ࡱ> #   I Stg jbjb e]    8N <  |A H ", , U0!9"\!A#A#A#A#A#A#A,zCnEBOA-"aU""OA!2,  !2!2!2" , p !A "!A!2!2 7?:!A> %|  G/@6MINISTRY OF THE RUSSIAN FEDERATION FOR ATOMIC ENERGY lead-bismuth cooled fast neutron reactors for usE in nuclear powEr PRODUCTION and as one of the trends of realizing President POUTINE's Initiative Zrodnikov,A.V., Chitaykin,V.I., Toshinsky,G.I., Grigoriev,O.G., (SSC RF-IPPE named after Acad. A.I. Leypunsky) Dragunov, U.G., Stepanov,V.S., Klimov, N.N. (EDOGidropress) Kopytov, I.I., Krushelnitsky, v.n., Grudakov, A.A. (fgup Atomenergoproekt) 2002 TABLE OF CONTENTS  TM \t "headAr;1" List of Acronyms  RENVOIPAGE _Toc17866491 \h 3 1. Introduction  RENVOIPAGE _Toc17866492 \h 4 2. Brief Description of LBC Using Experience  RENVOIPAGE _Toc17866493 \h 7 3. Basic Statements of RI SVBR-75/100 Concept  RENVOIPAGE _Toc17866494 \h 8 4. Safety Providing Concept  RENVOIPAGE _Toc17866495 \h 12 5. Concept of the Modular NPP Based on RISVBR75/100  RENVOIPAGE _Toc17866496 \h 15 6. Fuel Cycle  RENVOIPAGE _Toc17866497 \h 18 7. Long-lived RAW Management and Environmental Impact  RENVOIPAGE _Toc17866498 \h 20 8. Technological Maintenance of Non-proliferation  RENVOIPAGE _Toc17866499 \h 21 9. Possible Areas of Using RI SVBR-75/100  RENVOIPAGE _Toc17866500 \h 23 10. Conclusion  RENVOIPAGE _Toc17866501 \h 24 References  RENVOIPAGE _Toc17866502 \h 25  List of Acronyms BN sodium fast reactorCBR core breeding ratioCIS Commonwealth of Independent StatesEP emergency protection FN fast neutron FR fast reactor FSA fuel subassemblyHLMC heavy liquid metal coolantHPP heat power plant LBC lead-bismuth coolantLF loading factorLMC liquid-metal coolantLWR light water reactorMA minor actinidesMCP main circulation pumpMOX fuel mixed oxide ((PuO2+UO2) mixture) fuelNC natural circulationNFC nuclear fuel cycle NFM nuclear fissile materials NHPP nuclear heat power plantNP nuclear power NPP nuclear power plant NPT nuclear power technology NS nuclear submarine NSSS nuclear steam supplying systemNVNPP Novovoronezh NPPPGU steam-gas installationPHRS passive heat removal systemPI Russian President's InitiativeRAW radioactive waste RI reactor installationSG steam-generatorSNF spent nuclear fuelSVBR lead-bismuth cooled fast reactorTRU transuranium elementsTRUOX fuel oxide fuel based on mixture of uranium and plutonium with added neptunium and americium VVERwater cooled water moderated power reactor 1. Introduction At the Millennium Summit President of the Russian Federation VladimirV.Putin put forward an initiative (President's Initiative PI) that different countries must join forces to develop a nuclear power technology (NPT) ensuring long hundreds of years of mankind's steady development in conditions of non-limited fuel sources and elimination of materials that can be used for fabrication of the nuclear weapon (i.e. enriched uranium and pure plutonium). The essence of this idea is that it is only the nuclear power (NP) that can replace in the fuel-power balance the major part of fossil fuel burnt for both generation of electricity and heat supply of the cities and eliminate an increasing release of great amounts of greenhouse gases and other harmful pollutants into the environment. It can be expected that all the requirements resulted from the PI will be realized gradually, step-by step, when the NP is developed to achieve its mature phase with replacement of approximately 50% of fossil fuel sources for electricity production. To achieve the mature phase, it is projected that several NPT including the nuclear power plants (NPP) of a certain type and a corresponding nuclear fuel cycle (NFC) will be superseded. The NPT that best meets the requirements of a current stage of NP development in a certain country will dominate at each stage of NP development. Probably, the duration of each stage necessary to achieve the mature phase of the NP would take many decades, which is caused by significant sluggishness of development of any new NPT. The mature phase of the NP will be featured by: Domination of fast reactors (FR) operating in the entirely closed NFC; The most complete realization of the inherent safety principles; Finding the practical solution to the problem of handling long-lived radioactive waste (RAW); Maximum of technological maintenance of nuclear fissile materials (NFM) non-proliferation. As the alternative energy sources exist together with the NPPs, at each stage of NP development in conditions of a liberalized electricity market, the NPPs must be competitive with the heat power plants (HPP) using fossil fuel. The NPT will be superseded by the other one in case: The new NPT showing the better technical parameters appears; The NPP competitiveness is lost due to the higher paces of increase of the nuclear fuel costs at exhaustion of cheap nature uranium sources comparing with the paces of increase of the fossil fuel costs (that is low probable) or due to the higher costs of the NPPs and NFC caused by imposing the more stringent requirements. For example, if the regulation authorities specify the requirements for non-proliferation technological maintenance as well as high safety requirements, the NPT competitiveness would be noticeably affected. At the same time, increase of the cost of any source used by a certain NPT will stimulate both the exploration and development of this source (e.g. uranium) and launching works on changing over to the new NPT using this source more effectively (e.g. NFC closing) or using the new source (e.g. thorium). However, at the electricity market the NPP competitiveness with HPPs using fossil fuel is not sufficient enough to ensure self-financing of NP development with a pace providing increase of the NP share in the total production of electricity generated by all kinds of power sources. This is caused by the fact that at the liberalized electricity market in the developed countries the rate of electricity costs is decreasing due to excess of generating power capacities. In Russia the investment potentials of the NP have been limited by regulated cost rates, increase of an annual electricity production cost, and a value of total annual electricity production and they cannot cover the long-term investment needs consisting of the costs for: Increasing the loading factor (LF) of the NPP; Enhancing safety of the first generation units; Extending the life time of the units over the designed ones; Constructing the units of a high and medium extent of readiness; Decommissioning the NPP units with an expired lifetime; Constructing the new replacing power capacities compensating for the decommissioned units; Constructing the new NPPs providing a desired pace of NP development. Along with it, in conditions of a market economy the NP cannot rely on any noticeable support of the state financing. This fact put forward the requirement for the NPP competitiveness at the market investments (including those based on the credit repayment). Realization of this requirement needs for considerable reduction of specific capital costs of NPP constructing and construction terms making these parameters close to those of the modern steam-gas HPPs. Besides, to reduce the investor's risk, it is necessary to considerably improve the safety level in order to eliminate the severe accidents like Chernobyl one. This is also necessary for ensuring an acceptance of the NP by public opinion when its scale is considerably increased. It is very difficult to solve this problem on the basis of evolutionary improvement of the traditional NPP designs with thermal neutron reactors because an intrinsic conflict between economic and safety requirements is peculiar to such reactor installations (RI). This cause a necessity of constant increasing the unit power of the reactors that results in increasing the total investments, increasing the construction terms and reducing an investment attractiveness of the design. Besides, the existing thermal neutron reactors cannot provide long (hundreds of years) development and functioning of the NP due to the low efficiency of using natural uranium power potential even in the closed NFC that makes electricity production more expensive at low prices of natural uranium. The solution to this problem can be find by using an innovative NPT that uses a new type of the fast reactors (FR) which must not build up plutonium with a short doubling time (this task has lost its actuality). This enables chemically inert heavy liquid-metal coolants (HLMC) (instead of sodium) with high boiling point to be used for heat removal, i.e.: eutectic lead-bismuth alloy (45% of Pb, 55% of Bi) that has been mastered in conditions of operating the nuclear submarines' (NS) reactors of Russian Navy [1] and currently mastering lead coolant. Available reference information on explored bismuth resources has not allowed use of lead-bismuth coolant (LBC) in the large scale NP. However, just recently the specialized MINATOM enterprises OAO Atomredmedzoloto and VNIPI of industrial technology have carried out technical and economical investigations into an opportunity to organize large scale bismuth production in Russia and estimations of bismuth sources in the Commonwealth of Independent States (CIS). On the basis of the explored bismuth mines of the only Chita region in Russia, it is possible to produce bismuth in quantities sufficient enough to put n operation ~70GWe of NPPs with LBC cooled FRs. In addition, there are large bismuth sources in the North Caucasus. It is possible to put in operation ~300GWe by using the bismuth mines of Kazakhstan. Japanese explorers have determined the world's available bismuth sources to be ~5million tons[2]. After that Japan terminated works on lead cooled FRs. It should be highlighted that in compliance with a general geological and economical law, the quantity of the mineral raw ore increases as the square of the cost that the consumer would be ready to pay. At existing costs of bismuth in the world its contribution to the capital costs of constructing the large NPP on the basis of considered FRs is ~1%. For that reason, in practice the real technical and economical parameters of the NPP will not be noticeably worth even in case of the bismuth cost increases several times. In the future when cheap bismuth sources have been expired, it will be possible to change over to lead-bismuth alloy of a non-eutectic composition with reduced bismuth content and the higher boiling point. For example, when bismuth content in the alloy is reduced 5.5 times, the melting point is increasing from 125 to 250(C that is considerably low than a melting point of pure lead. This proposal is considering the NPT based on using small power lead-bismuth cooled FRs SVBR75/100 (Lead-Bismuth Fast Reactor of equivalent electric power 75100MWe depending on the steam parameters). In the nearest 1520years it can be implemented both in developed and developing countries with meeting the most requirements to the GenerationIV DOE reactor systems and International Project INPRO. Due to the higher technical and economical parameters of the NPP and the higher safety level[3], this technology can be considered as one of the possible ways of gradual replacement of the current NPT based on using the light water reactors (LWR). An effect of improving the economic parameters of the NPPs based on RIs SVBR75/100 isachieved due to lack of many safety systems necessary for the NPPs with LWRs, which make NPPs of this type considerably more expensive. As it can be seen further, at the minimal starting costs of industrial mastering that NPT in the process of its evolutionary improvement, all these enable to implement gradual meeting of all requirements resulting from the PI, which are peculiar to the mature phase of NP development. The fact that this technology is perspective has been verified because in recent years works on mastering this technology have been launched in many countries (the USA, Japan, Korea,etc.). 2. Brief Description of LBC Using Experience In the early 1950s, nearly at the same time the USA and the USSR launched their development programs on RIs for NSs. Both countries developed two types of RIs: pressurized water reactors and reactors cooled by liquid-metal coolants (LMC). In the USA sodium was selected as LMC because its thermo-physical characteristics were better than those of LBC. The ground-based test facility-prototype of the RI and experimental NS Seawolf were constructed. However, operating experience revealed that option for the coolant, which was fire- and explosion-dangerous in contact with air and water, did not prove itself. After several RI accidents which occurred at this NS, it was decommissioned together with the compartment and replaced by a pressurized-water RI. R&D works on mastering LBC were also carried out in the USA. However, a selected approach of finding the solution to the problem of structural materials corrosion resistance, control and coolant quality maintenance (coolant technology) did not give any positive results, and the works were stopped. In the USSR lead-bismuth eutectic alloy was selected as LMC in the very beginning. And in our country failures could not be prevented either. The history of engineering has revealed that, unfortunately, this is inevitable for the process of mastering any new technology. During the period of operating the experimental NS of Project645 in 1968, an accident with melting the part of the core occurred[4]. The cause of the accident was poor knowledge of physical and chemical properties of LBC. After this accident works on solving the technology problem of LBC were launched. For fifteen years the certain organizations had been carrying out these works under IPPE scientific supervision. As a result, the problem was solved successfully, and it was verified by further many-year experience of RIs operating at the NSs. When operating the second generation RIs, there were no problems caused by structural materials corrosion in the primary circuit and violating the circuit purity standards. The problem of ensuring radiation safety that was caused by forming polonium210 was solved in the same way. During the whole period of operating LBC cooled RIs, including the primary circuit equipment's repair period and removal of spilled LBC, there were no cases of personnel's extra-irradiation over the permissible limits in terms of this radionuclide. Altogether eight NSs with LBC cooled RIs were constructed. The first experimental NS of Project645 had two reactors. Each of the other seven NSs of Project705 (in terms of NATO Alpha) had one reactor. Due to its speed parameters this NS was entered into Guinness Book of Records. Besides, two full-scale ground reactor facilities-prototypes were constructed and operated in IPPE (Obninsk) and NITI (Sosnovy Bor). A total sum of operating time of the considered type RIs has been ~80reactor-years[1]. The new nuclear power technology that has no analogs in the world has been demonstrated in industry. Currently the conditions for introducing this technology into the civilian nuclear power have been formed. 3. Basic Statements of RI SVBR-75/100 Concept RISVBR75/100 was designed in compliance with a conservative approach. This approach allows: without exceeding the limits of the experimentally tested mode parameters of the primary and secondary circuits, to use to the maximal extent the already mastered fuel and structural materials and verified in practice the principal technical solutions to the equipment components and RI scheme. This approach ensures a high extent of succession of the RISVBR75/100 technical solutions, first of all, the technical solutions of LBC cooled NSs' RIs that has been favored by nearness of their scale factors. Adhering to this approach reduces the execution terms, R&D scopes and costs, investment risk, ensures reliability of the RI and its operation safety. These factors make it possible to avoid the mistakes of an initial stage of mastering the innovative NPT. Use of the conservative approach does not mean that the new technical solutions should not be used and an evolutionary way of NP development should be only followed. This would cause stagnation and hindrance of the scientific and technical progress. However, use of verified in practice technical solutions ensures the applicable technical and economical parameters of the NPP[3]. For that reason, the new, perspective technical solutions that considerably improve the parameters of the RI will be used when changing over to the next generation of the given type RIs after carrying out the necessary R&D. Expediency of this approach has resulted from the analysis of the technique development history. It shows that for successful introduction of new technologies, the share of new technical solutions in complicated systems should not be too high. Ignorance of this fact will result into considerable delay of start launching, unnecessary over-expenditure of materials and financing. For that reason, when RISVBR75/100 was being developed, priority was given to the already developed technical solutions even if they did not ensured achievement of the highest technical and economical parameters. With due account of all mentioned, the following basic approaches and technical solutions have been realized in the RISVBR75/100 design: A monoblock (integral) design of a pool type is used for the primary circuit equipment. Valves and LBC pipelines are completely eliminated; A two-circuit scheme of heat removal is used; The levels of coolants' natural circulation (NC) in the heat-removal circuits are sufficient enough to ensure reactor's heat decay removal without dangerous over-heating of the core; A reactor monoblock with a safeguard vessel is installed and fixed in the tank of the passive heat removal system (PHRS). The tank is filled with water and also performs the neutron protection function; A wrappless design of the fuel sub-assemblies (FSA) is used. This ensures high cross heat-mass-exchange in the core and eliminates unallowable over-heating of fuel elements at large blockages of flow rate at the core inlet; A steam-generator (SG) operating in compliance with a multiple NC scheme and producing saturated steam is used. This ensures the best lifetime and operating parameters, e.g. reliable RI operation at any power levels, simplicity of maintaining LBC in a liquid state at low power levels (including the mode of heat decay removal via theSG); A slow-rotating gas-tight uncontrolled electric engine of the main circulation pump (MCP), which power does not exceed 500kW, is used. This eliminates the necessity to seal the rotating shafts, enables to use the ball-bearings with greasing and provides the necessary against-cavitation condition at the suction of the MCP impeller due to coolant column's hydrostatic pressure; The RI equipment can be repaired or replaced; On ending the lifetime, refueling can be performed at once, FSA-by-FSA; It is possible to use different kinds of fuel (UO2, MOX fuel with weapon or reactor Pu, TRUOX fuel, nitride fuel) without changing the reactor design and at meeting the safety requirements. With due account of the relatively high cost of LBC, there have been developed the measures reducing the specific mass of LBC in the RI. The summarized analysis of experience of developing RIs of different power capacities[5] has revealed the LBC specific mass decreases at reducing the RI nominal power. Along with this, reducing the LBC specific mass is limited. It is caused by the fact that atsmall dimensions of the core, it is impossible to provide core breeding ratio (CBR)e"1. Computations have revealed that an optimal diameter of the core should be not less than 1600(1700mm at height 900mm. These core dimensions make it possible to achieve equivalent electric power of the reactor ~100MWe. In this case, CBRINCORPORER Equation.31 is provided not only for the mixed nitride fuel but also for the less dense but well mastered MOX fuel. This point can be carried out if the volumetric fuel fraction is not lower than 55(60%. Reduction of the LBC specific mass in small-sized FRs, for which the core power density is several times less than that of sodium cooled FRs, is also achieved by elimination of the inreactor storage of spent nuclear fuel (SNF) and in-reactor refueling mechanisms (rotating plugs, etc.). In this case, refueling is performed once during the core lifetime. For that purpose, a special refueling equipment is used, it is also used for refueling all reactors of the power unit. Therefueling technology is similar to that of LBC cooled NSs' RIs. Another way of reducing the LBC specific mass is increasing its average velocity in the RI and diminishing the length of the LBC circulation circuit. However, this way has its own constraints caused by the necessity to provide the safety requirements. The first requirement is caused by the necessity to provide the power level of the reactor with naturally circulating LBC to be not less than 5% of Nnom. This makes it possible to eliminate dangerous temperature increase in case of shutting down the MCPs. The second requirement is caused by the necessity to provide an effective separation of steam bubbles from LBC with steam surfacing to the LBC free levels in case of an accident with leaking SG tubes. This is necessary for elimination of steam ingress into the core and impermissible pressure increase in the monoblock vessel. The necessity to satisfy the highlighted requirements resulted into development of the LBC circulation scheme in which core hydraulic resistance equals to 90% of the total hydraulic resistance of the primary circuit and hydraulic resistance of the SGs, in which LBC flow rate is much less, only equals to 10%. With due account of the highlighted requirements, the specific mass of bismuth in RISVBR75/100 is ~1100t/GWe. It should be highlighted that the low values (2530%) of the LBC volumetric fraction in the core (tight lattice of fuel elements) and LBC specific mass do not deteriorate the safety parameters of RIsSVBR75/100 in cases of shutting down the MCP and leaking SG tubes (ascomputations have revealed) but in the case of unauthorized insertion of positive reactivity as well. The latter is caused by a sufficiently high negative feedback being typical of small power reactors and a low time of delaying its temperature component at the LBC core inlet (extending of the lower core-plate) coupled with sufficient heat-accumulation ability of the monoblock. The following have been provided at the selected power level (100MWe): The lifetime duration is ~53000eff. hours if mastered oxide uranium fuel is used (CBR=0.87); CBR(1 if MOX fuel is used, the reactor operates in the closed fuel cycle in the mode offuel selfproviding; CBR(1 if mixed nitride fuel is used, the reactor operates in the mode of fuel selfproviding at a burn-up reactivity margin being less than (eff or in the mode of extended breeding with CBR=1.13 at a plutonium doubling time being ~45years; A burn-up reactivity margin is less than (eff, the lifetime duration is ~80000eff. hours incase of using uranium nitride fuel; Reactor's heat decay removal is entirely passive, heat is removed through the monoblock vessel to the PHRS tank; Complete plant fabrication of the reactor monoblock, RIs are produced in large quantities that improve the quality of works and reduce the cost; The reactor monoblock can be transported by railway, truck or marine transport (with fuel in a nuclear and radiation-safe state due to LBC freezing in the monoblock vessel thatalso meets non-proliferation requirements); The term of constructing the NPP unit can be considerably reduced as modules are delivered in high plant readiness and the assembling scopes are sharply reduced. (Thisimproves the terms of receiving the NPP construction credits and reduces the period of capital investments recoupment); The NPP unit in which the RIs have been replaced by the new ones can be renovated in 5060years. This postpones the necessity to construct the replacing power capacities to 50years; The cost of decommissioning the unit can be considerably reduced as after removing the monoblock, no radioactive materials remain in the main reactor building; The NPP units with LWRs which RIs have exhausted their reactor lifetime can be renovated by installing the necessary number of RISVBR75/100 in the empty SG and MCP rooms. The basic parameters of RISVBR75/100, a longitudinal section of the reactor monoblock and reactor compartment are cited in Table1, Fig.1, Fig.2. TABLE 1. BASIC PARAMETERS OF RISVBR75/100 Name and Dimensions of the ParameterValue1 Heat power (nominal), MW2802 Steam-production, t/h5803 Pressure of generated saturated steam, MPa9.54 Feed water temperature, CARSPECIAUX 176 \f "Symbol"C240.95 Primary circuit coolant's flow rate, kg/s117606 Primary circuit coolant's temperature, outlet/inlet, CARSPECIAUX 176 \f "Symbol"C482/3207 Core dimensions: diameter CARSPECIAUX 180 \f "Symbol" height, m1.645 CARSPECIAUX 180 \f "Symbol" 0.98 The number of fuel elements121149 The number of CPS rods3710 Average power density of the core, kW/dm314611 Average linear load of the fuel element, kW/m~24.312 The time interval between refuelings, years~813 Uranium fuel (UO2) load: mass, kg/enrichment, %9144/16.114 The number of SG modules2 ( 615 The number of MCPs216 Power and head of the MCP, kW/MPa450/0.5517 The core lifetime, eff. hours5300018 LBC volume in the primary circuit, m31819 Dimensions of the reactor monoblock: diameter CARSPECIAUX 180 \f "Symbol" height, m4.53CARSPECIAUX 180 \f "Symbol"7.554. Safety Providing Concept Lead and bismuth natural properties, physical features of FRs coupled with an integral (monoblock) design of the primary circuit equipment make it possible to eliminate deterministically an opportunity of the certain severe accidents. High boiling point of coolant enhances reliability of heat removal from the core and safety due to lack of the heat removal crisis phenomenon and being coupled with a safe-guard vessel eliminates the accidents of the LOCA type. Low pressure in the primary circuit enables to reduce the thickness of the monoblock vessel walls and reduce the limitations imposed on the temperature change rate in compliance with the thermo-cycling strength conditions. LBC reacts with water and air very slightly. Development of the accident processes caused by primary circuit's tightness failure and SG intercircuit leaks occurs without hydrogen release and any exothermic reactions. There are no materials within the core and RI that release hydrogen as a result of thermal and radiation effects and chemical reactions with coolant. Therefore, the likelihood of chemical explosions and fires as internal events is virtually eliminated. In the case of failure of all active cool-down systems and total blacking out the unit, elimination of core melting caused by residual heat release and keeping the monoblock vessel intact are ensured by an entirely passive way due to heat accumulation in the in-vessel structures and coolant and heat removal to the PHRS water tank through the monoblock vessel with further water evaporation. The grace period necessary for achieving the safety operation limit is about five days' time. A scheme of heat removal to the PHRS tank is shown in Fig.3, Fig.4 shows how the maximal temperature of the fuel element's cladding and the water level in the PHRS tank depend on time. Core melting is also eliminated at postulated LBC freezing in the SG. In this case, NC of LBC with a flow rate being ~1% of the nominal one is performed over the continuously operated by-pass circuit past the SG from the central buffer chamber to the peripheral one via the holes in the shells, which have been provided for this purpose. In case of unauthorized insertion of positive reactivity at postulated failure of all emergency protection (EP) drivers, elimination of prompt neutron reactor runaway is ensured by a special algorithm of compensating rods control, which is the part of the automatic control system. In this case, when the reactor operates at nominal power, during a certain time (~4months) a reactivity margin controlled by an operator is much less than (eff. Besides, an efficiency of each rod is much less than (eff, a rate of moving the absorbing rods extracted gradually is technically limited. For that reason, the inserted positive reactivity has time for being compensated by negative feedbacks without dangerous increase of the core temperature. In the case of EP system failure caused by the events not specified in the regulatory documents (for example, damage of all servo-drivers), there are fusible locks connecting arod with a driver bar. When the coolant temperature exceeds 700(C, EP rods that are installed in the dry channels are separated from the bars and drop into the core due to their gravity. For considered fuel loads, the total void reactivity effect of the reactor is negative and thelocal positive void reactivity effect is less than (eff. and cannot be realized due to the coolant's very high boiling point and lack of the opportunity for gas or steam bubbles to arise in great quantities. Elimination of water or steam penetration into the core caused by a large SG leak and consequent overpressurization of the monoblock vessel designed to be resistant against the maximum possible pressure under this condition are ensured by the coolant's circulation scheme. This scheme provides that steam bubbles are thrown out on the free coolant level bythe moving up LBC flow. Then steam goes to the gas system condensers. In the event of their postulated failure, steam goes to the bubbler (PHRS tank) through the bursting membranes (see Fig.5). Carried out studies have revealed that no equipment failures, personnel's errors or their combinations may cause core melting. Negative reactivity feedbacks ensure power decrease down to the value that does not cause core damage even in case of failure of all reactor shutdown systems and total blacking out. The additional barriers of the safety providing system are the separate concrete cells of the RI (confinements) that restrict radioactivity release into the central reactor hall, and the protection shell of the central hall covering all RIs (containment) purposed to protect the reactor against the external impacts. In the event of accident tightness failure of the primary circuit, high pressure radioactive exhausts (which can happen in LWRs) do not occur in LBC cooled RIs. For that reason, there is no need to design the containment of the unit and RI compartment to be resistant against high excess internal pressure. There is also no need to design the double containment with a water cooling system and corium catch. An extremely simple design of RISVBR75/100, lack of plant safety systems caused by developed inherent safety properties of the RI, that made it possible to couple the functions of RI safety systems with those of normal operation systems, sharply reduce theprobability of personnel's errors. Theconsequences of any personnel's errors and their combinations do not affect the safety but only result in economical losses and the necessity to carry out an unscheduled repair. RI safety does not depend on the equipment and systems of the turbine-installation. Major safety viable systems operate passively being independent on either right or wrong personnel's actions. It should be highlighted that there are no valves or mechanical devices in the RI safety systems, which may cause failure of their operation in case of failure or switching off the valves or mechanical devices that may be caused by someone's error or malicious actions or over-standard external impacts. For that reason, RI safety systems' action has been assured by: Melting the locks of the EP rods and their free fall down to the core; LBC natural circulation, heat transfer via the main and safe-guard vessels, air convection in the gap and heat irradiation, water boiling in the PHRS tank in the modes of emergency heat decay removal; Rupture of the safety membrane that protects the monoblock from excess pressure at large SG leaks and failure of the gas system's steam condenser. As computations have revealed, an extremely high safety potential typical of the considered RI is characterized by the following: even in an event of the postulated combination of such initial events as containment destruction, damage of the RI compartment overlapping and primary circuit gas system's serious failure with direct contact of the LBC surface with atmospheric air in the monoblock vessel that is possible in the case of terror attacks, neither reactor runaway, nor explosion, nor fire occurs, and the radioactivity release is less than that requiring the population evacuation. Obtained results enable to conclude that the safety level of SVBR75/100 reactors is higher than that of LWRs and sodium cooled FRs. It can be practically demonstrated at the stage of experimental operation of RISVBR75/100 with controlled simulation of different designed initial events and their combinations. 5. Concept of the Modular NPP Based on RISVBR75/100 It is highlighted in [6] that Modular plant fabrication of nuclear power systems and their assembling on the site will replace the existing expensive construction methods. Economical advantages of modular principle of constructing the NPP are also highlighted in[7]: Measures on reducing the construction terms much affect the total capital costs especially at high record rates because in the course of construction, the credit payment may reach 25percent and more of the total scope of investments. Modular production that makes it possible to fabricate and assemble the units at the plant but not on the site reduces the construction term and, consequently, expenses on the credit payment during the construction period. Reduction of the investment cycle of constructing the NPP, that has been ensured by a modular structure of the NPP and delivery of ready fabricated modules, is extremely viable for the technical and economical parameters of the NPP to approach those of steam-gas HPPs with short investment cycles[8]. For developed countries, which power systems have high-voltage electric transfer lines with high transmission, it will be economically effective to use large modular power units. Maximal possible capacity of the modular type power-unit will not be restricted by maximal possible reactor capacity but by maximal possible (or optimal in technical and economical parameters) turbine-generator capacity. At the existing technical level of turbine-constructing factories in Russia, power-unit's capacity can be taken as 16001800MWe. SSCRFIPPE, EDO Gidropress, FGUP Atomenergoproekt have developed a conceptual design of the two-unit NPP, which power unit includes the nuclear steam-supply system (NSSS) consisting of 16 RIsSVBR75/100 (reactor modules) and one turbine-installation of 1600MWe[3]. This allows to compare correctly the technical and economical parameters of that NPP to those of the NPP based on RIVVER1500. When NPP unit's capacity was selected, it was taken into account that specific capital costs of the reactor compartment (nuclear island) would decrease at increasing the unit's capacity. Itis caused by the fact that at increasing the number of modules in the NSSS, the cost of the equipment and providing systems installed beyond the RI compartments increases slightly. For that reason, their contribution to the specific capital costs of the reactor compartment will decrease. Such systems and equipment include the refueling equipment, coolant's in-taking equipment, equipment for coolant's transferring to the monoblocks at initial filling, etc. So, the specific capital cost of constructions necessary for installing these systems will be reduced correspondingly. A modular principle of the NPP design is the most economically effective for reactors, inwhich the inherent safety properties against severe accidents have been realized to the maximal possible extent. First of all, this means the accidents with coolant's loss such as LOCA. To overcome these accidents, the LWRs need a lot of safety systems that are not necessary for RIsSVBR75/100. This considerably reduces the construction scopes of the reactor compartment. Control of the modular NSSS is carried out by an operator who uses the common power master unit. If there is any fault in the certain RI, it is automatically removed out of operation and can be cooled down autonomously with the turbine-installation systems. A simple scheme of the RIs and similarity of their types allow to reduce the number of the operation and maintenance personnel (a staff coefficient) at the modular NPP unit as compared with that at the NPP unit with one large-power RI with lots of safety systems, localizing accidents' systems, controlling and providing systems. For example, the safety systems of the AP1000 reactor have 184 pumps, 1400 driver valves, 40km of the pipelines and cables[9]. A modular design of the NSSS power unit makes it possible to provide LF to be not less than 90% under long reactor operation without refueling. When each RI is shut down for refueling, power unit's power reduces slightly. Once-moment sequential refuelings of each RI included into the NSSS are equivalent to the mode of partial refuelings of the large-power reactor (1600MWe) at annual refuelings of ~1/8 share of fuel each year). Duration and periodicity of scheduled maintenance and repair works are determined by requirements to the turbine-installation equipment. Licensing of constructing the modular type large power power-unit will be much simplified in the case of constructing one RI or the small power modular power-unit which RI has been certified. Small power of the RI determines a comparatively low cost of its construction. A plan and a longitudinal section of the SVBR1600 reactor compartment's main building with the NSSS are shown in Fig.6. The basic technical and economical parameters of the two-unit NPP based on RISVBR75/100 in comparison with those of the two-unit NPPs with RIVVER1500, RIVVER1000 (V392), RIBN1800 and HPP with ten steam-gas units PGU325 are summarized in Table2[3]. TABLE 2. COMPARABLE PARAMETERS OF DIFFERENT POWER PLANTS Name and Dimensions of the ParameterNPP with RI SVBR-75/100NPP with RI VVER-1500NPP with RI BN- 1800 [10]NPP with RI VVER-1000HPP with PGU- 3251. Set up power of the power-unit, MWe 1625 1479 1780 1068 3252. The number of the units at the plant  2 2 2 2 103. Electric power necessary for plant's own needs, %  4.5 5.7 4.6 6.43 4.54. Efficiency of the net plant (power unit), % 34.6 33.3 46.2 33.3 44.45. Specific capital investments in the industrial construction of the plant, $/kW  661.5 749.8 783.4 819.3 6006. Design cost of produced electricity, cent/kW-h  1.46 1.85 1.56 2.02 1.75The results of technical and economical computations have revealed that in compliance with the data obtained at the conceptual design stage, the technical and economical parameters of the NPP with two 1600MWe units, each based on the SVBR75/100 type RI, are better than those of the NPP based on the large power LWRs and than those of the HPP with ten units PGU325 operating by using natural gas. The term of constructing this NPP can be ~3.5 years. However, it should be highlighted that reliability of the sited economical parameters of LWRs, which have been developing during several generations, is higher than those of the NPPs with RIsSVBR75/100 because they have not had experience of practical realization. For that reason, the costs of the capital investments in the industrial construction of the NPP based on RIsSVBR75/100 has an additional margin of 17% over the standard one (60% of the RI equipment cost). Besides, this NPP project that is actually the first generation design based on the conservative approach has a great potential for its development. In the future, after carrying out corresponding R&D, the following technical solutions would improve the RI design and considerably improve the technical and economical parameters: Use of the one-through SG producing super-heated steam that makes it possible to increase an efficiency of the thermo-dynamic cycle and increase electric power by 1015%; Increasing the LBC temperature at the reactor outlet at increasing the maximal temperature of the fuel elements' cladding from 600 to 650(C that provides increasing the reactor heat power by 1520% without changing its design. 6. Fuel Cycle Due to the low current costs of uranium and its enrichment, use of oxide uranium fuel with postponed reprocessing and SNF storing on the NPP site is economically justified for RISVBR75/100. Duration of this stage depends on the available resources of cheap uranium and NP scales. In compliance with the [11] data, in Russia an estimated term of expiring the uranium resources will be 70 years at an average level of NPPs' total power being 45GWe and in the world this value will be 40 years at an average level of NPPs' total power being 750GWe. This implies that the increase of natural gas costs will overtake the increase of uranium costs. This will ensure competitiveness of the NP even at higher costs of uranium due to a lower share of the fuel component in the electricity cost for the NPPs as compared with HPPs. At this stage the major way of improving the economic parameters of the fuel cycle will be increasing the lifetime duration (fuel burn-up depth) as experience in the core elements operation ability is gained. Further, an economically expedient will be the stage at which the own SNF will be reprocessed, NFC will be uranium closed (at adding enriched uranium into the NFC), plutonium, MA, fission fractions will be extracted and then stored. Duration of the uranium stage can be increased when changing over to uranium nitride. At the same time, it could be expedient to only use uranium nitride for export in order to reduce the risk of unauthorized fissile materials proliferation at expanding the refueling interval to 15years. Actually, in the future it will be necessary to change over to the entirely closed NFC. The time period required for this change will be determined by appearing the developed in an industrial scale technology of SNF reprocessing that will be acceptable from the standpoint of RAW minimization and fissile materials non-proliferation. The existing technology of radiochemical reprocessing SNF do not meet these requirements. Besides, it will be only economically justified under stable operation of a large radiochemical factory (i.e. at reprocessing scales being 10001500t of SNF per year that corresponds to the total level of NPP set up power being ~6090GWe). One of the economically expedient variants of changing over to the entirely closed NFC that meets the necessary requirements is a technology based on using the dry methods of reprocessing SNF and a vibro-pack technology when the fuel elements are fabricated. SSCRFNIIAR have carried out the researches revealing that construction of power capacities on reprocessing SNF of SVBR75/100 reactors and fresh FSA fabricating increases the specific capital costs of constructing the NPP by not more than 1015% (about $76/kW of set up power). And it has been presumed that reprocessing is performed on the basis of the pyro-chemical processes in the chloride melts and the reprocessing rate is 120t of heavy metal per year (the highlighted reprocessing rate corresponds to the total NPP power on the basis of RIsSVBR75/100 being ~12GWe). Change over to the closed NFC for the SVBR75/100 reactors will have the lower cost if for fabricating the first fuel load from MOX fuel we use plutonium that has not been extracted from LWRs' SNF but has been extracted from the own SNF of uranium loads due to considerably lower scopes of reprocessing in terms of 1t of plutonium. The quantity of plutonium extracted from SNF of three uranium cores is enough for fabricating one core from MOX fuel. When reactors SVBR75/100 operate in the closed NFC, economically effective use of LWRs' SNF as make up fuel without separation of uranium, plutonium, MA, and fission fractions instead of waste pile uranium is possible (similarly to the DUPIC-technology for the CANDU-reactors). That is, instead of reprocessing SNF of thermal reactors (both from the VVER and RBMK reactors) for the purpose to only extract 1% of plutonium, after long storing during ~50years this kind of SNF will be step by step utilized in the FR. Due to the fact that the fraction of LWRs' SNF in fresh fuel of SVBR75/100 operating in the closed NFC is ~1012% and the plutonium fraction in LWRs' SNF does not exceed 1%, influence of the plutonium isotopic vector in LWRs' SNF on the isotopic vector of fresh fuel is negligible for SVBR75/100. Therefore, RISVBR75/100 makes it possible to develop a principally new strategy of the closed NFC that does not require expensive reprocessing SNF of thermal reactors for the purpose to extract plutonium for FRs' fuel supplying. Flexibility of RISVBR75/100 relative to the fuel cycle technologies that is realized in compliance with a principle: To operate using the type of fuel that is the most effective makes it possible to postpone a task of constructing a specialized enterprise to several decades after the first unit of the NPP with such reactors has been put in operation. For example, after introduction of about 10GWe of power capacities on the basis of RIsSVBR75/100 and repaying the costs of NPP construction, a certain share of profit could be spent on developing the industry on SNF reprocessing and fabricating FSA from MOX fuel. After launching that factory, the cost of the core would be only determined by current operating costs of SNF reprocessing and FSA fabricating. If the SSC RFNIIAR designs are used as a basis of that complex, contribution of fuel costs to the cost of the SVBR75/100 core would be even less than that of the basic variant using oxide uranium fuel. This will make it possible to considerably improve the NPP competitiveness. This approach to construction of power capacities on SNF reprocessing and FSA fabricating presumes the owner of the NPP units is also the owner of the fuel cycle enterprise. Along with this, on considering the prospects of economically substantiated closing the NFC of FRs caused by lack of uranium, some experts presume it might be required in 50100 years[12]. In this case it has been accounted that the cost of natural uranium is a very small fraction in the cost of electricity production. Even the opportunity of extracting uranium from the sea water (that seems exotic today) has been assessed and that cost will be 500$ per kg. 7. Long-lived RAW Management and Environmental Impact In the course of NPP operation, liquid RAW is produced in very low quantities. This fact has been verified by experience of operating LBC cooled NSs' RIs. The NPP design provides an installation for concentrating and solidifying the low quantities of liquid RAW. After expiring the RI lifetime, the radioactive LBC can be many times recycled in the new RIs. In 1000years of irradiation, slight residual long-lived radioactivity of LBC caused by Bi208 and Bi210m radionuclides will be lower than natural radioactivity of the uranium ore (in terms of U3O8). It will be only important at the final stage of NP functioning. In this connection, LBC in the form of solid radioactive waste being disposed in the deep geological formations will not disturb the natural radioactivity equilibrium. Low chemical activity of lead and bismuth rules out radioactivity release into the biosphere. Therefore, the radio-ecological consequences of this disposal will be of no risk for the population of the next generations. There is a similar problem for the LWRs as long-lived radionuclide zirconium93 is forming in the fuel elements' zirconium claddings and channels. The quantity of tritium release into the environment due to unavoidable water losses in the RI secondary circuit does not exceed ~50TBq/GWe*yr that is within the limits of normalized release of tritium with liquid wastes into the environment of the world's operating NPPs. Radioactivity release into the environment out of unloaded SNF is eliminated by a multi-barrier shielding against activity release out of the spent FSA. Being unloaded out of the reactor they are installed into the steel capsules filled with liquid lead. After lead solidifying four barriers are formed in the capsules: the fuel matrix, fuel element cladding, solidified lead, capsule vessel. When operating in the closed NFC, fission products management does not presume their transmutation because of the low efficiency of the process. Taking into account that the half life of majority of fission products does not exceed 30years (except for technetium99, iodine129, cesium135, and some others), it is supposed that after vitrifying they are placed into the dry control storage for about 300years of long storing. After that storing, their activity will be determined by long-lived nuclides of technetium, iodine and cesium. It is proposed to dispose these vitrified fission products in the deep geological formations with providing a multi-barrier shielding. (Instead of vitrifying a sinrock-technology may be used after verifying its advantages.) That method of fission products management rules out radioactivity release into the environment. Management of transuranium (TRU) elements presumes that their release beyond the fuel cycle will be eliminated (except for very low losses at the stage of RAW chemical reprocessing) as they are well fissionable in a hard neutron spectrum of FRs and their concentration achieves a saturation condition very quickly. To estimate the environmental impact caused by the NFC of SVBR75/100, a value of specific radiotoxicity of transuranium elements (neptunium, plutonium, americium and curium) and long-lived fission products (technetium99, iodine129 and cesium135) as a function of produced electricity was taken as a criterion. If this value decreases with energy production, the NFC environmental effect can be considered as a friendly one. The radiotoxicity standard was adopted as a volume of water necessary for diluting a built-up quantity of radionuclides to decrease its concentration down to the level meeting sanitary requirements to the drinking water in terms of specific radioactivity. Specific radiotoxicity is determined as SNF radiotoxicity divided by produced energy. The following assumptions were made to evaluate radiotoxicity: MOX fuel with plutonium extracted from LWRs' SNF was used as a first load of the reactor; At the end of each lifetime and three-year cooling, SNF was reprocessed; Radiotoxicity of the main bulk of fission products with half-lives being less than 30years was not accounted as after 300years of cooling their radiotoxicity would be very low; Curium was extracted and transported to the temporary storage (repository) for 100150-year cooling. After cooling, all radioactive isotopes of curium (except for curium245) were transformed into plutonium isotopes. Then this isotopic mixture was transported back to the reactor for its further incineration[13]; Mixture of plutonium, neptunium and americium with the rest uranium and necessary addition of depleted (waste pile) uranium was used for fabricating the fuel load for the next lifetime. Figure7 presents SNF long-lived specific radiotoxicity as a function of produced energy for reactor SVBR75/100 within the NFC. Specific radiotoxicity of technetium99, iodine129 and cesium135 in the final disposal is 0.014km3 /GWe*year that is nearly equal to that of natural uranium annually added to the fuel cycle in terms of GWe*year. The analysis of the obtained results shows the environmental-friendly effect of the NFC of SVBR75/100 as specific radiotoxicity of long-lived RAW reached the final disposal decreases at increasing the value of cumulative produced energy to the value of specific radiotoxicity of the extracted uranium ore. This is caused by the fact that the hard neutron spectrum in the reactor facilitates efficient incineration of both own MA, and MA built up in the LWRs. 8. Technological Maintenance of Non-proliferation Non-proliferation of fissile materials means creating the conditions when inappropriate use of fissile materials is least attractive for potential distributors of the nuclear weapon. It is evident that the problem of non-proliferation cannot be only solved by technological measures as despite the development of a new nuclear technology, there are the opportunities for illegal receiving the weapon materials and using the well-developed technologies of isotopic uranium separation and plutonium extraction out of spent fuel. For that reason, the complete solution to the problem of non-proliferation can be only achieved by coupling the technological and political measures. Relationship of these measures will be different for nuclear and non-nuclear countries. During the recent decades all nuclear countries, which legally possessed the nuclear weapon, have solved this problem successfully using the measures of physical protection, accounting, control and safeguard. For that reason, the additional measures of technological maintenance of non-proliferation will be justified in case they do not reduce the NP competitiveness. When using the NPP in developing countries, the additional measures of technological maintenance of NFM non-proliferation should be taken along with the political measures and international control. In case of the export deliveries, the reactor design should provide a fuel load for the whole lifetime and no access to fuel. A reactor seller should keep the property rights to the reactor and core and provide necessary maintenance caused by periodical replacement of the core/reactor. A reactor module should be designed as a wholly replaceable unit after expiring the core lifetime (the goal is 15years). This approach requires neither the refueling equipment, nor the spent fuel repositories located on-site. It is expedient to concentrate SNF reprocessing at the certain closed factories in nuclear or developed countries. In the far future, when FRs are used worldwide, it would be possible to give up gradually the technologies of uranium enrichment and pure plutonium extraction for NP needs. To reduce the risk of NFM unauthorized proliferation, the following measures realized in the RISVBR75/100 design and at the different NFC stages are considered below: The RI design eliminates uranium and thorium blankets that make it possible to accumulate fissile materials for the weapon purposes. Refueling is performed very seldom (once a 710year period) and can be inspected easily. Partial refueling is impossible. Refueling is only performed by using the special equipment kit. At the stage of fabricating the initial fuel load by using oxide uranium fuel, NFM non-proliferation is ensured by the core design due to use of uranium with U235 enrichment being less than 20%. This satisfies the IAEA requirements and allows to use these reactors in non-nuclear countries. At the stage of SNF storing, proliferation resistance is ensured by the fact that built-up plutonium together with high radiotoxic fission products is in the SNF (the spent fuel standard). Therefore, the possibility to steal SNF is eliminated and the SNF movements can be easily inspected by gamma-radiation. At the stage of SNF reprocessing, proliferation resistance is ensured by the fact that during technological reprocessing, built-up plutonium along with built-up MA is separated from uranium at non-deep purification from fission products. Therefore, plutonium stealing is impeded, and its applicability for fabricating the explosion devices becomes insignificant. At the stage of fabricating and transporting the MOX fuel, proliferation resistance is ensured by the fact that during fabricating re-fabricated fuel, 2% of fission products built-up in SNF and all MA remain in it. This requires remote management of that fuel, impedes its stealing and facilitates the inspection of its movements. This fuel can be delivered to any countries as fuel management is only possible by using the special large heavy equipment that facilitates the accounting and inspection of fresh fuel. Fuel transportation in the reactor monoblock with solidified LBC creates an additional technical barrier to the fuel thefts. The IAEA inspection is ensured at all stages of the NFC. The measures of physical protection and safeguard are used. 9. Possible Areas of Using RI SVBR-75/100 High technical and economical parameters of RISVBR75/100, ability of the reactor monoblocks to be transported by railway, inherent safety properties of the RIs make conditions for their multi-purpose usage when they have been produced in large quantities. First of all, it is renovation of the NPP units with LWRs which RIs have expired their lifetime. They can be renovated by installing the necessary number of RIsSVBR75/100 in the empty SG and MCP compartments. Results of technical and economical researches into technical opportunity and economical expediency of renovating the 2nd, 3rd, and 4th Novovoronezh NPP (NVNPP) units on the basis of RIsSVBR75 have revealed that renovation reduces two times the specific capital costs as compared with construction of the new replacing power capacities[14]. Installation of RIsSVBR75 in the SG/MCP cells of the 2nd NVNPP unit is shown in Fig.8. A similar renovation technology can be used for almost all LWRs units. In this case, the capital costs saving will be $400M per GWe (in Russian conditions) as compared with construction of the new replacing power capacities. Experience gained by operating RIsSVBR75/100 in conditions of the NVNPP will make it possible with a minimal investment risk to launch sequential renovation of the LWR units, which RIs have expired their lifetime and construction of modular NPPs with large power units in the countries, which power systems have high-voltage electric transfer lines with high transmission. Taking into account the high extent of inherent safety of these RIs, it is expedient to use them for the small power needs. In Russia: these are the regional NHPPs of 200600MWe, which are necessary to be located near the cities. The term of constructing the regional NHPPs and their total cost will be much less than those of large power NPPs. Their construction can be realized at the expense of the finance sources of the Russian Federation subjects, including the joint stock but this require for the legislation to be changed. Abroad: these are the power-complexes designed for producing electricity, heat and water desalination. Carried out by IAEA marketing studies have revealed that in developing countries that have no powerful electricity-transfer lines, there is a vast market for small power reactors of ~100MWe. Export potentials can be realized by granting on lease the transportable reactor unit for steam-supplying the power-complexes. In this case, the Supplier keeps the property rights, the Consumer needn't develop and maintain the complex infrastructure of fuel management, the Supplier takes all the possible risks. In this case, the requirements to fissile materials non-proliferation are ensured by using uranium enriched in low than 20%, lack of refuelings in the User-country with transportation of the reactor unit for refueling to Russia (once a 10-year period) in a nuclear and radiation-safe state due to freezing LBC and the core in the monoblock vessel. 10. Conclusion The inherent safety properties of RISVBR75/100, economic competitiveness of the NPPs with their usage, opportunity to operate in the closed NFC in the fuel self-providing mode or with low breeding, opportunity both to burn own MA in the reactor and use LWR's SNF as make up fuel, providing technological maintenance of non-proliferation make it possible to consider the proposed reactor technology as one of the most suitable ways for realization of the Russian Federation President's Initiative. RISVBR75/100 meeting most of the requirements to the GenerationIV DOE reactor systems and IAEA Project INPRO can be proposed as a basic installation of the collaborative International Project. The obtained results have revealed the technical opportunity and economical expediency of using the RIs of the SVBR75/100 type for finding the solution to the certain basic tasks of Russian NP both in the nearest and far future at minimal launching costs of industrial mastering. First of all, this refers to the opportunity of economical effective considerable expanding the NPP units' lifetime by renovating them. Use of the modular structure of the power-unit's NSSS makes credible an opportunity to change over in the future to the innovative technologies of the standard design of the various capacity power units on the basis of the standard modules produced in quantities and a conveyer system of carrying out assembling works that allows to considerably reduce the terms of NPP construction and technical maintenance of the reactor modules on the service basis. Nuclear power based on the considered type NPPs can compete with heat power based on the modern steam-gas HPPs not only at the liberalized power market but at the investment market that will ensure the necessary pace of its development. To realize the highlighted above potentials, it is expedient and substantiated to construct the first RISVBR75/100 as a part of the NPP unit by the year 2010. In Russian conditions the least cost of it will be if the proposed RI has been installed in the building of the shut down NVNPP second unit with using the existing constructions and some equipment. Carried out estimations have revealed that it will cost ~$100M. References Gromov,B.F., Toshinsky,G.I., Stepanov,V.S., et al., 1997, Use of Lead Bismuth Coolant in Nuclear Reactors and Accelerator-Driven Systems, Nuclear Engineering and Design, 173, pp.207217. Masakazu Ichimiya, 2000, A Conceptual Design Study on Various Types of HLMC Fast Reactor Plant, Lead-Bismuth Technology International Meeting (December 12-14, OEC/JNC/Japan). Zrodnikov,A.V., Chitaykin,V.I., Toshinsky,G.I., et al., 2001, NPPs Based on Reactor Modules SVBR75/100, Atomnaya Energiya, 91, Issue6. Gromov,B.F., Grigoriev,O.G., Toshinsky,G.I., Dedoul,A.V. (SSCRFIPPE), Stepanov,V.S. (EDOGidropress), Nikitin,L.B. (NavySC), 1998, The Analysis of Operation Experience of Reactor Installation Using Lead-Bismuth Coolant and Accidents Happened, Heavy-Liquid Metal Coolant in Nuclear Technology, (Proceedings, Conference, Obninsk, Russia), Vol.1, pp.6369. Grigoriev,O.G., Toshinsky,G.I., (SSCRF-IPPE), Leguenko,S.K (PKFRosatomenergoproekt), 1998, Bismuth Demand for Commercial Use of RISVBR75/100 for Solving Different Tasks, Heavy-Liquid Metal Coolant in Nuclear Technology, (Proceedings, Conference, Obninsk, Russia), Vol.2, pp.556563. RollandA.Lengly, 2000, Nuclear Industry of the USA in the Transition Period, Journal of Russian Nuclear Society, 56, pp.3436. SavellyF., 2000, Nuclear Power Economy in the OESR Countries, Journal of Russian Nuclear Society, 56, pp.3640. Sidorenko,V., Chernilin,Yu., 2000, Free Market of Electricity and Possible Consequences, Journal of Russian Nuclear Society, 56, pp.2633. Paulson,C.K., 2002, Westinghouse AP-1000 Advanced Plant Simplification Results, Measures and Benefits, ICONE1022784, 10th International Conference on Nuclear Engineering, (Proceedings, Arlington,VA (Washington,D.C.), USA). Poplavsky,V. M., Kiryushin,A.I., Suknev,K.L.,et al., 2002, Prospects of FN Reactors Development, 3rd Scientific Conference of MINATOM in Russia Nuclear Power. The State and Prospects, Report, (Proceedings, Moscow, Russia). Rachkov,V.,I., 2002, Nuclear Power Economy, 3rd Scientific Conference of MINATOM in Russia Nuclear Power. The State and Prospects, Report, (Proceedings, Moscow, Russia). Wilson, R., 2000, The changing need for a breeder reactor, Nuclear Energy, 39, No.2, pp.99106. Adamov,E.O., Ganev,I.H., Lopatkin,A.V., Muratov,V.K., Orlov,V.V., 1999, Transmutation Fuel Cycle in Large Scale Nuclear Power of Russia, Moscow, GUPNIKIET. 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