An Introduction to the Moltex Energy Technology Portfolio

Transcription

1 An Introduction to the Moltex Energy Technology Portfolio January 2018

2 Introduction Molten Salt Reactors (MSR) were conceived at the very start of the nuclear era. They have many advantages over conventional nuclear power plants but have never been commercialised. The main advantages of MSRs are discussed along with their challenges. The inventions of Moltex Energy overcome these challenges and enable the many advantages of molten salt fuelled reactors, resulting in the Stable Salt Reactor (SSR) that can produce power at a lower cost than fossil fuelled power generation, store energy at grid scale to complement intermittent renewable energy sources while largely eliminating the problem of long lived nuclear waste from current nuclear reactors. The contents of this report are covered by international patents. Contents 1. Generic Molten Salt Reactor Advantages Generic Molten Salt Reactor Challenges Solutions with Static Fuelled MSRs (Stable Salt Reactors) Other Features... 6 Compact Design... 6 Modular construction... 6 On power continuous refuelling... 7 Standard turbines and Baseload peaking power with GridReserve... 7 Coolant integral electrochemical conditioning and filtering... 8 Passive decay heat removal Walkaway Safety... 8 Generic Small Modular Reactor benefits SSR Plant Level Description (SSR-W300)... 9 The Reactor Fuel Handling Heat Transport Safety Systems Instrumentation and Control Balance of Plant Moltex Energy Technology Portfolio & Fuel Cycle Moltex s fuel conversion technology (WATSS) SSR-W Reprocessing Heat Battery GridReserve - Hybrid Nuclear & Renewables Detailed Technology Descriptions and Justifications (SSR-W) Development Challenges Ahead Independent Validation

3 1. Generic Molten Salt Reactor Advantages 1. Low volatile source term In conventional fuels, the two nuclear fission waste products that carry the greatest risk to human health are caesium and iodine. In molten salt fuel these hazardous fission products are not in the form of gases but as stable salts that cannot be dispersed through the air. This theoretically reduces the volatile source term by six orders of magnitude compared with an equivalent severe accident with an oxide fuelled reactor. 2. Low pressure operation Molten salts have a large liquid range, with a melting point at 400C+ and a boiling point of up to 1500C. High pressures are not required anywhere in the reactor island. This reduces stresses in all containment vessels which increases safety and reduces cost. 3. Excellent heat transfer and radiation resistance Sodium is deemed an effective coolant for nuclear reactors for its high heat capacity and neutron transparency. The volumetric heat capacity of molten salts is over three times that of sodium meaning compact cores and low coolant flow rates are possible. Molten salts have been demonstrated to be extremely tolerant to radiation fields. 4. Potentially simpler fuel cycle and reprocessing Reprocessing by aqueous means has proven to be uneconomical. Reprocessing with molten salts is more compact and simpler. The high purity and tight tolerances required for MOX fuel are not necessary for a liquid fuel. On line removal of U-233 bred from thorium is achievable in some configurations and makes thorium breeding viable. 5. Strong reactivity feedback Molten salt fuel expands far more than solid fuels when heated giving molten salt reactors strong negative temperature vs reactivity coefficients making them easier to control and safer. 3

4 2. Generic Molten Salt Reactor Challenges 1. Fission products and fission throughout the reactor system Historically, all molten salt reactor designs have mixed the fuel and coolant as one. All fission products (about a third of the periodic table of elements) therefore circulate through the plant, depositing on surfaces, creating reactivity perturbations, or introducing complex corrosion behaviour. This makes component specification and chemistry management extremely difficult. The fuel/coolant mix is also extremely radioactive with strong gamma emissions at the kw per litre level. As it is circulating through the plant it must be assumed to have the potential to leak and the reactor would not be accessible by humans. All maintenance and repair must be remote. 2. Material challenges Molten salts are highly corrosive to metals unless their chemistry is controlled to ensure they are strongly reducing. In many MSR designs, a balance has to be achieved between the salt being reducing enough to control corrosion, the salt having an acceptably low melting point and uranium salts being reduced to uranium metal by the salt. Enabling this difficult balance can require special metal alloys with low corrosion potential. High nickel alloys have been used to slow down corrosion, but it is still an ongoing challenge and long lifetime of alloys exposed to molten salts are difficult to achieve, and even more challenging to validate for a long life commercial nuclear reactor safety case. 3. Incompatibility with IAEA Safeguards procedures All fuel in the world today is tracked in accordance with IAEA Safeguards procedures. A liquid fuel circulating through a system would require new procedures, as this could be more easily-diverted without detection. IAEA safeguards experts have commented that such new procedures will take years to agree. 4. Off gassing needs active management Gaseous fission products are produced continuously and if salt chemistry is not appropriately controlled is a source of volatile radioactive iodine. This require systems to take the gasses out of the fuel salt and manage them. This has been done once in the 1960s but caused problems and will require extensive research before commercialisation. 5. Sophisticated linked neutronic and thermal hydraulic codes do not exist Today s nuclear simulation codes have been developed and validated over many years based on fission occurring in pins of fuel and a coolant taking it away to a heat exchanger. In a conventional MSR the heat is produced directly into the fuel/coolant mix and much more sophisticated models are required that simultaneously track the nuclear materials and heat transfer. In particular, the delayed neutrons that are critical to reactor stability may be swept out of the reactor core in reactors where the fuel and coolant mix is pumped through heat exchangers. This will require many years of development and would be difficult to fully validate without an operational reactor. 4

5 3. Solutions with Static Fuelled MSRs (Stable Salt Reactors) The Stable Salt Reactor boasts all the advantages of a generic MSR and overcomes the above challenges as follows: 1. The engineering complexity of pumping a highly radioactive fuel is avoided The nuclear material is contained in fuel assemblies in the core like all other reactors in operation. Moltex has granted international patents on the use of molten salt fuel in fuel assemblies. Standard industrial pumps and vessels can be used for the low radioactivity separate coolant salt. Fewer parts are required. 2. Simple passive redox control is possible and hence existing certified alloys Moltex has granted patents on a novel chemistry control mechanism for molten salt. For the fuel, a sacrificial anode (zirconium) is used to keep the molten salt in a strongly reducing state. This ensures there is no tendency for chromium or other elements to be leached out of the fuel pin steel into the salt. Standard steels can be used that already have the appropriate irradiation data (any steel with sodium fast reactor irradiation). The same redox control essentially eliminates release of radioactive iodine via volatile tellurium compounds. This is not achievable with other salts or when the fuel salt is pumped, like every other MSR. 3. Compliant with current IAEA safeguards protocols As the fuel is in conventional fuel assemblies, it is tracked as per existing international best practice. 4. Off gassing is passive Because caesium and iodine gas are not emitted in significant quantities, non-return gas release vents can be used at the top of the fuel assemblies, so no pressure build up in the fuel pin occurs. The gases collect first in the upper plenum of the fuel tubes, then in the reactor gas containment and are only released to atmosphere in a controlled manner through operation of the containment airlocks. This mechanism ensures that highly radioactive decay products of xenon are retained in the fuel pins and not released to atmosphere. 5. All materials exposed to a high neutron flux are in consumable fuel assemblies In Sodium Fast Reactors sodium is highly transparent to neutrons so many components beyond the core suffer severe damage. The molten fluoride coolant salt in the SSR contains hafnium and acts as an excellent neutron screen to protect all components beyond the core. Only the fuel assemblies which pass through the reactor as the fuel is burned up, are exposed to the high flux. 5

6 4. Other Features Compact Design The lower quantity of parts required, as highlighted above, has a clear and direct correlation with the overall cost of the SSR. Another consequence is the much smaller overall size/footprint. From an economic perspective, this is a key benefit leading to lower construction and operation costs. The low pressure and chemical stability of the fuel salt and coolant salt means that a simpler engineering system can be used in the SSR. This in turn means the SSR can be significantly smaller than a comparative reactor of equal generating capacity demonstrated by the to-scale diagrams below; on the left a comparison with a modern 1000MW Pressurised Water Reactor, and on the right a comparison of the 150MW SSR module vs. a 50MW module of NuScale s Integrated Pressurised Water Reactor, both perceived as the most innovative and cost effective on the market today. Both reactors illustrated are 1 GW in capacity Figure 1: To-scale comparison of the SSR-W1000 with AP1000 (left) and to-sclae compariosn of an SSR-W module compared with a a NuScale module. The dimensions of an 1000MW SSR unit are 30m x 17m x 8m. This compares with the AP1000 shown in the diagram on the left above with dimensions of 82m high x 40m diameter. Clearly such a materially smaller design means the construction costs are lower. It also means design costs are lower, as well as future transport costs for Moltex s customers. Informed readers will observe that this comparison is not quite a like to like comparison as both the PWR and the NuScale module include all the steam generating hardware inside the nuclear island and inside the reactor shield building. But this is actually the point, the SSR design permits all that complex and hazardous high pressure equipment to be outside the nuclear island, indeed outside the licensed nuclear site in many cases. It thus has costs similar to those in fossil fuel powered stations, which are dramatically lower than in nuclear stations. Modular construction Small Modular Reactors may benefit from low cost modularisation of their units. However they lose the economies of scale as many small reactors are now needed. The SSR combines these two concepts. Simple small modules combine together to form one large plant. The dual economies of modular construction and economies of scale are achieved. The below illustration shows how the modules are 6

7 placed in a long tank to form a 1200MW reactor. At only 18m long, this could be transported on a standard lorry. Figure 2: SMR-W module in isometric view and plan view illustration of 8 modules in a tank. On power continuous refuelling Refuelling is carried out in a similar manner to a CANDU reactor. A depleted assembly comes out one side and a fresh one goes in the other. This is illustrated in the image adjacent. No excess reactivity is in the core, mitigating the need for control rods and increasing safety. Thermal transients are avoided which impacts material lifetime. Fresh fuel assemblies are inserted at the edge of the coolant and once the fuel has melted, they are moved into the core where vertical movement is restricted. Once burned up, they are removed from the opposite side of the core to the edge of the coolant and left to cool before removal. Figure 3: SSR-W module in plan view with fuelling direction illustrated. Standard turbines and Baseload peaking power with GridReserve The primary coolant circulates between 525C and 650C. This temperature enables standard superheated steam turbines to be used from the fossil fuel power industry. The turbines and entire steam island are also outside the nuclear licensed area and are not linked to the safety case of the plant other than as a potential external hazard. This makes the capital cost substantially lower and reduces operating costs. The SSR outputs heat in the tertiary salt loop at 550C. This is stored in large low-cost salt storage tanks so that the thermal energy can be turned to steam for electricity only when it is needed. Any size storage and any size steam turbine can be used. A 1GWe SSR costs 1,330/kW (<$1,700/kW). A 1GWe SSR with 3 x 1GW turbines and 8 hours of storage costs circa 600/kW ($750/kW) as it is now a 3GWe power plant operating at 33% capacity. This capital cost is actually lower than a comparable Combined Cycle Gas Turbine plant operating at the same, quite typical, capacity factor. MIT and the University 7

8 of Berkeley carried out a study and the increased value of electricity produced when it is flexible in this manner is 42% in Texas and 67% in California 1. This need for flexible power will only grow in the future as renewables expand. Coolant integral electrochemical conditioning and filtering Several simple monitoring systems are in place to monitor and control the reactor. Hafnium levels are monitored and topped up as they are depleted in the coolant over months to years. This keeps the coolant acting as a neutron screen. Cadmium and zirconium tetrachloride vapour is taken out of the cover gas continuously. This is a simple removal using a condenser trap. Other gases are primarily the noble gases, and have sufficiently decayed that they can be released to atmosphere. The coolant salt is maintained in a chemically reducing, non corrosive, state and any contaminating oxygen is removed by a continuous electrochemical process in a replaceable module inserted into the coolant tank. The use of ZrF 4 as coolant salt ensures that any traces of oxygen are complexed prior to their electrochemical removal in the form of highly stable zirconium compounds. The existence of a low valence form of zirconium (ZrF 2) ensures that the chemically reducing state of the salt is maintained throughout the coolant salt system and cannot be overcome by local ingress of small amounts of air or water. Passive decay heat removal Walkaway Safety The emergency cooling system consists of air ducts around the reactor tank which continually circulate air past the tank purely by natural convection. This is similar to other reactors such as PRISM. The Moltex system is superior however because it can take away the decay heat from small high power reactors which is not possible for reactors like PRISM. This system relies on a novel method to capture heat radiated from the hot reactor tank wall on a large surface area of fins which in turn pass the heat to the circulating cool air. This system has been patented and allows passive air cooling to be used on much more powerful and compact reactors than was ever conventionally thought possible. Generic Small Modular Reactor benefits The SSR also benefits from the advantages of SMR designs, which are: Financing cost reduction a shorter design and construction period means lower overall borrowing and lower financing costs for this borrowing. Off-site construction the reactor is constructed offsite where there are fewer constraints (for example with tooling access, or physically with other companies operating on the same site). The constructor can also benefit from operating out of their typical factory set-up, i.e. with the equipment and skills in situ. Modularisation reactors are built in smaller modules that can be constructed several times over. This yields economies of scale both in terms of the knowledge and experience gained from one module to the next and in the materials used for the reactors, which can be brought in on a greater scale than if it were a bespoke reactor. Digitising line management the production line for the parts in the reactor is digitised, streamlining it and ensuring maximum efficiencies. 1 Fluoride-Salt-Cooled High-Temperature, Reactor (FHR) Commercial Basis and Commercialization Strategy, MIT, 2014, see page 63 8

9 5. SSR Plant Level Description (SSR-W300) The SSR-W300 is the physically smallest and simplest design in the SSR family (see Section 6 below). The reactor is rated at 750 MW th and is capable of providing a mean electrical output of 300MW. It is fuelled with uranium, plutonium and higher actinide trichlorides that have been derived from spent fuel from traditional thermal reactors, and hence is identified as a waste burner. An overview of the reactor facility is given in Figure 4. The SSR reactor and associated ancillary services buildings, along with the control building and offices and workshops, fits into a ground footprint of 150 m by 75 m. The turbine building, and associated ancillaries/electrical switch rooms do not have to be contained within the same site, although they can be if this is expedient. If they are included the total site will fit within a ground footprint of 150 m by 150. The SSR-W facility description is broken down into six high level plant areas to facilitate understanding. These are: The Reactor Heat Transport Safety Systems Instrumentation and Control Fuel Handling Balance of Plant Containment Shroud Reactor Service Room Decontamination Bay Containment Access Airlock Reactor Pit Reactor Tank Reactor Containment Figure 4: Overview of SSR-W Reactor Facility 9

10 The Reactor Figure 5 illustrates a section through a module and illustrates the key features. Figure 5: Section through the SSR-W module The reactor core itself follows a conventional configuration of a fuel material contained within pins that are themselves managed as assemblies and mounted in a lattice configuration. The core is supported on a diagrid arrangement that, along with the primary heat exchangers and coolant pumps form the reactor internals. The reactor internals are supported in the reactor tank which effectively provides a pool of reactor coolant in which the reactor core, heat transfer equipment and supporting structures are all immersed. The coolant itself is a eutectic mix of sodium, potassium and zirconium fluorides. The zirconium used for the coolant salt is not intended to be highly purified and hence will still contain hafnium. The presence of hafnium in the secondary coolant is used to ensure that adequate neutron shielding to components and structures away from the core can be provided by ensuring sufficient depth of coolant only. There is hence no requirement for additional components in the reactor to function as neutron shields. The reactor coolant pumps are mounted at the top of the primary heat exchangers and draw coolant through the core and pump it through the shell side of primary heat exchangers. The secondary coolant is pumped counterflow through the tube side of the primary heat exchangers and takes heat away. There is an alternative flow path that is available to provide heat rejection through the reactor tank walls, as required during Emergency Heat Removal Mode. This is driven by natural convection and ensures adequate transfer of decay heat from the core to the walls of the reactor tank. 10

11 The external surfaces of the reactor tank, and the concrete lining of the reactor pit which forms the building foundations, are cooled by ambient air. This also provides the ultimate heat sink when decay heat removal is through the tank walls. The fuel itself is in the form of a eutectic mix of sodium, uranium and plutonium / other transuranic trichlorides and fission product lanthanide trichlorides. It is maintained in a liquid state throughout operation and periods of shut down. It is contained in vented pins in fuel assemblies. The reactor is built up from individual core modules mounted in a common reactor tank. Each module consists of a 10 x 10 array of fuel assemblies and is held within a structure that includes primary heat exchangers and primary coolant pumps. The layout is illustrated in Figure 6. The principle module parameters are: Core Power o 375 MW(th) o 150MW(e) Power density o 135 kw/l peak o 105 kw/l average Linear heat rate o 10 kw/m peak o 7.3 kw/m average Fuel Salt nominal temperatures o centreline peak = 1075 C o fuel tube inner surface min = 525 C o fuel tube inner surface max = 650 C o average fuel salt across all tubes= 760 C Dimensions Individual Module Core o 2.05m length across fuel assemblies o 2.05m width across fuel assemblies o 3.7m height of fuel (incl. support frames) o 1.6 m active height of fuel assemblies Dimensions Reactor Tank External for 300 MW(e) o 6 m length (along modules) o 5.3 m length (across modules) o 4.2 m overall height Fuel Support Rails New Fuel Charging Positions Reactor Coolant Pump Motor Irradiated Fuel Discharge Positions Primary Heat Exchanger Diagrid Fuel Assemblies Figure 6: Overview of Stable Salt Reactor Core Module 11

12 The SSR-W300 is power reactor capable of supporting 300 MW(e) continuous power generation. The reactor is made up a central standard module and two end modules. Figure 7 shows the plan view. Shutdown Blades are mounted so that they can be moved into and out of the reactor in the channels between the modules. The SSR-W300 has two such channels, each with two independent blade mechanisms, so four Shutdown Blades in total. Larger reactors simply have more standard modules. Shutdown Blade Location Shutdown Blade Location Figure 7: SSR-W300 Core Module Configuration Fuel Handling As described above, new fuel is charged into the side of the reactor, between the Primary Heat Exchangers and the core itself. Irradiated fuel is also discharged from the side of the reactor. Fuel is moved progressively across the core as it burns up from the charge point to the discharge point. Adjacent channels are fuelled in opposite directions using this process to aid control of core reactivity and neutron flux shape. Figure illustrates the fuel charging and discharging movements and also the way in which fuel is progressively moved across the core. A fuel shuffle is defined as the full set of operations required to remove a fuel assembly that has been burned up to the target irradiation to the discharge position, increment each assembly in the row to the next position and charge a new fuel assembly from the charge position into the vacant space in the core. Refuelling operations are carried out at power to minimise the excess reactivity available in the core. This allows the core to be operated without the use of neutron absorbing assemblies partially inserted in the core, or soluble neutron absorbers dissolved in the reactor coolant. The reactivity change resulting from individual fuel row shuffles can be accommodated with changes in bulk reactor coolant temperature. Figure 8: Refuelling of the SSR looking at a section through an SSR module 12

13 The core has been designed such that the negative temperature coefficient of reactivity associated with doppler broadening for the fuel material cross-sections, expansion of the fuel salt and expansion of the coolant salt surrounding the core are always much greater than the positive temperature coefficient of reactivity associated with coolant salt expansion within, and between, the fuel assemblies. All changes in reactor power can be achieved through control of temperature, itself effectively controlled by rate of heat removal. The control system for this, combined with the inherent characteristics of the reactor and the wide operating margins, minimise the need for shutdown action. The design range over which the reactor temperature coefficient can be used is sufficient to ensure that a full reactor shutdown can be accommodated within acceptable temperatures even if the shutdown blades fail to operate. Reactor coolant conditioning requires filtration, control of redox state and replacement of the hafnium as it is burned out. The redox state of the reactor coolant is maintained through the use of an electrochemical cell that is positioned in containment in contact with the reactor coolant salt, which maintains the relative concentrations of zirconium difluoride and zirconium tetrafluoride at the desired value. Filtration is provided on the Reactor Coolant Pump inlet and also on each Fuel Assembly. Normally isolated fill and drain lines are provided, to facilitate any other coolant processing operations that may be found to be necessary, and the fill line can be used to feed additional hafnium fluoride salt into the Reactor Coolant. The Reactor Coolant salt is maintained under an argon atmosphere to prevent absorption of oxygen, nitrogen or moisture. A Reactor Atmosphere Conditioning system is placed adjacent to the containment building and gas is drawn through the system for filtration, purification, and scrubbing of minor gaseous radwaste products including Cd, ZrCl 4, ZrF 4 and tritium by condensation means. Heat Transport Heat generated by nuclear fission in the fuel salt is transferred by convection and conduction to the fuel tube inner surface, then conducted through the fuel tube and transferred again by convection to the reactor coolant. The reactor coolant is a sodium potassium zirconium fluoride salt that is pumped through the core and through the shell side of the Primary Heat Exchangers before returning to the core. Fuel Salt Reactor Coolant Secondary Coolant Hot Tertiary Salt Tanks Tertiary Coolant To Steam Turbine Cold Tertiary Salt Tanks Feedwater Supply Primary Heat Exchanger Intermediate Heat Exchanger Figure 9: Overview of SSR Heat Transfer Loops Steam Raising Plant 13

14 Heat is transferred to the secondary coolant and then, through the Intermediate Heat Exchanger, to the tertiary coolant, as shown in Figure 9. The secondary coolant is a similar composition to the reactor coolant and the tertiary coolant is a sodium potassium nitrate salt, similar in composition to commercial solar salt used in the concentrated solar power industry. The tertiary coolant system has hot and cold storage tanks. These are used to provide a heat storage capability that decouples the operation of the reactor plant from the balance of plant and electricity generation. The tertiary coolant system is therefore considered in two halves: the tertiary coolant (heat removal) system covers the removal of heat from the reactor to the tertiary coolant storage tanks whilst the tertiary coolant (boiler heat supply) system covers the transfer of heat from the tertiary coolant storage tanks to the balance of plant and electricity generation systems. The tertiary coolant is at a higher pressure than the secondary and the secondary is at a higher pressure than the primary to avoid leaks going outwards. Nominal temperatures around the systems at nominal full power are: Reactor Coolant Salt peak temperatures o Core outlet = 650 C (max) o Core inlet = 525 C Secondary Coolant Salt peak temperatures o Primary Heat Exchanger inlet / Intermediate Heat Exchanger outlet = 450 C o Primary Heat Exchanger outlet / Intermediate Heat Exchanger inlet = 595 C Tertiary Coolant Salt temperatures o Cold Storage Tank = 270 C o Hot Storage Tank = 565 C The respective overall coolant flows are: Reactor Coolant = 5700 kg/s Secondary Coolant = 4925 kg/s Tertiary Coolant = 3125 kg/s The Tertiary Coolant Storage tank size can be selected with considerable freedom. There is a requirement to maintain a small 35 MWh th storage margin at all times for use by the heat transfer system to manage decay heat removal in the early stages after a reactor trip, but aside from this the size selection is a commercial decision for the facility operator depending upon the extent of flexible operation that it is desired to accommodate. Safety Systems The IAEA defines three top level principle safety functions that must be provided. These are: Control of Reactivity Control of Heat Removal Control of Containment Each of these safety functions is provided by both an inherent safety system, based on intrinsic properties of the reactor plant systems, and a diverse safety system, based on engineered safety features. The diverse safety systems, however, are not dedicated safety systems and are effectively a safety role carried out by plant systems that are required for normal operations. The claims, and consequent qualification, against these roles reflects this philosophy. Control of Reactivity The inherent safety system for control of reactivity is based on the core physics, particularly the net reactivity temperature coefficient. The power reactivity coefficient is such that the reactor will always 14

15 be in a shutdown state when it is isothermal at a temperature around 800 o C. The contributing factors of the temperature coefficient are: Fuel temperature coefficient is large and negative due to doppler broadening and physical expansion of the fuel salt. Coolant temperature coefficient is small and positive due to the competing effects of reduction of moderation and parasitic absorption as it expands in the centre of the core (positive) and increase in neutron leakage, resulting from reduced reflection, as it expands around the edge of, and external to, the core. Diagrid expansion coefficient is negative due to increased leakage as the fuel assemblies move apart. This provides intrinsic protection against all at power overpower faults. The long term hold down capability is assured through the removal of some fuel assemblies from the edge of the core back into the fuel charging or discharging positions. This is a feature of SSR-W, thanks to the unpressurised online refuelling process. The diverse safety system for control of reactivity consists of four fast-action boron shutdown blades. Control of Heat Removal The inherent safety system for control of heat removal is based on passive heat removal through the Reactor Pit Cooling System operating in Emergency Heat Removal System (EHRS) mode. Heat transfer from the reactor tank walls to the atmospheric air in the cooling ductwork will remove heat by convection to the atmosphere. The decay heat loading profile, for a steady-state core results in it reaching a peak coolant temperature below 900 C, if emergency heat removal is required immediately post trip and there is total failure of heat removal through the secondary and tertiary coolant systems. This is through a combination of thermal inertia of the reactor coolant and establishment of stable operation of the EHRS. Since the reactor core is reliably sub-critical at this temperature, the fuel peak temperature is commensurately near-isothermal with the bulk coolant and the reactor is firmly shutdown even if the control blades have failed to operate. This peak temperature occurs after some 13-14h but the isothermal temperature remains in excess of 800 C for at least 35h post trip. After that time, if the control blades still fail to operate, the reactor is maintained subcritical by removal of a number of fuel assemblies to their storage location inside the reactor tank. The diverse safety system for control of heat removal makes use of the normal at power heat transfer system. The Reactor Coolant System operates in a natural circulation mode and the Secondary and Tertiary Coolant Systems operate in a low flow pumped circulation mode to remove heat to the Tertiary Coolant Heat Store. Control of Containment The inherent safety system for containment is based on the chemical composition of both the fuel and coolant salts and their very high affinity for volatile radioactive fission products. Both the fuel and the coolant salt will provide a very high retention capability for both caesium and iodine isotopes. This is only effectively breached as the salt approaches boiling point for which there is no conceivable mechanism. Only the noble gas fission products and cadmium are not effectively retained by either fuel or coolant salt. These are dealt with by the containment argon treatment system. The selection of low-tritium yield coolant salt greatly reduces the need for tritium management which, if necessary, is also carried out by the argon treatment system. The diverse safety system for volatile radioactivity containment makes use of a steel shell reactor containment structure. This is designed to be sealed, but it is not designed to withstand significant internal pressure as there is no postulated initiating event that is able to lead to containment 15

16 pressurisation. Access into and out of the containment is provided by a single equipment airlock structure. The containment structure is a thin walled metal construction to provide a gas tight building, but one where large-scale heat transfer is possible through the structure. The containment itself is surrounded by a thick concrete Containment Shield that provides the biological shielding and missile defence functions that are often associated with traditional reactor structures. The containment of the non-volatile radiological source term in the event of a severe accident is via a series of physical and chemical containment barriers comprising the fuel tubes the coolant salt which is miscible with the fuel salt and would radically dilute any released fuel salt and thereby limit the potential for temperature rise by increasing thermal inertia by a factor ~20. Additionally, dilution lowers vapour pressure of key hazardous isotopes. the reactor tank the steel lining of the concrete pit the concrete pit a frozen layer of salt that would form between the molten salt and cooler structures Instrumentation and Control The Instrumentation and Control systems on the SSR are split into three: The Reactor Protection System The Station Control System The Station Monitoring System The Reactor Protection System provides the detection and actuation logic to initiate the operation of the diverse protection systems which support the provision of the high level safety functions. The system monitors plant parameters and provides outputs to the switchgear for operation of the reactor shutdown blades and both primary and secondary coolant pumps. The logic functions are provided by high integrity hard wired electronic logic. An Innovate UK project is underway with the University of Bristol and Altran to assess the technical viability of wireless I&C systems to reduce costs and increase reliability further. The Station Control Systems provides integrated control of all the power station control systems on site. The systems are divided between three groups. The Reactor Control System covers the secondary / tertiary coolant pump speed control systems, primary coolant chemistry control and containment atmosphere control along with the sequencing control for the reactor fuel handling operations. The Balance of Plant Control System covers the boiler controls, main turbine governor and feed system. The Ancillary Services Control System covers plant room HVAC, fire protection equipment, the water treatment plant, etc. The Station Control Systems are provided by a standard industrial process controller network system. The Station Monitoring System is provided by an industry standard data acquisition and display system network. An overview of the Station Monitoring system is shown adjacent. Finally, site radiation monitoring, fire protection monitoring and status of all ancillary plant is provided through inputs to the Station Monitoring System from their respective control systems. The primary displays are presented in the central control room in real time. The data logger collects and stores information so that it is recorded for further use, if required. 16

17 Balance of Plant The balance of plant is shown in outline in Figure 10. Tertiary salt is drawn from the storage tank, by the boiler salt pumps at a rate to match the required steam generation demand. The boiler units are in four parts, economiser, evaporator, superheater and reheater, and there is an individual boiler unit for each steam turbine. Tertiary Coolant salt enters the boiler units at 565 o C and leaves at 270 o C. Steam generation provides steam to the main turbogenerator at conditions similar to traditional fossil fired turbines: 160 bar, 538 o C, main steam and 40 bar, also 538 o C, reheat. The tertiary nitrate salt system ensures that the nuclear island is effectively separated from the turbine plant, and constitutes a decoupled activity which does not affect the provision of key nuclear safety functions accordingly, any process heat or power conversion cycle can be retrofitted to best-suit the end user, without affecting the licenced activity. The low cost and ease of storage of nitrate salts permits easy provision of a heat store to further decouple operation of the reactor and turbine plant. The output from the SSR-W would be exactly matched to a 300 MWe steam turbine but the availability of the heat store means than it is possible to size the turbine independently and additionally to install more than one machine if needed to facilitate flexible operation. Steam is exhausted from the turbine to an air-cooled condenser and condensate returned via the condensate and feed systems to the boilers. Feedwater to the boilers is provided by a steam driven main feed pump that is integrated into the main turbine steam cycle. A separate Station Auxiliary Generator, operated in parallel with the main feed pump, provides power for the site electrical systems. This arrangement has been chosen to enable main turbine operation to be considered entirely separately from the reactor. Figure 10: Overview of the Balance of Plant 17

18 6. Moltex Energy Technology Portfolio & Fuel Cycle The understanding that static molten salt fuel in conventional assemblies is viable opens up an entire portfolio of reactor configurations. Given the initial model, the SSR-W as described above, is fuelled from converted spent oxide fuel of which there is a finite quantity, Moltex has developed to conceptual level two further reactor versions which have near limitless fuel. These will be deployed once the first model has been rolled out successfully. The various reactor configurations are: SSR-W : The Wasteburner fuelled by plutonium from nuclear waste SSR-U : The Uranium burner, thermal spectrum fuelled by low enriched uranium with graphite moderator built into the fuel assemblies SSR-Th : Either of the above with a thorium fluoride based coolant salt breeding uranium- 233 in the coolant to achieve higher breeding or conversion ratios These reactors have the same fundamental engineering design for the plant and modules. This will reduce the cost and time for deployment of these reactors. The differences are principally the fuel assemblies and fuel and coolant chemistries. The SSR-W has been progressed first as it has the biggest margins in the design and minimal need for research. Associated with these reactors are several other technologies being developed by Moltex Energy that also have potentially enormous values: WATSS - Moltex has designed and patented a radically simplified process to convert spent oxide fuel to chloride form suitable for a fast spectrum MSR. This is known as WATSS (WAste To Stable Salts). Heat Battery - A by-product of the WATSS process is a calcium chloride salt that contains very high concentrations of the heat producing fission products. This can generate high temperature heat in the kilowatt to hundreds of kilowatt range with no maintenance or refuelling for several years. This has many high value applications such as distributable power for remote communities who are reliant on fossil fuel generation, to cost-effectively decarbonise their energy requirements. GridReserve The concept of storing the thermal energy from the reactor and dispatching it onto the grid only when needed. This is not possible with PWR operating temperatures. These technologies are discussed further in the following sections. Moltex s fuel conversion technology (WATSS) Spent fuel from the 449 operational reactors in the world is currently managed in one of the following ways: 1. Held in temporary storage there are several methods by which this can be done. The most common is to store the fuel rods in deep water pools. Water acts as a barrier for radioactivity, safeguarding the surrounding atmosphere in case of any escape of radioactive material from the fuel rods. Several of these storage facilities, especially in the UK, have aged and are proving difficult and costly to manage. Sellafield in the UK is an example, with the UK s NDA spending 3bn p.a. for the management of the site. 2. Reprocessed into MOX fuel spent oxide fuel can be reprocessed into mixed oxide (MOX) fuel, which contains one or more oxide of a fissile element, e.g. plutonium blended with 18

19 uranium. This can be used in light water reactors as an alternative to low-enriched uranium (the conventional fuel), without needing to upgrade the reactor. There have been significant problems with reprocessing sites globally, however, with several suffering radioactive leaks and questions being asked about the risks to non-proliferation posed by reprocessing spent oxide fuel into high purity plutonium. 3. Placed in deep geological repositories similar to the storage options described in the first bullet, geological repositories differ in that they are purpose built to encase the nuclear waste for a much longer timeframe. The waste is stored in an excavated mine-like system of caverns, with subterranean depths ranging from just 50m to 1000m. None of these options represents an optimal spent fuel management approach for economic and environmental reasons. It is widely agreed that an optimal fuel management approach would recycle the long lived and highly toxic radioactive higher actinides into fresh nuclear fuel leaving a much shorter lived radiological hazard that could be managed at far lower cost. With that in mind, and in keeping with the environmentally responsible vision of the company, Moltex has designed the SSR so that the fuel can be generated from existing spent nuclear fuel stocks. The process will be carried out by a fuel conversion facility, which can be built and operated by the customer. The process is relatively simple and adds a low incremental amount to the LCOE/operating costs. It is anticipated in fact that the economic benefits shall significantly outweigh this incremental cost, as waste owners will pay a proportion, if not all, of the costs saved by no longer having to manage spent fuel. As such it is calculated that the net effect of operating a fuel fabrication facility alongside a power plant will be to increase profitability. The ability to simply and cheaply convert spent oxide fuel into SSR fuel is made possible by the fact that the SSR can be powered by highly impure plutonium - the fuel salt mixture can contain nearly as much lanthanide fission products as it does plutonium. The numerous complex steps required for conventional reprocessing to produce pure plutonium are simply not required. This reduces the risk of proliferation substantially. WATSS is a complete system that takes used fuel bundles, recovers fissile, fissionable and fertile materials and turns them into a fuel for the SSR-W. WATSS is a pyroprocessing technique and so is compact and doesn t have the problems of effluent management incumbent in the aqueous dissolution based techniques. Where the UK's Thermal Oxide Reprocessing Plant (THORP) has the footprint of a football pitch, WATSS would be the size of a billiard table. WATSS can be used to reprocess the fuel from any conventional reactor but the initial focus has been used CANDU fuel. A unit 2m 2 by 2m 2 would be capable of processing 170/t/yr of spent pellets to produce the start-up fuel for an SSR-W in less than 3 years. Two hundred fuel assemblies make up an SSR-W300 reactor core. One new fuel assembly is added, on average, about once a week. WATSS takes a CANDU Fuel Bundle that contains pellets of used nuclear fuel and treats them in a three-stage process: 1. Removal of the zirconium cladding from the used fuel (CANDU bundle decladding) the zirconium is then reused in the SSR coolant as ZrF Electro-reducing the fuel pellets to form a molten uranium alloy. All higher actinides and most lanthanides go into the uranium alloy along with the noble metal fission products. The major radioactive fission products such as caesium, strontium and iodine stay in the electrolyte salt. The process is illustrated in figure

20 Figure 8: Illustration of electroreduction cell 3. Contacting uranium alloy (containing higher actinides and lanthanides) with a clean sodium chloride/iron chloride salt to make the fuel salt through exchange of the iron chloride with plutonium and higher actinides/lanthanides in the alloy. The remaining uranium alloy is then sufficiently free of long lived actinides to be disposed of as ILW or (more likely) kept for future breeding use. Figure 12 shows the various separations in molar percent at each step. All of the long-lived actinides causing the spent fuel to be high level waste goes into the new Fuel Salt which is valuable in an SSR. The bulk of the mass stays as the uranium alloy, now Intermediate Level Waste. Figure 12: Sankey diagram showing the various constituents of spent fuel as they go through the various stages. All numbers are in mol%. 20

21 SSR-W Reprocessing The intention for the SSR is that the spent fuel produced will be reprocessed using a similar process to WATSS to reuse the remaining energy. It also substantially reduces the volume of disposal required. The process will be simpler than WATSS as the fuel salt is already in salt form and there are no oxide bonds to break. Heat Battery One of the waste products discussed above is the electrolyte salt. The salt can hold extremely high quantities of fission products by weight. As the fission products are the main heat producing elements in the medium term (primarily caesium and strontium) the salt is a valuable source of high grade heat that does not require maintenance and can be easily transported to its point of use. The quantity of heat available is in the kilowatt to hundreds of kilowatts range and can be extremely valuable for certain applications. These include remote military sites, space exploration, remote communities, hospitals and mines. Conventional fission product heat is too dilute to be usable. It is only now that it is in concentrated form that it becomes usable and valuable. GridReserve - Hybrid Nuclear & Renewables Need for flexibility One of the key criteria for an attractive modern energy generation method is flexibility. Historically, power generation was either baseload, operating continually, or intermittent, operating only at peak demand times. The advent of renewable energy sources such as wind and solar have introduced a new challenge, intermittent generation with zero marginal cost but which cannot be relied on to generate at times of peak demand. This has severely undermined the conventional baseload market since zero marginal cost renewables can on occasion provide all the power needed by the electricity system. Nuclear energy has been almost universally used as baseload power. While power ramping of reactors is technically possible, it rarely makes economic sense since the reduction of operating costs by reducing power output is minimal. This leaves electricity systems dependent on high carbon fossil fuel plants to provide the flexibility needed to compensate for the intermittency of renewables. As renewable power continues to grow its share of the world s energy generation, this need for additional flexible generation that can counter the intermittent nature of the sun and the wind will continue to grow Governments and energy industries have to-date struggled to manage this problem. They have created capacity and frequency response markets, whereby electricity generators are paid either a) for the availability of power that is to say it may not necessarily be used or b) for a rapid response to a rise or fall in demand i.e. by turning off or on their asset. The latter is possible with gas fired power stations and diesel generators but low carbon forms of generation struggle to respond at an appropriate speed. In the absence of a low carbon power source capable of flexible power generation, this approach puts a hard limit on the degree to which electrical generation can be made carbon free. In order for low carbon forms of generation to have the desired flexibility, they need to incorporate a storage element, so the baseload nature of the plant can continue but without necessarily exporting to the national grid constantly. Technology to enable economically viable grid scale storage of electrical energy has yet to be developed however. Batteries cost $100 s per kwhr capacity and even the most ambitious projections see 100 per kwhr as a holy grail target. The SSR can use a thermal 21

22 energy storage system that is already deployed in concentrated solar power stations at a capital cost below 100 per kwhr (Including the cost of the necessary increased steam turbine capacity). GridReserve technology With this in mind, and keen to meet the attractive criteria that would catapult the SSR to the forefront of the global generation market, Moltex have incorporated their GridReserve technology. GridReserve is a form of thermal energy storage. It employs the molten salt technology developed for the reactor itself, storing energy in the form of heat within a molten salt compound of 60% Sodium Nitrate (NaNO 3)/40% Potassium Nitrate (KNO 3). The molten salt mixture has a melting point of 220 o C and a maximum operational temperature of 600 o C. It is heated to 550 o C in the GridReserve facility, and is easily able to retain this heat energy for the period required i.e hours. Figure 13: Typical thermal salt storage tank from a Concentrated Solar Power plant. The GridReserve asset will comprise a plurality of insulated tanks containing the molten salt liquid. For convenience it will be located close to the SSR power station; the shorter the connection line the lower the line losses. GridReserve is envisaged to cater for several hour s generation from the SSR. For a 1GW plant, the storage will therefore be 1GWh for each hour of storage capacity, enabling the plant to export double or triple capacity at times of peak demand/frequency response requirement. To utilise this increased capacity, the SSR power plant will require additional turbine and generator capacity. This represents the majority of the incremental cost as a result of GridReserve, in addition to the cost of additional tank installation. However the incremental benefits are expected to be material and easily outweigh the cost. Each conventional-off-the-shelf GridReserve tank pair contributes ~200MWhe storage capacity. An example 2GWe, 50% load factor plant (with diurnal power variation) can consist of one SSR-W1000 installation, coupled to a 2GWe-rated power conversion island, with ~8 hectares of GridReserve installation. The energy conversion system can be a hybrid of multiple process heat requirements, so that a single installation can service multiple end-user markets, and can be simply retrofitted without relicensing, to service a rapidly-evolving global energy market. Figure 14: GridReserve to support intermittent renewables. 22

23 7. Detailed Technology Descriptions and Justifications (SSR-W) In this section, a number of key design decisions taken during development of the conceptual design of the SSR-W are set out with the reasoning behind those decisions. Fast reactor to avoid xenon and graphite problems A key early decision was to focus first deployment of the SSR on a fast spectrum reactor rather than a thermal spectrum reactor. Two key analyses underpinned this decision. First was the substantial challenge of accurately predicting the behaviour of the Xe135, which is a very potent neutron poison in any thermal spectrum molten salt reactor. It is known that even in the MSRE (where it was expected that helium purging would rapidly remove the xenon) unexplained and substantial reactivity transients were experienced that were ascribed to lack of proper understanding of the gas behaviour. While far from insoluble, this challenge was seen as one that would require substantial research effort to overcome and which would therefore delay commercial deployment of the reactor. This effect is insignificant in fast spectrum reactors. Second was the chemical challenge represented by using graphite as moderator. There was seen to be no practical alternative to graphite, since zirconium hydride was rejected, due to its violent reactivity with molten salts and beryllium due to its substantial tritium production. Graphite has been shown to react with molten salts which are strongly reducing, rendering them more oxidising. It was considered therefore that having graphite in contact with the molten salts would compromise efforts to reduce metal corrosion by those salts. Again, this problem may not be insoluble, but managing the fuel salt chemistry to find the narrow optimum region between graphite stability and metal corrosion was seen as a challenging and time consuming research program which would require in reactor tests since both graphite reaction and metal corrosion would be expected to be affected by radiation fields. Chloride fuel salt to allow highly reducing redox state and hence standard steel use The choice of chloride salts for the fuel was not driven by neutronic concerns. Chloride salts do permit a somewhat harder neutron spectrum and hence better neutron economy than fluoride salts but that was not seen as a compelling argument for a reactor where the overarching strategy was to enable rapid deployment. The first reason to choose chloride salts was that such salts allow high concentrations of actinides to be used at temperatures low enough to allow use of standard steels. With fluoride salts, the achievement of sufficiently low Figure 15: Chromium concentration in the fuel salt with and without Zr redox stabilisation temperatures usually requires use of lithium or beryllium salts. Both of those elements, even if isotopically enriched, generate large amounts of tritium in reactors orders of magnitude more than the unavoidable tritium from rare ternary fission events. Since tritium readily penetrates metals in molten salts the use of tritium generating base salts would require development of an effective tritium scavenging system. That technological challenge would inevitably delay reactor deployment. 23

24 The second reason to choose chlorides is because uranium trichloride is substantially more stable to disproportionation than is uranium trifluoride. This allows chloride salts to be rendered so strongly reducing, through contact with zirconium metal, that corrosion of metals is virtually completely thermodynamically prevented. That cannot be achieved with fluoride salts without causing uranium metal deposition. Fluoride salt use therefore drives reactor designers to require special high nickel alloys which have no recent nuclear experience of use. This represents a substantial technical barrier to rapid deployment. Moltex has a granted international patent on the use of chemistry control with molten salts in this manner. Fuel assemblies and ferritic steel using existing neutron data The entire fuel assembly is manufactured from 12%Cr-1%Mo ferritic/martensitic steel. The HT9 steel used in the Fast Flux Test Facility at Hanford, USA, has been down selected as most viable material for immediate deployment, based on thermodynamic potential of not only Cr and Fe, but also the carbide phases important to irradiation creep performance in HT9. As such, the ready-established code cases and irradiated experimental data for HT9 in sodium supplies practical qualification experience for SSR- W. Moltex are working with US National Labs and DOE experts to provide HT9 qualification evidence based on established manufacturable materials technology, to expedite deployment. The fuel tube is illustrated in Figure 16. The fuel pin tubes are 1900 mm long with a 10mm circular cross-section. The tube wall thickness is 0.3 mm. The fuel tube dimensions have been selected following an optimisation process between mechanical properties and heat transfer capability. The fuel assemblies are formed of 18 x 18 fuel pins. The fuel tubes are intended to operate at a pressure that is in equilibrium with the bulk reactor coolant pressure. The fission products that are released as gases (primarily krypton and xenon, cadmium and zirconium tetrachloride) rise to the top of the fuel tubes and are collected in the plenum area above the fuel salt. A vent at the top of this plenum leads to a second gas hold up plenum chamber. This functions in a similar fashion to a diving bell, as illustrated in Figure 16, with the gas hold up plenum being initially filled Fuel pins Support leg Figure 17: Fuel assembly wrapper and shroud. Figure 16: Fuel pin and gas vent Wrapper shroud with reactor coolant that is slowly displaced as the volatile fission products accumulate. Once the gas volume is sufficient, gas will escape through the vent ports into the reactor coolant. The seal chamber ensures that the gas hold up plenum remains sealed when the fuel is discharged from the reactor. The diving bell effectively provides a hold up point, to allow for decay of short lived isotopes, before these are released into the main reactor coolant. The fuel pins are held in a square array to form fuel assemblies with top and bottom tie grids along with additional spacers to prevent fouling of fuel pins. The assembly suspends from a support leg and 24

25 anchor head which mates with the module support structure above the surface of the coolant. A wrapper duct surrounds the pins, similar to a Boiling Water reactor assembly. The wrapper duct and fuel pins can be seen in Figure 17 as two separate components before they are fastened together after fuelling. With this design, and the use of the hafnium containing coolant, significant neutron damage is limited to the materials of the fuel assembly. No permanent structure in the reactor experiences sufficient neutron damage to limit its operational life. This eliminates one of the key problems that has limited the life of sodium fast reactors and would be expected to limit the life of pumped fuel salt fast spectrum molten salt reactors. The fuel assembly is currently designed to withstand the thermal and radiation creep that would be anticipated in the reactor operating up to a burn-up limit of 200 GWd/tHM. This is approximately equivalent to 200dpa according to our modelling using state of the art MCNP6 simulations. Irradiation data exists from sodium fast reactors for the various suitable steels up to 150dpa at our operating temperatures of 650C. 150dpa is therefore used as our design limit which leaves substantial margin for uncertainty in thermal and radiation creep behaviour. Some experimental results from previous operating reactors are shown in the table below. Candidate Material Fast Reactor Commercial Applications Fast Reactor Experimental Applications %Void %Diametral Change 650C HT9 EBR II, USA FFTF, PFR MPa 15-15Ti/12R72 Phenix/SuperPhenix, FR ASTRID, MYRRHA (@75GWd/t) 140MPa PE16 Nimonic Dounreay DFR & PFR, UK DFR MPa Creep rupture stress in HT9 at elevated temperature of 650C, over a period of 50000h, is characterised in irradiated environments due to extensive prior application, with a practical rupture limit of 50MPa. The peak stresses in the SSR fuel assembly are <30MPa as can be seen in Figure 18 Figure 18: Peak stresses in the fuel assembly wrapper under normal operating conditions. Moltex has carried out extensive studies to ensure that existing validated steels can be used in the SSR. This summary of results makes it clear that this is possible with substantial margin. 25

26 Neutronics analyses Preliminary analysis of the SSR-W300 has been carried out using the MCNP6 Monte Carlo software together with the ENDF/B-VII.0 cross section library. To perform these analyses, a full-core model of a start-up core including 200 fuel assemblies, coolant salt, heat exchangers, shutdown rods, supporting diagrid, gas plenum, and core surroundings has been implemented. The development of this model followed an iterative process considering engineering, neutronics, and thermal-hydraulics requirements, and aimed at optimising the margins for fuel shuffling, heat transfer, and neutron economy A number of key parameters of the SSR-W300 core have been provisionally quantified. Some of these numbers are reproduced in the table below. Fuel temperature coefficient (pcm/k) -8.8 Coolant temperature coefficient (pcm/k) +1.9 Diagrid expansion coefficient (pcm/k) -0.9 Reactivity worth of all shutdown blades (pcm) Effective delayed neutron fraction (pcm) 242 These calculations confirm the inherently safe behaviour of the SSR under power excursions. Although the coolant temperature coefficient is positive, the strongly negative fuel temperature coefficient will dominate transients and ensure that reactor power quickly returns to safe levels. Furthermore, the results clearly support a control mechanism based on the average fuel temperature. As can be seen, small perturbations to the coolant temperature, and thus to the fuel temperature, can be used to effectively control the power output of the reactor without the need for control rods. Additional neutronics work is currently advancing in different fields. In particular, the characterisation of the fuel evolution of the reactor is underway, in addition to neutron activation and neutron-induced damage to reactor internals. At the same time, system-scale analyses of transient reactor behaviour with input from both neutronics and thermal-hydraulics are under preparation and planned to start in the near future. ZrF4 coolant to enable soluble ZrF2 redox couple and standard steel use The SSR is unusual in using different halide salts for fuel (chloride) and coolant (fluoride). The choice of chloride for the fuel salt has been explained above. The absence of actinides in the coolant however made fluorides the preferred choice due to their low neutron capture rate and benign activation products. Zirconium tetrafluoride was selected as the base salt due to its low melting point (385 C) in eutectic mixture with NaF and KF. Figure 19: Chromium concentration in the coolant salt with and without 2%ZrF 2 redox stabilisation. A second reason to use a zirconium fluoride as the base salt was its chemistry. Zirconium has an exceptionally high affinity for oxygen and acts as a sink for any traces of oxygen or oxides in the coolant salt. That aids prevention of corrosion. More critically however, zirconium has two stable fluorides, ZrF 4 and ZrF 2. Even very small amounts of ZrF 2 render the salts almost as non corrosive as would the presence of metallic zirconium (as used in the fuel salt) but does so with a slightly soluble compound that provides redox buffering to the 26

27 whole salt volume without the tendency of zirconium metal to migrate from its point of contact with the salt to other locations such as pumps where it could deposit and create hazards. Moltex has robust patents on the control of salt chemistry using sacrificial metals. Experience with the Aircraft Reactor Experiment led many researchers to conclude that zirconium fluoride salts were unusable due to formation of ZrF 4 snow. In that reactor however, the level of NaF was too low to fully complex the zirconium fluoride. Vapour pressures where ZrF 4 is less than 42mol% are however very low due to complexation of the ZrF 4 in M xzrf (4+x) species. It is expected, but will be experimentally confirmed, that the small amounts of coolant salt vapour will condense as liquids dominated by the M xzrf (4+x) species instead of subliming solids and can hence be readily returned to the tank. Coolant activation/contamination, electrochemical oxygen and nitrogen removal with filtration (all in tank) Over the >60 year life of the reactor, it is inevitable that the coolant salt will change in composition and therefore require some form of chemical processing. The major changes will be ingress of air (oxygen and nitrogen) and addition of oxides dissolved from the oxide layers on the steel fuel assemblies. In addition, there will be a series of transmutations - hafnium to tantalum, sodium to magnesium, fluorine to neon and zirconium to niobium. These challenges are managed via a variety of approaches as follows. Oxygen, nitrogen and oxides. These form compounds with the zirconium salts, primarily oxyfluorides and nitrides. Provided the redox state of the salt is maintained at a strongly reducing level with the presence of ZrF 2 they do not create a corrosion challenge. However, molecular dynamics simulations suggest that oxygen can form bridged compounds with a resulting increase in salt viscosity. Their level must therefore be controlled. This is done via an electrochemical apparatus, immersed in the tank of coolant salt in a low temperature region. The apparatus electrolyses the salt depositing zirconium metal on the cathode and generating oxygen and nitrogen at the anode, from which the gas is swept by a stream of argon. The deposited zirconium, in contact with the coolest salt in the system, maintains the minimum level of ZrF 2 in the coolant salt and hence the redox potential. Tantalum and niobium produced by activation of zirconium and hafnium are metals at the redox state of the coolant salt and will deposit on other metal surfaces. Metal wool filters on the electrolysis apparatus and pump inlets capture those metals, along with any particulates formed. These filters (or in the case of the electrolysis apparatus the entire apparatus) are designed for simple replacement using the fuel assembly lifting apparatus. Conversion of fluorine to neon and sodium to magnesium results in change in redox state of the coolant which is simply managed by the redox control system. Small amounts of MgF 2 in the salt have minimal effect on its thermochemical properties and are therefore tolerated. The only chemical addition expected to be required over the course of the reactor life is hafnium tetrafluoride to maintain the neutron screening power of the coolant. This will be added via fixed, normally capped, stand pipes in the reactor service room when the reactor neutron detectors indicate that neutron fluxes at a distance from the core are rising above design levels. Heat exchanger and pump choice and design 27

28 The primary heat exchanger system consists of a nested pump and counter current single-pass heat exchanger bank. The system is designed to be a deployable unit, which can be modularised and replaced expediently by extracting vertically, without incurring substantial reactor downtime. A key design criterion for the heat exchanger system is to eliminate high pressure failure modes, ensuring containment of activated coolant in the event of a failure by operating the secondary coolant at higher pressure. This ensures the system is compact and lightweight, which results in a Reactor tank which can also be volumetrically-efficient, enabling off-site fabrication and reduced cost. The primary coolant pump is a triple-blade, cantilevered axial flow pump. A filter on the pump inlet protects the heat exchange system from damage due to insoluble debris. The pump operates at low power, <50kWe, and low angular velocity, ~1800rpm, with a pressure head of 100kPa. The pump drive system operates in the upper Argon containment zone, which is actively cooled and treated. The pumps are not actively controlled for power or refuel, and are run continuously during operation, with trip states only as part of the reactor protection logic. Each SSR module consists of 4 primary heat exchanger and coolant pump units. This provides redundancy, and permits conventional industry standard components and manufacturing processes to be employed. This radical simplification of primary reactor coolant systems enables greater tolerance to loss of primary coolant pump failures. Passive Emergency Heat Removal System, natural convection, peak temperatures In the event of a LOSCA (Loss of Secondary Coolant Accident), the SSR-W can manage residual decay heat build-up through an auxiliary cooling system. The emergency heat removal system (EHRS) consists of a passive bypass loop in the primary reactor system. This enables primary coolant to bypass the primary heat exchangers, and transfer heat directly to the tank wall instead. The relatively thin-walled reactor tank rapidly increases in temperature, which elevates the temperature of the outer surface. The Reactor Tank is cooled by a jacket of cool ambient air, brought to the lower tank surface by a downcomer duct. The effectiveness of this solution is increased by featuring of the riser duct, with thin fins spaced regularly, to provide a larger collection area for radiative heat transfer and subsequent convective transfer to the flowing air. This innovation greatly-increases the effectiveness of passive tank cooling systems, to enable a far greater power density, or greater fuel burn-up and hence decay heat loading, than contemporary reactor designs. As illustrated below, preliminary concept validation has shown that the primary coolant reaches approximately 860 C in a LOSCA event where all power is lost. This is a commercial design requirement so that the reactor tank can stay in service after a single <900 C Figure 20: Time (seconds) - temperature ( C) graph of the primary coolant in a LOSCA event. Figure 21: Peak temperature gradients across the EHRS, tank wall and coolant in a LOSCA event. 28