Atomic Energy, Vol. 89, No. 4, 2000

FUEL CYCLE FOR BREST REACTORS

A. V. Lopatkin,1 V. V. Orlov,1 A. G. Sila-Novitskii,1 A. I. Filin,1 Yu. K. Bibilashvili,2 B. D. Rogozkin,2 and V. F. Leont’ev3

In the 21st and subsequent centuries the contribution of nuclear power to the fuel balance in Russia will be much greater than it is today. Fast reactors operating in a closed uranium-plutonium cycle could be the basis for this growth. But, wide adoption of such reactors will be possible if the developers solve the following problems: – safety, ruling out catastrophes such as the Chernobyl accident, is proved convincingly; – economic indicators are no worse than for the light-water reactors competing with them; – a closed fuel cycle (in an integral part of fast reactors), which presupposes radiochemical reprocessing of fuel, does not result in the proliferation of fissioning materials suitable for nuclear bombs; – ecologically safe handling of radioactive wastes will be implemented. The concept of BREST-type reactors should, in the opinion of the developers, solve these problems. It is assumed that 1200 MW(e) BREST-1200 reactors will be the basis of future nuclear power. The technical design of a demonstration BREST-OD-300 reactor with 300 MW electric power and 700 MW thermal power is under development; this reactor should show the structural and conceptual features of BREST-1200 reactors and their fuel cycle. The sodium-cooled fast reactors burning oxide fuel, which have been developed and built thus far, were intended for intense production of excess (more than required by them) plutonium in rapidly developing nuclear power production with plutonium-makeup of thermal reactors. They require the following: – uranium-containing screens surrounding the core, which accumulate weapons-quality plutonium for replenishment of thermal reactors and partially the screens themselves, and a plutonium separation technology; – storage sites for temporary storage of the extracted plutonium. All this creates substantial difficulties for nonproliferation of fissioning materials. The problem of transmutation of long-lived radionuclides was not raised for these reactors. The fire danger of the coolant and the associated safety measures substantially degraded the economics of fast reactors compared with water-cooled reactors. The current situation of nuclear power in Russia makes it possible to change the current view of fast reactors substantially, primarily, to remove the requirement of expanded breeding of plutonium. Approximately four tons of the plutonium contained in the spent fuel from light-water reactors are produced in this country every year. The plutonium which has already been accumulated and the plutonium that will be accumulated in the next 10–15 years will be sufficient to double the capacity of nuclear power in Russia by introducing fast reactors. Consequently, the requirement of expanded plutonium breeding at a high rate can be removed with the new generation of fast reactors, quite simple breeding of fuel, and the developers can concentrate on solving the problems determined by the new requirements for reactors and the closed fuel cycle technology.

1

Scientific-Research and Design Institute of Power Engineering. State Science Center of the Russian Federation – A. A. Bochvar All-Russia Scientific-Research Institute of Standardization in Machine Engineering. 3 State Special Planning Institute. 2

Translated from Atomnaya Énergiya, Vol. 89, No. 4, pp. 308–314, October, 2000.

1063-4258/00/8904-0827$25.00 ©2000 Plenum Publishing Corporation

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In the last ten years BREST fast reactors with uranium–plutonium nitride fuel, which are cooled by liquid lead, have been under development by Scientific–Research and Design Institute of Power Engineering together with other enterprises. The BREST reactors are being developed on the basis of the following assumptions: – complete breeding of plutonium in the core without uranium-containing screens with BR ~ 1 and moderate energy intensity of the fuel; – natural reactor safety with deterministic elimination of the most dangerous accidents with rapid run up, coolant loss, fires, steam and hydrogen explosions with destruction of the fuel and radioactive emissions on a catastrophic level; – decrease of the radiation danger of radioactive wastes as a result of transmutation of the most dangerous long-lived actinides and fission products in the reactor and deep removal of actinides from the radioactive wastes, achieving a radiation balance between the stored wastes and the uranium extracted from the earth; – prevention of the use of the products of the closed fuel cycle for extracting plutonium from irradiated fuel (nonproliferation of nuclear materials); and, – economic competitiveness of nuclear power plants with a fast reactor compared with modern nuclear power plants with thermal reactors (VVÉR, RBMK, BWR, PWR, and CANDU type). The results of the investigations show that BREST reactors can satisfy the requirements indicated above. It is proposed that fuel reprocessing be performed directly in the nuclear power plant, in order to prevent shipment of a large mass of high-activity and fissioning materials. The main characteristics of the reactors are as follows [1]:

Power, MW: thermal electric Core: diameter, mm height, mm Fuel-element diameter, mm Spacing of fuel elements in a square lattice, mm Fuel Fuel load (U + Pu)N, tons Load Pu/(239Pu + 241Pu), tons Fuel run, yr Interval between reloadings, yr Lead temperature at entrance/exit, °C Maximum temperature of fuel-element cladding, °C Maximum lead velocity, m/sec Steam pressure at the exit from the steam generator, MPa Net efficiency of the power generating unit, %

BREST-1200

BREST-300

2800 1200

700 300

4750

2300 1100 9.1, 9.6, 10.4 13.6 UN + PuN 60 16 7.34/4.93 2.2/1.6 5–6 5 1 420/540 650 1.6 1.8 24.5 ~43

General Requirements for the Fuel-Cycle Technology. The development of nuclear power on the basis of a new technology should not result in opening new channels for obtaining weapon materials and should rule out the use of the technology itself for such purposes. The development of nuclear power based on fast reactors with an appropriately organized fuel cycle creates conditions under which the risk of the proliferation of nuclear weapons gradually decreases. Fast reactors do not need uranium enrichment, and this technology can be phased out with time. The fabrication of the first loads of fast reactors requires that the storage sites for extracted plutonium and the holding ponds for spent fuel at current nuclear power plants be gradually freed in order to reprocess the fuel and extract the plutonium. The primary extraction of plutonium and fabrication of the first fuel loads for fast reactors, in this case, must be done in enterprises in nuclear countries or in international nuclear-technology centers. The gradual transfer of the plutonium from storage sites and the plutonium contained in the holding ponds for spent fuel in modern nuclear power plants to the most highly protected conditions of fast reactors and their fuel cycle in time will close this channel for the proliferation of weapons materials. 828

The nonproliferation of plutonium is accomplished in the fuel cycle for BREST reactors in two ways: physical and structural features of the nuclear reactor, corresponding to the choice of the fuel-cycle technology. BREST reactor fuel containing 13.5% plutonium with the equilibrium composition in a uranium–plutonium mixture is unsuitable for nuclear weapons [2]. The plutonium contained in the fuel with a high 238Pu, 240Pu, and 241Pu (power plutonium) content is not in the weapons-plutonium class, but it is undoubtedly also dangerous from the standpoint of nonproliferation of the product. The relatively high energy release and other radiation characteristics make it difficult to prepare a nuclear bomb from it, but unfortunately this cannot be ruled out. The increasing turnover of power plutonium (spent nuclear fuel, storage in enterprises performing radiochemical reprocessing) outside nuclear programs with their specific measures of accounting, monitoring, and guarding, with time will become one of the substantial difficulties of maintaining a nonproliferation regime, which is now underestimated. The plutonium problem can be solved radically by eliminating the presence of plutonium in the form of a separate product at all stages of the fuel cycle, leaving only one place where plutonium could be present – the plant performing the radiochemical reprocessing of fuel from thermal reactors (1 or 2 in the country) and production of the first loads of fast reactors. The nonproliferation problem in the fuel cycle of fast reactors can be solved by eliminating the appearance of an extracted plutonium fraction or a uranium–plutonium mixture, whose multiplication properties are better than those of uranium with 235U 20% enrichment, at all stages of the fuel cycle. In the fuel cycle for fast reactors with BR ~ 1: – uranium-containing screens producing weapons-grade plutonium, are not required and are eliminated; – there is no need to separate plutonium in order to fabricate fresh fuel; – a radiochemical technology which cannot separate plutonium from the fuel during reprocessing should be used; – a low degree of removal of fission products from the recovered fuel is allowed (the residue of fission products in the fresh fuel is 10–1–10–3 of their content in the irradiated fuel), and the presence of Np and Am in the fuel gives rise to high radioactivity (the radiation barrier to misappropriation of the fuel); – all production of the fuel cycle can be placed at the site of the power plant in order to eliminate shipment and the risk of misappropriation or loss of the fuel. In the fuel cycle being considered the 238U added during reprocessing burns up in the reactor. Plutonium is an integral part of the fuel and is handled in the fuel cycle as part of the high-activity material. Nonseparation of uranium and plutonium during fuel reprocessing must be guaranteed by the character itself of the chemical processes and the equipment present in the process lines. Possible changes in the controllable process parameters (temperature, pressure, reagents, and so on) should not result in the extraction of plutonium or a large increase of the plutonium content in the fuel composition, i.e., the technology must be self-protective. If plutonium is not separated from the fuel, then, naturally, there is no problem with its proliferation. It should be noted that the nonproliferation of fissioning materials in principle cannot be guaranteed solely by technical measures, since there always remains the possibility of illegal use of the well-developed uranium enrichment technology or the technology for separating plutonium from the spent fuel of modern nuclear power plants, which has been standing for a long time in the holding ponds. Only the improvement of the international political conditions for nonproliferation and corresponding monitoring and protective measures can prevent this danger. In addition to the condition of nonseparation of uranium and plutonium, the reprocessing technology was faced with several additional conditions intended for improving the radiation balance between the fuel-cycle wastes and the natural uranium used. Radiation balance between the natural uranium consumed and the long-lived high-activity components of the radioactive wastes produced in a system of BREST-type nuclear reactors can be attained by transmutation of actinides in the fuel (U, Pu, Am) and long-lived products (Tc, I) in the blanket of BREST reactors and by controlled holding prior to burial of high-level wastes for approximately 200 yr in order to reduce their activity by approximately a factor of 1000. In the proposed conception of the fuel cycle it is suggested that the wastes to be stored should not contain more than 0.1% uranium, plutonium, amerecium, and curium, 1–5% cesium, strontium, technetium, and iodine. All other actinides and fission products can be stored after a holding period. Curium should be extracted from the fuel and held for approximately 100 yr until the short-lived isotopes decay into plutonium, which together with the long-lived curium isotopes is returned into the fuel. This measure will greatly reduce the energy release and other radiation characteristics of the recovered fuel. Cesium and strontium are stored for approximately 200 yr until they completely decay.

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Fig. 1. Contribution of Pu (2), U (3), Am (4), Cm (5) 90Sr + 90Y (6), Cs + 137mBa (7), and 14C (8) to the potential biological danger of irradiated BREST-1200 reactor fuel (1). 137

Plutonium and americium (Fig. 1) determine the long-term (holding period 102–105 yr) potential biological hazard (potential irradiation dose D) of the irradiated fuel. If no more than 0.1 mass% of the recycled actinides and 5% Cs, Sr, Tc, and I end up in the wastes which are to be stored, then the radiation balance of the fuel wastes and the natural uranium consumed will be attained after a 200–300 yr holding period for the wastes. Figure 2 shows the balance between the potential biological danger of radioactive wastes and the natural uranium employed. The wastes contain actinides and fission products of the irradiated fuel and the irradiated steel jacket of a fuel element. The results are normalized to 1 kg of irradiated actinides (1.06 kg of nitride fuel and 0.132 kg of ÉP-823 steel). The possible biological danger of the wastes is the same as the danger of 13.7 kg of natural uranium. This mass of natural uranium was necessary in order to produce in a thermal reactor 1 kg of BREST-1200 fuel for the first load and takes account of the fact that the first load will be recycled 12 times during the 60-yr period of operation of the BREST reactor. Stages of the Fuel Cycle. The fuel cycle for BREST reactors contains stages that are conventionally considered in the closed fuel cycle of fast reactors, with the exception of the cycle of breeding screens: – fuel irradiation in the reactor; – post-reactor holding period for irradiated fuel assemblies and their shipment to a recovery plant; – cutting up of the fuel assemblies, extraction of fuel, and separation of the steel elements of the assemblies; – radiochemical reprocessing; – adjustment of the composition of the fuel mixture; – fabrication of nitride pellets; – fabrication of fuel elements and fuel assemblies; – temporary storage; – transport into a reactor. The entire cyclic turnover of the fuel is concentrated in the reactor building and the adjacent fuel-cycle building. A storage site is present for holding irradiated fuel at the nuclear power plant site. We shall examine the stages of the fuel cycle for the BREST-OD-300 reactor, which have now been worked out in greatest detail as part of the design [2–5]. The volume of the annual reprocessing of BREST-OD-300 reactor fuel is 3.2 tons (29 fuel assemblies) with a 5-yr fuel run. For a 4-yr run the interval between reloadings decreased, so that five reloadings would be made in a run. For a nuclear power plant with two BREST-1200 units, the yearly reprocessing volume will be 24 tons. The fuel run in a reactor is 5 yr, the core-average burnup of the off-loaded fuel is 8.8% heavy atoms, and the fuel is reloaded once per year. For the BREST-OD-300 reactor, during the design period the run for the initial loads was decreased to 4 yr and the fuel burnup was decreased correspondingly, but five fuel reloadings in a run were retained.

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Fig. 2. The variation of the potential biological danger of radioactive wastes of a BREST-1200 reactor with the following composition: all fuel + 132 g of the alloy ÉP-823 (1), 5% each of Sr, Cs, Tc, I + 100% of all other fission products + 0.1% each of U, Pu, Am, Cm + 100% each of Th, Pa, Np, Bq, Cf + 132 g of the alloy ÉP-823 (2), 100% fission products + 0.1% each of U, Pu, Am, Cm + 100% each of Th, Pa, Np, Bq, Cf + 132 g of the alloy ÉP-823 (3), 132 g of the alloy ÉP-823 (4). The potential biological danger of 13.7 kg of natural uranium (4) is presented for comparison.

After the extraction from the core the irradiated fuel assemblies, raised no higher than the lead level, are rearranged for holding in a storage site located inside the reactor vessel. The lead circulating in the reactor cools the fuel assemblies. The irradiated fuel assemblies are extracted from the intrareactor storage site after the residual heat release in the fuel decreases to 30 W/kg (the shortest holding period is 120 days); this permits shipment through the gas-filled channel in a dry storage site. Two regimes are considered for the removal of fuel assemblies from the intrareactor storage site into an external site – a single reloading after a holding period with a duration of one microrun and partial removal from 120 days to 1 yr (duration of a microrun). The fuel-cycle building stands next to the reactor building. The reloading machine lifts an irradiated fuel assembly above the lead level in a gas atmosphere, turning the assembly into a horizontal position and moving it, along a special channel, into a receiving chamber and then into a dry storage area. All operations are performed in an argon atmosphere. The section where the fuel assembly is cut up, the fuel is extracted, and the steel components are separated includes dry storage of the fuel assembly, temporary storage of the nonfuel parts of the fuel assembly, a laser cutting setup, and setups for removing and catching gaseous, volatile, and aerosol radioactive fission products. Depending on the proposed recovery technology, various schemes are proposed for cutting up the fuel part of the fuel assembly – mechanical fragmentation and metallurgical methods for separating fuel and cladding. The fuel pellets separated from the cladding are transferred to radiochemical reprocessing. At the present time, electrochemical recovery in a melt consisting of chloride salts with precipitation of actinides on a solid rotating cathode has been adopted in the BREST-OD-300 design. Reprocessing is conducted in two electrolyzers, one electrolyzer serving as a backup. The cathode precipitate contains an admixture of the electrolyte. The actinides contain admixtures of rare-earth elements in amounts of 2–10% of their content in the irradiated fuel. A general scheme of the process, the volume of the equipment, and the consumption of reagents and electricity have been worked out. Scientific research work is required to substantiate the proposed scheme and to confirm that the technology meets the appropriate requirements. Other forms of radiochemical reprocessing in application to the fuel cycle for BREST reactors are also being considered as part of the investigations to substantiate the BREST-1200 fuel cycle: – water reprocessing with and without organic extracting agents; – electrolysis in a melt of chloride salts with reduction of actinides to nitrides; – metallurgical without destruction of nitrides at all stages of reprocessing; – in melts of fluoride salts; 831

– gas fluoride; – high-temperature annealing (as the first stage of fuel reprocessing); – electrolysis in a melt of fluoride salts; – recrystallization in molybdate and phosphate melts, and others. The possibility of not separating uranium and plutonium and the prescribed requirements regarding the degree of purification of the fuel and the fractionation of wastes are being investigated. The schemes for most processes, the equipment composition, and so on have been developed. The technical–economic investigations will be performed for the substantiated choice of the most appropriate technology. The technological scheme, developed at the Scientific-Research Institute of Nuclear Reactors, with electrolysis in a melt of chloride salts and reduction of actinides to nitrides presupposes that there is no need for the laborious operation of extracting fuel pellets from the claddings of the fuel elements. This operation is replaced by cutting of the fuel elements, after which their fragments are fed into an electrolyzer. Nitride fuel dissolves in the elctrolyte, and the cladding fragments accumulate on the bottom the electrolyzer. To remove impurities from the electrolyte, the cathode deposit (in the BREST-OD-300 design) is remelted in a vacuum at 1150–1200°C. Tailings uranium is added to the purified metallic alloy of actinides in order to compensate for the fission products removed from the fuel and the low quantity of actinides. Ingots with the required dimensions are cast from the alloy obtained. The section where nitride pellets are prepared is placed in a chain of protected chambers. The metal ingots are hydrated in a flow of argon-hydrogen mixture, after which the atmosphere in the working chamber is replaced by nitrogen and nitriding is performed. The main equipment consists of the horizontal synthesis apparatus, heated by moving electric furnaces. The nitride powder is graded, mixed with the binder, and granulated to a grain size of 0.5 mm. The granules obtained are fed into automatic molding machines to prepare pellets. There are four such molding machines, three for each size of pellet and one as a backup. The raw pellets are sintered in an induction furnance at 1250°C. The finished pellets are sorted, the acceptable pellets are stored temporarily, and the rejected pellets are pulverized and returned into the process line. To prepare the fuel elements and fuel assemblies the nitride pellets are placed in cladding, and the space between the pellet and the cladding is filled with lead. An absorbing element consisting of tungsten carbide is placed in a gas volume at the top of a fuel element. The gas volume is filled with helium at 0.05 MPa pressure. The fuel element is sealed, and a protective coating is deposited on the outer surface. In the fuel-element fabrication process the height of the fuel column, the dimension of the lead layer, the airtightness, and the uniformity of the protective layer are monitored. The rejected fuel elements are cut up and returned into the process line. Before the fuel assemblies are mounted the fuel elements are formed into sets according to type; special tools are used to secure the fuel elements and other structural components in space in accordance with the construction of the fuel assembly, and the spacing components, the head, and the tail are welded. The weld seams in the finished fuel assemblies are heat-treated, and the airtightness, the quality of the weld seams, the heating of the ends, and the degree of entry into the steel are checked. All structural components of the fuel assemblies (fuel element cladding, ends, and so on) are made of ÉP-823 steel. The BREST-OD-300 design provides for temporary storage for 30 fresh fuel assemblies – an adequate volume for reloading 1/5 of the core. The same transporting technology as that used to move the irradiated fuel assemblies from the reactor into the fuel-cycle building is used to transport the fuel assemblies from the fuel-cycle building into the reactor. In a closed fuel cycle the BREST fuel never leaves the reactor + fuel-cycle building system. Aside from the basic stages indicated above the fuel-cycle design contains systems for collecting and storing solid, liquid, and gaseous radioactive wastes, a storage site for holding radioactive wastes for a long period of time, a control and monitoring system, a storage area for reagents and materials, a decontamination system, maintenance services, and so on. Calculations of the cost of the entire nuclear power plant with two BREST-1200 reactors and a fuel cycle at the plant showed that the cost of the latter is 15% of the cost of the entire complex.

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E. O. Adamov (ed.), White Book of Nuclear Power [in Russian], NIKIÉT, Moscow (1998).

2.

3. 4. 5.

A. V. Lopatkin and V. V. Orlov, “Fuel cycle for a new generation of fast reactors based on the principles of nonproliferation of nuclear weapons and radiation-equivalent burial of radioactive wastes,” in: Reports at the International Seminar “Fast reactor and fuel cycle with natural safety for the next stage of nuclear power. Fuel balance, economics, safety, wastes and nonproliferation,” Moscow, May 29–June 1, 2000. B. D. Rogozkin, N. M. Stepennova, Yu. E. Fedorov, et al., “Mononitride mixed U–Tu fuel and its electrochemical recovery in fused salts,” ibid. E. I. Tyurin, V. F. Leont’ev, E. D. Spitsin, and M. T. Vorontsov, “Fuel cycle at the BREST-OD-300 nuclear power plant,” ibid. O. M. Borisov, V. V. Orlov, and A. G. Sila-Novitskii, “Requirements for the core,” ibid.

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