USA in particular) on gen II reactors has been directed towards increasing fuel burnup burnup at discharge, for example by use of burnable poison. One of the drivers of this program has been the decreased waste disposal problem. That's relevant to an article on burnup.

And IMO it's a shame that this information has been removed from the article. Your relatively sophisticated analysis is I'm sure of interest to many too, but it will go right over the head of others, and probably belongs in a different article anyway. Andrewa 23:00, 10 June 2007 (UTC)

What has been removed from the article? My one edit did not remove the information that amount of waste in terms of number of fuel elements per energy generated decreases with increased burnup. It added the information that amount of waste in terms of radiation and heat generation does not follow the same relation. I don't doubt that much work has been done on increasing burnup and hope you will add information on it if you have it. --JWB 01:14, 11 June 2007 (UTC)

I think it did remove the information... the point may have been still there on careful reading, but was obscured by the qualifications to the point that, if you didn't already know it, you wouldn't have been informed.

I've had a go at a refactor, to incorporate your (excellent) points without clouding the big picture (yuck, mixed metaphor). See what you think. Andrewa 04:47, 17 June 2007 (UTC)

Reducing fuel requirements

"Higher burnup allows more of the fissile ²³⁵U and of the plutonium bred from the ²³⁸U to be utilised, reducing the uranium requirements of the fuel cycle." Is there a reference for this? If anything, the use of burnable poison should reduce the neutrons available for breeding. --JWB (talk) 07:47, 11 May 2008 (UTC)

I am sure that there is, but it isn't hard to explain. While we often use one number of reactor criticality, reactors are much larger than the mean free path. You want reactivity to be somewhat equalized throughout. If you put new fuel in, it will obviously have more U235 than older fuel. So the burnable poison allows criticality to stay more uniform. The big problem is stability. You want the reactor to run at a nice constant power level throughout, both spatially and temporally. Gah4 (talk) 07:17, 13 November 2022 (UTC)

What is the maximum possible burnup?

Could someone please give the burnup values (GWd/MTU) for *pure* U-233, U-235, and Pu-239, to establish a ceiling on burnup performance? --HH 68.110.169.4 (talk) 09:56, 28 July 2008 (UTC)

I think you are talking about the maximum efficiency for nuclear weapons, which is classified :) --JWB (talk) 16:52, 28 July 2008 (UTC)

Not really. Back in the 1930s,

"...Lise Meitner... worked out that the two nuclei formed by the division of a uranium nucleus together would be lighter than the original uranium nucleus by about one-fifth the mass of a proton. Now whenever mass disappears energy is created, according to Einstein's formula E=mc2, and one-fifth of a proton mass was just equivalent to 200MeV. So here was the source for that energy; it all fitted!" (Nuclear fission)

http://hyperphysics.phy-astr.gsu.edu/Hbase/nucene/u235chn.html says 215 MeV, for U-235.

http://www.science.uwaterloo.ca/~cchieh/cact/nuctek/fissionenergy.html says 204 MeV.

http://nuclearweaponarchive.org/Library/Fission.html says 200±6 MeV.

http://www.world-nuclear.org/education/phys.htm says,

"The total energy released in fission varies with the precise break up, but averages about 200 MeV[*] for U-235 or 3.2 x 10^-11 joule. That from Pu-239 is about 210 MeV[*] per fission. (This contrasts with 4 eV or 6.5 x ^10-19 J per molecule of carbon dioxide released in the combustion of carbon in fossil fuels.)

[*] these are total available energy release figures, consisting of kinetic energy values (Ek) of the fission fragments plus neutron, gamma and delayed energy releases which add about 30 MeV."

http://www.ieer.org/reports/n-basics.html says,

"U-236 ===> fission fragments + 2 to 4 neutrons + 200 MeV energy (approx.)"

http://www.eoearth.org/article/Accelerator-driven_nuclear_energy says,

"These numbers compare with 200-210 MeV released by the fission of one uranium-235 or plutonium-239 atom."

http://www.kayelaby.npl.co.uk/atomic_and_nuclear_physics/4_7/4_7_1.html Oh!

U-233: 197.9 - 6.9 + 9.1 = 200.1 MeV

U-235: 202.5 - 8.8 + 8.8 = 202.5 MeV

Pu-239: 207.1 - 7.1 + 11.5 = 211.5 MeV

Using the last,

1 MeV = 1.602 e-13 J = 4.45 e-20 kW–hr = 1.854 e-27 GW–d; 1 amu = 1.661 e-30 t, so

U-233: 958 GW–d/MT; U-235: 962 GW–d/MT; Pu-239: 988 GW–d/MT.

So could we say ~960 GW–d/MT for uranium and ~990 for plutonium?

—WWoods (talk) 20:02, 28 July 2008 (UTC)

You're talking about the energy released per fission. Converting this to the same units as fuel burnup and noting it in the article is not a bad idea, but it is not itself called burnup. Burnup is mostly a function of fission product buildup, reactor design, fuel enrichment, breeding ratio, and fuel and cladding mechanical limitations. Burnup is also measured per original mass of not only fissile material but also fertile material; the mass fissioned in one fuel cycle may even exceed the mass of fissile material at the beginning of that cycle.

Also note a small amount of the energy is fission product decay energy that does not appear until after the fuel has been discarded, and is not usable for energy production. --JWB (talk) 21:53, 28 July 2008 (UTC)

But doesn't fission energy set the upper bound on burnup? Or an upper bound, anyway. Like the speed of light setting the maximum possible speed for a rocket.

—WWoods (talk) 23:03, 28 July 2008 (UTC)

Take a look at Mass–energy equivalence, which states the proportion of mass lost in fission of Pu-239 is almost exactly one-thousandth, and that one gram of mass is equivalent to 24.9 million kilowatt-hours (≈25 GW·h) --JWB (talk) 21:53, 28 July 2008 (UTC)

And 24.9 GW–hr/kg = 1,037.5 GW–d/t; right. —WWoods (talk) 23:03, 28 July 2008 (UTC)

Burnup in FIMA

The relation between burnup in GWd/t and FIMA is approximately a factor of 10. Source: CEA Glossary.--137.138.4.20 (talk) 08:30, 19 May 2010 (UTC)

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Graphic

I just ran across this, used at the German Wikipedia article on burnup at de:Abbrand (Kerntechnik). Should be useful here. Andrewa (talk) 01:54, 8 September 2020 (UTC)

U235

I was trying to find the amount of U235 left in spent fuel, and thought that would be related to burnup. That is, if you extract all the uranium, how much U235 is still there? This would be easier to understand than some other ways to show it. Gah4 (talk) 07:24, 13 November 2022 (UTC)

EBR-II Location

EBR-II is labeled as being "at" Argonne National Laboratory. This is partially true, as the location of EBR-II was in the Idaho National Reactor Testing Station, run by Argonne. However the reactor is in Idaho and has been part of Idaho National Laboratory since the creation of the lab. Given that Argonne is in Illinois it may be worth distinguishing the physical location of EBR-II. 141.221.55.90 (talk) 20:07, 4 April 2023 (UTC)