CONTENTS CO-GENERATING WATER-DESALINATING FACILITY POWERED BY SVBR-75/100 NUCLEAR REACTOR DESIGN ORGANIZATIONS INVOLVED IN THE PROJECT

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Job Allen Stevenson
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3 CONTENTS CO-generating water-desalinating facility powered by SVBR-75/100 NUCLEAR 1 REACTOR Design organizations involved in the project 1 Layout of co-generating nuclear-powered water desalinating facility 2 Performance of co-generating nuclear-powered water desalinating facility 2 Transportable reactor unit 3 SVBR-75/100 REACTOR 4 SVBR-75/100 REACTOR SAFETY CONCEPT 5-6 Coastal power-generating facility 7 Water-desalinating facility 7-8 Economic indices of nuclear-powered water desalinating facility 8 CO-GENERATING WATER-DESALINATING FACILITY POWERED BY SVBR-75/100 NUCLEAR REACTOR PURPOSE electric power generation and desalination of sea water FIELD OF APPLICATION regions with low level of power-generation infrastructure DESIGN ORGANIZATIONS INVOLVED IN THE PROJECT FSUE OKB GIDROPRESS - General Designer of SVBR -75/100 reactor Russian Research Centre IPPE - Scientific Supervisor FSUE SPbAEP Architect Designer of nuclear-powered water desalinating facility FSUE SPMBM «Malakhit» - General Designer of transportable reactor unit FSUE Central Research Institute named after A.N.Krylov Designer of water desalinating facility comprising a distillation water desalinating plant (DDP) and reverse osmosis waterdesalinating plant (RODP) 1

LAYOUT OF CO-GENERATING NUCLEAR-POWERED WA- TER DESALINATING FACILITY replaceable transportable reactor unit stationary coastal power-generating facility 1 transportable reactor unit 2 protective dry

4 LAYOUT OF CO-GENERATING NUCLEAR-POWERED WA- TER DESALINATING FACILITY replaceable transportable reactor unit stationary coastal power-generating facility 1 transportable reactor unit 2 protective dry dock 3 building for steam-turbine plant 4 building for desalinating plant pumps 5 water-desalinating plant modules 6 desalinated water storage tanks 7 platform for reactor coolant solidification prior to transportation 8 office building PERFORMANCE OF CO-GENERATING NUCLEAR-POWERED WATER DESALINATING FACILITY Thermal power of SVBR-75/100 reactor, MW 280 Service life of transportable reactor unit to elapse before replacing, years 8 Maximum capacity of fresh water, m 3 /day Production cost of fresh water, $/m 3 0,74 Electric power of nuclear-powered water desalinating facility with TG operating in the 80 mode of condensing, MW Power output into grid at maximum capacity of fresh water production, MW 9,5 Production cost of electric power, $/kw*h 0,035 2

5 TRANSPORTABLE REACTOR UNIT TRANSPORTABLE REACTOR UNIT resembles a replaceable nuclear storage battery. The transportable reactor units (TRU) are replaced as soon as reactor fuel cycle terminates and they are supplied according to the principle: Construction Ownership Leasing for a period determined by reactor core cycle (at least 8 years). Supplier runs all the financial and radiation risks of TRU construction, transportation, operation and probable accidents. The transportable reactor unit: houses SVBR -75/100 reactor and support systems can be shipped from the Supplier-country by sea to the coastal stationary facilities and back after reactor fuel cycle has terminated can be transported inside the water area of water-desalinating plant, to be installed in a permanent coastal protective structure can be operated as a part of co-generating nuclear-powered water desalinating facility till the end of fuel cycle Layout of TRU on board a carrier 1 transportable reactor unit 2 floating dock 3

6 SVBR-75/100 REACTOR The TRU contains a multi-purpose fast-neutron reactor of SVBR- 75/100 type with lead-bismuth coolant. SVBR-75/100 reactor has been developed within the framework of conversion program for a unique Russian reactor technology applied in nuclear submarines. The well-proven engineering basis for SVBR-75/100 reactor design is created by the 50-year experience in designing and operation of lead-bismuth reactors for nuclear submarines and Russian experience in elaboration and operation of fast sodium reactors. The design of multi-purpose SVBR-75/100 reactor makes use of the following approaches and engineering solutions: fast reactor with chemically inert heavy liquid-metal coolant eutectic lead-bismuth alloy with a very high temperature of boiling and a low temperature of melting; integral layout of reactor primary equipment as a single unit (monoblock): no valves and pipelines for liquid-metal coolant, additional unit shielding is provided by placing the unit (monoblock) inside water tank of passive heat removal system; a possibility of applying different types of fuel (UO2, MOX-fuel with warhead or reactor Pu, mixed oxide fuel with minor actinides - TRUOX-fuel, nitride fuel) without changes in reactor design and meeting safety requirements; side; two-circuit heat removal with natural circulation on the steam generator secondary normal operation and safety functions combined in reactor systems as much as possible; reactor main components are of modular design, a possibility is provided for their replacement and repair; small mass and overall dimensions of the reactor provide a possibility to fabricate it at the Manufacturer s, supply to the NPP site and collect it by any kind of transport. 4

SVBR-75/100 RP SAFETY CONCEPT SVBR-75/100 reactor meets the most stringent safety requirements (human error proof, fail-safe, proof against sabotage and other illintentioned human actions) due to

provides for a possibility to maintain low primary pressure under normal operating conditions and in case of hypothetical accidents, if any; all primary equipment is housed inside a strong vessel

7 SVBR-75/100 RP SAFETY CONCEPT SVBR-75/100 reactor meets the most stringent safety requirements (human error proof, fail-safe, proof against sabotage and other illintentioned human actions) due to reactor inherent safety resulting from reactor design and primary coolant properties: very high temperature of lead-bismuth coolant boiling (~ C) prevents accidents due to DNB in the core and provides for a possibility to maintain low primary pressure under normal operating conditions and in case of hypothetical accidents, if any; all primary equipment is housed inside a strong vessel with a protective housing to provide an integral (single-unit) layout. Small free space between the main vessel and protective housing prevents loss of coolant in case the integrity of reactor main vessel is lost (a postulated accident); the level of natural circulation of the primary and secondary coolant is sufficient for passive heat removal under cooldown conditions; reactor is located inside a tank filled with water. Passive heat transfer via vessel to the tank water provides passive reactor cooldown in case all active heat removal systems fail (a postulated combination of a series of initiating events) within at least 5 days of human non-intervention; negative reactivity feedbacks provide for power decrease in case of spurious CPS rod withdrawal in case of a postulated scram failure to a level that would not result in the core melt; SVBR-75/100 reactor 5

8 a possibility of chemical explosions and fires due to internal causes is ruled out thanks to inherent safety as the lead-bismuth coolant keeps chemically inert in case of loss of circuit integrity and possible contact with water and air. The capability of leadbismuth coolant to retain fission products (iodine, caesium, actinides - except for inert gases) can considerably mitigate the radiological consequences of a postulated lossof-coolant accident; reactor coolant system design and a great difference in the density of steamwater mixture and lead-bismuth coolant prevent the former from getting into the core and provide effective steam separation in case of primary-to-secondary leaks in SG. Gas system condensers and pipelines with rupture disks connect the gas cavity above the level of lead-bismuth coolant and the water storage tank, limit pressure in reactor vessel in case of SG leak and provide steam-water mixture condensation. The tank serves as a relief tank in this case; design of reactor in-vessel shielding, accepted values of boiler water quality parameters and assumed primary-to-secondary pressure ratio, its value being permanently higher in the secondary circuit, prevent the possibility of radioactive contamination of steam to be generated by the system not only under normal operating conditions but also in case of primary-to-secondary leaks. In case the tubes lose their tightness, leaks in SG are repaired by tube plugging with reactor taken out of service. Ingress of lead-bismuth coolant into tight tubes is prevented by SG design; low potential energy accumulated in the primary circuit (low primary pressure) only restricts the scale of possible reactor damage to external impacts. Protection against external impacts is ensured by placing SVBR-75/100 reactor inside a tight and strong vessel in reactor hall, and the transportable reactor unit, in its turn, is located inside a shielding coastal structure ( dry dock); no materials are applied that could evolve hydrogen either under normal operating conditions or in accidents; due to high safety inherent to SVBR-75/100 reactor, even a postulated combination of such initiating events as reactor hall shielding destruction, damage of dry dock ceiling and a large break of primary gas system followed by a direct contact of leadbismuth coolant surface with atmospheric air, does not bring about reactor runaway, explosion and fire. Possible radioactive release is predicted to be below the level that might require evacuation of the local population; 6

9 COASTAL POWER-GENERATING FACILITY receives the transportable reactor unit operates for as long as the nuclear-powered water desalinating facility does and is the property of the Customer Country employs local personnel, uses local resources and manufacturing facilities at construction, equipment installation and operation as much as possible The coastal structures of nuclear-powered water desalinating facility include: a protective dry dock with the system for transportable reactor unit on-site mounting, protective platform for reactor cooling after the core life is over, turbine hall, control room, switchgears, water intake and spillway systems, water-desalinating plant, fresh water storage, startup-backup boiler room, infrastructure buildings, structures for physical protection of desalinating facility site. WATER-DESALINATING FACILITY The nuclear-powered water desalinating facility consists of two types of plants: distillation water-desalinating plant (DDP) of multi-stage evaporation and reverse osmosis water-desalinating plant (RODP). Installed capacity of water-desalinating plants of both types is the same and amounts to 50 % of the desalinating facility installed power, which is m 3 /day. Application of a combined scheme of water-desalinating plants (DDP+RODP) improves efficiency of desalinating facility with a simultaneous assurance of the required quality of produced water. Salt content in DDP desalinated water is 20 mg/l, salt content in RODP desalinated water is 200 mg/l. The water intake and spillway structures are shared by DDP and RODP. Water, which cools the TG and DDP condensers, is used to dilute the discharged brine to ecologically acceptable parameters (< 50 g/l and 32 0 С). The DDP and RODP systems for preliminary cleaning of sea water are combined to a necessary extent. The warm brine, discharged from DDP, is used to heat up water supplied to RODP. Depending on the terms of the Contract, the water processed at the water-desalinating plants can: be subjected to additional treatment to be turned into potable water; additionally pass through the stages of purification to obtain distilled and twice distilled coolant for power plants; be used for irrigation with nitrogen, phosphorus, potassium added. 7

10 As the quality of potable water shall correspond to the national standards or the standards of World Health Organization, the desalinated water to be turned into potable water, if necessary, shall be disinfected and saturated with compounds of calcium to meet the sanitary standards. It can also be saturated with carbon dioxide to meet the Customer s requirements etc. Water is disinfected by chlorination (or by ozonization, which is more expensive), by ultra-violet radiation and treatment with argentum ions. The required backup for water-desalinating plant equipment will be defined as agreed with the Customer. ECONOMIC INDICES OF NUCLEAR-POWERED WATER DE- SALINATING FACILITY For the Customer the cost of construction and operation of nuclear-powered water desalinating facility amount to: capital costs ~ 260 M$, including: - coastal structures $ 60 million; - DDP equipment $ 120 million; - RODP equipment $ 80 million; annual costs ~ 30 M$/year, including: - rent for the transportable reactor unit with account for shipment ~ 12 M$/year; - cost of NUCLEAR-POWERED water desalinating facility operation and maintenance ~ $18 million/year. Estimated cost of transportable reactor unit construction for the Supplier will be ~ $ 44 million including the cost of the first core loading. Pay-back period for the site-specific nuclear-powered water desalinating facility and crediting rate are determined by the Customer depending on the local tariffs for fresh water. For example, the price of fresh water being 1 $/m3, the project pay-back period being 12 years, crediting rate to 10% a year can be acceptable The accepted rent for transportable reactor unit of 12 M$/year will make it possible: for the Supplier to attract investments for transportable reactor unit construction on the basis of commercial credit; for the Customer to provide competitive price of the products produced (for example, fresh water ~ 1 $/m3 and electric power ~ 0,035 $/kw*h) and commercial attractiveness of the project (for example, the pay-back period of the project is ~12 years at loan interest rate of ~ 10%) 8

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