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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The key stages of HLW treatment are:

1.Monitored interim storage:

∙Liquid HLW is kept in stainless steel tanks; monitoring covers heat release and gas blanket composition (hydrogen resulting from water radiolysis is removed);

∙IFAs are kept in water pools on NPP sites;

2.Evaporation of liquid HLW (with its volume reduced by a factor of 150 to 200). This process is accompanied by:

∙a rise in specific activity of concentrated HLW;

∙intensified gas formation due to water radiolysis, with the ensuing risk of hydrogen–air mixture explosion;

∙increase in specific heat release caused by natural decay of nuclides, with the ensuing rise of HLW temperature;

∙intensification of corrosive activity of HLW as concentration and temperature rise.

3.Solidification of HLW. The purpose of this stage is to implant HLW into a stable matrix which will prevent waste migration into the environment. Incorporation of HLW into glass (vitrification) is believed to be the most suitable form of its immobilization.

There exist two vitrification processes.

Single-stage process. Liquid concentrated HLW is poured into a crucible, and glass-forming agents are added to it. The mixture heating is accompanied by complete evaporation of moisture, baking of the resultant dry waste, and melting of the glass mass.

Once it is cooled down, the crucible together with its contents will be sent off for disposal.

Two-stage process:

a) calcination of the starting HLW at 300–400 оС;

b) mixing of the calcination product and the glass-forming additives, with the resulting mass transferred to a melting furnace;

c) heat-up and vitrification of the mass at 1100–11 50 оС;

d) periodic discharge of the glass mass into steel containers; e) interim storage and disposal of the containers.

There are alternative technologies of HLW immobilization. They consist in incorporation of this waste into other stable materials, such as ceramics, glassceramics, mineral-like materials, e.g., SYNROC.

SYNROC is an abbreviation of Synthetic Rock. Creation of synthetic rock for HLW immobilization is based on the assumption that such material will be as stable and durable as natural rock.

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Treatment of liquid intermediate-level waste

Intermediate-Level Waste (ILW) includes solutions from extraction cycles (except for the first cycle), condensate resulting from evaporation of low-level waste, and vapor produced in evaporation of high-level waste.

The main stages of ILW treatment are:

1)precipitation and removal of ILW from liquid phases (settling and filtering with the use of coagulants);

2)ion-exchange treatment of the remaining liquid;

3)evaporation to produce solid residue;

4)immobilization by bituminization (the material is mixed with bitumen mass and is left to solidify);

5)containerization of the bituminous mass with ILW;

6)interim storage and final disposal.

Bitumen as an immobilization material has the following merits:

·low water leaching;

·suitability for all chemical forms of ILW (salts, hydroxides, organic compounds);

·good radiation resistance.

The disadvantages of bitumen are its combustibility (as a product of oil refining) and softening when heated (asphalt).

An alternative to ILW immobilization is its cementation, i.e. incorporation into concrete. Concrete as an immobilization material is appreciated for its:

·low cost and simplicity of handling;

·high radiation resistance;

·high heat conductivity;

·being neither combustible nor softening when heated.

However, concrete does not have sufficient chemical resistance to water. Various rates of water leaching are given below for comparison of several materials:

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That is why glass and SYNROC are predominantly used for immobilization of high-level waste and bitumen and concrete for immobilization of intermediateand low-level waste.

A fuel cycle ends with final disposal of special casks with RW in deep, stable geological formations. Common candidates for this purpose are: salt deposits, clayey sedimentary and other rocks.

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CHAPTER 3

INTERNATIONAL NON-PROLIFERATION SAFEGUARDS

3.1. IAEA activities

The global problem of non-proliferation is being dealt with by the International Atomic Energy Agency (IAEA [1]). Using a well-developed system of measures, referred to as the International Safeguards, the IAEA seeks to ensure that nuclear materials (NM) are used exclusively for peaceful purposes.

The system of International Safeguards is based on the Treaty for the Non-Proliferation of Nuclear Weapons (NPT) and is described in the IAEA document INFCIRC/153. It relies largely on the national systems of safeguards, making it possible to exercise control over all NM in the states

– signatories to the NPT.

Organization of international control over nuclear materials

The IAEA, established in Vienna in 1957, is an international intergovernmental institution related to the United Nations Organization through a special agreement and is thereby made part of the UNO system. This agreement states that the Agency acts as an independent international organization having a working relationship with the UNO. The IAEA reports its activities to the UN General Assembly and cooperates with the Security Council by providing it with the information and assistance that may be required for the Council to perform its functions of keeping peace and security in the world.

The Agency’s objectives are defined in its Statute:

∙It shall seek to accelerate and enlarge the contribution of atomic energy to peace, human health and prosperity throughout the world;

∙The Agency shall encourage research and development in the field of peaceful energy uses, and shall provide materials, services, equipment and facilities to this end;

∙The IAEA shall establish and apply safety rules in connection with the use of nuclear energy and radiation;

∙The Agency shall be authorized to establish and implement safeguards for exclusively peaceful uses of nuclear materials.

As of 2007, the IAEA had 144 states as its members. Its steering entities are the General Conference and the Board of Governors.

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The General Conference is the highest authority of the Agency. Its sessions are held once a year. Most of its decisions and resolutions are made by a simple majority vote.

The Board of Governors guides the IAEA work in between the General Conference sessions. Its decisions on the main aspects of the Agency’s activities are subject to approval by the General Conference. The Board of Governors has 35 members. Its permanent members are ten states whose nuclear science and engineering are best developed, namely: Russia, China, USA, UK, France, Germany, Japan, Italy, Canada, and India. Besides, every year, three more states most advanced in the field of nuclear technology are appointed to the Board for a term of one year. All other members are elected for two years to represent states, other than the above ten, from seven regions: Latin America, Western Europe, Eastern Europe, Africa, Middle East and South Asia, South-East Asia and the Pacific, and Far East.

Besides, the Board of Governors will periodically set up ad hoc committees to address major specific matters. Thus, in view of the NPT coming into force in 1970, the Board appointed a committee for development of a standard safeguards agreement, which came to be referred to as document INFCIRC/153.

All regular practical activities of the Agency are carried out by its Secretariat. The IAEA Secretariat relies on the guidance and leadership of its Director General who is responsible for fulfillment of the Agency’s work program. At present, the post of Director General is held by Yukiya Amano from Japan.

The administrative structure of the IAEA is similar to that of a large research center (Fig. 3.1) and is tailored to its purposes. Its six major departments deal with: technical cooperation; nuclear energy; nuclear safety and security; management; nuclear sciences and applications; and safeguards. The departments consist of divisions, each subdivided into sections to address specific subjects.

Main areas of IAEA activities

All the variety of IAEA activities can be divided into three general areas: control over nuclear materials; technical assistance to countries; and scientific programs. Out of the six departments, three are responsible for implementation of scientific programs:

The departments for nuclear energy and for safety and security pursue studies in two directions:

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