May 2010 FEDERAL SCIENTIFIC PRODUCTION CENTER JSC “Afrikantov OKBM” EFFECTIVENESS EVALUATION FOR THE FAST SODIUM- COOLED REACTOR DESIGN SOLUTIONS AND THEIR EVOLUTION IN NEW DESIGNS B.A. Vasilyev 7th International Scientific and Technical Conference "Nuclear Power Safety, Effectiveness and Economy", MNTK-2010

BN FAST REACTOR TECHNOLOGY STATUS In February 2010, the Government of the Russian Federation approved the Federal Target Program “New Generation Nuclear Power Technologies for the Period and for the Future to 2020” (FTP NGNT) As part of the FTP NGNT, R&D work is provided for the BN-1200, next generation sodium-cooled fast reactor with the electric power of 1200 MW BN-1200 is being developed such that the task be resolved for serial construction of the BN-1200 reactors after 2020 The BN-800 reactor is under construction; the planned completion date for the construction is 2013 On April 8, it was 30 years since the BN-600 reactor commissioning. The BN-600 is the only operating fast power reactor in the world

PROVEN DESIGN SOLUTIONS. BN-600 DESIGN Activity on fast reactors in Russia was started in 1960 by designing the first pilot industrial BN-350 power reactor. The reactor was commissioned in 1973 and was in operation until 1998 In 1980, the next, more powerful BN-600 reactor was commissioned in Beloyarsk NPP In April 2010, the reactor completed its assigned service life of 30 years. Its service life extension to 45 years has been validated

BN YEAR OPERATION EXPERIENCE Capacity factor78% over the last 5 years (close to the capacity factor of serial VVERs equal to 79.9% over the same period) Reactor scrams The average number of reactor scrams for 7000 operating hours is 0.2 (for NPPs across the world, it is 0.5–0.7). No reactor scram was in the period 2000 through 2009 Average radioactive gas release over the last 6 years 1% of the allowable level (4 times lower as compared to VVER NPPs over the same period) Collective personnel dose rate over the last 5 years 0.54 man·Sv per year (2.2 times lower than the same for VVER NPPs)

SODIUM LEAKS IN BN-600 There have been 27 outside leaks (5 of them were radioactive sodium leaks) and 12 SG leaks. The main reason for them is deviations in the manufacture quality of auxiliary pipelines An only radioactive sodium leak (~1 m 3 ) resulted in a radioactive substance release into atmosphere below the allowable limits for normal operation of the NPP The last sodium leak took place in the BN-600 reactor in 1994 Over the last 24 operational years, only one small leak took place in the SG Capacity factor reduction due to the leaks is negligibly small The reliability of design measures to prevent and localize inter- circuit and outer sodium leaks has been convincingly demonstrated

BN-600 CAPACITY FACTOR VARIATION Capacity, %

BN-600 REACTOR CORE MODIFICATIONS Characteristics Modification 0101M2 Operation period for the core option from 2005 Core height, mm Axial blankets height, mm - upper - lower Number of fuel enrichment zones2 3 Core structural materials: - Fuel cladding - FA shroud EI-847 Cr16Ni11Mo3 ChS-68cw EP-450 Maximum fuel rod linear heat rating, kW/m 54.0  48.0 Maximum fuel burnup, % h.a Maximum core life, eff. days (central/peripheral FAs) 200/300560/720 * Average fuel burnup, MW·d/kg U Maximum damage dose, dpa

TASKS TO BE SOLVED BY THE BN-800 CONSTRUCTION AND OPERATION  Breeding mode of operation with MOX-fuel  Pilot-scale demonstration of key closed fuel cycle components  Developing innovative technologies for future LMFBR: advanced fuel and structural materials testing and qualification MA burn-out technology demonstration testing novel technical solutions sustaining competency in the LMFBR technology BN-800 – important milestone in evolution to next generation nuclear power technology

THE BN-800 UNIT DESIGN EVOLUTION AND KEY FEATURES  1984 – initial design– evolutionary up-rated version of the BN-600  1994 – updated design approval Distinctive design features: much higher unit capacity passive safety systems MOX-fuel

 Additional passive emergency protection system is provided for  Emergency decay heat removal system is introduced that utilizes air heat exchangers  A tray is provided for localization of molten core debris in a postulated accident with a failure of all reactor protection equipment IMPROVEMENTS IN BN-800 DESIGN SOLUTIONS(1)  Introduce technical solutions to enhance safety, efficiency and reliability of the power unit:

IMPROVEMENTS IN BN-800 DESIGN SOLUTIONS (2)  Hand operations are eliminated from the refueling system to ensure the possibility to handle fresh FAs with highly radioactive MOX-fuel  The design lifetime is increased from 30 years (BN-600) to 45 years with the possibility of future extension to 60 years  MOX-fuel burnup is increased through the replacement of ChS-68 austenitic steel (burnup is up to 10% h.a.) by EK-164 c.w. (up to 13% h.a.); and then by ferritic- martensitic steel (up to 15% h.a.)

VIEW OF THE BN-800 CONSTRUCTION SITE. MAY 2010 Support belt Vessel bottom

EVOLUTION OF DESIGN SOLUTIONS IN BN-1200 (1)  The primary circuit layout is integral with the safety vessel and lower vessel support  The rotating plugs of the in- reactor refueling system have sealing hydraulic locks based on tin-bismuth alloy  Separate suction cavities for the primary circuit pumps with check valves in the discharge nozzles make it possible to isolate one of the three primary loops without a reactor shutdown in case of equipment failure  There is an in-reactor spent FA storage  Main design solutions, which proved to be successful in BN-600 and used in BN-800: CPS column Intermediate heat exchanger (IHX) Pressure chamber Core Support belt Tray Pressure pipeline MCP-1 Refueling mechanism Rotary plugs Safety vessel Reactor vessel ECS AHE

 Improved reactor and SG designs (reduced materials consumption)  Bellows-type compensators in the secondary circuit pipelines (reduced length and material consumption)  Substantially simplyfied refueling system as compared to BN-600 and BN-800 (reduced materials consumption)  New Solutions: EVOLUTION OF DESIGN SOLUTIONS IN BN-1200 (2)  The emergency heat removal system uses autonomous heat exchangers in-built into the reactor vessel (enhanced reliability)  Primary sodium cold traps are located in the reactor vessel (pipelines containing radioactive sodium and their auxiliary systems are eliminated) Secondary MCP Buffer tank Emergency dump tank Steam generator Intermediate heat exchanger Emergency dump tank Leak-tight cover for above-the- reactor space Expansion tank Air heat exchanger Autonomous heat exchanger Reactor BN-1200 RP Layout

CharacteristicsBN-1200 Nominal thermal power, MW2900 Electric power, gross, MW1220 Number of heat removal loops4 Primary coolant temperature,  C: - IHX inlet/outlet 550/ 410 Secondary coolant temperature,  C: - SG inlet/outlet527/355 Third circuit parameters: - live steam temperature,  C - live steam pressure, MPa - feedwater temperature,  C BN-1200 DESIGN DATA

DESIGN OF THE MAIN EQUIPMENT Technical solutions for MCP-1, MCP-2, IHX in the BN-800 and BN-1200 are basically the same as those in the BN-600 The BN-800 CRDM design was upgraded through simplification of the kinematic chain and enhancement of its reliability; the specific metal intensity was reduced. A similar technical solution will be used in the BN-1200 The BN-800 steam generator design is characteristic of the less number of modules (20 per loop instead of 24 per loop utilized in the BN-600) due to elimination of sodium intermediate steam superheaters. The BN-1200 SGs are substantially enlarged: 2-4 moduls per loop. Technical solutions for the AHX are the same as those in the BN-800 – finned tubes, heat removal by air natural convection.

 Enlarged fuel rods (  6.9 mm   9.3 mm, reduced average fuel heat rating, increased FA life)  Enlarged FA (S=96 mm  S=181 mm, reduced number of FAs)  Increased fuel volume fraction (0.43  0.47, increased breeding factor)  Increased gas cavity in a fuel rod, T clad < 670 °C (to ensure high fuel burnup)  Use of a single fuel enrichment zone instead of three ones (simplified fuel production)  In-reactor storage area that ensures 2-year fuel delay (simplified refueling) NEW SOLUTIONS FOR THE CORE BN-1200

ParameterValue FA core residence, eff. day1320→1650→1980 Maximum fuel burnup, % h.a.14.3→17.8→21 Average fuel burnup for unloaded FAs, MW  day/kg 93→116→138 Maximum damage dose per FA, dpa140→170→200 Maximum fuel rod linear power, kW/m46.5 Breeding factor~1.2 The reactor core design is being developed to ensure possible transition to mixed nitride fuel (breeding factor is up to 1.45) CORE SPECIFICATIONS

REDUCTION IN SPECIFIC METAL INTENSITY FOR THE BN REACTORS ParameterBN-800BN Specific RP materials consumption, t/MW(e), including: Reactor Steam generators External fuel handling system Other equipment

ParameterBN-600BN-800BN-1200 Specific RP materials consumption, t/MW(e) Refueling interval, day110… Capacity factor0.77 – Lifetime, years45 60 BN-1200 economic performance will be comparable with that of VVERs having the same power. In perspective, the cost of electricity generated by BN-1200 should be lower than that of VVER due to expected growth in natural uranium prices. COMPARISON OF BN REACTORS TECHNICAL AND ECONOMIC PERFORMANCE

Properties and technical solutions to ensure safetyBN-600BN-800BN Main properties: - low pressure - low corrosion activity - high boiling temperature + 2 Technical solutions 2.1 Emergency protection: - active - active + passive based on hydraulically suspended rods - active + passive based on hydraulically suspended rods + passive temperature-actuated system Emergency heat removal system: - belongs to the third circuit - air heat exchanger in the second circuit - air heat exchanger connected to the primary circuit Molten fuel tray system Emergency release localization system--+ EVOLUTION OF SOLUTIONS FOR BN REACTOR SAFETY (1)

 Thanks to the solutions adopted in the BN-1200 design, safety parameters are planned to be significantly improved: EVOLUTION OF SOLUTIONS FOR BN REACTOR SAFETY (2)  Core severe damage probability is by an order of magnitude lower than that required by regulatory documents  The exclusion area is within the NPP site for any design accidents  A target criteria has been established: the area for protective measures planning shall coincide with the NPP site boundary for severe beyond-design basis accidents of which occurrence does not exceed reactor/year

 Experience gained in development and operation of fast sodium- cooled reactors demonstrates effectiveness of basic design solutions adopted in BN-600 reactor, their reliable operation and high safety level  The basic design solutions evolved in the BN-800 and BN-1200 designs. The new design solutions for the BN-1200 will have to be tested by analytical and experimental investigations  The BN-1200 design may be related to Generation IV NPPs due to:  optimal combination of reference and innovative solutions  enhanced safety characteristics  high technical and economic performance  possibility of extensive fuel breeding CONCLUSIONS