State of the Art for Fuel-Coolant Interactions Research for LFRs Teodora Retegan, Christian Ekberg, Gunnar Skarnemark tretgan@chalmers.se
Outline Chalmers Lead Cooled (Fast) Reactors Physical bariers and potential release of radionuclides State of the art The Goal of FP7 Project SEARCH The Goal of FP7 Project MAXIMA Ongoing work
Sweden q Forsmark (3 BWR s) q Oskarshamn (3 BWR s) q Ringhals (1 BWR, 3 PWR s) q Barsebäck ( 2BWR s)* * Presently shut down. 23 of March 2010, the Swedish government launched the proposal for building new reactors at present sites; The number will not be higher than 10; The decommissioning law is abolished; 10 of August 2010 the proposed change should be valid.
Sweden Göteborg... situated on the beautiful west coast of Sweden... with two pleasant campuses in the centre of Göteborg
Nuclear Chemistry and Industrial Materials Recycling
Competence areas Aqueous chemistry Chemistry of the nuclear fuel cycle Reactor chemistry (in operation and accidents) Detection of ionising radiation Radioactive work Radiation chemistry and radiation biology Separation processes Organic synthesis Design of equipment for separation processes Processes for materials recycling Statistics and uncertainty analysis
Equipment Fully equipped alpha laboratorium (only university in northern Europe): - Alpha laboratory with permit to work with 4 g Pu or 20 GBq other alpha activity in 3 small and 1 large glove box - Infrastructure to handle large amounts of alpha emitters - We are allowed to possess 50 g Pu, 50 g U (enrichment >20 %), unlimited amounts of LEU and Th - Max 200 GBq of toxicity class A + B (=Pu, Am, Cm, etc.) Fuel laboratorium (to be finalized) Gamma-laboratorium with hot cell and manipulators Several radiochemical laboratories Separation equipment for lab- and pilot processes Detection equipment for alpha-, beta-, gamma- and neutron radiation Highly sensitive equipment for metal concentration determination: ICP-MS, ICP-OES (few in EU that can handle radioactive samples) XRD, SEM, Autoradiograph.
The lead cooled reactor
Pros and cons of heavy liquid metals K705 Soviet submarine + No rapid exothermal reaction with water. Heat exchanger may be located in primary circuit + High boiling temperature (2016 K for lead) + High absolute expansion coefficient. Buoyancy forces may be used for decay heat removal + Lead is good for retaining iodine & caesium (chemically) and gamma radiation (physically) Coolant technology only used in military reactors with intermediate neutron spectrum Costs for oxygen control and surface protections Erosion of pump materials
First commercial LFR: SVBR-100 SVBR-100 " Based on silicon oxide protected ferritic-martensitic steel cladding, the Soviet submarine reactor design has been converted to a commercial concept " SVBR-100: lead-bismuth cooled reactor with 100 MWe power, using MOX or nitride fuel. MOX gives breeding ratio ~ 0.84, nitride ~ 1.0 " Development financed by consortium between Rosatom and private investors " Construction of prototype to start 2017 in Dimitrovgrad " http://www.akmeengineering.com/svbr100.html
Other LFRs (planned) SVBR, 100 MWe Russian LBE cooled reactor with MOX fuel, for commercial electricity production in remote areas. Construction to start in 2017. BREST, 300 MWe Russian LFR with (U,Pu)N fuel. MYRRHA, 100MWth Belgian LBE cooled multi- purpose facility, developed by SCK CEN Hyperion, 25 MWe LBE cooled battery with UN fuel. Prototype planned for Savannah River site. ALFRED, 130 MWe European LFR demonstrator. Romania official host candidate.
Test bed for LFR technology (1st pure LFR!) Research on fast reactor dynamics R&D on fuel recycle & manufacture Training of LFR operators (MYRRHA, ALFRED) Education of nuclear engineering students Specifications ELECTRA - FCC 397 fuel pins, Dclad = 12.6 mm (Pu0.4,Zr0.6)N fuel with Pu from spent UOX Fuel column height: 30 cm Active core dimensions: ~ 30 x 30 cm Reactivity compensation using 6 rotating drums with B4C initially facing core
Physical bariers and potential release of radionuclides Three physical bariers: Fuel matrix itself The cladding The boundary of the primary coolant system (SKB)
If the matrix is changing (during normal operation) enought to breach the cladding, than: A primary crack can occur, thus allowing the coolant to potentially react with the fuel and form another chemical compound, Due to this reaction the volume of fuel will increase resulting in further cracking, Eventually the radionuclides will leave the pin Three types of groups of elements can be distinguished in fuel after irradiation: long-lived fission products (LLFPs) such as: 129 I (t 1/2 =1.57 10 7 y), 99 Tc (t 1/2 = 2.1 10 5 y), 135 Cs (t 1/2 = 2 10 6 y), etc., transuranium (TRU) elements, which mainly include plutonium isotopes and minor actinides (MAs), i.e. isotopes of Np, Am and Cm activation products such as 59 Ni (t 1/2 =7.5 10 4 y), 93 Zr(t 1/2 =1.5 10 6 y), 94 Nb (t 1/2 =1.57 10 7 y) etc. THESE MUST BE CONTAINED
Scarce literature data 1. An Innovative Fuel Design Concept for Improved Light WaterReactor Performance and Safety, J.S. Tulenko, R.G. Connell, Final Technical Report: US Department of Energy Grant April 24, 1992 -April 14, 1995, DE-FGOS-93ER75880 - The purpose: to explore a technique for extending fuel performance by thermally bonding LWR fuel with a non-alkaline liquid metal alloy Filling the gap between the fuel and clad with a high conductivity liquid metal - The experiments: An assessment of the technical feasibility of this concept for LWR fuel is presented, including the results of research into materials compatibility testing and the predicted lifetime performance of Liquid Metal Bonded LWR fuel. A lead-bismuth eutectic alloy (44.8wt% Pb-55.2wt% Bi) was then studied, along with a lead-bismuth-tin alloy (33wt% Pb-33wt% Sn-33wt % Bi) at normal operational conditions: 750 F (398 C) for 100-3,500 hours and limiting accident condition 1,200 0 F (648 C) to 1,500 F (815 C) for time periods up to 24 hours - The results: liquid metal bonded BWR peak fuel temperatures are 400 F (204 C) lower at beginning-of-life, and 200 F (94 C) lower at end-of-life compared to conventional fuel.
2. Extraction of uranium nitride pellets during extraction of lead from Brest reactor fuel-element simulators by melting, L. P. Sukhanov, O. A. Ustinov, V. I. Sorokin, A. V. Markov,* and V. I. Zherikov, Atomic Energy, Vol. 99, No. 1, 2005 - The purpose: Spent fuel from the BREST-OD-300 fast reactor, which is now under development, is supposed to be reprocessed directly at the nuclear power plant by a water-free technology (the cladding gap is filled with Pb, fuel: uranium and plutonium mononitride pellets). - The method: A method of separating the jacket from the fuel, including opening up of the fuel elements from the ends and extracting the pellets together with lead to be melted out by heating to 500 C, has been checked experimentally; - The results: The uranium content in the lead melted out did not exceed 6 10 3 mass%. (!) Essentially no uranium is transfered into the lead. 3. The Russian testing program ( as presented by Dr. V. Smirnov) however, no data are available.
SEARCH To study the LBE-fuel interaction and characterization of the reaction products and their potential effect on the fuel element in order to qualify the fuel for MYRRHA and additionally experiments with UO2/PuO2 pellets/powder; To understand the phase relations Pb-U-O, Bi-U-O, Pb-Pu-O and Bi-Pu-O for which there is no published available data; To investigate two MOX manufacturing route: - sol-gel technique (wet route) JRC-ITU - powder pressing (dry route) NRG To acquire knowledge on phase relation between the different species involved in order to extent the current knowledge the phase diagrams.
MAXIMA Methodology, Analysis and experiments for the "Safety In MYRRHA Assessment WP5 Addresses severe accidents scenarios like: MOX/LBE interaction under different atmosphere MOX/LBE/Cladding interaction under different atmosphere Fuel-coolant interaction up to 1700 C (CHALMERS) Fuel dispersion experiments (KTH)
ELECTRA-FCC Study the coolant-fuel or coolant fuel-cladding interaction and characterization of the reaction products and their potential effect on the fuel element in order to qualify the fuel for the future reactor system (ELECTRA-FCC) for different temperatures (e.g. 500, 800 and above 1500 0 C). - KTH investigates already UN with pure lead made at KTH on the dry route: powder pressing (D. Grishchenko, A. Konovalenko, A. Karbojian, P. Kudinov and S. Bechta) in GENIUS project; - CHALMERS will start in short time investigations on Pu nitrides interactions with pure lead Investigate the fuel manufacturing route (in the CHALMERS fuel laboratory under ASGARD project) influence on interraction: - sol-gel technique (wet route) - powder pressing (dry route)
Thank you for your attention!