Atomic Energy, Vol. 95, No. 3, 2003
MONONITRIDE FUEL FOR FAST REACTORS
B. D. Rogozkin,1 N. M. Stepennova,1 and A. A. Proshkin2
UDC 621.039.526
Substantiation is given for the development of nuclear power based on inherently-safe fast reactors with a mononitride core. Fundamental studies and design work on the development of such reactors with lead (BREST-OD-300), lead–bismuth (SVBR-75/100), and sodium coolant (BN-800) are being performed. The development of nuclear power in our country is based on organizing a closed fuel cycle. The results of experimental investigations of the properties of mononitride fuel are correlated. Mononitride fuel meets all requirements for fast-reactor fuel.
Large-scale development of nuclear power requires solving several basic problems including the following: • development of inherently safe fast reactors, including with excess reactivity and a reactivity effect at a level of fractions of delayed neutrons, which eliminates the possibility of serious reactivity accidents; • increasing the utilization efficiency of natural uranium approximately by a factor of 100; • efficient utilization of plutonium, accumulated in spent nuclear fuel and released from nuclear weapons, and uranium tailings; • burning of actinides, transmutation of long-lived fission products, and burial of radioactive wastes without destroying the natural level of radioactivity; • achievement of cost-effectiveness; • resistance of nuclear power production technology to the proliferation of nuclear weapons. These problems can be solved by nuclear power based on fast reactors because of their unique neutron excess. As a result of the accumulation of thousands of metric tons of plutonium in spent nuclear fuel from nuclear power plants, it is not important to reach in the next 30 yr high breeding ratios and multiplication rates. This makes fast reactors more cost effective and safer. Fast reactors with a mononitride core are the key to the solution of these problems. In our country, fundamental research and design work is being done to develop inherently safe fast reactors with mixed nitride fuel with lead (BREST-OD-300), lead–bismuth (SVBR-75/100), and sodium (BN-800) coolants [1–5]. The use of mononitride uranium–plutonium fuel with high density – 90 ± 5% of the theoretical value – and high thermal conductivity – 10–15 times higher than for oxide fuel in fast reactors – gives a breeding ratio in the core ≈1.06 without a uranium blanket, decreases the power effect, the burnup effect, and the total excessive reactivity to a safe level, stabilizing the power distribution and eliminating weapons-grade plutonium production. The low fuel temperature prevents overheating and destruction of the claddings in a loss-of-coolant accident and when reactivity is introduced. When high 1
Federal State Unitary Enterprise A. A. Bochvar All-Russia Scientific-Research Institute of Standardization in Machine Engineering. 2 Institute of Nuclear Reactors, Russian Science Center Kurchatov Institute. Translated from Atomnaya Énergiya, Vol. 95, No. 3, pp. 208–221, September, 2003. Original article submitted November 22, 2002. 624
1063-4258/03/9503-0624$25.00 ©2003 Plenum Publishing Corporation
Mononitride fuel fabrication from oxides
Mononitride fuel fabrication from Pu, U, and U–Pu alloy
Assembly of fuel elements and fuel assemblies Uranium tailings BREST-OD-300 and BN-800 fast reactors
Uranium and plutonium oxides
Fuel reprocessing
Hydrometallurgical reprocessing, purex process
U–Pu–Np, Am, Cm
Electrochemical reprocessing in fused salts
Spent fuel
VVÉR
Fig. 1. Closed fuel cycle scheme.
breeding rates are not needed, moderate energy density of the fuel will increase the reliability of the fuel, the degree of burnup, and the stability during accidents. In the BREST-OD-300 reactor, the amount of plutonium burned will be approximately the same as that produced from 238U. Consequently, when fuel is refabricated, after reprocessing only 238U with no added plutonium needs to be added to the final product. Removal of the uranium blanket and replacing it with a lead reflector make the void effect of reactivity negative and eliminate the production of weapons-grade plutonium [1–4]. The use of a blanket-free nitride core in BN-800 also makes this reactor safer, makes the control system more reliable, and satisfies the requirements for nonproliferation of fissioning materials [5]. In our country and abroad, processes are being developed to obtain mixed mononitride fuel from initial alloys of uranium with plutonium [6] and their oxides. In other countries, preference is given to obtaining such fuel from the initial oxides [7–21]. The cores fabricated using as initial materials oxides and metals are characterized by a high degree of homogeneity. Investigations of the processes leading to the production of mixed nitrides from the initial oxides and metals (alloys) have shown that high-purity mononitride fuel can be fabricated by periodic and continual methods on a laboratory scale. The results obtained serve as a basis for developing a technology and the required equipment for remote-control fabrication of fuel from initial oxides and metals (alloys) obtained after reprocessing. It is suggested that a mixed nitride, obtained from oxides of power-generation plutonium and uranium, be used for the first few loads of BREST-OD-300 and BN-800 cores. The stores of oxide of power-generation plutonium, already separated and accumulated in the spent fuel, will make it possible to operate nuclear power plants wtih a total capacity of hundreds of GW with fast reactors with no fuel limitations. The strategy for development of nuclear power in Russia is based on organizing a closed fuel cycle (Fig. 1). In some works of American, Russian, and Japanese specialists, it is indicated that from the standpoint of safety, ecology, and nonproliferation a nuclear fuel cycle at the plant is desirable [1–4, 6, 22–26]. This will make it possible to decrease radically the 625
duration of the fuel cycle, storage volumes, and oppositely-directed shipments of high-level spent and fresh fuel, to decrease the risk associated with these operations, to localize fuel production, and to make possible combined circulation of uranium, plutonium, and highly-active actinides and burial of wastes, and thereby to make nuclear power safer. A drawback of mononitride is the production of ecologically dangerous 14C via the reaction 14N(n, p)14C. The contribution of 14C to the total radiotoxicity of the wastes is about 1%, which is not significant when carbon is bound into stable compounds. The production and atmospheric emission of volatile carbon compounds need to be prevented when spent fuel is reprocessed. Nitrogen enriched with 15N can be used in nitride fuel; this will decrease the amount of 14C which is produced and will conserve neutrons and fuel and substantially compensate the cost of nitrogen enrichment. We shall now present the basic properties of mononitride fuel. In modern nuclear power, oxide fuel has been studied much more than nitride fuel. Consequently, there is less knowledge about oxide fuel. However, the basic data on the properties and technology and from radiation tests have made it possible to start the design of fast reactors with such fuel. This has been facilitated by more than 18 years of successful operation of the BR-10 reactor with a mononitride uranium core [27–29]. It should be noted that nitride fuel has for a long time been regarded as a means for achieving high energy density, which together with a high breeding ratio results in a short plutonium doubling time. Consequently, tests have been performed primarily under high heat loads (up to 1000 W/cm and higher) and high core temperature – up to 1600°C and higher. At the present time, for the moderate heat loads which are currently used the maximum core temperature depends on the use of a gas (~1400°C) or metal (≤900°C) sublayer. In these cases, the fuel temperature which is reached in accidents for a short time differs substantially. Since the properties and behavior of nitride fuel strongly depend on the temperature, the correct use of experimental data in designs requires that the temperature conditions of the tests be carefully satisfied. The choice of a temperature regime largely determines the structural implementation and efficiency of the fuel elements. In the fast lead- and sodiumcooled reactors (BREST-OD-300 and BN-800, respectively) which are now being developed it is proposed that fuel elements with mononitride fuel and liquid-metal (Pb, Pb–Bi, Na) and helium sublayers be used. Of course, the temperature conditions for fuel operation in these fuel elements will be different. Consequently, in the present paper an attempt is made to correlate data on the basic properties in a wide temperature range. The most important characteristics of mononitride fuel are its creep, swelling, and gas release under irradiation, compatibility with structural steel and coolants, and thermal stability. Creep. Thermal creep of uranium mononitride and mixed mononitride has been studied in [2, 27, 30–33]; radiation creep has been much less studied [8, 34–41]. The investigators noted that the content of oxygen and carbon in the nitride, grain size, and porosity influence creep. The equation for the rate of steady creep for U0.8Pu0.2N with density 13–13.25 g/cm3, where the mass fraction of oxygen and carbon does not exceed 0.15%, is ε0 = 308σ1.35exp(–40000/RT),
(1)
where σ is the stress ranging from 10 to 60 MPa [31, 37]. The influence of porosity on the rate of thermal creep of fuel is given by the relation ε0 = ε00(1 + 0.125ρ2),
(2)
where ε00 is the creep velocity of pore-free nitride and ρ is the porosity of the fuel (%). It should be noted that according to this equation a change in fuel porosity from 5 to 20% increases the creep rate by approximately a factor of 12. On the basis of relation (2) and the fact that Eq. (1) was obtained for fuel with 9% porosity, the equation of thermal creep for U0.8Pu0.2N becomes ε0 = 27.7σ1.35exp(–40000/RT)(1 + 0.125ρ2).
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Fig. 2. Radiation creep of uranium mononitride fuel with density 96% of the theoretical value and stress 20 MPa, normalized to 2.5·1014 cm–3·sec–1: é, ÿ) [8], [36], respectively.
Fig. 3. Creep of uranium mononitride (1) and oxide (2) fuel with stress 20 MPa, normalized to 1·1013 m–3·sec–1.
The dependence of the rate of thermal creep of uranium mononitride on temperature and stress with σ = 20–35 MPa can be described by the equation [30] ε0 = 2.05·10–3σ4.5(–39395/RT). Radiation creep of fuel can be represented in the form ε0 = Aƒσ, where A is a constant and ƒ is the fission intensity in the fuel (cm–3·sec–1). According to data for uranium mononitride, the constant A = (0.4–0.6)·10–21 [(fissions/cm3)MPa]–1 [8, 36, 41]. Figure 2 shows the experimental data on the in-reactor creep rate of mononitride fuel with density equal to 96% of the theoretical value for fission rate 2.5·1014 cm–3·sec–1 and stress 20 MPa [8]. Figure 3 illustrates the in-reactor creep of UN and UO2, normalized to f = 1·1013 cm–3·sec–1 and stress 20 MPa [36]. According to Fig. 3, the radiation component of the creep of oxide fuel is 10 times higher than that of nitride fuel; this is explained by the higher rate of suppression of 627
Fig. 4. Swelling of nitride fuel versus burnup at 1680°C and initial fuel density 87% of the theoretical value [43] (1), 1400°C and 93% [15] (2), 1190°C and 95% [44] (3), 550–900°C and 84–94% [28, 29] (4).
Fig. 5. Swelling of nitride fuel versus burnup, temperature, and oxygen mass fraction: ÿ, é) <0.05 and >0.3% by mass at <1260°C; ñ) 1520, 1470, 1550, 1455, and 1390°C.
thermal peaks in the nitride fuel, whose thermal conductivity is higher. Increasing the fission intensity from 1·1013 to 2.5·1014 cm–3·sec–1 increases the rate of creep of nitride fuel by more than a factor of 10 (from ~5·10–7 to 1·10–5 h–1) [8, 36]. Other studies confirm this result [2, 37]. Radiation Swelling. It depends on the temperature, burnup, carbon and oxygen content, density, and open and closed porosity. The investigations were performed under various conditions taking account of the possibility of using fuel in fast reactors and space-based reactors [2, 8, 12, 15, 17, 27–29, 31, 37, 38, 41–54]. In fuel elements with a helium sublayer, the temperature at the center of a core is much higher than in fuel elements with a liquid-metal sublayer (Pb, Pb–Bi, Na, and others); this also predetermines the different swelling [8–12, 18, 27, 31, 42, 49, 50–51].
628
A special feature of nitride and carbide-nitride fuel, in contrast to oxide fuel, is the higher degree of confinement of gaseous fission products and chemically active elements, such as, cesium, iodine, selenium, tellurium, and others. This decreases the chemical interaction of corrosive fission products with fuel-element claddings. There are grounds for believing that this is due to the lower temperature of nitride fuel and easy occurence of reactions leading to the formation of chemical compounds with other elements of fission products. Figure 4 shows the swelling of irradiated nitride fuel, containing from 0.3 to 0.5% oxygen and carbon each, as a function of burnup at various temperatures at the center [15, 28, 29, 43, 44]. Figure 5 shows the swelling of nitride fuel as a function of burnup, temperature, and oxygen content [8, 18, 45]. The average rate of free swelling increases with fuel temperature. The experimental results are explained by the fact that at high temperature the creep rate of nitride fuel is quite high. For example, at Tc ~ 1600 C it is approximately 1000 times higher than at Tc ~ 1000°C, and adding gaseous fission products into fuel pores produces greater changes in the volume of the plastic fuel matrix with low burnup. This is confirmed by computational estimates of free swelling of fuel based on the spherical-pore model [55]. Thus, the rate of free swelling of nitride fuel until fuel–cladding contact occurs depends strongly on temperature. At contact, the rate of gaseous swelling of the fuel decreases. The decrease in swelling of the fuel and contact pressure are determined by the characteristics of creep, the swelling of the cladding and fuel material, and the mechanical properties of the latter. In almost all works on irradiation of nitride fuel, it is noted that the oxygen and carbon content strongly influences swelling if the mass fraction of each element exceeds 0.1–0.15%. Some experimental data on radiation swelling of nitride fuel were obtained at fuel-element cladding temperatures ~1300°C for setups to be used in space. On the basis of the data of [46], the following relations are recommended for the dependence of fuel swelling on temperature, burnup, density, and average rate of swelling S on ∆V/V at 1% burnup h.a.: ∆V/ V = 1.16·10–8T 2.36B0.82ρ0.5; S 0 = 1.16·10–8T 2.36ρ0.5B–0.18, where B is the fuel burnup, % h.a.; T is the maximum temperature of the fuel core, K; ∆V/V is the fuel volume change, %, ρ is the density, % of the theoretical value. These relations hold in the ranges 700°C < T < 1650°C, 80 ≤ ρ ≤ 95% of the theoretical value, 1% h.a. ≤ B ≤ 10% h.a. These relations can be used to determine the swelling of fuel up to the moment that fuel–cladding contact occurs. At contact, the fuel swelling follows the spherical-pore model, taking account of the external contact pressure, which is determined by solving the thermomechanical problem of fuel–cladding interaction [55–57]. In tests of mixed fuel U0.8Pu0.2N, containing 0.1% oxygen and carbon each, in a BOR-60 reactor with heat density 1045 W/cm, maximum temperature at the center of the fuel up to 2000°C, and burnup 4% h.a., the swelling was 1.5–1.7% at 1% burnup [31]. The presence of a sodium sublayer (BN-800) or lead sublayer (BREST-OD-300) in the fuel elements makes it possible to lower the temperature at the center of the fuel to 1000°C and lower; this makes it possible to ensure, over the planned service life of a fuel element, a fuel–cladding technological gap. At this temperature the free swelling of fuel at the center of the nitride fuel does not exceed 1–1.2% ∆V/V at 1% burnup h.a. This makes it possible in principle to reach burnup up to 15% h.a., since the fuel-element cladding in this case will be loaded only by the resulting pressure of the coolant and the gaseous fission products beneath the cladding. Generalizing the experimental results, it can be concluded that at temperatures up to 1200°C at the center of the fuel the average rate of swelling does not exceed 1.5%/% burnup. The temperature at the center of nitride fuel of a BREST-OD-300 reactor with a lead sublayer will not exceed 1000°C, so that the swelling will be only 1%/% burnup. Tests of fuel elements with cores with a special structure consisting of mixed mononitride fuel U0.8Pu0.2N with density 85% of the theoretical value and helium sublayer in a JMTR reactor with heat density 700 W/cm up to burnup 5.5% h.a., have shown that the radiation swelling was less than 1%/% burnup [42]. Japanese specialists attribute the low swelling to the special structure of the cores, where a high-density matrix surrounds large pores.
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Fig. 6. Emission of gaseous fission products with initial fuel density 87% of the theoretical value versus burnup with irradiation temperature at the center 900°C [28, 29] (1), 1680°C [43] (2), 1400°C [15] (3), 1190°C [44] (4).
Emission of Gaseous Fission Products. The results of investigations of gas emission from nitride fuel, which were performed in our country, France, Japan, the US, and Great Britain, on the whole agree with one another [2, 7–13, 27–29, 31, 37, 40–42, 47, 48, 58]. Gas emission depends, first and foremost, on the average temperature at the center of nitride fuel, burnup, oxygen and carbon content, density, and open porosity. Tests of two zones in the BR-10 reactor with mononitride uranium fuel up to burnup 8–9% h.a. showed the following: • gas emission from fuel with oxygen mass fraction 0.4–0.5%, carbon mass fraction 0.35–0.45% with temperature 900°C at the center of the fuel core and a linear power density 450 W/cm up to burnup 8–9% h.a. was about 25% of the total amount of gaseous fission products formed; • with temperature 900°C at the center, the gas emission from the fuel, where the mass fraction of oxygen and carbon is 0.1%, was 20–22%; • no iodide and cesium corrosion of the ÉI 847 steel cladding was observed; cesium becomes concentrated at the top of the reactor core. As the temperature at the center of the fuel increases to 1400°C, the gas emission up to burnup 7.5% h.a. is 26–27% [47, 48]. Radiation tests in the BOR-60 reactor [2, 31] of mixed mononitride fuel, obtained from metals (oxygen in carbon ≤0.15%) with linear power density 1000 W/cm and maximum temperature at the center of the fuel cores up to 2000°C up to burnup 4% h.a. showed that gas emission in these experimental conditions was 42% of the total amount of gaseous fission products produced. The composition of the gas under the cladding (in mass fractions) was as follows (%): He 16, CO + N2 0.44, Ar 0.22, O2 0.11, CO2 0.025, Kr 5.69, Xe 77.26. Radiation studies of nitride fuel, irradiated in thermal and fast reactors, did not show any appreciable iodine, cesium, tellurium, and selenium corrosion of the interior surfaces of the fuel element claddings consisting of austenitic and ferrite steels [2, 8–12, 28, 29, 31, 42]. For other conditions remaining the same, the emission of gaseous fission products from nitride fuel is less than oxide fuel. As a result of irradiation of U0.8Pu0.2N with density 85 ± 2% of the theoretical value, in a JMTR reactor with heat density 700 W/cm up to burnup 5.5% h.a., the gas emission did not exceed 3% [42], which is much less than that obtained in other investigations and it is explained by the special structure of the fuel cores. In our opinion, the low carbon and oxygen content also had an effect on the decrease of gas emission. Figure 6 displays the generalized curves of the gas emission from nitride fuel versus burnup with irradiation in different reactors [15, 43, 44]. An athermal relation is proposed in [47] for the emission of gaseous fission products at low fuel temperature (below 1000°C): F = 0.331 – 0.496B + 0.409B 2, where B is fuel burnup, % h.a. (up to 10% h.a.). 630
Fig. 7. Emission of gaseous fission products versus burnup at fuel temperature 500 (1), 700 (2), 1000 (3), 1200 (4), 1500°C (5).
However, this relation neglects the influence of temperature. The authors, using as a basis an analysis of experimental data on the emission of gaseous fission products from nitride fuel during radiation tests in foreign reactors and in the BR-10 reactor, have made the first attempt to represent their emission as a function of burnup and temperature. The emission of gaseous fission products is a complex process, which depends on the burnup, porosity, and fuel temperature. The experimental data are an integral characteristic, obtained with variable fuel temperature along the radius and the height of a fuel element, and they show a substantial variance under identical average thermal loads and burnup; this is due to the different technology used to fabricate fuel, the special structural features of a fuel element, and, most importantly, the operating conditions of the experimental reactors, determining the temperature regime. Analysis of data on the irradiation of fuel elements with nitride fuel in BR-10, BOR-60, and EBR II and in reactors used in space gave the relation F = 3.05B 1.92exp(–4130/RT), where R = 1.98 cal/(mole·deg). This relation holds at 600°C < T < 1600°C. 2% ≤ B ≤ 10% h.a., oxygen and carbon mass fraction 0.2–0.4%. It makes it possible to calculate the emission of gaseous fission products taking account of temperature gradients along the height and radius of the fuel. Specifically, it describes satisfactorily the kinetics of emission of gaseous fission products from fuel elements with nitride fuel irradiated in a BR-10 reactor, taking account of the special operating features of this reactor (Fig. 7). As additional data are obtained, this equation can be adjusted. It can be recommended for preliminary estimates of the emission of gaseous fission products from nitride fuel in BREST fast reactors. The following conclusions can be drawn on the basis of the experimental radiation studies of mononitride fuel: • the emission of corrosive fission products Cs, Se, I, and Te is much less than from oxide fuel; this is probably one reason for the much weaker corrosion of the cladding on the fuel side; • gas emission depends on the temperature at the fuel center and varies from 22–25% at 900–950°C up to 70% at 1600°C, attainment of burnup 9–10% h.a., and the oxygen and carbon concentrations; • the gases emitted under the cladding at temperatures at the center of nitride fuel from 900 to 1800°C during reactor tests in BR-10, BOR-60, EBR II, Rhapsody, Phoenix (France), and DFR up to burnup 4–10% h.a. consisted primarily of xenon, krypton, and helium; the mass fraction of nitrogen did not exceed 0.4%. Thermal Conductivity. The thermal conductivity has been studied and continues to be studied to determine the influence of fuel composition, impurities (carbon, oxygen), and porosity on it [8, 14, 30, 34, 46, 52, 59, 60].
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Fig. 8. Thermal conductivity of fuel UN (1), NpN (2), PuN (3), U0.8Pu0.2N (4) [30, 34, 14] versus temperature, composition UN–NpN (5), UN–PuN (6), NpN–PuN (7) [14, 34, 59] at 1000°C, porosity 0 (É), 0.05 (8), 0.1 (9) [8, 9, 46].
Figure 8 displays measurements of the thermal conductivity of uranium mononitride and mixed fuel. They show that the thermal conductivity increases with temperature. The following relation expressing the dependence of the thermal conductivity on the temperature (≤1620°C), the plutonium content, and the initial porosity (≤0.2) of the fuel has been proposed in [46] on the basis of a correlation of experimental results: λ = (1.37 – 1.6C + 1.142C 2)T 0.41(1 – P0)/ (1 + P0), where λ is the thermal conductivity of the fuel (W/(m·deg)), P0 is the porosity of the fuel core (relative units), and C is the mass fraction of plutonium nitride in the mixed fuel (relative units). This relation can be used up to temperature 1620°C and porosity of the fuel up to 0.2. Increasing the plutonium nitride fraction in mixed nitride decreases the thermal conductivity. For example, the thermal conductivity of U0.8Pu0.2N is 25% less than the thermal conductivity of UN under otherwise the same conditions (temperature, porosity). The change in thermal conductivity during burnup can be estimated on the basis of porosity changes resulting from swelling. To take account of the change in thermal conductivity correctly with fuel burnup, the irradiated or model fuel must be studied. Compatibility with Structural Steel, Sodium, and Lead. The compatibility of mononitride fuel with structural steel and alloys (304, 316, 1.4919, ÉI 847, ÉI 68, ÉP 823, ÉP 450) has been investigated under laboratory and reactor conditions at 550–900°C [2, 7–16, 27–29, 38, 42, 49, 51–54, 61, 62]. To simulate accident situations, the short-time compatibility of UPuN fuel with ÉP 823 and ÉP 450 for 5 h at 1200 and 1300°C was studied [51]. The investigations established that there is no interaction between the mixed and uranium mononitride fuel with the types of steel listed above under laboratory and reactor conditions with a holding time of 10000 h or longer (EBR II, MTR, ETR, JMTR, Rhapsody, Phoenix, BR-10, BOR-60, DFR, and others). At the same time, comparative radiation experiments in BR-10 showed that oxygen and carbon impurities influence the compatibility of uranium mononitride fuel with ÉI 847 steel cladding. Under identical conditions (the experiment was continued until the burnup reached ~7% h.a.) with oxygen and carbon mass fractions 0.3–0.45% carbonization of the inner surface of the cladding was three times greater than that arising with oxygen and carbon mass fractions 632
below 0.15%. The carbonization is probably due to successive reactions in the fuel-element core at temperature from 900–1000°C at the center up to 600–650°C on the surface of the cladding: MeO2 + MeC → MexOyCz + CO; 2CO → C + CO2. The carbon released on the inner surface of the cladding results in carbonization and embrittlement of the cladding. Investigators in other countries have obtained similar results; they have recommended, taking account of the combined influence of these elements on the creep, swelling, gas release, and compatibility with structural materials, that the mass fraction of oxygen and carbon in the nitride fuel be no more than 0.15% when these elements are present simultaneously [2, 7, 8–12, 14, 20, 21, 51]. In our country, the investigations of the compatibility of fuel cores consisting of UN and UPuN with sodium (under laboratory and reactor conditions), Pb, and Pb–Bi were conducted at coolant melting temperatures up to 700°C with sodium and up to 800°C with lead and for a short time (5 h) at 1300°C. It was established that no change occurs in the surface layers of the fuel washed by the lead, and no change was found in the near-surface zone of the steel washed by lead. Prolonged radiation tests of fuel elements with UN, UPuN, and a sodium sublayer, which were also performed in Great Britain and the US, up to burnup 8–16.8% h.a. showed no fuel–sodium interaction [7–12, 28, 29, 31, 42]. In summary, investigations performed over many years have shown that the mononitride fuel is highly compatible with austenitic and ferrite–martensite class steel and less compatible with liquid-metal coolants (Na, Pb) up to 800°C. Conclusions. The development of fast reactors with mixed uranium–plutonium mononitride fuel is a key element in the development of large-scale nuclear power in the future. The use of nitride fuel with high thermal conductivity, compared with oxide fuel, in fast reactors makes it possible to increase substantially the safety and technical-economic performance of nuclear power plants with these reactors. Experimental data were used to obtain a relation expressing the rate of thermal and radiation creep of nitride fuel as a function of the temperature, burnup, and porosity. The creep rate of mononitride fuel increases with temperature, fission rate, and porosity. For temperature at fuel center less than 1200°C, the swelling rate does not exceed 1.5%/% burnup. The emission of fission products from mononitride fuel, including I, Cs, Se, and Te, is much less than from oxide fuel, and depends on the temperature at the center of the fuel core, the porosity, and the content of oxygen in carbon impurities. Mononitride fuel is compatible with austenitic and ferrite–martensite structural steel up to 800–900°C and the coolants sodium, lead, and lead–bismuth alloy. The oxygen and carbon mass fractions in mixed mononitride fuel should be maintained at a level not exceeding 0.15% for reliable operation of fuel elements. We thank V. V. Orlov, A. G. Sila-Novitskii, V. S. Smirnov, A. I. Filin, and O. A. Ustinov for suggestions and interest in this work.
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