Prospects of Nuclear Power Industry with Sodium-Cooled Fast Reactors

Transcription

1 XI International Public Dialogue Forum NUCLEAR ENERGY, ENVIRONMENT, SAFETY 2016 Prospects of Nuclear Power Industry with Sodium-Cooled Fast Reactors B.A. Vasilyev Chief Designer of BN Reactor Plants November 22-23, 2016, Moscow

2 Introduction The development of fast neutron reactor technology was started by Enrico Fermi in the USA (reactor Clementine, 1946) and Alexander Leypunsky in USSR (reactor BR-5, 1955). Sodium as coolant was firstly applied in fast reactor by A.Leypunsky (reactor BR-5 in 1959). All the fast reactors built worldwide (11 experimental and 9 power reactors) have sodium as coolant. Russia is, nowadays, the only country having commercial powerproducing fast neutron reactors: BN-600 and BN-800 (although research and development of such reactors was carried out actively in USA, France, UK, Germany, and Japan and continued in India, China, and Korea) 2

3 Possibilities for Uranium Application as Nuclear Fuel Natural uranium contains 0.71% U-235 (fissile isotope) and ~99.3% U-238 U-238 is a raw material to produce plutonium (artificial fissile isotope) n,γ -β 238U + n 239Np 239Pu Breeding ratio (BR) = Thermal (moderated neutrons) reactors: Вreeding ratio < 1 (~ 0.5 at VVER) Quantity of the produced plutonium Quantity of the burnt up plutonium (uranium-235) Theoretically, in case of fuel cycle closure, instead of 0.7% natural uranium, at least 1.5% will be available for economy application Fast neutron reactors: BR > 1 (~1.2 with MOX fuel, ~1.4 with nitride fuel) In case of fuel cycle closure, all natural uranium can be used 3

4 Estimation of Energy Potential of Various Nuclear Fuels Nuclear energy with thermal reactors Nuclear energy with fast reactors Coal х 9 Uranium Gas Coal Oil Gas Oil Complete implementation of uranium energy potential by transforming U-238 in fast neutron reactors into plutonium and reapplication of plutonium for production of new fuel (fuel cycle closure) will provide humanity with energy for thousands of years 4

5 Stages of Sodium-Cooled Fast Reactor Development in Russia Main developers of fast reactors: Technical design, R&D stage Scientific Supervisor A.I. Leypunsky Institute of Physics and Power Engineering, Obninsk Chief Designer of RP Afrikantov OKBM, Nizhny Novgorod Chief Designer of steam generator OKB GIDROPRESS General Designer VNIPIET, SPbAEP, ATOMPOROEKT, Saint Petersburg Chief Designer/Technologist of fuel assemblies A.A. Bochvar VNIINM 5

6 Parameters of BN Power Units Parameter BN-350 BN-600 BN-800 BN-1200* Thermal power, МW Electrical power, МW 150** Sodium temperature, ºС at the outlet of reactor Steam temperature, ºС Steam pressure, МPа * BN-1200 power is specified considering application of VVER-1200 generator (NPP-2006) ** electricity generation of МW and production of desalinated water up to t/day. 6

7 Selection of Integral Layout for the Primary Circuit (BN600 Design) To ensure its safety, sodium-cooled NPP is provided with a three-circuit scheme of heat transfer : 1.Core intermediate heat exchanger (IHX) (sodium) 2.IHX SG (sodium) 3. SG turbine (water/steam) IHX Integration of primary circuit equipment into the reactor vessel ensures minimum materials consumption s/output ratio of the RP and minimizes risks of radioactive sodium leakage Core Core 7 Reactor BN-350 (1973) Loop-type layout Reactor BN-600 (1980) Integral layout

8 Main Results of BN-600 Operation (36 Years of Successful Operation) Capacity factor (CF) Sodium leakages The most serious disturbance in the RP operation Emergency shut downs of the reactor Average release of radioactive gases during the period of Collective dose of personnel irradiation in % within the period of commercial operation (since 1982). More than 80% during the recent 5 years Non-scheduled CF losses are ~ 1% 27 leakages of sodium from auxiliary systems (five of the leakages of radioactive sodium) and 12 leakages in SG. Leakages were mostly caused by quality deviations in manufacturing of pipelines and equipment. The last outward leakage was registered in 1993, and the last leakage in SG in The maximum leakage of radioactive sodium is ~ 1 ton; it did not result in any significant radioactive consequences; that event was rated as level 1 by INES scale Average number of emergency reactor shut downs within 7000 hours of operation during the period of is ~0,2 (~0,6 for NPPs worldwide), during there were no emergency shut downs <1% of permissible level 0,408 mansv/a, less than at other NPPs

9 Peculiarities of BN-800 Design BN-800 is significantly improved as compared to BN-600: Power is increased by 1.45 times One turbine, instead of three, is used Additional safety systems are introduced The design was developed on the basis of usage of uranium-plutonium fuel (UO 2 + PuO 2 MOX). Meanwhile enriched uranium fuel UO 2 was used in BN-350 and BN-600 reactors. Construction was recommenced in 2006 (it was commenced in 1984) Introduction into commercial operation is on October 31,

10 BN-800 is an Exceptionally Important Stage in Mastering the BN Reactor Technology Almost all the equipment of BN-800 RP is produced by Russian companies. More than 25 large-scale production factories were involved. Total number of involved companies exceeds 60. BN-800 design s implementation prevented critical breakdown in mastering the fast reactor technology Introduction of BN-800 into operation ensured accumulation of competencies of developers and operational personnel 10

11 Requirements for the BN-1200 Project BN-1200 power unit is developed as a serial facility with focus on safety and economy: Exclusion of necessity to evacuate population, which lives close to the NPP site, in all types of technically possible emergencies, including hardly possible scenarios with failure of all active safety systems and single failures of passive safety systems foreseen by the project for limitation of consequences of such emergencies Ensuring competitiveness of the power unit by simplification of reactor design and confidence in possibility to design reactor with service life of 60 years. The development started in 2007 and is performed as Federal Target Program (Project Proryv ( The Breakthrough ) supported by R&D activities of State Concern Rosenergoatom. In 2016 there was a revision of the project and based on the results of conducted expertise new suggestions for project improvement were given 11

12 BN-800 and BN-1200 Reactor Plant Layouts BN-800 BN-1200 Pipelines of the Emergency Heat Removal System (EHRS) Secondary sodium pump (incorporates the secondary circuit buffer tank) Secondary pipelines (with expansion joints) (connected to the secondary circuit) Primary sodium pump Primary sodium purification system inlet/outlet (BN-1200 does not have these pipelines) In BN-1200 as compared to BN-800: The number of SG modules reduced from 60 to 8 75 DN300 sodium valves eliminated Secondary piping length reduced from 770 to 400 m External primary pipelines (auxiliary systems) eliminated 12 Autonomous heat exchanger (AHX) Intermediate heat exchanger (IHX) Steam generator modules (enlarged, without shutoff valves) Air heat exchanger (connected to the primary circuit)

13 Improvement of BN Designs Characteristics Characteristic Reactor BN-350 BN-800 BN-1200 Technical parameters CF Service life, years 30 (45) Duration of reactor continuous operation between refuelings, eff. days Technical-and-economic parameters Specific volume of reactor housing, m Specific materials consumption of the RP, t/мw(e) The conducted estimations showed that BN-1200 project is close to AES-2006 with regard to specific capital investments, and if to consider additional improvements suggested on the basis of the expertise results of expertise, it is not inferior to VVER-TOI 13

14 Development of Safety Ensuring Solutions in BN Designs Technical solutions to ensure safety BN-600 BN-800 BN-1200 Emergency protection: Active Passive, based on hydraulically suspended rods Passive with temperature principle of actuation System of emergency heat removal: as part of third circuit air heat exchanger is connected to the second circuit air heat exchanger is connected to the primary autonomous heat exchanger + +* In-vessel retention system System of emergency discharges localization * One channel after modernization in 2012

15 Safety Level Increase in BN Designs The probability of severe accidents is ensured at the level of normative requirements for BN-600 reactor and it is significantly lower for new designs: BN-600 ~10-5, BN-800 ~2 10-6, BN-1200 ~ The predicted maximum dose in severe accidents for BN-600 and BN-800 is significantly lower than the limit dose of msv within the first year, that requires obligatory resettlement of the population. The predicted maximum dose in a severe accident in BN-1200 project at the border of construction site will not exceed 50 msv within the first year; population evacuation and resettlement definitely will not be required. 15

16 Mastering the Nuclear Fuel Reactor Type of fuel Average burnup BN-600 BN-600 experimental FSA BN-800 BN-1200 UO 2 enriched MOX (UPu O 2 ) Initial loading of UO 2 enriched. (16% MOX) from % MOX Nitride or MOX fuel Achieved 74 МW day/kg Planned up to 90 МW day/kg Achieved ~ 70 МW day/kg Planned 66 МW day/kg and above Planned МW day/kg* and above * Parameters will be established on the basis of experimental research results which are being carried out using nitride fuel as well Fuel burnup increase is the main area to increase fuel cycle economic indices. Achieved by nowadays average burnup value for VVER reactors is ~ 55 МW day/kg (for UO 2 ). 16

17 Mastering the Closed Fuel Cycle 17 Production facilities of uranium plutonium fuel There is a testing research center for production of FSAs with tablet MOX fuel at PO MAYAK and vibro-mox fuel at Research Institute of Atomic Reactors (RIAR) There is a testing research center for production of FSAs with nitride fuel at Siberian Chemical Combine Production facility of tablet MOX fuel is established at the Mining and Chemical Combine to ensure 100% fuel provision for BN-800 (flexible one in term of quality of plutonium to be used) Stage of feasibility study for plot selection and parameters of uranium-plutonium fuel production for reactors of BN-1200 type Spent fuel processing facilities In operation since Factory RТ-1 (spent fuel from VVER-440, BN-600, Research Reactors, etc.) Production capacity is up to 400 t/year. ~ 50 t. of plutonium are accumulated A pilot batch of EFSAs BN-600 with MOX fuel was processed at factory RT-1 Under construction (commissioning before 2020) Experimental Demonstration Center at the Mining and Chemical Combine for processing of VVER-1000 spent fuel. Production capacity is 250 t/year (~2.5 t Pu/year) Feasibility study is performed to select site and parameters of BN spent fuel processing factory

18 ratio of isotope Pu i, % Fast Neutron Reactor is Capable to Maintain Energy Quality of Plutonium Pu VVER Pu BN WG Pu Pu-238 Pu-239 Pu-240 Pu-241 Pu-242 While uranium-plutonium fuel burns up (in thermal reactors like VVER, PWR), there is an increase of non-fissile (uneven) isotopes content, that makes only one recycle possible. Fast neutron reactor can operate with plutonium of any isotopic composition. Number of recycles is unlimited. In the process of burnup in the fast neutron reactor plutonium is seeking an equilibrium Pu VVER (burnup is 33 МW day/kg) 2 - Pu VVER (burnup is 60 MW day/kg) 3 - Pu VVER (MOX after 1-st recycle) 4 - Pu of its own composition (after multiple recycling in BN) 5 - Pu of weapon quality 18

19 PBH, m 3 Н 2 О per 1 kwh t*yaer (t) Actinides Disposal in Fast Neutron Reactors Fission products Fission products + actinides Actinides: plutonium and its companions neptunium, americium, curium cause radiotoxicity of the spent fuel for a long time and their minimization in the fuel cycle is required Fast neutron reactors are capable of efficient disposal of actinides, i.e. to use plutonium for production of fuel and to burn up neptunium and americium Potential of biological hazard (PBH) of the spent fuel 19

20 Conclusion (1) 1. The gained in Russia experience in development and operation of sodium cooled fast neutron reactors enables their commercialization on the basis of BN-1200 project. 2. Closing of nuclear fuel cycle using fast reactors is a strategic streamline of nuclear energy development in Russia. Therefore, The territorial planning of energy generation in Russian Federation, approved by the Russian Government in its directive 1634-Р of , foresees construction of BN-1200 power units at Beloyarsk NPP and Yuzhno-Uralsk NPP. 3. Readiness of BN-1200 projects for their implementation must be ensured in terms of their improvement to reach required technical and economic indices under conditions of the selected construction sites and to take optimum solutions for establishment of fuel cycle facilities. 20

21 Conclusion (2) 4. The objective of the first stage of BN-1200 operation must be efficient usage of plutonium, both accumulated and newly generated at PT-1 plant as well as new supplies of plutonium from the MCC Experimental Demonstration Center. 5. In future, during transfer to serial construction of BN-1200 reactors and arrangement of spent fuel processing and bulk-scale processing of spent fuel from VVER reactors, there will come a possibility to establish an optimum infrastructure of two-component nuclear energy system using thermal and fast reactors. During operation within the system, owing to their unique features BN reactors will ensure capability of multiple plutonium recycle simultaneously providing thermal reactors with high-quality fuel. 21

22 Thank you for your attention! NPP with BN-800. Autumn

23 Attachment Rationales for Selection of Sodium as Coolant for Fast Neutrons Reactors (1) Presence of significant (decisive) beneficial parameters High boiling temperature 883 С Low specific weight ~ 0.9 g/cm 3 Capability to have more advantageous parameters of reactor plants with pressure in the equipment close to atmospheric Thin-wall structures Excellent thermal physics properties (heat capacity, heat conductivity, heat release) Efficient cooling of fuel rods, including emergency modes Capability to use compact thermal exchanging equipment Low corrosion activity with a simple method to maintain sodium purity (cold filtering traps) Absence of corrosion susceptibility of structures Capability of long term RP operation 23 Efficient in-vessel retention of volatile fission products cesium and iodine (complete bounding due to their chemical reaction) Considerable mitigation of radiation consequences in case of fuel rod cladding failure and damage

24 Attachment Rationales for Selection of Sodium as Coolant for Fast Neutrons Reactors(2) Possibility to compensate problematic features efficiently Melting temperature is above room temperature 98 С There is a vast experience of heating sodium to its liquid state and maintaining in this condition by relatively simple means (electric heating) Interaction with air (burning of hot sodium when leaking from the circuit) Chemical interaction with water (in steam generators) There are proposed technical solutions excluding possibility of radioactive sodium leakage from the first contour Possibility of localization and passive extinguishing of sodium from other systems which is proved by practice There is an intermediate sodium circuit foreseen Potential incident is not radiation hazardous one, its scale and consequences effect only steam generator protection systems. 24

25 Attachment Backgrounds for Success of Russian Sodium-Cooled Fast Reactors Development Continuality and efficiency in R&D arrangement demonstrated by top-management of the nuclear industry and leading technological companies (BR 5/10 BOR-60 BN-350 BN-600 BN-800); Close cooperation of scientists and engineers, as well as high level of their responsibility; Comprehensive substantiation of designs with their compulsory experimental validation at test facilities including sodium ones or under irradiation: - Testing of prototypes (MCP, CRDM, etc.), full-scale mockups (FSAs) or presented models (SG, etc.) were made possible at sodium test facilities developed in IPPE, OKBM, OKB GIDROPRESS and other companies. - Continuous taking into account experimental data obtained using commissioned reactors (especially important in terms of fuel). 25