Passive Boiling Water SMR Simulator

The BWRX-300 is a direct-cycle light-water boiling water reactor. Water boils in the reactor pressure vessel, steam passes directly to the turbine, exhaust steam is condensed, and condensate is pumped back as feedwater. The tall reactor vessel and chimney create buoyancy head for natural circulation, eliminating normal recirculation pumps and their external vessel loops.

The model represents the reactor pressure vessel, 240-assembly core, 57 cruciform control rods driven by fine-motion control rod drives (FMCRDs), steam separators and dryer, main steam isolation, turbine control and bypass valves, turbine-generator, condenser, feedwater pumps, reactor level and pressure controls, three isolation-condenser trains, the Passive Containment Cooling System (PCCS) at a highly simplified level, and boron injection.

Reactor power is evolved with a six delayed-neutron-group point-kinetics model. The precursor groups use representative educational U-235 delayed-neutron fractions and decay constants. Reactivity is the sum of calibrated core excess reactivity, negative control-rod absorber worth, Doppler feedback, moderator/void feedback, xenon poisoning, boron injection and automatic power-control bias. Rod withdrawal removes negative absorber worth; it is not displayed as positive rod reactivity.

The prompt generation time, feedback coefficients and aggregate rod-worth curve are calibrated educational values. They are selected for numerically stable, qualitatively realistic response across shutdown, startup and power operation; they are not BWRX-300 safety-analysis constants.

Radioactive decay heat is represented by a separate multi-time-scale fission-product inventory driven by operating chain power. The equilibrium groups sum to about 6.5% of rated heat immediately after a long full-power run and retain a long tail through hours and days; the fit is generic and educational rather than a core-specific ANS-5.1 calculation.

The public Boron Injection System (BIS) is a diverse emergency reactivity-control path intended to make the reactor subcritical from full power down to a cold 20 °C condition without relying on control-rod motion. The current General Description states that BIS can be initiated manually or automatically; the simulator exposes only a manual Boron inject action and compresses the system to a lumped buildup of negative reactivity. It does not model boron concentration, mixing, injection-pump hydraulics, automatic initiation parameters or real plant actuation timing.

The reactor thermal-hydraulic model uses lumped fuel and coolant energy balances. Natural-circulation flow increases with core power and density difference across the core/chimney path. Steam generation follows the heat transferred above the saturation threshold, while separator inventory and steam discharge determine vessel pressure and indicated level.

At rated conditions, model steam flow is calibrated to approximately 430 kg/s from an 870 MWth direct-cycle heat balance using representative BWR steam/feedwater enthalpies. A public rated main-steam mass-flow value was not identified in the audited BWRX-300 sources, so 430 kg/s is treated explicitly as a model estimate. It should not be confused with the public nominal reactor-coolant circulation value of 1827 kg/s.

Reactor level is driven by mass inventory, feedwater inflow, steam outflow and a simplified void-related shrink/swell term. Public PSR Chapter 7 states that below 10% reactor power the level-control system uses a low-flow control valve, while above roughly 10–15% it controls adjustable-speed feedwater-pump drives. Chapter 10 adds that the fuller multi-element heat-balance demand logic—using level, steam flow, reactor pressure and feedwater temperature—is typically applied above about 25% power; at lower power with the low-flow valve, control is based on level only. The simulator smooths these transitions instead of reproducing vendor control logic exactly.

The feedwater temperature model is also dynamic rather than an instantaneous load correlation. It represents stored heat in the LP/HP heater train, turbine extraction-steam heating and the public CFS control concept in which the C31 system modulates steam flow to Feedwater Heater #6 to control final feedwater temperature. During a feedwater line break, pump/header flow is separated from flow actually reaching the RPV; the RPV inlet temperature trace is discontinued after feedwater reactor isolation because no inlet stream then exists.

Opening steam valves or increasing steam demand briefly increases voiding and indicated level (swell); reducing steam demand collapses voids and can temporarily lower indicated level (shrink). The effect is deliberately restrained to preserve stability in a browser-based educational model.

When the generator is off-line, reactor pressure is controlled primarily by the turbine bypass valves discharging to the condenser. When the generator is on-line, the turbine control valves regulate steam flow and reactor pressure while the load controller sets electrical output. Public PSR Chapter 10 specifies turbine-bypass steam-flow capacity of no less than 25% of rated steam flow and states that the bypass/reactor combination can accommodate a 25% load rejection without reactor SCRAM. The reduced-order simulator conservatively fixes its simplified maximum bypass capacity at exactly 25%; this should not be read as a published BWRX-300 maximum.

The turbine can be rolled only with adequate steam pressure, available condenser vacuum and open steam isolation valves. The public General Description characterises a typical BWRX-300 turbine as a 3000/3600 rpm, single-shaft, tandem-compound, impulse-reaction, two-stage-reheat condensing turbine, coupled to a direct-driven three-phase 50/60 Hz synchronous generator. The simulator represents those public functional characteristics and does not assume a proprietary turbine model designation. Closing the breaker then allows electrical loading. On turbine trip, the TCV/TSV path closes rapidly while the TBVs initially open and then modulate, so full-load rejection produces an initial reactor-pressure rise before SCRAM and subsequent cooldown.

The simulator includes automatic trip logic for high power, short reactor period during startup, high/low vessel pressure, low or high level, high containment pressure, high condenser pressure, turbine-trip demand, main-steam isolation, low electrical bus voltage, sustained low feedwater flow and validated line-break indications. The public BWRX-300 I&C architecture uses both parameter trips and anticipatory event/status signals through the Automatic Protection System (APS) and associated protection functions. Most numerical thresholds and qualification times in this trainer are calibrated educational values rather than plant setpoints. Two public Step 2 protection values are used directly where the pressure datum is unambiguous: the generic high-RPV-pressure SCRAM fallback is 7.787 MPa(g), approximately 7.89 MPa(a), and the high-containment-pressure protection reference is 18.5 kPa(g). The simulator uses MSIV as compact operator-interface shorthand; public BWRX-300 documentation distinguishes Main Steam Reactor Isolation Valves (MSRIVs) at the reactor pressure boundary from Main Steam Containment Isolation Valves (MSCIVs) farther out on each main-steam line. For generic post-SCRAM pressure escalation, the three ICS trains are staged at the public HP2/HP3/HP4 pressure thresholds rather than at invented lower pressures.

Failure scenarios no longer command an automatic SCRAM merely because a scenario button was pressed. The initiating event first changes equipment or process conditions, the relevant monitored signal must develop and persist for a finite qualification time, and only then is the protection action generated. On a normal SCRAM, stored accumulator energy in the hydraulic control units rapidly inserts the control rods hydraulically. Public I&C documentation also describes a simultaneous FMCRD “scram follow” signal whenever APS demands hydraulic SCRAM: the motor drives the ball nut upward to just below the inserted and latched piston. That follow motion is distinct from the diverse electric motor-driven rod run-in function, which can shut the reactor down if hydraulic SCRAM fails or is delayed. The trainer collapses these mechanisms into one rod-insertion state rather than resolving them separately. Fission power collapses, but decay heat remains.

The public UK GDA describes three independent isolation-condenser trains. Each heat exchanger is submerged in a dedicated inner pool and linked to outer pool inventory. Steam flows from the reactor to the condenser, condenses, and returns by gravity. In the trainer an automatic ICS actuation is latched until reset; actual heat-transfer duty then falls naturally as vessel pressure and temperature decrease rather than by artificial rapid valve cycling.

Public PSR Chapter 6 specifies three independent ICS trains, each rated at 33.75 MW(th). For safety-analysis credit, one train can generally mitigate AOOs, two trains are required for LOCA mitigation with one of the three assumed unavailable by single failure, and all three trains are credited for beyond-design-basis events. The same chapter states that the smallest inner-pool inventory combined with outer-pool inventory provides at least three days of decay-heat removal, while the two-smallest inner-pool inventories combined with the outer pool support at least seven days; the outer-pool arrangement is also described as retaining a minimum seven-day capability with one ICS train unavailable. ONR separately reports a conservative 72-hour one-train endurance calculation leaving about 50.6% of the smaller Pool A volume and a seven-day extended calculation leaving about 7.3% in the larger Pools B/C. The trainer compresses these public capacity/endurance anchors into one aggregate pool meter. Generic pressure-driven actuation now uses the three public Step 2 high-pressure thresholds—8.289, 8.504 and 8.719 MPa(g), corresponding to approximately 8.390, 8.605 and 8.820 MPa(a) in the trainer's pressure basis. Public I&C also lists low RPV water level, high containment pressure and qualified line-break indications as ICS actuation signals. Because the audited public material does not disclose the numerical L2 water-level setpoint, the trainer's generic deep-low-level mapping remains explicitly educational; the dedicated DBA scenario branches reproduce their published signal sequence separately. The aggregate pool indication is not an individual BWRX-300 pool-level, boil-off, makeup-water or licensing-credit calculation.

The Passive Containment Cooling System (PCCS) is a separate passive containment heat-removal function. The current public design describes three passive PCCS trains that are continuously available/in service and become effective as accident steam and heat load the containment. The trainer represents PCCS only as a lumped containment heat sink and availability indication; it does not model PCCS loop natural circulation, pool boil-off, condensation surfaces, containment heat structures or individual-train performance.

The guided sequence is a teaching workflow assembled from public descriptions of BWR controls and the BWRX-300 plant automation functions: establish systems, select STARTUP mode, approach criticality under wide-range neutron monitoring, heat and pressurize through bypass control, establish feedwater level control, roll and synchronize the turbine, close the breaker, transfer to RUN mode, and raise load.

For normal STARTUP/SHUTDOWN only, the trainer limits bulk-coolant heat-up/cooldown to about 55.6 °C/h (100 °F/h) as a generic BWR vessel thermal-stress teaching constraint. This is not a published BWRX-300 operating limit and it is not imposed on accident or passive-safety cooldown transients.

It is not an operating procedure and omits required verifications, surveillance, chemistry controls, radiation protection controls, technical specifications, staffing and hold points. Actual plant startup would be performed only by licensed personnel using approved plant-specific procedures.

Turbine trip / load rejection: trips the turbine first; an APS turbine-trip demand is then qualified before SCRAM, and the normal bypass path limits the initial pressure rise. The trainer does not initiate ICS for this AOO unless a separate ICS actuation signal is actually reached. Loss of feedwater: trips all feedwater pumps while the bounding pressure-control assumption is held failed-as-is; RPV inventory and pressure fall, low water level initiates SCRAM, low RPV pressure later isolates main steam, and a lower-level condition starts all three ICS trains. Loss of condenser vacuum: progressively raises condenser backpressure; turbine trip/SCRAM occurs first, the bypass initially controls pressure, and continued vacuum degradation can inhibit/close the bypass so pressure later rises toward one-train ICS high-pressure actuation. Station blackout: removes normal AC support first; low-bus-voltage protection then initiates the reactor trip while passive systems remain available, with generic ICS actuation following the public high-pressure sequence rather than an artificial immediate start.

Small LOCA + LOPP: follows the public bounding small-break sequence: feedwater coasts down, pressure control is unavailable, the RPV depressurizes, SCRAM follows qualified low steam/RPV pressure, main-steam isolation follows later, and ICS starts on low level. Main steam line isolation: allows finite valve travel before APS closure logic trips the reactor. Inadvertent rod withdrawal: demonstrates the public Automatic Thermal Limits Monitor (ATLM) / Multi-Channel Rod Block Monitor (MRBM) rod-block concept; abnormal withdrawal stops before a large power excursion unless that protection were separately assumed failed. ICS train failure: latches one passive heat-removal train unavailable without causing SCRAM; it may be combined with one initiating event, and automatic logic selects the remaining available trains. Feedwater line break: uses a bounding large-break + LOPP training sequence: blowdown and pump coastdown begin at time zero, SCRAM follows validated break indication, and feedwater/main-steam isolation valves close over several seconds rather than instantaneously. Heat-sink degradation: raises condenser pressure gradually until a protection threshold is exceeded.