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Use of lead-bismuth coolant in nuclear reactors and accelerator-driven systems
ELSEVIER
Nuclear Engineering and Design 173 (1997) 207 217
Nuclear Engineering and Design
Use of lead-bismuth coolant in nuclear reactors and accelerator-driven systems B.F. G r o m o v a,,, Yu.S. Belomitcev a, E.I. Yefimov a, M.P. Leonchuk ~, P.N. M a r t i n o v a, YU.I. Orlov ~, D.V. P a n k r a t o v a, Yu.G. Pashkin ~, G.I. Toshinsky a, V.V. C h e k u n o v a, B.A. Shmatko a, V.S. Stepanov b " State Scientific Centre RF, Institute of Physics and Power Engineering, Bondarenko Sq. 1, Obninsk, Kaluga Region 249020, Russia b RDB Gidropress, Ordzhonikidze st. 7, Podolsk, Moscow Region, Russia
Abstract
Experience of using lead-bismuth coolant in reactors of Russian nuclear submarines is briefly presented. The salient points of the concept providing the safety of reactor facilities cooled by a lead-bismuth eutectic are covered. The key results of developments for use of a lead-bismuth coolant in nuclear reactors and accelerator-driven systems are presented. © 1997 Elsevier Science S.A.
1. Introduction
In the beginning o f the 1950s the U S A and U S S R began nearly simultaneously a nuclear power facility ( N P F ) development for nuclear submarines (NS). In b o t h countries the w o r k was carried out on two types o f N P F - - w i t h pressurized water reactors and with liquid metal cooled ones ( L M R ) .
Abbreviations: NPP, nuclear power plant; NHPP, nuclear heating power plant; RF, reactor facilities; TES-M, index of small power nuclear heating plant; RDIPE, Research and Design Institute of Power Engineering; RAW, radioactive waste; PRISM, power reactor inherantly safe modular; BRUS, index of fast reactor facilities; ALMR, advanced liquid metal reactor; LANL, Los Almos National Laboratory; MSRE, molten salt reactor experiment. * Corresponding author.
In the U S S R , where the works were begun in 1952, a l e a d - b i s m u t h eutectic was chosen as a coolant for L M R ( G r o m o v et al., 1992a). A scientific supervision o f this w o r k was provided by academician A.I. Leipunsky (IPPE, Obninsk) and after his death in 1972, by Prof. B.F. G r o m o v . In the U S A , sodium, as possessing better thermophysical features, was chosen as a coolant for L M R . The g r o u n d test facility o f N P F and an experimental submarine 'Sea W o l f ' have been constructed. Yet the operation experience has revealed that the choice o f a fire-and-explosion dangerous at contact with air and water coolant did not justify itself. After a n u m b e r o f accidents the reactor facility ( R F ) o f this N S was dismantled together with the c o m p a r t m e n t and replaced by a pressurized water R F .
0029-5493/97/$17.00 © 1997 Elsevier Science S.A. All rights reserved. PII S0029-5493(97)00110-6
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In the USA there were also conducted research and development on using a lead-bismuth coolant (LBC). But the alternative of solving the corrosion resistance problem of structural materials, control and maintenance of the coolant quality (the coolant technology) gave no positive results and those works were stopped. In our country as a result of systematic work of a number of organizations in the course of about 15 years there was solved a complex problem of the coolant technology, the corrosion of structural materials and the mass transfer, that ensured a long and reliable operation of N P F with LBC under NS conditions. Proceeding from the experience, obtained under development and operation of N P F with LBC, both the concept of maintaining a new RF generation safety for civil purposes and the proposals on using LBC in the nuclear power (NP) have been developed.
2. Experience of using lead-bismuth coolants The expediency of using a lead-bismuth eutectic (Pb is ~ 44%, Bi is ~ 56%) as a coolant of the primary circuit of nuclear reactors is caused by its physical, chemical and thermodynamic properties which allow to meet the safety requirements to the most complete degree. Due to a very high boiling point ( ~ 1670°C) it is allowable to have a low pressure in the primary circuit when the outlet coolant temperature is equal to 400-500°C. This simplifies a RF design and increases its reliability. A low pressure in the primary circuit enables to reduce a wall thickness of the reactor vessel and introduce no restrictions to the temperature change rate in compliance with a thermocyclic strength, that provides R F operation in a manoeuvre mode. The LBC low chemical activity allows to use a two-circuit scheme of the heat removal. Emergency processes related to the primary circuit loss of tightness and intercircuit leaks in a steam generator (SG) occur without hydrogen formation, fires and explosions.
In the course of operations a number of engineering problems, related directly to LBC features, has been solved. The LBC melting point accounts for ~ 125°C. LBC maintaining in a liquid state under all RF operation conditions is provided by using SG with a multiple circulation of a steam-water mixture over the secondary circuit, therefore the inlet temperature of the water supplied to SG is higher than the LBC melting point. The system of a steam or electrical heating can be used for the initial heat-up and keeping the primary circuit in a hot condition at a low level of heat releases in the core. An important practical problem is the substantiation of an opportunity of the multiple 'solidification-remelting' of LBC in R F that could be required in prolonged shut-down regimes. A low shrinkage of LBC under solidification and rather a high plasticity with a low strength in a solid state promote realization of these regimes without a structural material damage at LBC transition from a liquid into a solid state and its further cooling down up to the ambient temperature. A special technology of a safe remelting has been developed. When carrying out maintenance and reactor refuelling, there is no need in the primary circuit decontamination, which is associated with assembling, storing, transporting and reprocessing of a plenty of liquid radioactive wastes (LRW). A specific character of LBC, when bismuth is exposed by neutrons, is the formation of ~-active radionuclide of polonium-210 with a lifetime of 138 days. A major reason of its radiation danger is the formation of radioactive aerosols and a volatile polonium hydride when a hot LBC contacts with air. It may take place under conditions of the primary circuit tightness loss and the coolant leakage. In this case, as the RF operation experience at the NS has displayed, the release of Po aerosols and the air radioactivity (according to the thermodynamics laws) reduce strongly with the temperature decrease and solidification of the coolant leakage. A fast solidification of the LBC leakages restricts the area of radioactive contamination and simplifies its removal in the form of solid radioactive wastes.
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A low concentration of Po in the coolant (at the level of 10 - s at.%) and the formation of a thermodynamically p r o o f chemical compound of polonium with lead, decrease polonium volatility by eleven orders of magnitude in comparison with a sole metal polonium. In addition, for safety ensuring there were developed means of an individual and collective personnel protection, methods of an equipment decontamination and radioactivity fixation on surfaces. All the above has resulted in the fact that for a long-term period of N P F operation with a lead-bismuth coolant, including the primary circuit repairs and removal of coolant leakages, there were no cases of personnel extrairradiation by this radionuclide above the permissable limits. This positive practical result coincides with the conclusions of foreign experts who have experimentally investigated the Po danger problem when radioactive lead-bismuth is used as a nuclear reactor coolant (Tupper et al., 1991). Among basic problems, which have been solved during designing, development and operation of this type of facilities, it is necessary to highlight the problem of LBC technology, i.e. the development of the systems and devices, which ensure controlling and supporting the LBC required quality during a long operation under normal conditions of a tight circuit and in the event of SG seal failure, as well as at the circuit opening for repairs and the reactor refuelling. The application of such a technology is necessary for eliminating a structural material corrosion and slagging the circuit by lead oxides. To settle these problems the following has been developed: the special filters for cleaning LBC from insoluble impurities, the devices to ensure chemical reduction of a lead oxide, the controlling devices to sustain in LBC a necessary concentration of a corrosion inhibitor-dissolved oxygen, the appropriate sensors to indicate the quality of LBC and the protective inert gas. The corrosion resistance of materials is provided by the alloying of steels, a preliminary formation of protective films on their surfaces and maintaining the necessary concentration of a dissolved oxygen in LBC.
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A nuclear reactor with an intermediate neutron spectrum was applied as an energy source in NPF. A metal beryllium served as a neutron moderator. Beryllium available in the core afforded a neutron flux high level in a subcritical reactor because of a photo-neutron reaction with y-rays of fission products that in turn resulted in a safe and rapid reactor start-up. The following characteristics were obtained in the course of N P F tests and operation: the facility capacity and parameters, the lifetime, the reactivity margin, reactivity coefficients, poisoning effects, temperature distributions, dynamic parameters, the coolant radioactivity, dose rates of neutron and y-radiations behind the shield and some other parameters. They were in a sufficiently good agreement with the calculation results. A high operation capability was demonstrated by the cores and control rods. During last 10-15 years of N P F operation there were no problems with either corrosion of structural materials in the primary circuit or the deviations from the standards of a circuit purity. Experience of N P F with LBC development and operation at NS allows to make a number of practically important conclusions and recommendations, concerning the LBC reactor primary circuit at civil NPPs and transportable NPFs. The best technology parameters should be expected in integral design reactors. The most convenient design scheme of a steam generator is the one in which a liquid metal circulation takes place in the space between pipes, and that of water or steam does in pipes. At such a design the possibility of repairing SG is provided by plugging a separate pipe, which has lost its tightness, without dismantling SG or opening the primary circuit. Shut-down regimes, the regimes of start-up and heat decay removal are more easy realized when SG operates with a multiple circulation in a steam-water circuit. Mechanical vertical pumps with a turbine or an electrical driver as well as the electromagnetic ones may be used for the LBC circulation. For the past period of time, there were constructed eight NSs having R F with LBC (Nilsen and Bemer, 1994). The first experimental NS of
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the project 645 had two reactors, other seven NS of the project 705 (in NATO's terms, 'Alpha') had one reactor. Due to its high-speed qualities this NS was put into the Guinness Book. Besides, two full-scale ground reactor test facilities were constructed and put into operation: N P F prototype in IPPE (Obninsk) and in NITI (Sosnovy Bor). A general operating time of R F was about 80 reactor-years. The R F were designed by RDB 'Gidropress' (Podolsk) and EDBM (Nizhniy Novgorod), the relevant NS were designed by SPMBM 'Malakhit' (St. Pertersburg). The experience of RF design and operation at NS has been used for the safety concept development of R F with LBC.
3. Concept of safety ensuring Realization priority of safety requirements in comparison with other requirements has become obvious after the Chernobyl accident. An intensive development of NPP new concepts for the future has been begun all over the world, their safety is based on a R F inherent self-protection to a considerable degree. It allows to exclude deterministically the possibility of the most severe accidents, which results in catastrophic consequences not only in the event of safety technical system failure and personnel errors, but also in the event of ill-intentioned actions, a subversive activity, an influence of usual explosive substances (terrorism, military attacks). The facilities with a L M C fast reactor (FR) are referred to RF, whose safety is maintained principally due to an inherent self-protection. It is associated with a number of their internal features. An absence of poisoning effects in a FR, a low value of a negative temperature reactivity coefficient, the compensation of fuel burn-up processes and slagging by plutonium production, as well as partial refuellings allow to ensure the operative reactivity margin (in an operating reactor), which is less than a share of delayed neutrons, and to exclude in this way the runaway on prompt neutrons.
Sodium is now widely used as LMC in NP. The choice of this L M C for fast reactors was caused by an intensive heat removal due to good thermal and physical properties. Hence, a short doubling time of plutonium was achieved that appeared an indispensable requirement at early development and construction stages of fast breeder reactors in the sixtieth-seventieth. Just for this reason academician. A.I. Leipunsky, who considered various LMC for cooling F R in the fiftieth, preferred sodium, though from a safety view-point LBC was originally considered for these purposes (Leipunsky, 1990). At present and in prospect, such a short doubling time of plutonium which can ensure sodium cooled FR is not required. Besides, the necessity of developing breeder reactors is postponed for many decades (von Hippel, 1995). It makes reasonable to return to using LBC in NP, taking into account the experience obtained. A lead-bismuth coolant due to its chemical inertness and a higher boiling point in comparison with sodium allows to reach even a better FR safety than in the case of sodium. Since the energy accumulated in the LBC (thermal, chemical and a potential energy of compression) is the minimum one in comparison with other coolants and, taking into consideration the FR features mentioned above, it should be expected that the RF design with LBC meets the requirements of the concept of the maximum achievable inherent safety (Adamov and Orlov, 1994). At present, there are two approaches of ensuring the RF safety. The first approach, being traditional, is based on extending the number and increasing the efficiency of various protective and localizing systems and devices, which decrease the probability of heavy accidents and reduce a danger of their consequences. A practical realization of this approach results in a more complex and expensive facility, a deterioration of its other characteristics and does not exclude, in principle, an opportunity of a severe accident with catastrophic consequences because internal reasons of its realization have not been eliminated. The second approach is based on the concept of RF with an inherent (natural) safety ensuring the
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R F self-protection, in which the reasons for severe accidents by the nature laws are deterministically excluded. With this approach, the construction of protective and localizing systems is not required. There is also no need in a complex substantiation of the safety by means of many calculational and experimental investigations in a framework of abstract scenarios of a severe accident progression and by constructing large-scale test facilities. This approach is most consistently developed by Prof. A. Weinberg in the USA, and then by Prof. V. Orlov in Russia. Neither first nor the second approaches are realized in a pure mode as each of them possesses elements of the other. Nevertheless, proceeding from the safety modern philosophy at developing the advanced RF, it seems necessary to apply the approach, basing at most on the inherent self-protection, as far as the most severe accidents are concerned. This conclusion is justified by the fact that the probabilistic safety analysis (PSA), whose results are taken for substantiation of a R F safety in the first approach, considers failures of technical devices and errors of operational personnel as casual events, which probability may be estimated only with a large uncertainty because of insufficient statistical data. A small probability of a severe accident is neither proof of its impossibility nor the evidence that it might happen not earlier than thousands or tens of thousands of years later. Besides, with ill-intentioned actions of people, that should be taken into account, these events will be not casual but programmed beforehand. In such a case the PSA conclusions lose their validity. The R F safety with LBC is provided by so natural properties as a very high boiling point and an evaporation latent heat. It excludes practically a possibility of the primary circuit overpressurizing and the reactor thermal explosion at a coolant emergency overheating, because there is no pressure increase. Besides, a low pressure allows to reduce the reactor vessel thickness and to use for its construction a less strong austenitic steel, being resistant to a radiation embrittlement under operating conditions. Owing to the impossibility of the coolant boiling, the heat removal reliability from
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the core and the safety are increased as there is no heat removal crisis. Loss of a coolant with its circulation interruption through the core at failure of the reactor vessel tightness (beyond design accident) is ruled out due to a guard shell availability and a small free volume between the reactor vessel and the guard shell. In the event of the primary circuit gas system failure it is ruled out due to the coolant boiling out impossibility. In case of all heat exchanger failure in the independent emergency cooling system (beyond design accident) the core melting-down under a decay heat effect and a reactor vessel break for facilities of a small and average capacity are prevented due to a large difference between an arising temperature and the coolant boiling point. It occurs because of the heat accumulation in the internal reactor structures and in the coolant itself and because of the heat removal through the reactor vessel by means of a natural circulation of the ambient air. The reactor power decrease in the case of emergency overheating and simultaneous failure of the emergency protection systems is provided by reactivity negative feedbacks. In the core and within the reactor vessel there are no materials releasing hydrogen under irradiation or as a result of chemical reactions with the coolant. The coolant itself reacts very slightly with water and air, their contact is possible in case of the circuit failure, and no explosions and fires take place. The operation experience of R F with LBC at nuclear submarines has revealed R F safe operation during some time under conditions of a small SG leak which results in no significant deviations of engineering parameters from the designed ones. This fact gives the possibility to realize necessary repair measures not urgently but in a convenient time. The elimination of a water or steam penetration into the core and consequent overpressurizing the reactor vessel, designed for the pressure occurring at a large SG leak, are provided by a coolant circulation scheme. With this scheme, bubbles of steam and drops of water are thrown out on a free level by a coolant ascending flow. A steam
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effective separation occurs therefore in the gas volume above the coolant level, whence steam goes to the emergency condensers operating on a passive principle. The coolant does not boil at the primary circuit failure and has the property to retain iodine, whose radionuclides represent, as a rule, a major factor of a radiation danger just after an accident. This reduces sharply a scale of radiation consequences of so accident in comparison with the pressurized water reactors. The containment serves as an additional safety system barrier, which mainly protects against external actions. A low storage of a potential energy in the primary circuit restricts a R F destruction scale under external action conditions. An extremely high safety potential, which is peculiar to R F of this type, is characterized by the fact that when such initial events as the containment destruction and the primary circuit failure coincide (it is possible in the course of subversive or military activities), neither reactor runaway, nor explosion and fire occur, and the radioactivity release is essentially lower than the level requiring the population evacuation. Thus, the use of such a type of reactors is possible and expedient not only for an electric power generation at NPP but also for simultaneous production of heat at N H P P with locating them near large cities, as well as for the sea water desalination. It should be mentioned that in association with the factors presented, the interest in using lead bismuth or lead coolants has been lately displayed in other countries (Tupper and Wett, 1992; (Sekim o t o and Su'ud, 1995; Carte et al., 1993; T a k a n o et al., 1993).
4. Proposals on using lead-bismuth coolant A number of proposals on civil N P P of a small and average power (from 1 to 600 MW) in transportable, floating and ground performances has been lately developed on the basis of the concept presented and the experience obtained. These proposals were developed in the frameworks of defensive work conversion under a scientific
supervision of the State Scientific Centre Russian F e d e r a t i o n - - t h e A.I. Leipunsky Institute of Physics and Power Engineering (SSC RF-IPPE) and R D B 'Gidropress'. A brief description of them is presented below. 4.1. Modular-transportable N H P P T E S - M with electric power o f 1 M W
The use of nuclear stations of small power might appear expedient for far regions having no centralized electric energy supply, and with complicated conditions of an organic fuel delivery. As the principal requirements for small power NPP (SPNPP) are safety, reliability and simplicity, the following basic concepts were accepted in SPNPP TES-M conceptual design ( G r o m o v et al., 1992b), (the engineering development of this SPN P P was carried out in C D B M St. Petersburg); • the circulation of a coolant in a primary circuit in all regimes is realized by natural convection; • all the equipment of a primary circuit is concentrated in the same vessel (integral design); • the air is used as a coolant of a secondary circuit and simultaneously drives a gas turbine; • the transportation of a nuclear heating unit ( N H U ) to SPNPP site is realized at a complete factory readiness with an assembled reactor, filled by a solidificated coolant; • the elimination of the core refuelling on SPNPP site. For this purpose the core lifetime is provided to be 90000 eft. h (10-15 years); • the refuelling of the spent core is designed in the central maintenance factory where N H U with a solidificated coolant is delivered. The coolant features and the use of the air in the secondary circuit exclude N P P equipment failure and release of radioactive products in the environment in the event of N H U temperature decrease below zero that may occur in emergency situations, accompanying by loss of internal and external heat and electric energy. Realization of the core emergency cooldown is provided by using only passive ways of heat removal. SPNPP simplicity and reliability provide for its operation under automatic operation conditions without a constant maintenance.
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Maximum weight of the transport unit is about 50 tons. 4.2. Modular-transportable N H P P 'Angstrem'
N H P P 'Angstrem' is to provide by heat and electric power the settlements with up to 5000 inhabitants. N H P P 'Angstrem' of a stationary ground performance is to be assembled out of factory functionally completed units on a preliminary prepared site over 1 month. N H P P can be transported by 9 12 railway platforms and by other transport vehicles as well. The maximum weight of the unit is 220 tons. The primary circuit consists of two loops with a compulsory circulation. A steam turbine design allows to take steam for heat supply. The air-radiator cooling of turbine condensers is designed so that it makes possible to use N H P P 'Angstrem' in waterless regions (Gromov et al., 1993a). The electric power is 6 MW, the heat supply is equal to 12 Gcal per h, the operation term without a refuelling is ~ 6 years. 4.3. Floating N P P 'Cruise-50'
Many coastal regions have no reliable sources of power supply. For economical or ecological reasons the use of the floating NPP with a high safety potential may show promise. For this purpose, the conceptual design of the floating NPP 'Cruise-50' with the electric power of 50 M W has been developed with participation of SPMBM 'Malakhit' and the 'Turbine factory' in Kaluga. This NPP may be also used for the sea water desalination or the heat production (Gromov et al., 1993b). This reactor facility has an integral structure with a compulsory circulation (two pumps). Steam generators produce a saturated steam at the pressure of 5.6 MPa. The refuelling is carried out at the central service factory, where the floating N P P is transported once over 6 years.
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4.4. Modular N P P with R F L B F R - 7 5
In the course of 2003 2005 the third and fourth units with reactors VVER-440 of the Novovoronezh NPP (NVNPP) are to be decommissioned. The first and second units have been already shut-off. For replacing the units decommissioned, the construction of the sixth and seventh units is proposed that will require large capital investments. In this association, SSC RF-IPPE, RDB 'Gidropress' and 'Atomenergoproekt' have carried out a technological investigation for determining a technical possibility and economical feasibility of the N V N P P 1, 2, 3 and 4 units renewal when their service life is exhausted. This renewal is proposed to carry out by means of a dismantled SG replacing with nuclear steam generating modules of an integral design with LBC fast reactors (Gromov et al., 1996a). Such modules (the power of each is 75 MW) allow to replace the removed capacities and to use the existing capital structures and the equipment of the Novovoronezh NPP. Thus, the regulation safety requirements will be met, the radioecological conditions will be improved due to a nearly total absence of liquid radioactive wastes (LRW), the NPP operation in a manoeuvre mode will be provided. It should be expected a significant saving of the expenditures for a N P P renewal as compared to the construction of new VVER units, when taking into account the circumstances and an essential reduction of the auxiliary systems and the equipment, the safety systems, a nearly total LRW absence, a complete factory production of the module carried by rail to the NPP site, as well as a high serial production. 4.5. Reactor facilities o f B R U S type
The following group of the R F considered with a higher power of 150-300 M W (BRUS-150, BRUS-220, BRUS-300), being in conformity with above presented safety concept, can solve the problem of modular N P P construction, which is similar to the PRISM concept of the USA
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(Berglund and Tippets, 1989) and 4S concept of Japan (Hattori and Minato, 1995). But an inherent safety principle is more consecutively realized in these reactors than in the A L M R concept with a sodium coolant. Except for the modular NPP or NHPP, this type of R F may be used in nuclear power-technological plants, special NPP for solving the problems of utilizing the reactor and weapon grade plutonium, as well as the actinide transmutation. For example, SSC RF-IPPE and RDB 'Gidropress' have carried out a preliminary consideration of the possibility of using these R F as a source of mean potential heat and electric power at the enterprises for reprocessing a brown coal into a liquid motor fuel. This revealed (Gromov et al., 1992c) that the application of three RF of BRUS-150 type (Myasnikov et al., 1992) of total thermal power of 1500 MW relieved the enterprise, which annual production of 1.3 million tons of synthetic liquid fuel, from 6 million tons of coal consumption per year for energy and technology purposes. 4.6. Fast power reactor operating in open nuclear fuel cycle with slightly enriched or depleted uranium make-up The duration of NP operation stage in an open nuclear fuel cycle (NFC) with sufficient capacities of enriched uranium producing is determined by the total power of NPP, the resources of comparatively cheap natural uranium and the efficiency of its energy potential use. This efficiency does not exceed ~ 0.5% in reactors of a traditional type. However, as for FR its operation may be performed in a non-traditional way, i.e. with a makeup by a slightly enriched or depleted uranium at a partial refuelling. In this case the efficiency of a natural uranium use is increased several times, the time of the FR operation in an open N F C increases accordingly (Subbotin et al., 1994). In such a reactor core, the start-up loading with a comparatively highly enriched uranium fuel (10-12%) is realized only once (U-Pu fuel may be used, as well). During the reactor operation in the course of partial refuellings the fuel subassemblies (FSA) of a start-up loading are gradually replaced
by earlier loaded ones, in which plutonium has been already built-up, and fresh make-up FSAs are introduced instead of them. There are no separate breeding blankets in such a reactor. The fast reactor operation in such a regime is not in conformiry with the accepted view-point on the FR role in NP, as in this case a built-up plutonium is neither extracted nor reused, but utilized directly inside the reactor to a great extent. From the view-point of a fuel use, this regime is much more effective in comparison with that of NPP with light-water reactors. If there is a purpose to achieve the most high parameters of the R F safety, it is expedient to use, as a primary circuit coolant, chemically inert (at contact with water and air) liquid metal coolants with a very high boiling point, i.e. the lead bismuth eutectic or the lead only. The operation of NPP reactors in an open N F C without reprocessing a spent nuclear fuel (SNF) and the related construction and operation of close NFC plants provide for a higher level of NP system safety ( N P P - N F C plants), the ecological tolerance and NP social acceptability, and allow also to reduce expenses and exclude, in principle, the risk of plutonium proliferation. Use of this type of reactors (LBFR-600) allows to develop NP without reprocessing SNF for a long period (200 300 years). However, the development and commissioning of the reactors considered would require a long period, since to maintain the criticality due to plutonium built-up in FSA make-ups, the fuel element reliability should be provided at a deep burn-up ( ~ 20% b.a.), a high radiation damage dose of fast neutrons ( ~ 350 dpa) and a long lifetime ( ~ 25 years). A number of engineering problems should be solved as well. The factor, which restricts the use of such reactors with LBC in the future large-scale NP, is an insufficient contemporary bismuth production in comparison with that of lead, being widely used in the industry, therefore, it is necessary to increase the bismuth production in case of NP significant development with using LBC. Bismuth cost, which is ten times that of lead, makes a very small share (about 1%) of capital expenses for NPP
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construction. It should be taken into consideration that LBC is not spent and may be reused in other RFs. In the future a pure lead coolant proposed by R D I P E (Adamov and Orlov, 1994) may be used after an appropriate development period. Yet a temperature level of a lead coolant should be considerably increased due to a higher lead melting point (327°C). It aggravates the problems of the coolant technology and corrosion resistance of structural materials. It required about 15 years to solve these problems for LBC. Besides, R F operation becomes considerably complicated due to a great probability of a coolant solidification in the primary circuit at accidents, transients, repairs and refuelling. At present the development works on a lead coolant study are at the initial stage. 4. 7. Accelerator-driven systems with liquid-metal target
One of the fast-developing areas of the nuclear engineering is associated with using high-current proton accelerators for realization of intensive neutron sources, which can be used in scientific researches, in medicine, as well as in a subcritical blanket with a high neutron flux for Pu utilization and longlived RAW transmutation (Rubbia, 1994), (Bowman et al., 1991), (Bauer, 1994) and (Takahashi, 1985). The expediency of using heavy metals as a target material for a proton beam is caused by a high neutron yield in spallation reactions for these materials. This necessity to apply a liquid metal target is associated with the heat removal problem at a high energy release density. Thus, lead-bismuth or lead seem to be the best materials for liquid metal targets. The use of a molten-salt with dissolved minor actinides, plutonium and thorium in a subcritical blanket with a high neutron flux is considered to be one of the advanced directions of the nuclear transmutation (Furukawa et al., 1980), that allows to solve the problem of longlived a-active RAW. The main advantage of the accelerator-driven systems is elimination a of the runaway on prompt neutrons in the blanket with a sufficiently large sub-criticality (Keff ~ 0.95).
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At present, within the frameworks of the ISTC project No. 559, IPPE and RDB 'Gidropress' are carrying out the development of a lead-bismuth target circuit for a L A N L linear accelerator with a beam power of 1 MW. The target circuit has a thin 'window' from ferritic-martensitic steel, which separates a gas cavity from LBC. The target circuit includes a pump, a heat exchanger, a system of heating, sensors of the monitoring and control system. A lead-bismuth temperature at the target outlet is equal to 320°C. A flow rate of a liquid metal is about 15 m 3 per h. An important problem is ensuring radiation damage stability of the target unit materials, being in intensive fields of protons and secondary neutrons. A specific feature of the target circuit is a formation of new chemical elements by a proton beam as a result of spallation reactions on lead and bismuth which can influence on the coolant technology. However, this influence does not seem to be essential (Gromov et al., 1996b). As a result of nuclear reactions in the lead-bismuth eutectic there are formed rather long-lived radionuclides of 2°8po and 2°9po, which are practically absent in the R F coolant, where 21°po is a dominant a-active nuclide. However, the radiation danger from ~-active nuclides will be also determined by 21°po in the course of the target circuit operation and the year after the accelerator shut-off (Gromov et al., 1996b). In 1999 the target circuit is proposed to be put for tests at the accelerator beam in L A N L after manufacturing and testing in IPPE. Further it is assumed that a much more powerful target of 20 M W will be developed with 'windowless' technology applying, probably. 4.8. Molten salt reactor cooled by L B C
An alternate option for plutonium utilization and highly active R A W transmutation, which is also considered to be promising, is the use of molten salt reactors (MSR) (Lecocq and Furukama, 1994). The runaway on prompt neutrons is eliminated in MSR due to a very low reactivity margin, which is peculiar to reactors with a circulating liquid fuel. In case of heat removal failure
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the safety in MSR is provided by negative feedbacks. In comparison, an accelerator-driven system with a sub-critical molten-salt blanket, requires an urgent shut-sown of the accelerator at such an accident. However, despite of a successful operation experience of an experimental reactor MSRE of 8 M W power (Oak-Ridge), there is a certain apprehension that in the event of a significant increase (hundred times or more) of the reactor power and its lifetime, some problems might arise concerning the component maintenance of the primary circuit, in which a high radioactive molten salt circulates. These problems may be avoided or their influence is essentially decreased if the molten salt circulation system fulfils only the function of maintaining a molten salt necessary composition (a make-up by fissionable and transmutation elements and removal of fission and transmutation products). The heat removal is ensured by LBC which is pumped through the core. In this case the flow rate and the required molten salt amount in the primary circuit are reduced by a few hundred times. Accordingly, the sizes and power of the molten salt circuit components decrease, their repair and replacement are simplified. The risk of a large molten salt leak decreases at an emergency loss of tightness in pipelines and facilities, the problem of the secondary criticality is nearly excluded. Besides, under such a reactor cooling circuit, the safety increases due to the growth of a delayed neutron effective share, since a molten salt circulates through the core very slowly, it allows to increase concentration of minor actinides in a molten salt and the efficiency of their transmutation. As the calculations reveal, it is possible to obtain a temperature outlet inlet difference about 30°C at a sufficiently large flow rate of LBC. Thus, the maximum temperature of heat transfer tubes is not higher than 600°C that allows to use steel mastered marks. The conceptual development of an experimental molten salt reactor cooled by LBC with the thermal capacity of 60 MW has been begun together with RSC 'Kurchatov Institute' and RDB 'Gidropress'.
It should be noted that the proposed molten salt reactor cooling scheme will be also effective for cooling a subcritical molten salt blanket in the accelerator-driven system. In this case the LBC target circuit may be included into the structure of the blanket cooling circuit.
5. Conclusion
The presented experience of using a lead-bismuth coolant, the concept of the safety provision of a new RF generation, being developed on the basis of the experience accumulated, as well as the proposals of using this coolant in the future NP display this N P development direction as showing promise. RF with LBC allows to ensure an essentially higher level of safety, which excludes deterministically severe accidents. Due to internal self-protection properties the amount of protecting and localizing safety systems is sharply reduced. It decreases capital and operational expenses and increases NPP competitiveness. A new approach, proposed for NP fuel supply during a long period, is provided by the use of non-traditional FR with LBC and allows to develop NP, postponing the fuel reprocessing by 200-300 years that excludes in practice the risk of plutonium proliferation. The LBC using appears effective in acceleratordriven systems and molten salt reactors for transmutation of long-lived RAW and for other purposes. A gradual introduction of LBC reactors in NP, and in the future, if necessary, of lead cooled reactors might allow to overcome the present crisis of N P (Stadie, 1996). Certainly, the adaptation of accumulated experience of a lead bismuth coolant use to R F operation conditions at NPPs is required accompanying by a necessary performance of some appropriate research and development. Some of the proposals developed (RF is of a low power, less than 75 MW) can be realized in sufficiently short terms (5 7 years), since they are based on the engineering approaches having been already settled. Development and deployment of
B.F. Gromov et al./Nuclear Engineering and Design 173 (1997) 207-217 RFs of a greater power would require more time owing to some existing problems. The ways of solving these problems are rather clear.
References Adamov, E.O., Orlov, V.V., 1994. Requirements to a New Nuclear Technology for the Large-Scale Power Industry. Proc. ARS '94 Int. Topical Meet. on Advanced Reactors Safety. Hyatt Regency, Pittsburgh, PA, 2, 636-642. Bauer, G., 1994. The Spallation Neutron Source SINQ. Switzerland, PSI. Berglund, R.C., Tippets, F.E., 1989. PRISM, the plant design concept for the US advanced liquid metal reactor program. 6th American Power Conference, Chicago. Bowman, C.P., Arthur, E.D., Cappiello, M.W. et al., 1991. Nuclear Energy Generation and Waste Transmutation Using an Accelerator-Driven Intense Neutron Source. Report, LANL. LA-UR-91-2601. Carre, F., Lemercier, G., Delpech, M., 1993. Brief and preliminary assessment of lead cooled fast reactors option. Centre d'Etudes de Cadarache. Franco-Russian exchange meeting on Liquid Metal fast Reactors, France, Paris. Furukawa, K., Tsukada, K., Nakahara, Y. 1980. M.S. Target and Blanket Concept. Proc. 4th ICONS, pp. 349-354. Gromov, B.F., Subbotin, V.I., Toshinsky, G.I., 1992. Application of Lead-Bismuth Eutectic and Lead Melts as Nuclear Power Plant Coolant. ISSN 0004-7163 Nuclear Energy, Moscow, 73, 19-24. Gromov, B.F., Toshinsky, G.I., Zelensky, V.N. et al., 1992b. Small Power nuclear power plant with lead-bismuth coolant. The paper on the Conference of Nuclear Society (Moscow), St. Petersburg. Gromov, B.F., Kozlovcev, V.G., Chekunov, V.V. et al. 1992. Energieproviding large-scale plants synthetic liquid fuel from nuclear power-technology plant on fast neutrons, Energy building, Moscow, 6, 34-39. Gromov, B.F., Checunov, V.V. et al., (IPPE), Stepanov, V.S., Kolik, M.V. et al., 1993. (RDB 'Gidropress'). Lead-bismuth cooled reactor plants with inherent safety. Radiation and Ecological Problems Related to Nuclear Power and Production Conversion. Symp. XV Mendeleev Cong. on General and Applied Chemistry. Obninsk, Russia. Gromov, B.F., Toshinsky, G.I., Stepanov, V.S. et al., 1993. Floating nuclear power plant electrical capacity of 50 MW-e with lead bismuth cooled reactor of high safety. Conf. of Nuclear Society (Moscow), NE-93. Nizhniy Novgorod. Gromov, B.F., Orlov, Yu.I., Toshinsky, G.I., Chekunov, V.V., 1996. Nuclear Power Facility with Lead-Bismuth Coolant. ISSN 0004-7163, Nuclear Energy, Moscow, 81, 340-347.
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Gromov, B.F., Leonchuk, M.P., Yefimov, E.I. et al., 1996. Flow Lead-Bismuth Target for High Energy Protons as an Intense Neutron Source in Accelerator-Driven Systems. ISSN 0004-7163, Nuclear Energy, Moscow, 80, 400 407. Hattori, S., Minato, A., 1995. A Large Modular LMR Power Station which Meets Current Requirements. Proc. of ICONE-3, Kyoto, Japan, 2, 787-790. von Hippel, Frank, 1995. Evolutionary Approach to Fission Power. Princeton University, USA. Int. Conf. on Evaluation of Emerging Nuclear Fuel Cycle Systems, Global-95. Versailles, France, 1, 380-387. Lecocq, A., Furukama, K., 1994. Rational Pu-Transmutation for U-233. 1AEA-Puu Tech. Meeting. Obninsk, Russia. Leipunsky, A.I., 1990. Elected labour. Memoirs, Naukova dumka, Kiev, pp. 225-228. Myasnikov, V.O., Stekolnikov, V.V., V.S. Stepanov, V., (RDB 'Gidropress'), Gromov, B.F., Pashkin, Yu.G., Chekunov, V.V. et al. (IPPE), 1992. Conceptual design of modular NPP of ultimate safety with lead-bismuth cooled fast reactor. Conf. Nuclear Society, St. Petersburg, Moscow. Nilsen, T., Bemer, N., 1994. Sources of radioactive pollution in Murmansk and Arkhangelsk areas. Books 8874, Youngstorget N-0028, Oslo, Norway. The report of association 'Bellona', Version 1. Rubbia, C., 1994. Status Report on the energy Amplifier. CERN, Geneva. Sekimoto, H., Su'ud, Z. (Japan), 1995. Design study of leadand lead-bismuth-cooled small long-life nuclear power reactors using metallic and nitride fuel. Nucl. Technol. (109) 307-313. Stadie, K., 1996. The rise and flaws of nuclear power. Nucl. Eng. Int. 502 (41), 23-24. Subbotin, V.I., Matveev, V.I., Toshinsky, G.I., 1994. Fast reactors with lead bismuth coolant in future nuclear power. Proc. of ARS '94 Int. Topical Meet. of Advanced Reactor Safety. Hyatt Regency, Pittsburgh, PA, USA, pp. 524-529. Takahashi, H., 1985. Actinide Transmutation by the Spallation Process. Workshop on the Feasibility of Res. Prog. in Actinide Transmutation by Spallation Process, Euratom, Ispra, Varese, Haly. Takano, H., Akie, H., Handa, M. et al., 1993. A Concept of Self completed Fuel Cycle based on Lead-cooled Nitridefuel Fast Reactors. Report on a seminar in RDIPE (Russia). Tupper, R.B., Minushkin, B., Peters, F.E., Kardos, Z.L., (Westinghous Electric), 1991. Polonium Hazards associated with Lead Bismuth used as a Reactor Coolant. Int. Conf. on Fast Reactors and Related Fuel Cycles. Kyoto, Japan. Proc., 4, 5 6. Tupper, R.B., Wett, J.F., (Westinghouse Electric, USA), 1992. The Market Potential for Heavy Metal Cooled reactors. 3rd Ann. Sci. Conf. Nuclear Society International,, Book of Abstracts. St. Petersburg, Moscow.
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Nuclear Engineering and Design
Use of lead-bismuth coolant in nuclear reactors and accelerator-driven systems B.F. G r o m o v a,,, Yu.S. Belomitcev a, E.I. Yefimov a, M.P. Leonchuk ~, P.N. M a r t i n o v a, YU.I. Orlov ~, D.V. P a n k r a t o v a, Yu.G. Pashkin ~, G.I. Toshinsky a, V.V. C h e k u n o v a, B.A. Shmatko a, V.S. Stepanov b " State Scientific Centre RF, Institute of Physics and Power Engineering, Bondarenko Sq. 1, Obninsk, Kaluga Region 249020, Russia b RDB Gidropress, Ordzhonikidze st. 7, Podolsk, Moscow Region, Russia
Abstract
Experience of using lead-bismuth coolant in reactors of Russian nuclear submarines is briefly presented. The salient points of the concept providing the safety of reactor facilities cooled by a lead-bismuth eutectic are covered. The key results of developments for use of a lead-bismuth coolant in nuclear reactors and accelerator-driven systems are presented. © 1997 Elsevier Science S.A.
1. Introduction
In the beginning o f the 1950s the U S A and U S S R began nearly simultaneously a nuclear power facility ( N P F ) development for nuclear submarines (NS). In b o t h countries the w o r k was carried out on two types o f N P F - - w i t h pressurized water reactors and with liquid metal cooled ones ( L M R ) .
Abbreviations: NPP, nuclear power plant; NHPP, nuclear heating power plant; RF, reactor facilities; TES-M, index of small power nuclear heating plant; RDIPE, Research and Design Institute of Power Engineering; RAW, radioactive waste; PRISM, power reactor inherantly safe modular; BRUS, index of fast reactor facilities; ALMR, advanced liquid metal reactor; LANL, Los Almos National Laboratory; MSRE, molten salt reactor experiment. * Corresponding author.
In the U S S R , where the works were begun in 1952, a l e a d - b i s m u t h eutectic was chosen as a coolant for L M R ( G r o m o v et al., 1992a). A scientific supervision o f this w o r k was provided by academician A.I. Leipunsky (IPPE, Obninsk) and after his death in 1972, by Prof. B.F. G r o m o v . In the U S A , sodium, as possessing better thermophysical features, was chosen as a coolant for L M R . The g r o u n d test facility o f N P F and an experimental submarine 'Sea W o l f ' have been constructed. Yet the operation experience has revealed that the choice o f a fire-and-explosion dangerous at contact with air and water coolant did not justify itself. After a n u m b e r o f accidents the reactor facility ( R F ) o f this N S was dismantled together with the c o m p a r t m e n t and replaced by a pressurized water R F .
0029-5493/97/$17.00 © 1997 Elsevier Science S.A. All rights reserved. PII S0029-5493(97)00110-6
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In the USA there were also conducted research and development on using a lead-bismuth coolant (LBC). But the alternative of solving the corrosion resistance problem of structural materials, control and maintenance of the coolant quality (the coolant technology) gave no positive results and those works were stopped. In our country as a result of systematic work of a number of organizations in the course of about 15 years there was solved a complex problem of the coolant technology, the corrosion of structural materials and the mass transfer, that ensured a long and reliable operation of N P F with LBC under NS conditions. Proceeding from the experience, obtained under development and operation of N P F with LBC, both the concept of maintaining a new RF generation safety for civil purposes and the proposals on using LBC in the nuclear power (NP) have been developed.
2. Experience of using lead-bismuth coolants The expediency of using a lead-bismuth eutectic (Pb is ~ 44%, Bi is ~ 56%) as a coolant of the primary circuit of nuclear reactors is caused by its physical, chemical and thermodynamic properties which allow to meet the safety requirements to the most complete degree. Due to a very high boiling point ( ~ 1670°C) it is allowable to have a low pressure in the primary circuit when the outlet coolant temperature is equal to 400-500°C. This simplifies a RF design and increases its reliability. A low pressure in the primary circuit enables to reduce a wall thickness of the reactor vessel and introduce no restrictions to the temperature change rate in compliance with a thermocyclic strength, that provides R F operation in a manoeuvre mode. The LBC low chemical activity allows to use a two-circuit scheme of the heat removal. Emergency processes related to the primary circuit loss of tightness and intercircuit leaks in a steam generator (SG) occur without hydrogen formation, fires and explosions.
In the course of operations a number of engineering problems, related directly to LBC features, has been solved. The LBC melting point accounts for ~ 125°C. LBC maintaining in a liquid state under all RF operation conditions is provided by using SG with a multiple circulation of a steam-water mixture over the secondary circuit, therefore the inlet temperature of the water supplied to SG is higher than the LBC melting point. The system of a steam or electrical heating can be used for the initial heat-up and keeping the primary circuit in a hot condition at a low level of heat releases in the core. An important practical problem is the substantiation of an opportunity of the multiple 'solidification-remelting' of LBC in R F that could be required in prolonged shut-down regimes. A low shrinkage of LBC under solidification and rather a high plasticity with a low strength in a solid state promote realization of these regimes without a structural material damage at LBC transition from a liquid into a solid state and its further cooling down up to the ambient temperature. A special technology of a safe remelting has been developed. When carrying out maintenance and reactor refuelling, there is no need in the primary circuit decontamination, which is associated with assembling, storing, transporting and reprocessing of a plenty of liquid radioactive wastes (LRW). A specific character of LBC, when bismuth is exposed by neutrons, is the formation of ~-active radionuclide of polonium-210 with a lifetime of 138 days. A major reason of its radiation danger is the formation of radioactive aerosols and a volatile polonium hydride when a hot LBC contacts with air. It may take place under conditions of the primary circuit tightness loss and the coolant leakage. In this case, as the RF operation experience at the NS has displayed, the release of Po aerosols and the air radioactivity (according to the thermodynamics laws) reduce strongly with the temperature decrease and solidification of the coolant leakage. A fast solidification of the LBC leakages restricts the area of radioactive contamination and simplifies its removal in the form of solid radioactive wastes.
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A low concentration of Po in the coolant (at the level of 10 - s at.%) and the formation of a thermodynamically p r o o f chemical compound of polonium with lead, decrease polonium volatility by eleven orders of magnitude in comparison with a sole metal polonium. In addition, for safety ensuring there were developed means of an individual and collective personnel protection, methods of an equipment decontamination and radioactivity fixation on surfaces. All the above has resulted in the fact that for a long-term period of N P F operation with a lead-bismuth coolant, including the primary circuit repairs and removal of coolant leakages, there were no cases of personnel extrairradiation by this radionuclide above the permissable limits. This positive practical result coincides with the conclusions of foreign experts who have experimentally investigated the Po danger problem when radioactive lead-bismuth is used as a nuclear reactor coolant (Tupper et al., 1991). Among basic problems, which have been solved during designing, development and operation of this type of facilities, it is necessary to highlight the problem of LBC technology, i.e. the development of the systems and devices, which ensure controlling and supporting the LBC required quality during a long operation under normal conditions of a tight circuit and in the event of SG seal failure, as well as at the circuit opening for repairs and the reactor refuelling. The application of such a technology is necessary for eliminating a structural material corrosion and slagging the circuit by lead oxides. To settle these problems the following has been developed: the special filters for cleaning LBC from insoluble impurities, the devices to ensure chemical reduction of a lead oxide, the controlling devices to sustain in LBC a necessary concentration of a corrosion inhibitor-dissolved oxygen, the appropriate sensors to indicate the quality of LBC and the protective inert gas. The corrosion resistance of materials is provided by the alloying of steels, a preliminary formation of protective films on their surfaces and maintaining the necessary concentration of a dissolved oxygen in LBC.
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A nuclear reactor with an intermediate neutron spectrum was applied as an energy source in NPF. A metal beryllium served as a neutron moderator. Beryllium available in the core afforded a neutron flux high level in a subcritical reactor because of a photo-neutron reaction with y-rays of fission products that in turn resulted in a safe and rapid reactor start-up. The following characteristics were obtained in the course of N P F tests and operation: the facility capacity and parameters, the lifetime, the reactivity margin, reactivity coefficients, poisoning effects, temperature distributions, dynamic parameters, the coolant radioactivity, dose rates of neutron and y-radiations behind the shield and some other parameters. They were in a sufficiently good agreement with the calculation results. A high operation capability was demonstrated by the cores and control rods. During last 10-15 years of N P F operation there were no problems with either corrosion of structural materials in the primary circuit or the deviations from the standards of a circuit purity. Experience of N P F with LBC development and operation at NS allows to make a number of practically important conclusions and recommendations, concerning the LBC reactor primary circuit at civil NPPs and transportable NPFs. The best technology parameters should be expected in integral design reactors. The most convenient design scheme of a steam generator is the one in which a liquid metal circulation takes place in the space between pipes, and that of water or steam does in pipes. At such a design the possibility of repairing SG is provided by plugging a separate pipe, which has lost its tightness, without dismantling SG or opening the primary circuit. Shut-down regimes, the regimes of start-up and heat decay removal are more easy realized when SG operates with a multiple circulation in a steam-water circuit. Mechanical vertical pumps with a turbine or an electrical driver as well as the electromagnetic ones may be used for the LBC circulation. For the past period of time, there were constructed eight NSs having R F with LBC (Nilsen and Bemer, 1994). The first experimental NS of
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the project 645 had two reactors, other seven NS of the project 705 (in NATO's terms, 'Alpha') had one reactor. Due to its high-speed qualities this NS was put into the Guinness Book. Besides, two full-scale ground reactor test facilities were constructed and put into operation: N P F prototype in IPPE (Obninsk) and in NITI (Sosnovy Bor). A general operating time of R F was about 80 reactor-years. The R F were designed by RDB 'Gidropress' (Podolsk) and EDBM (Nizhniy Novgorod), the relevant NS were designed by SPMBM 'Malakhit' (St. Pertersburg). The experience of RF design and operation at NS has been used for the safety concept development of R F with LBC.
3. Concept of safety ensuring Realization priority of safety requirements in comparison with other requirements has become obvious after the Chernobyl accident. An intensive development of NPP new concepts for the future has been begun all over the world, their safety is based on a R F inherent self-protection to a considerable degree. It allows to exclude deterministically the possibility of the most severe accidents, which results in catastrophic consequences not only in the event of safety technical system failure and personnel errors, but also in the event of ill-intentioned actions, a subversive activity, an influence of usual explosive substances (terrorism, military attacks). The facilities with a L M C fast reactor (FR) are referred to RF, whose safety is maintained principally due to an inherent self-protection. It is associated with a number of their internal features. An absence of poisoning effects in a FR, a low value of a negative temperature reactivity coefficient, the compensation of fuel burn-up processes and slagging by plutonium production, as well as partial refuellings allow to ensure the operative reactivity margin (in an operating reactor), which is less than a share of delayed neutrons, and to exclude in this way the runaway on prompt neutrons.
Sodium is now widely used as LMC in NP. The choice of this L M C for fast reactors was caused by an intensive heat removal due to good thermal and physical properties. Hence, a short doubling time of plutonium was achieved that appeared an indispensable requirement at early development and construction stages of fast breeder reactors in the sixtieth-seventieth. Just for this reason academician. A.I. Leipunsky, who considered various LMC for cooling F R in the fiftieth, preferred sodium, though from a safety view-point LBC was originally considered for these purposes (Leipunsky, 1990). At present and in prospect, such a short doubling time of plutonium which can ensure sodium cooled FR is not required. Besides, the necessity of developing breeder reactors is postponed for many decades (von Hippel, 1995). It makes reasonable to return to using LBC in NP, taking into account the experience obtained. A lead-bismuth coolant due to its chemical inertness and a higher boiling point in comparison with sodium allows to reach even a better FR safety than in the case of sodium. Since the energy accumulated in the LBC (thermal, chemical and a potential energy of compression) is the minimum one in comparison with other coolants and, taking into consideration the FR features mentioned above, it should be expected that the RF design with LBC meets the requirements of the concept of the maximum achievable inherent safety (Adamov and Orlov, 1994). At present, there are two approaches of ensuring the RF safety. The first approach, being traditional, is based on extending the number and increasing the efficiency of various protective and localizing systems and devices, which decrease the probability of heavy accidents and reduce a danger of their consequences. A practical realization of this approach results in a more complex and expensive facility, a deterioration of its other characteristics and does not exclude, in principle, an opportunity of a severe accident with catastrophic consequences because internal reasons of its realization have not been eliminated. The second approach is based on the concept of RF with an inherent (natural) safety ensuring the
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R F self-protection, in which the reasons for severe accidents by the nature laws are deterministically excluded. With this approach, the construction of protective and localizing systems is not required. There is also no need in a complex substantiation of the safety by means of many calculational and experimental investigations in a framework of abstract scenarios of a severe accident progression and by constructing large-scale test facilities. This approach is most consistently developed by Prof. A. Weinberg in the USA, and then by Prof. V. Orlov in Russia. Neither first nor the second approaches are realized in a pure mode as each of them possesses elements of the other. Nevertheless, proceeding from the safety modern philosophy at developing the advanced RF, it seems necessary to apply the approach, basing at most on the inherent self-protection, as far as the most severe accidents are concerned. This conclusion is justified by the fact that the probabilistic safety analysis (PSA), whose results are taken for substantiation of a R F safety in the first approach, considers failures of technical devices and errors of operational personnel as casual events, which probability may be estimated only with a large uncertainty because of insufficient statistical data. A small probability of a severe accident is neither proof of its impossibility nor the evidence that it might happen not earlier than thousands or tens of thousands of years later. Besides, with ill-intentioned actions of people, that should be taken into account, these events will be not casual but programmed beforehand. In such a case the PSA conclusions lose their validity. The R F safety with LBC is provided by so natural properties as a very high boiling point and an evaporation latent heat. It excludes practically a possibility of the primary circuit overpressurizing and the reactor thermal explosion at a coolant emergency overheating, because there is no pressure increase. Besides, a low pressure allows to reduce the reactor vessel thickness and to use for its construction a less strong austenitic steel, being resistant to a radiation embrittlement under operating conditions. Owing to the impossibility of the coolant boiling, the heat removal reliability from
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the core and the safety are increased as there is no heat removal crisis. Loss of a coolant with its circulation interruption through the core at failure of the reactor vessel tightness (beyond design accident) is ruled out due to a guard shell availability and a small free volume between the reactor vessel and the guard shell. In the event of the primary circuit gas system failure it is ruled out due to the coolant boiling out impossibility. In case of all heat exchanger failure in the independent emergency cooling system (beyond design accident) the core melting-down under a decay heat effect and a reactor vessel break for facilities of a small and average capacity are prevented due to a large difference between an arising temperature and the coolant boiling point. It occurs because of the heat accumulation in the internal reactor structures and in the coolant itself and because of the heat removal through the reactor vessel by means of a natural circulation of the ambient air. The reactor power decrease in the case of emergency overheating and simultaneous failure of the emergency protection systems is provided by reactivity negative feedbacks. In the core and within the reactor vessel there are no materials releasing hydrogen under irradiation or as a result of chemical reactions with the coolant. The coolant itself reacts very slightly with water and air, their contact is possible in case of the circuit failure, and no explosions and fires take place. The operation experience of R F with LBC at nuclear submarines has revealed R F safe operation during some time under conditions of a small SG leak which results in no significant deviations of engineering parameters from the designed ones. This fact gives the possibility to realize necessary repair measures not urgently but in a convenient time. The elimination of a water or steam penetration into the core and consequent overpressurizing the reactor vessel, designed for the pressure occurring at a large SG leak, are provided by a coolant circulation scheme. With this scheme, bubbles of steam and drops of water are thrown out on a free level by a coolant ascending flow. A steam
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effective separation occurs therefore in the gas volume above the coolant level, whence steam goes to the emergency condensers operating on a passive principle. The coolant does not boil at the primary circuit failure and has the property to retain iodine, whose radionuclides represent, as a rule, a major factor of a radiation danger just after an accident. This reduces sharply a scale of radiation consequences of so accident in comparison with the pressurized water reactors. The containment serves as an additional safety system barrier, which mainly protects against external actions. A low storage of a potential energy in the primary circuit restricts a R F destruction scale under external action conditions. An extremely high safety potential, which is peculiar to R F of this type, is characterized by the fact that when such initial events as the containment destruction and the primary circuit failure coincide (it is possible in the course of subversive or military activities), neither reactor runaway, nor explosion and fire occur, and the radioactivity release is essentially lower than the level requiring the population evacuation. Thus, the use of such a type of reactors is possible and expedient not only for an electric power generation at NPP but also for simultaneous production of heat at N H P P with locating them near large cities, as well as for the sea water desalination. It should be mentioned that in association with the factors presented, the interest in using lead bismuth or lead coolants has been lately displayed in other countries (Tupper and Wett, 1992; (Sekim o t o and Su'ud, 1995; Carte et al., 1993; T a k a n o et al., 1993).
4. Proposals on using lead-bismuth coolant A number of proposals on civil N P P of a small and average power (from 1 to 600 MW) in transportable, floating and ground performances has been lately developed on the basis of the concept presented and the experience obtained. These proposals were developed in the frameworks of defensive work conversion under a scientific
supervision of the State Scientific Centre Russian F e d e r a t i o n - - t h e A.I. Leipunsky Institute of Physics and Power Engineering (SSC RF-IPPE) and R D B 'Gidropress'. A brief description of them is presented below. 4.1. Modular-transportable N H P P T E S - M with electric power o f 1 M W
The use of nuclear stations of small power might appear expedient for far regions having no centralized electric energy supply, and with complicated conditions of an organic fuel delivery. As the principal requirements for small power NPP (SPNPP) are safety, reliability and simplicity, the following basic concepts were accepted in SPNPP TES-M conceptual design ( G r o m o v et al., 1992b), (the engineering development of this SPN P P was carried out in C D B M St. Petersburg); • the circulation of a coolant in a primary circuit in all regimes is realized by natural convection; • all the equipment of a primary circuit is concentrated in the same vessel (integral design); • the air is used as a coolant of a secondary circuit and simultaneously drives a gas turbine; • the transportation of a nuclear heating unit ( N H U ) to SPNPP site is realized at a complete factory readiness with an assembled reactor, filled by a solidificated coolant; • the elimination of the core refuelling on SPNPP site. For this purpose the core lifetime is provided to be 90000 eft. h (10-15 years); • the refuelling of the spent core is designed in the central maintenance factory where N H U with a solidificated coolant is delivered. The coolant features and the use of the air in the secondary circuit exclude N P P equipment failure and release of radioactive products in the environment in the event of N H U temperature decrease below zero that may occur in emergency situations, accompanying by loss of internal and external heat and electric energy. Realization of the core emergency cooldown is provided by using only passive ways of heat removal. SPNPP simplicity and reliability provide for its operation under automatic operation conditions without a constant maintenance.
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Maximum weight of the transport unit is about 50 tons. 4.2. Modular-transportable N H P P 'Angstrem'
N H P P 'Angstrem' is to provide by heat and electric power the settlements with up to 5000 inhabitants. N H P P 'Angstrem' of a stationary ground performance is to be assembled out of factory functionally completed units on a preliminary prepared site over 1 month. N H P P can be transported by 9 12 railway platforms and by other transport vehicles as well. The maximum weight of the unit is 220 tons. The primary circuit consists of two loops with a compulsory circulation. A steam turbine design allows to take steam for heat supply. The air-radiator cooling of turbine condensers is designed so that it makes possible to use N H P P 'Angstrem' in waterless regions (Gromov et al., 1993a). The electric power is 6 MW, the heat supply is equal to 12 Gcal per h, the operation term without a refuelling is ~ 6 years. 4.3. Floating N P P 'Cruise-50'
Many coastal regions have no reliable sources of power supply. For economical or ecological reasons the use of the floating NPP with a high safety potential may show promise. For this purpose, the conceptual design of the floating NPP 'Cruise-50' with the electric power of 50 M W has been developed with participation of SPMBM 'Malakhit' and the 'Turbine factory' in Kaluga. This NPP may be also used for the sea water desalination or the heat production (Gromov et al., 1993b). This reactor facility has an integral structure with a compulsory circulation (two pumps). Steam generators produce a saturated steam at the pressure of 5.6 MPa. The refuelling is carried out at the central service factory, where the floating N P P is transported once over 6 years.
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4.4. Modular N P P with R F L B F R - 7 5
In the course of 2003 2005 the third and fourth units with reactors VVER-440 of the Novovoronezh NPP (NVNPP) are to be decommissioned. The first and second units have been already shut-off. For replacing the units decommissioned, the construction of the sixth and seventh units is proposed that will require large capital investments. In this association, SSC RF-IPPE, RDB 'Gidropress' and 'Atomenergoproekt' have carried out a technological investigation for determining a technical possibility and economical feasibility of the N V N P P 1, 2, 3 and 4 units renewal when their service life is exhausted. This renewal is proposed to carry out by means of a dismantled SG replacing with nuclear steam generating modules of an integral design with LBC fast reactors (Gromov et al., 1996a). Such modules (the power of each is 75 MW) allow to replace the removed capacities and to use the existing capital structures and the equipment of the Novovoronezh NPP. Thus, the regulation safety requirements will be met, the radioecological conditions will be improved due to a nearly total absence of liquid radioactive wastes (LRW), the NPP operation in a manoeuvre mode will be provided. It should be expected a significant saving of the expenditures for a N P P renewal as compared to the construction of new VVER units, when taking into account the circumstances and an essential reduction of the auxiliary systems and the equipment, the safety systems, a nearly total LRW absence, a complete factory production of the module carried by rail to the NPP site, as well as a high serial production. 4.5. Reactor facilities o f B R U S type
The following group of the R F considered with a higher power of 150-300 M W (BRUS-150, BRUS-220, BRUS-300), being in conformity with above presented safety concept, can solve the problem of modular N P P construction, which is similar to the PRISM concept of the USA
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(Berglund and Tippets, 1989) and 4S concept of Japan (Hattori and Minato, 1995). But an inherent safety principle is more consecutively realized in these reactors than in the A L M R concept with a sodium coolant. Except for the modular NPP or NHPP, this type of R F may be used in nuclear power-technological plants, special NPP for solving the problems of utilizing the reactor and weapon grade plutonium, as well as the actinide transmutation. For example, SSC RF-IPPE and RDB 'Gidropress' have carried out a preliminary consideration of the possibility of using these R F as a source of mean potential heat and electric power at the enterprises for reprocessing a brown coal into a liquid motor fuel. This revealed (Gromov et al., 1992c) that the application of three RF of BRUS-150 type (Myasnikov et al., 1992) of total thermal power of 1500 MW relieved the enterprise, which annual production of 1.3 million tons of synthetic liquid fuel, from 6 million tons of coal consumption per year for energy and technology purposes. 4.6. Fast power reactor operating in open nuclear fuel cycle with slightly enriched or depleted uranium make-up The duration of NP operation stage in an open nuclear fuel cycle (NFC) with sufficient capacities of enriched uranium producing is determined by the total power of NPP, the resources of comparatively cheap natural uranium and the efficiency of its energy potential use. This efficiency does not exceed ~ 0.5% in reactors of a traditional type. However, as for FR its operation may be performed in a non-traditional way, i.e. with a makeup by a slightly enriched or depleted uranium at a partial refuelling. In this case the efficiency of a natural uranium use is increased several times, the time of the FR operation in an open N F C increases accordingly (Subbotin et al., 1994). In such a reactor core, the start-up loading with a comparatively highly enriched uranium fuel (10-12%) is realized only once (U-Pu fuel may be used, as well). During the reactor operation in the course of partial refuellings the fuel subassemblies (FSA) of a start-up loading are gradually replaced
by earlier loaded ones, in which plutonium has been already built-up, and fresh make-up FSAs are introduced instead of them. There are no separate breeding blankets in such a reactor. The fast reactor operation in such a regime is not in conformiry with the accepted view-point on the FR role in NP, as in this case a built-up plutonium is neither extracted nor reused, but utilized directly inside the reactor to a great extent. From the view-point of a fuel use, this regime is much more effective in comparison with that of NPP with light-water reactors. If there is a purpose to achieve the most high parameters of the R F safety, it is expedient to use, as a primary circuit coolant, chemically inert (at contact with water and air) liquid metal coolants with a very high boiling point, i.e. the lead bismuth eutectic or the lead only. The operation of NPP reactors in an open N F C without reprocessing a spent nuclear fuel (SNF) and the related construction and operation of close NFC plants provide for a higher level of NP system safety ( N P P - N F C plants), the ecological tolerance and NP social acceptability, and allow also to reduce expenses and exclude, in principle, the risk of plutonium proliferation. Use of this type of reactors (LBFR-600) allows to develop NP without reprocessing SNF for a long period (200 300 years). However, the development and commissioning of the reactors considered would require a long period, since to maintain the criticality due to plutonium built-up in FSA make-ups, the fuel element reliability should be provided at a deep burn-up ( ~ 20% b.a.), a high radiation damage dose of fast neutrons ( ~ 350 dpa) and a long lifetime ( ~ 25 years). A number of engineering problems should be solved as well. The factor, which restricts the use of such reactors with LBC in the future large-scale NP, is an insufficient contemporary bismuth production in comparison with that of lead, being widely used in the industry, therefore, it is necessary to increase the bismuth production in case of NP significant development with using LBC. Bismuth cost, which is ten times that of lead, makes a very small share (about 1%) of capital expenses for NPP
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construction. It should be taken into consideration that LBC is not spent and may be reused in other RFs. In the future a pure lead coolant proposed by R D I P E (Adamov and Orlov, 1994) may be used after an appropriate development period. Yet a temperature level of a lead coolant should be considerably increased due to a higher lead melting point (327°C). It aggravates the problems of the coolant technology and corrosion resistance of structural materials. It required about 15 years to solve these problems for LBC. Besides, R F operation becomes considerably complicated due to a great probability of a coolant solidification in the primary circuit at accidents, transients, repairs and refuelling. At present the development works on a lead coolant study are at the initial stage. 4. 7. Accelerator-driven systems with liquid-metal target
One of the fast-developing areas of the nuclear engineering is associated with using high-current proton accelerators for realization of intensive neutron sources, which can be used in scientific researches, in medicine, as well as in a subcritical blanket with a high neutron flux for Pu utilization and longlived RAW transmutation (Rubbia, 1994), (Bowman et al., 1991), (Bauer, 1994) and (Takahashi, 1985). The expediency of using heavy metals as a target material for a proton beam is caused by a high neutron yield in spallation reactions for these materials. This necessity to apply a liquid metal target is associated with the heat removal problem at a high energy release density. Thus, lead-bismuth or lead seem to be the best materials for liquid metal targets. The use of a molten-salt with dissolved minor actinides, plutonium and thorium in a subcritical blanket with a high neutron flux is considered to be one of the advanced directions of the nuclear transmutation (Furukawa et al., 1980), that allows to solve the problem of longlived a-active RAW. The main advantage of the accelerator-driven systems is elimination a of the runaway on prompt neutrons in the blanket with a sufficiently large sub-criticality (Keff ~ 0.95).
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At present, within the frameworks of the ISTC project No. 559, IPPE and RDB 'Gidropress' are carrying out the development of a lead-bismuth target circuit for a L A N L linear accelerator with a beam power of 1 MW. The target circuit has a thin 'window' from ferritic-martensitic steel, which separates a gas cavity from LBC. The target circuit includes a pump, a heat exchanger, a system of heating, sensors of the monitoring and control system. A lead-bismuth temperature at the target outlet is equal to 320°C. A flow rate of a liquid metal is about 15 m 3 per h. An important problem is ensuring radiation damage stability of the target unit materials, being in intensive fields of protons and secondary neutrons. A specific feature of the target circuit is a formation of new chemical elements by a proton beam as a result of spallation reactions on lead and bismuth which can influence on the coolant technology. However, this influence does not seem to be essential (Gromov et al., 1996b). As a result of nuclear reactions in the lead-bismuth eutectic there are formed rather long-lived radionuclides of 2°8po and 2°9po, which are practically absent in the R F coolant, where 21°po is a dominant a-active nuclide. However, the radiation danger from ~-active nuclides will be also determined by 21°po in the course of the target circuit operation and the year after the accelerator shut-off (Gromov et al., 1996b). In 1999 the target circuit is proposed to be put for tests at the accelerator beam in L A N L after manufacturing and testing in IPPE. Further it is assumed that a much more powerful target of 20 M W will be developed with 'windowless' technology applying, probably. 4.8. Molten salt reactor cooled by L B C
An alternate option for plutonium utilization and highly active R A W transmutation, which is also considered to be promising, is the use of molten salt reactors (MSR) (Lecocq and Furukama, 1994). The runaway on prompt neutrons is eliminated in MSR due to a very low reactivity margin, which is peculiar to reactors with a circulating liquid fuel. In case of heat removal failure
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the safety in MSR is provided by negative feedbacks. In comparison, an accelerator-driven system with a sub-critical molten-salt blanket, requires an urgent shut-sown of the accelerator at such an accident. However, despite of a successful operation experience of an experimental reactor MSRE of 8 M W power (Oak-Ridge), there is a certain apprehension that in the event of a significant increase (hundred times or more) of the reactor power and its lifetime, some problems might arise concerning the component maintenance of the primary circuit, in which a high radioactive molten salt circulates. These problems may be avoided or their influence is essentially decreased if the molten salt circulation system fulfils only the function of maintaining a molten salt necessary composition (a make-up by fissionable and transmutation elements and removal of fission and transmutation products). The heat removal is ensured by LBC which is pumped through the core. In this case the flow rate and the required molten salt amount in the primary circuit are reduced by a few hundred times. Accordingly, the sizes and power of the molten salt circuit components decrease, their repair and replacement are simplified. The risk of a large molten salt leak decreases at an emergency loss of tightness in pipelines and facilities, the problem of the secondary criticality is nearly excluded. Besides, under such a reactor cooling circuit, the safety increases due to the growth of a delayed neutron effective share, since a molten salt circulates through the core very slowly, it allows to increase concentration of minor actinides in a molten salt and the efficiency of their transmutation. As the calculations reveal, it is possible to obtain a temperature outlet inlet difference about 30°C at a sufficiently large flow rate of LBC. Thus, the maximum temperature of heat transfer tubes is not higher than 600°C that allows to use steel mastered marks. The conceptual development of an experimental molten salt reactor cooled by LBC with the thermal capacity of 60 MW has been begun together with RSC 'Kurchatov Institute' and RDB 'Gidropress'.
It should be noted that the proposed molten salt reactor cooling scheme will be also effective for cooling a subcritical molten salt blanket in the accelerator-driven system. In this case the LBC target circuit may be included into the structure of the blanket cooling circuit.
5. Conclusion
The presented experience of using a lead-bismuth coolant, the concept of the safety provision of a new RF generation, being developed on the basis of the experience accumulated, as well as the proposals of using this coolant in the future NP display this N P development direction as showing promise. RF with LBC allows to ensure an essentially higher level of safety, which excludes deterministically severe accidents. Due to internal self-protection properties the amount of protecting and localizing safety systems is sharply reduced. It decreases capital and operational expenses and increases NPP competitiveness. A new approach, proposed for NP fuel supply during a long period, is provided by the use of non-traditional FR with LBC and allows to develop NP, postponing the fuel reprocessing by 200-300 years that excludes in practice the risk of plutonium proliferation. The LBC using appears effective in acceleratordriven systems and molten salt reactors for transmutation of long-lived RAW and for other purposes. A gradual introduction of LBC reactors in NP, and in the future, if necessary, of lead cooled reactors might allow to overcome the present crisis of N P (Stadie, 1996). Certainly, the adaptation of accumulated experience of a lead bismuth coolant use to R F operation conditions at NPPs is required accompanying by a necessary performance of some appropriate research and development. Some of the proposals developed (RF is of a low power, less than 75 MW) can be realized in sufficiently short terms (5 7 years), since they are based on the engineering approaches having been already settled. Development and deployment of
B.F. Gromov et al./Nuclear Engineering and Design 173 (1997) 207-217 RFs of a greater power would require more time owing to some existing problems. The ways of solving these problems are rather clear.
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