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

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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.

Objectives of the Presentation

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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.

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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)

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UK Fast Reactor Heritage

ZEBRA ZEPHYR, ZEUS (R & D)

DFR PFR

R&D strongly underpinned by UK domestic fast reactor operational data

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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 high-

level 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.

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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.

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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

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HTRs - Current Status

VHTR design is based on a helium-cooled, graphite-moderated

thermal neutron flux reactor with target outlet temperatures of 1000 0C. 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).

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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: http://www.kontentkonsult.com/pbmr.html

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 burn-

up 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.

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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.

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MONK PBMR Modelling

Core region

Different colours designate different fuel batches

Fuel spheres Fuel grains

Carbon spheres

Explicit representation of individual fuel batches

Explicit modelling of fuel and carbon sphere mixing zone

Explicit representation of fuel grains in all spheres

All coatings on every fuel grain explicitly modelled in every fuel sphere in the core

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WIMS PBMR Modelling

1.05 1.1

1.15 1.2

1.25

1.35

0 5 10 15

Carbon

Iteration to Equilibrium core loading (Radial Geometry, 6 batch fuel cycle).

Iteration

K-eff

Fuel kernel

Silicon Carbide

Pyrolytic Carbon

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.

Heterogeneity Level 1, Fuel Batch: - Each core region is modelled as a mixture of fuel spheres and carbon spheres (if present). - Spheres of each fuel batch are represented separately to follow individual batch burnups. - A region is represented as a supercell of interacting fuel spheres, representing individual fuel batches, and carbon spheres. Using the WIMS sub-group theory approach allows the effect on resonance self shielding to be calculated due to the presence of the different fuel batch burnups.

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WIMS PBMR Modelling (Cont.)

1.05 1.1

1.15 1.2

1.25 1.3

1.35

0 5 10 15

Heterogeneity Level 2 Sphere

Heterogeneity

Level 3 Grain

Carbon

Heterogeneity Level 1

Fuel Batch

Iteration

K-eff

Fuel kernel

Silicon Carbide

Pyrolytic Carbon

- 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. - After each cycle, batch 1 -> batch 2, batch 2 -> batch 3 etc, batch 6 is discharged and batch 1 is replaced with fresh fuel. - Equilibrium burnups are in principle achieved after 5 iterations, however, due to the shielding interaction between batches additional iterations are required to achieve the equilibrium core k-effective.

WIMS has a specific fuel management code for the PBMR:

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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. Modelling using the MONK code of

different critical assemblies. Perform analysis of critical k-

effective and also the measured shut-down rod (SDW) worths. NB The partially inserted control

rods were not modelled, leading to an over-prediction of k-effective

Core k-effective Standard Deviation

Experimental Value

5 1.0112 0.0004 ~1.008

7 1.0088 0.0004 ~1.008

9 1.0095 0.0004 ~1.008

10 1.0059 0.0004 ~1.008

PROTEUS Plan

Section

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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 5 -3.45 -3.57 -7.25 -7.5 -11.04 -11.45 -15.16 -15.13

Core 7 -9.11 -9.5

Core 9 -3.68 -3.68 -7.62 -7.85 -11.57 -11.61 -16.01 -16.43

Core10 -2.46 -2.61 -5.39 -5.52 -8.20 -8.63 -11.20 -11.76

Ref: Franklin, B M, Newton, T D, PROTEUS Modelling using MONK, ICNC, St Petersburg, 2007.

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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.

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Sodium Cooled Fast Reactor (SFR)

Main characteristics Sodium cooled Reference power (150-500 MWe

(small), 500-1500 (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

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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 pool-

type 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).

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SFRs – R&D Requirements

Main R&D requirements concern the development of advanced fuels and MA-

bearing 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.

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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-of-

pile 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.

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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

‘restructuring’ grain growth clad corrosion solid-state thermo-migration chemical equilibrium and kinetics

structure, chemical composition

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

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TRAFIC Modelling Capabilities

Structure of TRAFIC calculation

Examples of Pins that TRAFIC can model

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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)O2 fuel to include (U,Am)O2 and (Pu,Am)O2 pins and compared with CAPRA test data. The TRAFIC pin model has also been developed to model PuN fuel

and (Pu,Zr)N fuel.

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Gas Cooled Fast Reactor (GFR)

Main characteristics Helium cooled Coolant outlet temperature (1000

0C) 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

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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.

Radial reflector

Outer core subassembly

Inner core subassembly

Diverse shutdown rods Control rods

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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.

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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 multi-

recycling 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, 2003.

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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.

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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

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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).

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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).

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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).

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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, 2012.

Shippingport (LWBR)

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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 non-electricity 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.

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‘Small’ Light Water Reactor Applications

Extensive legacy experience of ‘Small’ Light Water Reactor modelling and methods validation experience to call upon.

PWR related experiments in the DIMPLE Reactor.

Shippingport

UK Naval Programme

Integral PWR Development.

Light Water moderated and reflected rod lattices, NEA/NSC/DOE (2006)1

DIMPLE Reactor cross-section

Safe Integral Reactor (SIR)

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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.