Advanced Reactor Technology and Recent R&D in AMEC

Transcription

1 Advanced Reactor Technology and Recent R&D in AMEC Presentation: Nuclear Energy Lecture Series Venue: University of Cambridge Author: Dr John Lillington Date: 6th February 2014

2 Objectives of the Presentation Highlight the novel features of advanced reactor concepts: Discuss some our (AMEC s) interests; Illustrate with some particular examples. Review the current status of the Generation IV (Gen IV) designs: Very High Temperature gas cooled Reactor (VHTR); Sodium & Gas cooled Fast Reactors (SFR & GFR); Molten Salt Reactor (MSR). Consider the R&D requirements for these various designs: Particular reference in this presentation to those pertinent to the reactor fuel and fuel cycle. Discuss some of AMEC s code developments and applications pertinent to advanced reactor designs. In particular: ANSWERS radiation transport codes. 2

3 Evolution of Advanced Reactor Technologies Most of the advanced reactor and Generation IV concepts are evolutions of current generation concepts. These include: Water reactors Gas reactors Liquid metal cooled fast reactors. Over the years, all have received extensive support particularly from UK, France, Germany, US and Russia in terms of: R&D investment Operations support. In order to get the benefit from advantages that advanced reactors can offer, their operational conditions are more challenging in terms of one or more of : Higher temperatures; Higher pressures Greater material challenges with respect to corrosion potential etc, but Nevertheless the lessons learned from current generation plants performance are very valuable and relevant. 3

4 UK Gas Reactor Heritage R&D strongly underpinned by UK domestic gas reactor operational data DIMPLE (R&D reactor) DRAGON (HTR) Magnox (Wylfa) AGR (Torness) 4

5 UK Fast Reactor Heritage R&D strongly underpinned by UK domestic fast reactor operational data ZEPHYR, ZEUS (R & D) ZEBRA DFR PFR 5

6 Gen IV Applications Gen IV designs can meet various energy and future market demands: High-efficiency electricity generation, process heat and co-generation (VHTR); Sustainable fuel and fuel cycle options, improved management of highlevel waste and spent fuel, possibly including recycling; management of stockpiled plutonium from earlier weapons programmes (SFR & GFR (in principle)); Alternative thorium fuel cycle (MSR). Much experience of SFRs for electricity generation exists. The Gas cooled Fast Reactor (GFR) offers some advantages over the SFR, combining VHTR and SFR attributes. The MSR is the least studied of the Gen IV designs but offers the best conversion ratios, the potential for thorium utilisation and other fuel options. 6

7 AMEC Interests and Activities Participation in the UK s Advanced Reactor R&D Programme since the 1980s. Initially a strong fast reactor focus in core neutronics, nuclear data and fuel modelling and behaviour: EC CAPRA and CAPRA/CADRA programmes on Pu burning management and Minor Actinide (MA) destruction. Evolved to cover work relevant to all the Gen IV systems of interest to the UK (VHTR, SFR and GFR): EC advanced fuels programmes such as CONFIRM, FUTURE, RAPHAEL, AFTRA and PuMA; OECD/NEA international benchmarks through working groups IRPhE, JEFF and ICSBEP; IAEA Co-ordinated Research Programmes and international benchmarks. 7

8 Very High Temperature Reactor (VHTR) Main characteristics Helium cooled, Reference power (600 MWt ) Coolant outlet temperature (1000 deg C) 50% efficiency (at this efficiency it could produce 200 metric tonnes of hydrogen/ day Thermal spectrum, Once-through fuel cycle Very high temperature reactors, evolution on earlier HTR technology Applications: High efficiency electricity generation/ hydrogen production/ process heat. Source: USDOE 8

9 HTRs - Current Status VHTR design is based on a helium-cooled, graphite-moderated thermal neutron flux reactor with target outlet temperatures of C. HTR designs have been put forward from the 60s to the present day: UK (Dragon ); Germany (AVR & THTR); US (Peach Bottom & Fort Saint Vrain). Two fundamental designs in HTR technology: Prismatic: Japan (HTTR); France (ANTARES); US(GTMHR); Pebble-bed: South Africa (PBMR), China (HTR). Latest status of HTR; China (HTR-10 currently operational, HTR- PM - under construction). 9

10 HTRs R&D Requirements R&D is centred on fuel and fuel cycle, materials, the thermal conversion cycle, BOP and process heat and hydrogen production technologies. Modelling developments for fuel performance, reactor physics, coolant system. Fuel and fuel cycle Basic HTR (TRISO) fuel is based on UO2 particles coated with four layers: porous carbon, a dense inner layer of pyrolytic carbon, a ceramic layer of SiC and an outer layer of pyrolytic carbon; Ref: Fuel kernel of UCO with a ZrC layer (in place of SiC) also proposed. USDOE advanced fuel programme on ZrC at INL and ORNL, USA; Th and Pu-based fuel could be utilised within a closed fuel cycle; Design objectives are good structural integrity and fission product retention for high burnup at high temperature; R&D is required to confirm these design requirements NB. HTR coated particle fuel can be utilised for Pu and MA management. 10

11 HTRs ANSWERS Codes Developments ANSWERS codes have been developed to model advanced HTR fuel; in particular, reactor physics modelling capabilities for the PBMR using both Monte Carlo and deterministic methods: MONK (Monte Carlo) models randomly-placed particles within a pebble and the pebbles within the core using algorithms developed to give the required core packing fraction. Each pebble can be ascribed an individual burn up. WIMS (Deterministic) modular code allows a triple heterogeneity modelling of resonance self shielding. Sub-group methods are used to treat the heterogeneity of the fuel particles, the particles within the fuel pebbles and the interaction between pebbles of different burn up or composition. 11

12 MONK PBMR Modelling Explicit modelling of fuel and carbon sphere mixing zone Different colours designate different fuel batches Core region Carbon spheres Explicit representation of fuel grains in all spheres Fuel grains Fuel spheres Explicit representation of individual fuel batches All coatings on every fuel grain explicitly modelled in every fuel sphere in the core 12

13 WIMS PBMR Modelling Heterogeneity Level 3 Grain: - Fuel grains and their coatings are treated by collision probabilities and the WIMS sub-group method to account for grain heterogeneity Heterogeneity Level 2, Sphere: -Individual shells within the fuel spheres are modelled. Fuel kernel Carbon Silicon Carbide Pyrolytic Carbon Heterogeneity Level 1, Fuel Batch: - Each core region is modelled as a mixture of fuel spheres and carbon 1.35 spheres (if present). K-eff Iteration to Equilibrium core loading (Radial Geometry, 6 - Spheres of batch fuel cycle) each fuel batch are represented separately to follow individual batch burnups A region is represented 1.15 as a supercell of interacting fuel spheres, representing individual 1.1 fuel batches, and carbon spheres. Using the WIMS subgroup theory 1.05 approach 0 allows 5 the effect on 10 resonance self 15 Iteration shielding to be calculated due to the presence of the different fuel batch burnups. 13

14 WIMS PBMR Modelling (Cont.) WIMS has a specific fuel management code for the PBMR: Heterogeneity - Iteration to Equilibrium core loading (Radial Geometry, 6 batch fuel cycle). - Initial core has all unburnt fuel. - Immediate drop in k-effective due to Xenon buildup. Level 2 - Equilibrium burnups are in principle achieved after 5 iterations, Heterogeneity Sphere Carbon however, due to the shielding interaction between batches additional iterations are Silicon required Carbide to Level 3 Grain Level 1 Fuel Batch - After each cycle, batch Heterogeneity 1 -> batch 2, batch 2 -> batch 3 etc, batch 6 is discharged and batch 1 is replaced with fresh fuel. Fuel kernel achieve the equilibrium core k-effective. Pyrolytic Carbon K-eff Iteration 14

15 PROTEUS Experiments and PBMR Modelling using MONK IAEA programme on Validation of Safety Related Reactor Physics Calculations for LEU HTRs. PBMR simulation experiments in the Proteus facility at PSI. PROTEUS Plan Section Modelling using the MONK code of different critical assemblies. Perform analysis of critical k- effective and also the measured shut-down rod (SDW) worths. Core k-effective Standard Deviation Experimental Value NB The partially inserted control rods were not modelled, leading to an over-prediction of k-effective ~ ~ ~ ~

16 PROTEUS Experiments and PBMR Modelling using MONK (Cont.) Comparisons of MONK SDR worth calculations with experiment, in $, cores 5 to 10. The results show a tendency to under-predict the SDR reactivity by up to 0.4$. This is roughly proportional to the number of SDRs inserted. The PSI worths for graphite filled channels are based on calculation. The statistical uncertainty on the MONK reactivities is 0.08$. Core SDR 5 Inserted SDR 5 & 6 Inserted SDR 5, 6 & 7 Inserted SDR 5, 6, 7, & 8 Inserted MONK EXPT MONK EXPT MONK EXPT MONK EXPT Core Core Core Core Ref: Franklin, B M, Newton, T D, PROTEUS Modelling using MONK, ICNC, St Petersburg,

17 HTRs Other ANSWERS Computer Codes R & D Safety analysis modelling methods for HTR: Coupled Computational Fluid Dynamics (CFD) and ANSWERS WIMS code capabilities have been developed for PBMR and prismatic HTR within the EC Euratom framework. Applicable to normal operation and accident conditions. For GFR, the methodology is based on the European fast reactor physics computer code ERANOS in place of WIMS. AMEC fuel performance modelling capability exists but requires development for HTR fuel: TRAFIC computer code (discussed later) has detailed mechanistic modelling of fission gas, temperatures, stresses and strains in a fuel matrix and could be developed for HTR fuel. 17

18 Sodium Cooled Fast Reactor (SFR) Main characteristics Sodium cooled Reference power ( MWe (small), (large) ) Fast spectrum Closed fuel cycle Prototype sodium cooled fast reactors have been built and operated successfully in US, UK, France, Russia and Japan Concept fallen out of favour at the present time due to high capital cost, proliferation concerns. Applications: Electricity generation; future - actinide waste and plutonium management. Source: USDOE 18

19 SFRs - Current Status SFR is characterised by liquid sodium coolant at relatively low pressure but with high core power density. Three reference system designs: a large loop-type, a medium pooltype and smaller modular design. SFR concept has been established in various countries for many decades (390 reactor-years over 5 decades in 8 countries): Prototypes have successfully operated for many decades: UK (DFR, PFR); US (EBR I, Enrico Fermi, EBR II, FFTF); France ( Rapsodie, Phénix, Superphénix); Russia (BR5, BOR 60, BN 350, BN600); Japan (Joyo, Monju) & Germany (KNK II). Latest status: France (ASTRID SFR demonstrator by early 2020s); Russia ( BN800 under construction). 19

20 SFRs R&D Requirements Main R&D requirements concern the development of advanced fuels and MAbearing fuels, systems integration, component design and materials and safety issues associated with sodium chemical and neutronic reactivity. Modelling requirements: advanced fuels fuel performance, optimisation of core design to improve natural resilience in the event of abnormal transients, and how to limit energy release in the case of sodium-water interaction. Fuel and fuel cycle Advanced fuel designs include mixed oxide, nitrides and carbides, metal alloys (small modular designs). Performance of oxide fuel including cladding, oxide strengthened steel and metal alloy fuel and ferritic-martensitic SS cladding. Research work is examining the consequences of embedding low thermal conductivity fuel/ MA fuel particles in an inert matrix that has higher conductivity. Fuel cycle options associated with the above processes, the recycling of highly radioactive fuel and avoidance of the separation of pure plutonium. 20

21 SFRs Fuel Performance Developments ANSWERS TRAFIC code includes the major underlying mechanisms and models for fast reactor fuel performance in normal and accident conditions. Developed against various fuel-related research programmes: Experimental test programmes such as CABRI which grew out of the EFR project and PFR/TREAT. Other programmes including routine post-irradiation examination, out-ofpile annealing studies of fission gas, void swelling under irradiation in neutron and electron beam facilities ([ISIS, Diamond]. Theory and modelling work from other sources (e.g. atomistic modelling of metals and UO2) have also contributed. OECD/ NEA International uranium and plutonium benchmarks. 21

22 TRAFIC: Underlying Models mechanics fuel and cladding elastic deformation creep, plastic yield, solid swelling friction between fuel and cladding thermal expansion stress, displacement heat flow heat generation by fission conduction in fuel and clad conduction across gap heat transfer to coolant temperature fission gas and helium dissolved intragranular gas atoms trapping in intragranular bubbles face/edge bubbles interlink sintering of porosity porosity; gas in fuel and plenum restructuring grain growth clad corrosion solid-state thermo-migration chemical equilibrium and kinetics structure, chemical composition 22

23 TRAFIC Modelling Capabilities Structure of TRAFIC calculation Examples of Pins that TRAFIC can model 23

24 TRAFIC Code Validation and Development for Advanced Fuels The TRAFIC code developed over many years for predicting fast reactor fuel pin behaviour for steady state and transient conditions Validated against the SCARABEE experiments which simulated the behaviour of fast reactor fuel in reduced coolant flow conditions Also validated against TREAT reactor transient over-power conditions. It was extended from modelling the behaviour of traditional (U, Pu)O 2 fuel to include (U,Am)O 2 and (Pu,Am)O 2 pins and compared with CAPRA test data. The TRAFIC pin model has also been developed to model PuN fuel and (Pu,Zr)N fuel. 24

25 Gas Cooled Fast Reactor (GFR) Main characteristics Helium cooled Coolant outlet temperature ( C) Gas cooled fast reactors considered in the past, e.g. by UKAEA Combines the benefits of gas reactor and fast reactor technology Applications (additional to high efficiency electricity generation), high temperature process heat. actinide waste and plutonium management. Source: USDOE 25

26 Novel Fast Reactor Developments CAPRA project - plutonium management. CAPRA/CADRA - minor actinide (MA) destruction. AMEC (NNC) - a major partner in these EC projects - lasted about 10 years. Aim to investigate the use of fast reactors to manage the back-end of the fuel cycle Included gas cooled fast reactor (GFR) technology evolved from proven UK and other international gas reactor and LMFR experience. European Allegro GFR (100MWt) proposed to be built in Eastern Europe, starting in 2018 and operational from 2025, to be followed by a larger demonstration unit called GoFastR. Inner core subassembly Diverse shutdown rods Outer core subassembly Radial reflector Control rods 26

27 GFR versus SFR Both GFR and SFR are important for the sustainability of nuclear power: More efficient use of fuel; Reduced volumes and radiotoxicity of high level waste. SFRs are the shortest route to FR deployment, but: Sodium coolant has some undesirable features: Chemical compatibility, void coefficient of reactivity, restricted core outlet temperature to avoid sodium boiling. GFRs do not suffer from any of the above: Chemically inert, void coefficient is small (but still positive), single phase coolant eliminates boiling. Allows high temperature operation without the corrosion and coolant radiotoxicity problems associated with heavy liquid metal reactors. But: Gaseous coolants have little thermal inertia rapid heat-up of the core following loss of forced cooling; Compounded by the lack of thermal inertia of the core structure + very high power density. Arguably GFRs offer: enhanced safety, low radiotoxicity and improved performance compared with SFRs. 27

28 Example: Studies in GFR systems to reduce the Need for Heterogeneous MA Recycling Potentially, multi-recycling of the target fuel is required to reduce to a minimum the mass of MAs entering the waste stream. To remove the need for multirecycling it is necessary to achieve a very high level of mass destruction within the target during a single irradiation. The introduction of a moderator within the target has the potential to extend the target residence time and increase the burnup, while retaining the existing design limit on the fuel clad damage. Core Layout with Ex-Core Moderated Targets Ref: T D Newton and P J Smith, "A Moderated Target Sub-Assembly Design for Minor Actinide Transmutation", Proc Global 2003: Atoms for Prosperity, - New Orleans, Louisiana, November 16-20,

29 Studies in GFR systems to reduce the Need for Heterogeneous MA Recycling - Results Neutronic modelling of all core configurations was performed using the European fast reactor neutronics code scheme ERANOS, with the adjusted ERALIB1 nuclear cross section data library. Broad group resonance self-shielded cross sections were produced for each core material using the ECCO cell code. Coupled neutronic and thermal hydraulic analysis was performed to optimise fuel, moderator and burnable poison contents, as well as a moderated target geometry that satisfies the design criteria on thermal and neutronic performance. The optimised results indicated that a mass destruction rate of 70% and a peak burnup of approximately 85% could be achieved during a single irradiation. 29

30 Molten Salt Reactor (MSR) Main characteristics Liquid mixture of sodium, zirconium and uranium fluorides Reference power (1000 MWe) Low pressure/ coolant outlet temperature (700 deg C) Fast spectrum, closed fuel cycle Other design concepts Applications (additional to high efficiency electricity generation), actinide and plutonium management. Source: USDOE 30

31 MSRs - Current Status MSRs (Gen IV design )have the unique feature that fuel is dissolved in a liquid salt coolant, typically a mixture of lithium/ beryllium fluoride. Lithium/ sodium/ potassium and sodium/ zirconium fluorides are other possible salts. In the 1950/ 60s, two thermal spectrum reactors were operated at ORNL, US ( Aircraft Reactor Experiment and the Molten Salt Reactor Experiment (MSRE)) and there were also significant programmes in Russia starting in the 1970s. Early work of molten salt fast reactors was done in the UK. No fast spectrum reactors have been built though there has also been early work on conceptual fast spectrum designs in the US (ORNL, ANL) and in Switzerland (Institute of Energy Research). Recent work has been mainly in China that is launching a programme to develop a Thorium Molten Salt Reactor (TMSR), also referred to as the Liquid Fluoride Thorium Reactor (LFTR). 31

32 MSRs R&D Requirements The R&D requirements are concerned with molten fuel and salt technology, the fuel cycle, the chemistry and other properties of liquid salt, and safety. There is one fewer layer of defence in depth than for the other fuel designs in that there is no cladding. Fuel and fuel cycle: Focus on the fuel and coolant performance in the reactor Gaseous fission product extraction Fuel reprocessing Thorium fuel cycle. We have not modelled MSR directly, but have done some work on nuclear data for the thorium fuel cycle. (Some work on criticality in fissile solutions). 32

33 Thorium Fuel ANSWERS Developments The ANSWERS codes have been developed to allow modelling of general geometries and material compositions. Recent developments include optimization of the physics modelling for thorium fuel applications: Modelling a Th232/U233 thermal breeder core in the Shippingport facility using the deterministic WIMS code and the Monte Carlo MONK code. Comparison of continuous energy and broad group representations of nuclear data, and comparison of deterministic and Monte Carlo methods. Study of multiplication factor and depletion included comparison of the Th232/U233 data from the latest nuclear data libraries (including ENDF/B- VII, JEFF3.1 and JEF2.2) and identifying differences. Differences have been found in the capture cross sections across the thermal to fast energy spectrum. The differences are greater than those found for conventional fuel data (which have been extensively studied). 33

34 Shippingport LWR Modelling Shippingport Light Water Breeder Reactor (LWBR), thermal Th-232/U-233 fuelled reactor. WIMS & MONK analysis to compare through-life and fuel isotopic compositions generated by various data libraries. Acknowledge the contribution to this work from: Harrington, C., Reactor Physics Modelling of the Shippingport LWR, University of Cambridge, Shippingport (LWBR) 34

35 Features of SMRs and Modelling Requirements With respect to electricity generation, Light Water Small Modular Reactors (SMRs) offer flexibility with respect to interfacing with the external electricity grid systems and for meeting different grid requirements. Since SMRs have smaller size than large established power generating reactors, there are more opportunities for new nonelectricity generating applications. More modular construction enables less on-site construction and therefore potential for reduced building capital cost. Therefore we can take advantage of experience gained from earlier LWR, particularly small research and prototype reactors of varied relevant technologies. 35

36 Small Light Water Reactor Applications Extensive legacy experience of Small Light Water Reactor modelling and methods validation experience to call upon. DIMPLE Reactor cross-section PWR related experiments in the DIMPLE Reactor. Shippingport UK Naval Programme Integral PWR Development. Safe Integral Reactor (SIR) Light Water moderated and reflected rod lattices, NEA/NSC/DOE (2006)1 36

37 Summary Advanced reactors open the door to new applications, providing a more sustainable fuel cycle and improved waste management possibilities. However all require significant R&D, particularly in relation to advanced fuels developments. Of the Gen IV designs, the SFR concept has received the most attention and has the most operational experience, followed by the HTR. The MSR concept is the least mature. SMRs are largely based on current LWR technology and offer more flexibility for electricity generation for different capacity grids. Strong support in the US. The ANSWERS codes capabilities are being extended for modelling fuel and core design for all the above reactor concepts. This presentation gives some examples; the work continues. 37