Материал: Крючков Фундаменталс оф Нуцлеар Материалс Пхысицал Протецтион 2011

Radiation factor f3 describes the radiation hazard of a Pu–bearing material as compared with Pu metal. The radiation factor of plutonium metal is taken to be unity.

The factors of density f1(Vsp) and activity f3(А) account for the difficulty of producing a Pu–bearing material, while the facto r of time f2(t) characterizes the difficulty of converting it to a nuclear explosive device.

The generalized factor of attractiveness of Pu–bear ing materials is defined as a product of the three above factors: f1(Vsp), f2(t), and f3(А). The attractiveness factors for different Pu-bearing materials are given in Table 2.4.

Advanced proliferation-resistant aqueous INF reprocessing technologies

SAFAR process of INF reprocessing. The key proliferation resistance idea of this technology lies in incomplete separation of uranium, plutonium and fission products.

SAFAR process as distinct from PUREX process:

1. Pu is not fully separated from U and FP. Pu and U are separated together only in two cycles of extraction, i.e. U and FP “impurities” are deliberately left in Pu ( 1 % of the starting quantity).

2. Pure U and Pu dioxides are not separated. MOX fuel particles are produced by the sol-gel method.

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3. High radioactivity of the end product. It is unattractive for diversion and affords easy control over fuel, but additional measures are required for radiation safety of personnel.

Sol–gel process

The starting material is an acid solution of INF after two cycles of FP separation. Subsequent operations are:

1.Addition of reagents to the solution to reduce acidity.

2.Addition of the solution to an inorganic material, e.g. to ethyl benzoate, for moisture absorption. The nitrite mixture (U,Pu)O2(NO3)2 is treated to produce (U,Pu)O2(OH)0.4(NO3)1.6, which is a colloid.

3.Colloid injection into an ammonia-based organic mixture which continues removing water from the colloid. The operation produces jelly-

like spherical granules (40–100 μm in size).

4. Heat treatment of granules at higher temperatures:

∙at 95 оС – ammonia detachment;

∙at 125–200 оС – water detachment and formation of (U,Pu)O 2(OH)4;

∙at 300–400 оС – evaporation of organic substances, granulation;

∙at 400–500 оС – baking of MOX fuel granules.

Non-aqueous (“dry”) INF reprocessing technologies

Pyrochemical fluoride-gas process. This process relies on the difference in boiling temperature, volatility and sorption capacity of U, Pu and FP fluorides. The boiling temperatures of U and Pu hexafluorides at atmospheric pressure are 56 оС and 62 оС. At such temperatures, the main fission products form non-volatile or low-volatile fluorides.

The main stages of the fluoride-gas process are:

1)thermal stripping of fuel claddings at 1600 оС;

2)fuel fluorination at 400 оС:

(U,Pu)O2 + 4 F2 + 3 H2 → (U,Pu)F6 + 2 HF + 2 H2O;

the bulk of FP fluorides remains in the non-volatile precipitate. The substances released are fluorides of U, Pu and some FP, and as fission gas (Xe, Kr, I);

3) freezing-out of FP fluorides at a desuperheating coil at 27 оС. A gas flow is fed into a cylindrical tank from the top, at an angle to its axis. Solid particles hit its walls and settle out;

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4)passing of a gas flow through sodium fluoride (NaF) granules. Use is made of the difference in the sorption capacity of NaF relative to U, Pu and FP fluorides; U, NP and Tc fluorides are sorbed on NaF granules at 100оС and Pu, Ru, Zr and Nb fluorides, at 400оС;

5)desorption of uranium hexafluoride (UF6) from granule surfaces by a mixture of “fluorine(10 %)–nitrogen(90 %)” at 400 оС.

The disadvantages of the fluoride-gas technology lie in the following:

∙removal of FP fluorides from UF6 is incomplete;

∙Pu is inferior to uranium in changing into volatile fluorides, and spreads over surfaces;

∙this technology is unfit for MOX fuel reprocessing due to a high concentration of plutonium.

Pyrometallurgical reprocessing. One of the options with this technology is the method of electrochemical refining.

The bottom part of an apparatus for electrochemical refining is filled with liquid cadmium (anode), which is overlaid by a molten salt layer (mixed K, Na, Ca and Ba chlorides). An iron cathode is inserted into the salt melt from above.

Electrochemical refining includes the following stages:

1)fuel rods are cut and loaded into a perforated graphite basket, which is then immersed in a layer of liquid cadmium;

2)fuel dissolves in cadmium, while the claddings remain in the basket;

3)fuel and fission products are distributed between Cd and molten salt:

∙fission gas and volatile FP pass from the melt to the gas blanket;

∙solid FP pass into molten salt;

∙U and Pu are found in both layers;

4)current passing. U, Pu and Zr go from liquid cadmium and molten salt to the iron cathode.

The cathode deposit is removed and remelted to become fresh fuel. Removal of FP from uranium and plutonium is improved by halide slagging. U and Pu are transformed into chlorides:

2(U,Pu) + 3MgCl2 → 3Mg + 2(U,Pu)Cl3,

which return into the molten salt. The coefficients of FP removal from fuel are small (102–10 3 as against 106–10 8 in the PUREX process).

DUPIC process. DUPIC (Direct Use of spent PWR fuel in CANDU) is one of the INF reprocessing options which features an improved proliferation resistance.

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Its purpose is to reuse INF of light water reactors, such as PWR, in heavy water reactors, such as CANDU. The irradiated fuel of PWRs is enriched in U to 0.9 % (h.a.) and has 0.6 % of Pu, which contains 70 % of fissionable isotopes. So, the content of all fissionable isotopes in INF is about 1.3 % (in terms of h.a.). CANDU reactors can run on natural uranium (with 0.7 % of 235U), i.e. the irradiated fuel of PWRs can be reused in CANDUs.

The DUPIC process includes:

1)dismantling of irradiated fuel assemblies, separation of fuel rods;

2)lateral cutting of fuel rods into pieces of 20 cm;

3)longitudinal cutting of claddings;

4)voloxidation, i.e. heat treatment in oxygen environment at 400 оС. Uranium dioxide (UO2) changes to U3O8, the fuel core becomes 30 % larger, and the fuel sheds its cladding. The fuel becomes porous and releases fission gas;

5)treatment by the OREOX process. OREOX is a redox process in which oxidation alternates with reduction of uranium oxides:

a) air oxidation at 450оС; UO2 is transformed into U3O8, as it is in voloxidation;

b) reduction in the environment of Ar+4 % H2 at 700оС; uranium octaoxide (U3O8) turns into UO2; repeated oxidation–reduction cycles lead to formation of dispersed UO2 powder and to release of all fission gas;

6)UO2 powder sintering to produce pellets;

7)fabrication of fuel rods and fuel assemblies by a standard process. The DUPIC process is distinguished by the following.

1. Absence of solvents leads to:

∙small volume of radioactive waste;

∙compact reprocessing facilities, which can share the site with a nuclear power plant.

2.No separation of U from Pu; incomplete removal of FP.

DUPIC offers improved proliferation resistance due to:

∙high radioactivity of fuel materials;

∙no separation of U from Pu;

∙freedom from long hauls, with a reprocessing facility built on the same site with the NPP.

Radioactive waste treatment technologies

All nuclear technologies are associated with radioactive materials either by using or by generating them. Fresh fuel assemblies of nuclear reactors

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contain radioactive isotopes of uranium, while irradiated fuel incorporates isotopes of uranium, plutonium, transuranic elements and fission products. Part of these isotopes can be extracted and put to use. The remaining radioactive material, the utilization of which seems unfeasible today, is regarded as radioactive waste (RW).

A feature specific to RW is the impossibility of eliminating it by traditional methods, such as incineration or conversion to another chemical form. Whatever such a form may be, RW retains its radioactivity. The only thing traditional methods can offer is to convert waste to a form suitable for its final disposal in geological repositories.

The greatest hazard for the biosphere comes from RW produced by chemical reprocessing of irradiated nuclear fuel. Fission products are removed from INF during its reprocessing. The content of FP is 30–40 kg/t in INF of thermal reactors and 100 kg/t in INF of fast reactors, with the corresponding FP activity making 6 MCi/t INF and 20 MCi/t INF, respectively.

For comparison:

∙the total release of radioactive material during the Chernobyl accident is estimated at 90 MCi;

∙the total radioactive release during the Kyshtym accident (explosion of liquid HLW storage facility) is estimated at 20 MCi;

∙RW activity on Rosatom sites (in 1990) was 2.3 GCi;

∙as of 1995, Russia had about 9400 t of INF with the total activity of 4.65 GCi (its average specific activity being 0.5 MCi/t). Considering that the INF will be reprocessed sooner or later and its activity will be passed on to waste, the potential total activity of RW in Russia is 7 GCi.

Radwaste treatment is undertaken to protect man and environment against its adverse impacts.

RW is classified by its aggregative state into liquid, gaseous and solid waste and by the specific activity into low-level, intermediate-level and high-level waste. Bearing in mind the main purpose of this chapter, discussion will be confined to the specifics of treating high-level and intermediate-level waste.

Treatment of high level waste (HLW) will be discussed in more detail. There are two major types of HLW:

∙HLW resulting from INF reprocessing. This is mainly liquid waste, as industrial fuel reprocessing is based on aqueous processes of NM extraction from solutions;

∙irradiated fuel assemblies (IFA) of power reactors.

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