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Status of SNF Pyro Reprocessing RIAR, Dimitrovgrad, Russia

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Presentation on theme: "Status of SNF Pyro Reprocessing RIAR, Dimitrovgrad, Russia"— Presentation transcript:

1 Status of SNF Pyro Reprocessing RIAR, Dimitrovgrad, Russia
SSC Research Institute of Atomic Reactors. RUSSIA Status of SNF Pyro Reprocessing RIAR, Dimitrovgrad, Russia Mikhail Kormilitsyn Research Institute of Atomic Reactors (RIAR) Radiochemical Complex

2 Closing of FC RIAR General Goal Fuel Supply of BOR-60/MBIR
Radionuclide Productions Research Reactors PIE Fuel Supply of BOR-60/MBIR Advanced Fuel development and testing MA recycling Fundamental Studies RAW Treatment Demo of Closed Fuel Cycle R&D for MSR Fuel Cycle

3 New Generation Technological Package “Pyro-Chemistry for CFC”
Universal technological platform for decisions in the field of Closed Fuel Cycle of Nuclear Power: No limitation for: fuel types (oxides, nitrides, metal, carbides, cermet, MSR, IMF) burn up, cooling time No limitation for requirements of decontamination factor (DF up to 106)

4 New Generation Technological Package “Pyro-Chemistry for CFC”
The Base System – Molten chlorides The Base processes Dissolution of initial SNF ( chlorination or anodic dissolution) Electrolysis on solid and/or liquid cathodes Precipitation Purification of the melt Option – Technology of fluoride volatility Option – Partitioning in fluoride melt Option – chemistry and technology of molten fluoride fuel of MSR

5 R&D on SNF Pyro Reprocessing in a World
Oxide SNF reprocessing into Oxide – RIAR (Russia), JNC/JAEA (Japan) Oxide SNF reprocessing into Metallic – CRIEPI (Japan), KAERI (Korea) Nitride SNF reprocessing – JAERI/JAEA (Japan), RIAR (Russia) Metallic SNF reprocessing – INL, ANL (US), CRIEPI (Japan), RIAR (Russia) SNF metallization – KAERI (Korea), RIAR (Russia) HLW partitioning in molten salts – CRIEPI (Japan), RIAR (Russia), KAERI (Korea), CEA (France), ITU (EU) Fluoride volatility processes – CRIEPI, Hitachi/TEPCO (Japan), Kurchatov Inst., RIAR (Russia), INR (Czech. Rep.) MSR Fuel Cycle – RIAR, Kurchatov Inst., CNRS (France), Institute for Applied Physics, Shanghai (China) Other application

6 RIAR Experience

7 RIAR activities in the field of CFC
Since 1964 RIAR has been pursuing large-scale investigations in the following research lines: Pyrochemical production technology of vi-pack U and MOX fuel Pyrochemical reprocessing of SNF from nuclear reactors of various types . Fluoride volatility reprocessing of SNF

8 Milestones of Experience in Closed Fuel Cycle
Pyro R&D- from early 1960-s Demo of fluoride volatility reprocessing – 1970s Pilot facility for pyro/vi-pack MOX fuel production for fast reactor – from late 1970-s BOR-60 full scale fuel supplying only on the base of own RIAR pyro/vi-pack fuel production facilities – from 1980 Pyro reprocessing experience – from 1991 Study on transmutation cycle, nitride fuels and others – from 1992 Start of industrial implementation of pyro/vi-pack MOX technology – 2012 Start of so called “high density” FR SNF (nitride, metal) Creation of “Poly-functional Radiochemical Complex” (PRC)

9 RIAR experience in reprocessing of spent fuel of the BOR-60 and BN-350 reactors
Fuel type Burn-up % h.a. Cooling time, Yrs Weight, kg Date Reactor UO2 7,7 5 2,5 BN-350 MOX 4,7 10 4,1 1991 21..24 1-2 6,5 1995 BOR-60 15 2000 12 2000…2001 16 4-6 2004 U-Pu /Na 6,4 19 0,13 2010 (U,Pu)N /Pb 0,53 8 0,28 U-Zr / Na 9,7 9,5 0,12 U-Pu-Zr /Na 0,10 4 2011 9

10 Dimitrovgrad Dry Process (DDP) – MOX Fuel Pyro processing
Basic research of the molten salt systems allowed for the development of technological processes for production of granulated U and Pu dioxides and MOX. A distinctive feature of the Pyro technology is a possibility to perform all the deposit production operations in one apparatus - a chlorinator-electrolyzer Pyrochemical reprocessing consists of the following main stages Dissolution of initial products or spent nuclear fuel in molten salts Recovery of crystaline Pu dioxide or electrolytic MOX from the melt Processing of the cathode deposit and production of vi-pack

11 Production and testing of vi-pack MOX fuel
Fuel type Number of FAs Burn-up, max.% Load, kW/m Temperature, 0С Reactor (U, Pu)O2 Civil grade/ or weapon quality 330 30,3 51,5 720 BOR-60 UO2 + PuO2 132 14,8 45 705 Weapon grade 26 11,1 46 680 BN-600 Civil grade 4 development of the production technique

12 DDP MOXPuO2 flow sheet DDP MOXMOX flow sheet
Cathode (pyrographite) Stirrer UO22+ O2 Pu4+ PuO22+ + UO2 PuO2 UO2 + NpO2 Cl2 Ar (Cl2) Cl2+O2+Ar Na3PO4 Fuel chlorination 700 оС pyrographite bath, NaCl - KCl Preliminary electrolysis 680 оС Precipitation crystallization Electrolysis-additional 700 оС Melt purification Cl- (MA,REE) RW4 MA,REE NpO2+ DDP MOXPuO2 flow sheet Na3PO4 Cl2 Cathode (pyrographite) UO22+ UO2 + NpO2 Ar (Cl2) + Preliminary electrolysis 630 оС NpO2+ Cl2+O2+Ar Cl2+O2+Ar Stirrer Cathode Stirrer pyrographite bath, NaCl -2CsCl + + MOX MOX MA,REE Pu4+ Cl- UO22+ UO22+ PuO2+ UO22+ UO2 PuO2 PuO2+ (MA,REE) RW4 Fuel chlorination 650 оС Main MOX electrolysis 630 оС Electrolysis-additional 630 оС Melt purification 6500 оС DDP MOXMOX flow sheet

13 MOX-MOX Reprocessing 2000 Year 2004 Year MOX - 3 200 g MOX - 3 400 g
Pu content - 10 % wt. 2004 Year MOX g Pu content - 33,5 % wt.

14 * - TOSHIBA estimation for DDP
Pyro HLW treatment Na3PO4 Radioactive Cs Salt purification Pyroreprocessing Salt residue Fission products NaCl CsCl Phosphates NdPO4CePO4 Waste Phosphates Salt residue Special features contain fission products Alkaline metal chlorides, high activity, significant heat release Basic elements 11 wt.% Nd 4,4 wt.% Ce 81,96 wt.% CsCl 18,04 wt.% NaCl Quantity* <0,15 kg/kg of fast reactor SNF <0,03 kg/kg of fast reactor SNF * - TOSHIBA estimation for DDP

15 Vitrification of HLW from pyro process
Characteristic HLW type Phosphate precipitate Spent salt electrolyte Phosphate precipitate + spent salt electrolyte Glass matrix type Pb(PO3)2 NaPO3 NaPO3, AlF3 Al2O3 Introduction method vitrification, Т=9500С vitrification without chloride conversion, Т=9500С Introduced waste amount, % 28 20 36 137Cs leaching rate as of the 7th day, g/cm2 * day 7*10-6 4*10-6 Thermal resistance, 0С 400 Radiation resistance 107 Gr (for  and ) -decay/g

16 Type of high-level wastes Spent salt electrolyte
Ceramization of HLW arising from pyro process Characteristics Type of high-level wastes Phosphate deposit Spent salt electrolyte Type of ceramics monazite Cosnarite (NZP) Method of introduction into ceramics pressing, calcination , Т=8500С Conversion to NZP from the melt or aqueous solution, pressing, calcination , Т=10000С Quantity of waste introduced into ceramics, % 100 30..40 Leaching rate of 137Cs on 7-th day, g/cm2 * day 1*10-6 3*10-6 Thermal stability, 0С 850 1000 Radiation resistance 5*108 Gy( for  and ) - decay/g

Since 1992 Dry technologies Oxide fuel with MA Vi-pack Integrated disposition on the same site with the reactor TA Transmutation of Actinides


19 Experience in DOVITA Program
Pyrochemical technology of adding Np into oxide fuel (5-20%) has been developed Performance of vi-pack fuel with (U,Np)O2 fuel has been validated experimentally to ~20% burnup in BOR-60 No evidence of significant difference in performance of fuel rods with (U,Np)O2 fuel compared with UO2 or MOX fuel rods has been noticed Pyrochemical process of codeposition of Am with MOX fuel (2-4%) has been developed Methods of Am/REE separation in melts has been tested Special vi-pack targets containing Am oxide with UO2 or inert matrix have been developed Transmutation of Np, Am, Cm is being studied in BOR-60 Irradiated (U,Np)O2 fuel, % burn-up

20 New times consideration: DOVITA DOVITA-2
1992 Dry technologies Oxide fuel with MA Vi-pack Integrated disposition same site with the reactor TA Transmutation of Actinides 2007+ Dry technologies On-site reprocessing Various type of fuel with MA Integration of MA recycling into FR Closed Fuel Cycle TA - Transmutation of Actinides

21 DOVITA-2 Fuel type/ Stages Oxide vi-pack pellet Nitride Metal Molten
salt Concept Studies + +/- R&D -/+ - Fuel Production Irradiation Testing PIE ---- Reprocessing DOVITA-1

22 Current R&D

23 In the Frame of Federal Target Program
For the First Time in a World – pyrochemical reprocessing of FBR spent U-Pu nitride fuel and metal fuel ~ 0,6 kg SF 2 Cd ingots for fabrication of fuel Oxide concentrate FP for wastes preparation

24 Experimental Tests of Spent Nitride Fuel Reprocessing Methods
The empty pies of cladding after anodic dissolution of nitride SNF in chloride melt The sample of fluoride-phosphate glass with real immobilized FPs after reprocessing

25 100% PuO2 Pyro Pellets PuO2 Pellets Characteristics PuO2 Pellets
Melted salt 3LiCl-2KCl NaCl-2CsCl Pellets density, g/sm3 Visual view no cracks no cracs PuO2 Pellets (3LiCl-2KCl, T=450oC) PuO2 Pellets (NaCl-2CsCl, T=550oC)

26 80%UO2 + 20%PuO2 MOX Pyro Pellets
MOX Pellets Characteristics Batch № 1 Technical Requirements* Pellet density, g/sm2 Deviation of pellet density, g/sm2 +0.1 O/Me 1.98 Impurities content, % масс. <0.3 <0.4 Average size of crystalline granules, microns 30-40 < 50 Visual view no cracks Porosity uniform В том числе и только из диоксида плутония, что представляет трудность для порошков оксалатного происхождения *Reshetnikov Ph.G. and et al. Working out, production and operation of fuel rods of power reactors-M:.Energoizdat, p.

27 Pyro MOX pellets (80%UO2 + 20%PuO2)
Microstructure of MOX pellets EPMA μm Content, wt% Pickled pellet Unpickled pellet Solubility test in 8M HNO3 at 95-96оC , duration – 10 hrs. Composition of pellet Medium of powder production Insoluble residue, % wt. Residue composition, wt% 80wt%UO2 +20wt%PuO2 3LiCl-2KCl 0.14 (Pu0.81, U0.19)O2 (Pu0.49, U0.51)O2 NaCl-2CsCl 0.40 (U0.65, Pu0.31,Am0.04)O2 Состав частиц- непрерывный ряд твердых растворов оксидов урана и плутония

28 Fuel pin #TM0-01 with pyro MOX pellets
Dismountable BOR-60 FAs

29 Curium containing salt for spectroscopy studies
Fundamental Studies Curium containing salt for spectroscopy studies NaCl-2CsCl-CmCl3(0.115mol/kg)

30 Spectrum of curium-containing melt under atmosphere of Ar- HCl-H2O
Cm3+ Am3+ CmO+ NaCl-2CsCl-CmCl3(0.115mol/kg) 750oC log(P2HCl/PH2O)=-7.14

31 Synchronized potentiometric titration by oxygen pump and spectroscopy
Cavity from optical quartz Oxygen sensor Oxygen pump CE RE Gas-supply tube

32 MA and FP Partitioning Run # T,oC Weight, g Content in melt, % E, V
(Li,K,Cs)Cl Ga Am Ce 1 350 200 95 0.6 3 -1.2 2 -1.4 -1.6 Ga cathod

33 Experimental Facilities till 2020

34 CFC Pyro technology for International RIAR-based Center of Excellence
RIAR Site Post Irradiation Examination electricity Heat power radionuclides Post Irradiation complex R&D on Reprocessing and Refabrication of Advanced Fuels MBIR MOX or other fuel RadioWaste complex Vi-pack Вибро Radiochemical Complex Отходы на хранение RAW MOX Production Facility Refabricated Fuel Pyro Chemical-Technological complex Пиро RIAR Radiochemical Complex

35 Federal Tasks-oriented Program “New Generation Nuclear Power Technologies”
Large Poly-functional Radiochemical Complex (PRC) Molten salt Reprocessing Facility (1st hot cell line) capacity – up to 1-2 tons of FR SNF per Year (fuel type: oxide, nitride, metallic, IMF) Advanced water-technology Facility, (2nd hot cell line) capacity – up to 1-2 tons kg of SNF per Year New Lab for Experimental and Innovative Fuel Production – (incl. Fuel and Targets with MA) New facility for HLW treatment Demonstration of Closing Fuel Cycle Testing and Demonstration of Closing FR Fuel Cycle for MA Develop the full scale Design of Industrial plants for FR SNF Reprocessing

36 List of Advanced R&D for PRC
Testing of prototypes of technological equipment Development and testing of automatic and robotics systems Comparative FS for different technologies Advanced thermo-chemical decladding technologies Voloxidation Pyrochemical molten salt technologies MOX fuel Mixed Nitrides Metallic IMF, MSR Remote control fabrication technologies Vi-pack Pelletizing RAW treatment Vitrification Ceramization

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RIAR Radiochemical Complex

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