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The ReCore-Reactor is the nuclear-island part of an integrated engineering concept. The reference unit is sized around 720 MWth, ~300 MWe class when coupled to ReCore-Power at a ~45% net cycle target.
The design is deliberately conservative in materials and physics but more ambitious in plant arrangement: modular multi-loop layout, metallic fuel suited to recycle, heat rejection to nitrogen rather than water. For the broader rationale — supply chains, construction, and what is still open — see Why?.
TRL figures below are estimates for building blocks (NASA/EU-style 1–9 scale), not regulator-certified scores.
| Technology | Est. TRL | Notes / heritage |
|---|---|---|
| SFR sodium primary systems | 8–9 | Decades of operation and R&D: EBR-II (US), Phénix / Superphénix (France), BN-600 / BN-800 (Russia); industrial Na chemistry and loop practice |
| Metallic U-Pu-Zr / U-TRU-Zr fuel | 7–8 | EBR-II / IFR metal-fuel campaigns; demonstrated high conductivity and passive feedback behaviour |
| HT9 ferritic–martensitic cladding | 7–8 | Proven SFR metal-fuel cladding path (e.g. FFTF / EBR-II-class experience) |
| ALIP / EM sodium pumps | 7–8 | Used and developed for liquid-metal service; no shaft seals into primary sodium |
| 316LN vessel and primary piping | 8–9 | Standard nuclear/process austenitic steel; plate, pipe, and weld practice at ~600 °C class |
| Na→N₂ primary HX in an SFR plant | 4–5 | Single smooth tube, shrouded bayonet, one top tubesheet. Screening reference is Alloy 709; Alloy 800H is the commercial alternative and is not yet the baseline. Nuclear Na–N₂ coupling at this scale is FOAK engineering |
| Modular prefabricated vessel sectors | 5–7 | Prefabrication and high-integrity welding (e.g. EBW/LVEB candidates) exist industrially; nuclear modular SFR assembly is project-specific |
| Integrated modular SFR FOAK | 4–5 | Full plant integration, licensing, and demonstration remain open work |
Many historical SFRs used an intermediate sodium loop and water/steam generators — partly for sodium–water chemistry risk, partly for radioactivity isolation, thermal buffering, and other plant reasons. Intermediate loops also add inventory, cost, and leak paths.
This concept explores:
| Parameter | Target |
|---|---|
| Reactor type | Modular sodium fast reactor |
| Thermal power | 720 MWth |
| Electrical class | ~300 MWe (docs pin; screening ~316 MWe, under the 45% target) |
| Net cycle efficiency (target) | ~45% (recuperated, intercooled N₂ Brayton) |
| Net cycle efficiency (screening) | ~43.9% (current screening, under the 45% target; not a FOAK guarantee) |
| Primary pressure | near atmospheric (~1 bar) |
| Hot sodium (order of) | ~600 °C |
| Fuel | metallic U-(TRU)-Zr; U-Pu-Zr bridge cores |
| Breeding intent | near iso-breeder (conversion ratio ~1) |
| Primary Na inventory | order of 70 m³ (analytical ~73.1 m³; HX shells + vessel dominate; compact vs large pool SFRs, not “below 50”) |
| Primary heater | Vertical shrouded bayonet-tube Na→N₂; 6 × 120 MWth modules; smooth Alloy 709 tube, ~4.2 m active, single top tubesheet in the argon plenum |
Metallic fuel supports:
Iso-breeder operation is a design goal: produce roughly as much fissile material as is consumed, so equilibrium cores are less dependent on continuous enrichment. The screening model does not calculate a conversion ratio. That ratio is a result of the final core, not a property of the flowsheet.
The first cores are metallic U-Pu-Zr from separated civilian plutonium and depleted uranium, not HALEU. HALEU supply is thin. Separated civilian plutonium and depleted uranium already exist. That choice sets both the physics and the security. The delayed-neutron fraction of Pu-239 is about a third of that of U-235, so the gap between delayed-critical and prompt-critical is narrower than in a uranium thermal core. Doppler broadening, axial expansion of the metal, and a fast shutdown system are the response. A Curie-point latch, of the kind studied as a self-actuated shutdown on some SFR programmes, is a candidate, not installed hardware. See Physics.
Fresh fuel that contains separated plutonium is Category I material. The guard force and the accountancy do not shrink with the electrical rating. Putting the plant inside an existing Category I site is a siting hypothesis, not a feature of the vessel. Irradiated fuel, or fuel that still carries fission products, can be self-protecting by dose rate. Clean cast U-Pu-Zr is not. Making that first core is a glovebox or hot-cell plant. The pyro hub is the later multi-recycle plant. It is not the prerequisite for the first core. See ReCore-Processing and Fuel Science.
Primary sodium is intended to transfer heat to nitrogen in six vertical shrouded bayonet-tube modules, one 120 MWth module on each starburst loop. ALIP remains on the cold leg. Because nitrogen is chemically inert toward sodium, a circulating intermediate sodium loop is not the FOAK baseline.
Sodium is on the shell: inlet ~600 °C, outlet ~400 °C, pressure ~1 bar. Nitrogen at ~80 bar flows coaxially inside the tube. Cold gas (~328 °C) descends in a removable cartridge, turns under a hemispherical end cap, and returns up an annulus to ~550 °C. A shroud keeps shell-side sodium axial along the tubes, which is the arrangement for pure counterflow (LMTD correction F = 1.0 in the ideal case). That F is a geometric intent, not a measured plant result.
The pressure boundary is the outer tube only:
Tubes hang from a single Alloy 709 tubesheet in the dry argon cover-gas plenum, above the sodium free surface. There is no submerged lower tubesheet, so there is no lower crevice where stratified sodium would sit against a fixed sheet. Axial growth of the tube is not reacted by a second sheet. The intended axial stress from that restraint is about zero. The top tubesheet and any lateral guides still carry load.
The gas-side cartridge is a removable thin-walled stainless insert. It is not the 80 bar boundary. An insulated double-wall core limits thermal short-circuit between cold downward nitrogen and hot upward nitrogen. Helical ribs or wire-wrap spacers on the outside of the insert drive swirl in an annulus of about 2.5 mm. Both sides of the insert see the same ~80 bar gas; the differential is friction only (target under 0.5 bar).
Withdrawing the cartridge is what opens a straight, smooth bore to eddy-current and ultrasonic probes. Cartridge withdrawal mechanisms, seals in the argon plenum, and seismic lateral tube guides remain active engineering and qualification work.
The concept aims for defense in depth: inherent physics, passive heat removal, and engineered barriers. Reactor physics provide strong inherent negative feedbacks that assist safety; they do not make meltdown “impossible” or remove the need for engineered shutdown, decay-heat removal, and licensing.
Nitrogen at ~80 bar is inside the tube. Sodium is on the shell. The tube is closed by a free hemispherical cap and hung from a single top tubesheet in the dry argon cover-gas plenum. There is no lower header.
In a tube shear, buoyancy of the escaping nitrogen is intended to drive gas up into that plenum and its rupture disks, rather than into the sodium that feeds the downcomer, the ALIP pumps, and the core. Break size, sweep-out, and relief capacity are still open analysis. Cartridge withdrawal, the plenum seals, and the seismic guides are the same open qualification as the heater itself.
A direct Na–N₂ boundary still has to answer what an intermediate sodium loop was built to answer: keeping activated sodium (Na-24) and tritium off the turbine island, and showing that high-pressure gas does not reach the core. Two further pieces of hardware are candidates, not the barrier:
Valves are active components. Getters do not stop a break. Neither one replaces a scaled sodium–gas test. A circulating secondary sodium loop remains the licensing fallback.
A simplified in-vessel handling concept (triangular fuel wedges and a single-axis hoist, rather than complex multi-axis in-sodium robotics) is planned as a way to reduce mechanical risk relative to some historical designs. It remains concept-level until engineered and demonstrated.