Fuel reprocessing

Fuel reprocessing has three objectives (a) to recover U or Pu from the spent fuel for reuse as a nuclear reactor fuel or to render the waste less hazardous, (b) to remove fission products from the actinides to lessen short-term radioactivity problems and in the case of recycle of the actinides, to remove reactor poisons, and (c) to convert the radioactive waste into a safe form for storage. Fuel reprocessing was/is important in the production of plutonium for weapons use. 

The Purex process is used for almost all fuel reprocessing today. Irradiated UO2 fuel is dissolved in HNO3 with the uranium being oxidized to U02(N03)2 and the plutonium oxidized to Pu(NC 3)4. A solution of TBP in a high-boiling hydrocarbon, such as n-dodecane, is used to selectively extract the hexavalent U02(N03)2 and the tetravalent Pu(NC 3)4 from the other actinides and fission products in the aqueous phase. The overall reactions are 

The only fission fragments that extract during the Purex process are Zr, Ru, Nb, and Tc, with the most troublesome being Zr and Ru. Zr forms a number of complex species with the most important being [Zr(N03)4 2TBP]. The formation of this complex is inhibited by the addition of F whereby 

Feed to solvent extraction High-level aqueous waste 

In the dissolution step, the fuel pieces are dissolved in near boiling 10 M HNO3. This step, which takes a few hours, dissolves the uranium, plutonium, and fission products, leaving the cladding to be recovered. The Kr and Xe are recovered from the off-gas of steam, air, and NO. The chemical reactions for the dissolution of uranium involve processes like 

Because of the importance of reactor fuel reprocessing in nuclear power technology, some further discussion of this topic is warranted in this introductory chapter. 

In addition to fissionable isotopes ( U, or plutonium) and fertile isotopes ( U or thorium), spent fuel from a reactor contains a large number of fission product isotopes, in which all elements of the periodic table from zinc to gadolinium are represented. Some of these fission product isotopes are short-lived and decay rapidly, but a dozen or more need to be considered when designing processes for separation of reactor products. The most important neutron-absorbing and long-lived fission products in irradiated uranium are listed in Table 1.4. 

Technetium Rhodium Xenon Neodymium Samarium Europium Gadolinium 99 103 131, 133, 135 143, 145 149, 151 155 155 Molybdenum Technetium Ruthenium Rhodium Tellurium Iodine Xenon Cesium 99 99 103, 106 106 129 129, 131 133 137 

The principle of the Purex process, now commonly used for processing irradiated uranium by solvent extraction, is illustrated in Fig. 1.18. The solvent used in this process is a solution of tributyl phosphate (TBP) in a high-boiling hydrocarbon, frequently n-dodecane or a mixture of similar hydrocarbons. TBP forms complexes with uranyl nitrate [U0i(N03)2] and tetravalent plutonium nitrate [Pu(N03)4] whose concentration in the hydrocarbon phase is higher than in an aqueous solution of nitric acid in equilibrium with the hydrocarbon phase. On the other hand, TBP complexes of most fission products and trivalent plutonium nitrate have lower concentrations in the hydrocarbon phase than in the aqueous phase in equilibrium. 

In the Purex process, irradiated UO2 is dissolved in nitric acid under such conditions that uranium is oxidized to uranyl nitrate and plutonium to Pu(N03)4. The resulting aqueous solution of uranyl, plutonium, and fission-product nitrates is fed to the center of countercurrent solvent extraction contactor I, which may be either a pulse column or a battery of mixer-settlers. This contactor is refluxed at one end by clean solvent and at the other by a dilute nitric acid scrub solution. The solvent extracts all the uranium and plutonium from the aqueous phase and some of the fission products. The fission products are removed from the solvent by the nitric acid scrub solution. Fission products leave contactor I in solution in aqueous nitric acid. 

---

If the spent fuel is processed in a nuclear fuel reprocessing plant, the radioactive iodine species (elemental iodine and methyl iodide) trapped in the spent fuel elements ate ultimately released into dissolver off gases. The radioactive iodine may then be captured by chemisorption on molecular sieve 2eohtes containing silver (89). 

Nuclear Fuel Reprocessing. Spent fuel from a nuclear reactor contains Pu, Th, and many other radioactive isotopes (fission... 

J. T. Long, Pngineeringfor Nuclear Fuel Reprocessing, Gordon and Breach, Inc., New York, 1967. 

Hafnium neutron absorption capabilities have caused its alloys to be proposed as separator sheets to allow closer spacing of spent nuclear fuel rods in interim holding ponds. Hafnium is the preferred material of constmction for certain critical mass situations in spent fuel reprocessing plants where hafnium s excellent corrosion resistance to nitric acid is also important. 

Lead bricks are generahy used as temporary shields for radiation sources at nuclear power stations, research institutes, hospitals, and fuel reprocessing plants. Plat, rectangular bricks requite a double layer with staggered seams whereas the interlocking bricks requite only one course. Lead shot can be poured into inaccessible areas like a Hquid. 

As of 1995, there were no nuclear fuel reprocessing plants operating in the United States. Other nuclear nations have constmcted second- or third-generation reprocessing faciUties. These nations have signed the nuclear nonproliferation treaty, and the faciUties are under the purview of the International Atomic Energy Agency (IAEA). 

By contrast, HLW from LWR fuel reprocessing is stored ia cooled, well-agitated, stainless steel tanks as an acidic nitrate solution having relatively few sohds. Modem PUREX flow sheets minimise the addition of extraneous salts, and as a result the HLW is essentially a fission-product nitrate solution. Dissolver soHds are centrifuged from the feed stream and are stored separately. Thus the HLW has a low risk of compromising tank integrity and has a favorable composition for solidification and disposal (11). 

D. W. HoWid.2cy, A Eiterature Survey Methodsfor the Removal of Iodine Spedafrom Off-Gas andEiquid Waste Streams of Nuclear Power and Nuclear Fuel Reprocessing Plants, with Emphasis on Solid Sorbents, ORNL/TM-6350, Oak Ridge National Laboratory, Oak Ridge, Term., 1979. 

The preparation of used oil for appHcation as a fuel. Reprocessing may involve settling, filtration, and blending. This term has not been defined by... 

Carbonates. Actinide carbonate complexes are of interest not only because of their fundamental chemistry and environmental behavior (150), but also because of extensive industrial appHcations, primarily in uranium recovery from ores and nuclear fuel reprocessing. 

Fullwood and Erdman, 1983 circumvent this problem by comparing risk as cubes in which linear dimensions are the cube root of the volume/risk. Figure 1.4.3-5 compares the risks associated with nuclear fuel reprocessing, refabrication and waste disposal with nonnuclear risks. 

It is also invaluable in separating U from Pu and other fission products during nuclear fuel reprocessing, since Pu reacts only to give the (involatile) PUF4 and most fission products... 

Since the amount of fissile material in the fuel assemblies is only about 3 percent of the uranium present, it is obvious that there cannot be a large amount of radioactive material in the SNF after fission. The neutron flux produces some newly radioactive material in the form of uranium and plutonium isotopes. The amount of this other newly radioactive material is small compared to the volume of the fuel assembly. These facts prompt some to argue that SNF should be chemically processed and the various components separated into nonradioac-tive material, material that will be radioactive for a long time, and material that could be refabricated into new reactor fuel. Reprocessing the fuel to isolate the plutonium is seen as a reason not to proceed with this technology in the United States. 

Reported plant applications of a.c. impedance and electrochemical noise are rare, but include stainless steels in terephthalic acid (TA) plant oxidation liquors , nuclear fuel reprocessing , and fluegas desulphurisation (FGD) scrubber systems . 

Other reasons for investigating plutonium photochemistry in the mid-seventies included the widely known uranyl photochemistry and the similarities of the actinyl species, the exciting possibilities of isotope separation or enrichment, the potential for chemical separation or interference in separation processes for nuclear fuel reprocessing, the possible photoredox effects on plutonium in the environment, and the desire to expand the fundamental knowledge of plutonium chemistry. 

The possible application of aqueous plutonium photochemistry to nuclear fuel reprocessing probably has been the best-received justification for investigating this subject. The necessary controls of and changes in Pu oxidation states could possibly be improved by plutonium photochemical reactions that were comparable to the uranyl photochemistry. 

A second source of plutonium, dispersed more locally, is liquid effluent from fuel reprocessing facilities. One such is the fuel reprocessing plant at Windscale, Cumbria in the United Kingdom where liquid waste is released to the Irish Sea(6). Chemical analysis of this effluent shows that about one percent or less of the plutonium is in an oxidized form before it contacts the marine water(7). Approximately 95 percent of the plutonium rapidly adsorbs to particulate matter after discharge and deposits on the seabed while 5 percent is removed from the area as a soluble component ). Because this source provided concentrations that were readily detected, pioneering field research into plutonium oxidation states in the marine environment was conducted at this location. 

An interesting aspect of the characterization of plutonium as Pu(V) in the Irish Sea, Lake Michigan, and Pond 3513 is that the origins of the radionuclides are different in each system, i.e., fuel reprocessing waste, fallout, and laboratory effluents, respectively. 

Early experimental work in electrorefining at Los Alamos by Mullins et-all ) demonstrated that americium could be partitioned between molten plutonium and a molten NaCl-KCl salt containing Pu+3 ions, and Knighton et-al(8), working at ANL on molten salt separation processes for fuel reprocessing, demonstrated that americium could be extracted from Mg-Zn-Pu-Am alloys with immiscible molten magnesium chloride salts. Work... 

Bond, W. Leuze, R. "Feasibility Studies of the Partitioning of Commercial High-Level Wastes Generated in Spent Nuclear Fuel Reprocessing Annual Progress Report for FY-1974," ORNL-5012, Oak Ridge, Tennesse, January 1975. 

Research should continue on traditional separation methods. For example, there is a continuing need for more selective extraction agents for liquid-liquid and ion-exchange extractions. High-temperature processes that use liquid metals or molten salts as extraction agents should have potential in nuclear fuel reprocessing and... 

---