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This web page sets out details of reprocessing CANDU spent reactor fuel to make liquid sodium cooled Fast Neutron Reactor (FNR) fuel. For an overview of the Ottensmeyer Plan please review OTTENSMEYER PLAN.
The reprocessing involves initial selective U3O8 extraction, transport to a remote fuel reprocessing site, fission product extraction, zirconium recycling, fuel rod fabrication and fission product interim storage.
The process of recrystalization of uranyl nitrate hexahydrate [UO2(NO3)2.6H2O] is used to selectively extract U3O8 from spent CANDU fuel. This process is ineffective at rejecting the elements Np and Cs. The Np simply stays mixed with the uranium. It is not a radioactivity problem unless it contains the isotope Np-237. Np-237 arises from fast neutron n > 2n reactions in U-238. In CANDU reactors the fast neutron flux is very low, so the Np-237 production is low so the contribution of Np-237 to spent fuel radioactivity is very low. This statement is not true for spent fuel from fast neutron reactors. Hence recrystalization of [UO2(NO3)2.6H2O] will not achieve comparable radioactivity reduction in spent fuel from fast neutron reactors.
The process of recrystalization of [UO2(NO3)2.6H2O] is also ineffective at rejection of CsNO3.2H2O. After the recrystalization steps Cs is rejected by heating the U3O8 to over 650 degrees C at which point Cs2O, Cs2O2 and Cs2O3 vaporize leaving behind U3O8 with some contaminant Np. As indicated above the contaminant Np is not a problem unless it contains Np-237 from fast neutrons interacting with U-238.
DEFINITION OF "LOCAL" AND "REMOTE":
To minimize spent CANDU fuel transportation costs selective uranium oxide extraction is performed at the CANDU reactor sites. The term "local" refers to the reactor site. The term "remote" refers to the main fuel reprocessing site which for certainty of public safety should be remote from any major urban population center. A remote site suitable for fuel reprocessing is Chalk River, Ontario.
CANDU SPENT FUEL SELECTIVE URANIUM OXIDE EXTRACTION:
The purpose of selective uranium oxide extraction from spent CANDU fuel at CANDU reactor sites is to reduce the spent fuel shipping cost by dividing the spent fuel mass into a large minimally radioactive portion and a small highly radioactive portion:
a) Portion "A" is 90% of the spent CANDU fuel mass and consists of nearly pure uranium oxide. Pure uranium oxide has the advantage that its radio activity is low so it can be transported with minimal gamma ray shielding. The required amount of shielding is set by the Cs-137 and Np-237 concentrations in the uranium oxide. These isotopes may be present at low concentrations due to non-idealities in the U3O8 extraction process.
b) Portion "B" is the remaining 10% of the spent fuel mass that is highly radioactive and that for bio-safety must be transported in lead containers that have walls that are about 30 cm thick. The cost of transporting portion "B" is dominated by the cost of transporting the required heavy lead containers. Hence from a transportation cost perspective the smaller portion "B" is with respect to portion "A" the better.
REPROCESSING AT REMOTE SITE:
For reasons of public safety in the face of a potential terrorist attack the main fuel reprocessing is performed at a remote site such as Chalk River. The core rod material reprocessing must be done in a manner that ensures no accidental formation of a critical mass.
SPENT CANDU FUEL REPROCESSING SEQUENCE:
Selective extraction of U3O8 should be done at existing CANDU reactor sites to realize about 97.55% of the spent CANDU fuel weight as pure U3O8 and the remainder of about 2.45% of spent CANDU fuel weight as a mixture of (fission products + TRUs + remaining uranium oxide). Enough uranium must be left in this mixture to meet the uranium metal content requirement of FNR fuel. The extracted U3O8 must be sufficiently pure to reduce its radio activity sufficiently to enable low cost transportation.
1. From the spent CANDU fuel inventory that has been out of a CANDU reactor for at least 10 years withdraw spent CANDU fuel bundles as required.
2. Mechanically chop up the spent CANDU fuel bundles into small pieces.
3. Dissolve the spent CANDU fuel bundle pieces in warm HNO3.
4. Remove any undissolved zirconium and transport it to the remote irradiated zirconium store.
5. Cool the solution.
6. Grow and then physically extract urynal nitrate hexahydrate ([UO2(NO3)2-6H2O] + contaminant CsNO3.2H2O) crystals from this solution.
7. Heat the remaining solution to boil off and recycle the HNO3.
8. Condense, collect and recycle the HNO3.
9. Collect the remaining solution dry residue as feedstock for further core rod reprocessing (fission products + TRUs + uranium oxide + zirconium) material.
10. Wash the urynal nitrate hexahydrate [UO2(NO3)2-6H2O + contaminant CsNO3.2H2O] crystals with clean HNO3 to remove surface contamination.
11. Evaporate, condense, recover and recycle the washing HNO3.
12. Collect the dry residue remaining after evaporation of washing HNO3 as feedstock for further core rod reprocessing.
13. Mildly heat the ([UO2(NO3)2-6H2O] + contaminant CsNO3.2H2O crystals to drive off the NO2 and H2O to realize [U3O8 + contamination CsO].
14. Condense, collect and recycle the evaporated nitric acid.
15. Collect the dry U3O8 + Cs2O, Cs2O2, Cs2O3 residue.
16. Dissolve the collected U3O8 + Cs2O, Cs2O2, Cs2O3 residue in warm clean HNO3.
17. Cool the solution.
18.Grow and then physically extract urynal nitrate hexahydrate [UO2(NO3)2-6H2O] + CsNO3.2H2O crystals from this solution.
19. Heat the remaining solution to boil off the HNO3.
20. Cool, condense, collect and recycle the HNO3.
21. Collect the remaining solution dry residue as feedstock for further core rod reprocessing.
22. Wash the crystals with clean HNO3 to remove surface contamination.
23. Evaporate, condense, collect and recycle the washing HNO3.
24. Collect the remaining dry residue from the HNO3 wash as feedstock for further core rod reprocessing.
25. Heat the [UO2(NO3)2-6H2O + CsNO3.2H2O] crystals to drive off the H2O and NO2 to realize a residue of pure [U3O8 + contamination Cs2O,Cs2O2, Cs2O3]
26. Condense, collect and recycle the HNO3.
27. Move the pure [U3O8 + contamination Cs2O,Cs2O2, Cs2O3] to a suitable furnace.
28. Further heat the [U3O8 + contamination Cs2O,Cs2O2, Cs2O3] residue above 650 C to drive off the Cs2O, Cs2O2, Cs2O3 as a condensable gas.
29. Condense and collect the radioactive Cs2O, Cs2O2, Cs2O3.
30. Send the radioactive Cs2O, Cs2O2, Cs2O3 to fission product 300 year storage. After 30 years in storage the radioactivity should be dominated by Cs-137 (30 year half life) and Cs-135 (3.0 X 10^6 year half life). Use a segragated store in case this material contains other contaminants.
31. Collect the pure U3O8 residue and send it to supposedly pure U3O8 storage. The U3O8 may have to remain in this store for centuries until it is required for blanket rod fabrication. Use a segragated store in case this material contains unspecified contaminants. Since this material is reserved for future use as reactor fuel such contaminants have minimal import except for storage biosafety and contanination of the future reducing agent.
32. Combine the various FNR core rod feedstock material streams.
33. Transport the core rod feedstock material (fission products + TRUs + uranium oxide + zirconium) to the remote site for further reprocessing.
REMOTE SITE REPROCESSING:
Oxide Reduction Process:
1. Withdraw U3O8 material as required from interim storage..
2. Reduce the U3O8 to uranium metal. Use either Na, Mg or Al to avoid introduction of Ca into the process. Neutron activation of stable isotopes of Na, Mg and Al lead to other stable isotopes. Neutron activation of Ca forms long lived Ca-41.
3. Send the uranium metal to the pure uranium store at the remote site.
4. Withdraw core rod feedstock material from the remote store.
5. Reduce the core rod feedstock material to uranium metal plus fission products plus transuranium actinides. Use either Na, Mg or Al to avoid introduction of Ca into the process. Neutron activation of stable isotopes of Na, Mg and Al lead to other stable isotopes. Neutron activation of Ca forms long lived Ca-41.
6. Separate the Na2O, Al2O3 or MgO from the metallic fuel.
7. Send the oxide residue for electrolytic recovery of the reducing metal.
8. Electrolytically recycle the reducing metal. Note that the released oxygen is not radioactive. Even O-19 will fully decay to stable F-19 by about 300 seconds after neutron absorption.
9. Send the metallic residue to the fission product separation process.
Fission Product Separation Process (PYRO PROCESS):
In the following section a pyroprocess is used to separate lower atomic weight fission products and zirconium from the high atomic weight elements.
Conceptual diagram of an electrorefiner at the Argonne Lab circa 1999
Photo of an electrorefiner at the Argonne Lab circa 2012
Electrorefining at the Argonne Lab
Electrometallurgical Techniques for Spent Fuel Treatment
1. Separate (fission products plus zirconium) and the (heavy metals) using the liquid cadmium based pyro process. The seoaration ratio in this process is critical. It is crucial to minimize the concentration of high atomic weight atoms in the low atomic weight material discharge stream whereas a small concentration of low atomic weight atoms in the high atomic weight material discharge stream is tolerable. Multiple pyroprocessing steps may be required to achieve the required very low concentration of high atomic weight material in the low atomic weight material discharge stream.
2. Send the heavy metals to the core rod formation process.
3. Send the fission products + zirconium to the zirconium recovery process.
Zirconium Recovery Process (DRY CHLORIDE) and Fission Product Interim Storage:
In the below material flow description ZrCl4 distillation is used to extract zirconium from the fission product discharge stream. This extracted zirconium is recycled back into the fuel rod material. To the extent that the zirconium extraction process is non-ideal a small fraction of the zirconium remains in fission product chlorides in the 300 year fission product storage and a small fraction of the fission products flows back into the fuel rod material.
The dry fission product chlorides are stored in engineered porcelain containers placed within a naturally dry DGR to isolate the fission products from the environment for at least 300 years. The required chlorine is readily available as a by product of liquid sodium production. After 300 years SeCl4 and SnCl2 should be selectively extracted from the other chlorides and placed in long term dry DGR storage because the fission products Se-79 and Sn-126 are long lived radio isotopes. The remaining fission product chlorides can then be released to the environment.
1. Dry chlorinate the fission products plus the zirconium to form a spectrum of chlorides.
2. Distil these chlorides at about 331 degrees C to selectively remove ZrCl4.
3. Send the remaining chlorides to 300 year interim storage.
4. Determine the weight of fission products flowing to 300 year storage.
5.Transmit the weight of fission products flowing into 300 year storage back to previous steps so that the weight of uranium flowing into the process equals the weight of fission products flowing into 300 year interim storage.
6. Reduce the ZrCl4 with sodium to obtain NaCl + Zr.
7. Send the recovered zirconium to the remote site zirconium store.
8. Electrolytically separate the NaCl into Na and Cl. Send the recovered Na and Cl to the appropriate remote site stores.
9. Note that NaCl recovered from the contents of used fuel tubes is contaminated by long lived Cl-36 and should be kept separate from the Cl used for Zr extraction.
Core Rod Formation and Transport:
1. Obtain high atomic weight metallic elements from the fission product separation process discharge stream.
2. Draw pure zirconium from the remote site zirconium store as required to achieve the desired core rod mass ratio.
3. Draw pure uranium from the remote site uranium store as required to achieve the desired core rod mass ratio.
4. Alloy together the high atomic weight transuranium elements, uranium and the zirconium.
5. Cast the FNR core rods.
6. Transport the FNR core rods to the FNR core rod store on the FNR site.
7. The transport of FNR core rods must be safely done in a manner such that no matter what transportation accident occurs the core rods cannot form a critical mass. The transportation container must be filled with enough B4C to prevent a nuclear reaction if water penetrates the transportation container.
Blanket Rod Formation and Transport:
1. Draw uranium from the remote site uranium store. This uranium may be cointaminated with Np.
2. Draw pure zirconium from the remote site zirconium store.
3. Form a 90% uranium 10% zirconium alloy. This alloy may contain contaminant Np.
4. Cast the uranium-zirconium alloy into blanket rods.
5. Transport the uranium-zirconium blanket rods to the blanket rod store on the FNR site. Note that if the uranium was previously in a FNR the blanket rods will have significant radioactivity.
FNR PASSIVE FUEL TUBE ASSEMBLY:
1. Draw passive fuel tube components from the local fuel tube store, local new blanket rod store, local sodium store.
2. Weld on fuel tube bottom plug. Pressure test bottom weld and tube.
3. Assemble the passive fuel tubes. Add a measured quantity of liquid sodium.
4. Orient and weld on top plug. Pressure test top weld and tube.
5. Send the assembled blanket type fuel tubes to the local finished passive fuel tube store.
FNR ACTIVE FUEL TUBE ASSEMBLY:
1. Draw active fuel tube components from the local tube store, local new blanket rod store, local new core rod store, local sodium store.
2. Weld on bottom plug. Pressure test bottom weld and tube.
3. Assemble the active fuel tubes. Add a measured quantity of liquid sodium.
4. Orient and weld on top plug. Pressure test top weld and tube.
5. Send the assembled active fuel tubes to the local finished active fuel tube store.
FUEL BUNDLE ASSEMBLY AND LOADING:
For utility size FNRs fuel bundle assembly is done on the reactor site. For low output SMRs fuel bundles may be assembled at a central location and then transported to the reactor site.
1. Draw fuel bundle components from their respective stores.
2. Assemble active fuel bundles and passive fuel bundles. Use an assembly jig that has sets of parallel steel sheets at 90 degrees to each other to position fuel tubes on the bottom grating.
3. Transport the fuel bundles to the FNR.
4. Load the fuel bundles into the reactor zone of the FNR primary sodium pool.
5. Slip a 15 inch square steel float over the indicator tube.
1. Run the reactor. While the reactor is running neutrons are net emitted by the core zone and are net absorbed by the blanket zones.
2. After each year in the reactor zone move (1 / N) of the fuel bundles to the liquid sodium pool perimeter zone, where N is the number of years per fuel cycle.
3. After a fuel bundle has been in the perimeter cooling zone for ~ 10 years extract the fuel bundle for reprocessing.
FNR FUEL BUNDLE RECYCLING:
1. Disassemble the irradiated fuel bundles in an argon atmosphere at a temperature above the melting point of sodium. Mechanically sort the fuel bundle into its miscellaneous steel, fuel tube, sodium, sodium salt, core rod, blanket rod and inert gas components. Each of these components is reprocessed differently.
2. Extract the inert gas.
3. Store the inert gas to allow it to naturally decay.
4. Send the steel components to interim storage prior to metal recycling. Note that the steel may require additional processing to remove surface residue.
5. Transport the FNR core rods to the remote site for further processing.
6. Transport the FNR blanket rods to the remote site for further processing.
7. After a suitable decay period vent the residue inert gas to the atmosphere.
8. Send the liquid sodium to sodium recycling.
9. Send the sodium salts for interim storage. After 300 years the fission product I-129 will have decayed but Cl-36 will remain. After 300 years it may be economic to separate the NaCl from the other salts to minimize the cost of long term DGR storage.
This web page last updated June 23, 2018
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