Drawing No. EH–NR–005 // Nuclear Engineering & Radiation

Advanced BWR Reactor Simulator

Reviewed August 2026

Operate a detailed educational model of a large forced-circulation boiling water reactor using the ABWR as the public reference design. Control reactivity with fine-motion control rods, regulate core flow with ten reactor internal pumps, manage pressure, level, turbine-generator and feedwater systems, and test active emergency core cooling through normal startup and abnormal transients.

3,926 MWth reference thermal power ~1,350 MWe reference electrical output 872 fuel assemblies 205 control rods 10 RIPs forced recirculation 16 operating & failure scenarios Public-data checked educational physics model

Public ABWR design basis

Reference characteristics used by the model

ReactorDirect-cycle BWR

Primary flowForced recirculation

Core flow~52,200 t/h

Steam flow~7,640 t/h

FuelGE14 reference

Control rods205 FMCRD blades

Internal pumps10 RIPs

Operating pressure~7.17 MPa(a)

ECCSRCIC / HPCF / ADS / LPFL

Public values are drawn principally from U.S. NRC ABWR design-certification material and Hitachi-GE UK ABWR GDA documents. This is an independent educational model; proprietary protection setpoints, plant procedures, core-loading data and vendor analysis models are not reproduced. Protection thresholds, ECCS demand curves, break areas and indicated-level setpoints are educational approximations; public rated-flow anchors are used where available. The model combines six-group point kinetics, multi-timescale ANS-style decay heat, and lumped vessel mass/energy/void/containment feedback; it does not calculate MCPR/CPR, local cladding/PCT, 3-D power shape, ATWS spatial stability, detailed two-phase break flow or licensing margins.

Reactor controls

FMCRD / neutron control

Reactor mode switchSHUTDOWN

Aggregate FMCRD withdrawal0.0%

Automatic power control

Emergency reactivity actions

Rod withdrawal blockCLEAR

Reactor recirculation

10 reactor internal pumps

RIP speed command0%

Automatic flow/power control

RIP speed changes core flow and void fraction. Increasing flow generally reduces voids and inserts positive reactivity; decreasing flow does the opposite.

Pressure & steam

pressure regulator / bypass

Pressure setpoint0.10 MPa(a)

Automatic pressure control

Turbine control valve0%

Turbine bypass valve0%

Main steam isolation valvesOPEN

Feedwater & level

three-element educational model

Level setpoint0.0 m

Automatic level control

Feedwater command0%

Turbine-generator

HP / MSR / LP train

Grid frequency50 Hz

Electrical load demand0 MWe

ECCS & containment

3 safety divisions

Automatic ECCS logic

RCIC0 kg/s

HPCF B0 kg/s

HPCF C0 kg/s

LPFL A0 kg/s

LPFL B0 kg/s

LPFL C0 kg/s

Total ECCS water injection0 kg/s

Off-site ACAVAILABLE

Safety ACAVAILABLE

DC battery model24.0 h

Wetwell30.0°C

Failure scenarios

Inject / clear

Plant process mimic

ABWR-style mass, energy & safety paths

Plant trends

Last 15 simulated minutes

PowerPressureLevelElectrical

Combined normalized view0–120% display scale

Live instrumentation

Reduced control-room view

Active alarms

0 active

Operator event console

Newest first

Startup procedure trainer

Educational workflow · left-to-right sequence

Guided startup automates this simplified teaching sequence. It is not an operating procedure.

Diagnostics & engineering views

See how forced-flow BWR physics changes the plant

These views link operator actions to simplified core behaviour. They are conceptual educational displays, not replicas of proprietary ABWR core-monitoring or protection systems.

Conceptual 872-assembly core power map

Stable radial power shape with local control-rod suppression and global void/flow feedback. Geometry is deliberately schematic.

Power / core-flow operating map

Public ABWR documentation defines the normal operating area using minimum-pump-speed, constant-speed, steam-separator, natural-circulation and rod-pattern/flow-control boundaries. The redrawn chart separates the operating envelope from startup/low-flow and avoid regions, with continuous boundary lines identified in a clean legend below the plot. Coordinates remain educational approximations, not plant limits.

Reactivity balance

Approximate contributions from FMCRDs, voids, fuel temperature, moderator, xenon and standby liquid control.

ECCS and heat-removal margin

Compares decay heat and inventory support with condenser/RHR status and active RCIC, HPCF and LPFL functions.

Model architecture

What is being calculated

The simulator couples simplified neutronics, fuel/coolant heat transfer, forced recirculation, void feedback, steam/feedwater inventory, turbine-condenser response, containment, electrical availability and active emergency core cooling.

1. Point kinetics

Six delayed-neutron groups are integrated with a stiff implicit scheme. Reactivity combines aggregate FMCRD position, void feedback, Doppler feedback, moderator temperature, xenon and SLC boron.

2. BWR flow control

Ten modeled reactor internal pumps establish core flow. Changing flow changes average void fraction; void feedback then changes reactivity and power. This is the major dynamic difference from a natural-circulation reactor, which relies on density-driven flow instead of forced pumping.

3. Steam and feedwater

Generated steam is matched against turbine and 33%-capacity bypass flow. Feedwater control uses steam/feed flow plus vessel-level error, with shrink/swell represented through the void state.

4. Power conversion

The model includes HP turbine, moisture separator/reheater, LP turbine train, condenser vacuum, generator synchronization, load control and condenser heat-sink degradation.

5. ECCS

Three safety divisions are represented functionally: Division I RCIC plus ADS/LPFL-A, Division II HPCF-B plus ADS/LPFL-B, and Division III HPCF-C plus ADS/LPFL-C. RCIC, HPCF and LPFL flow curves are anchored to public ABWR rated-flow values; automatic demand modulation, initiation thresholds and interpolation remain educational.

6. Containment & wetwell

During normal operation the wetwell contains an upper gas space and a lower suppression pool maintained at its normal water level; it is not completely water-filled. LOCA and SRV/ADS discharge can raise drywell pressure and heat the suppression pool. The wetwell is used as the simplified sink for depressurization steam and safety injection recirculation.

7. Startup sequence

The teaching sequence establishes condensate/feedwater support, starts internal recirculation pumps at low speed, selects STARTUP, approaches criticality with fine-motion rods, raises temperature and pressure using bypass control, establishes power-range operation, rolls and synchronizes the turbine, transfers to RUN, and then raises power with a combination of rod trim and core-flow increase.

8. Automatic protection and ECCS logic

Illustrative trips cover short reactor period, high neutron flux, high pressure, low/high water level, high drywell pressure, turbine trip, main-steam isolation, very low core flow at high power and flow-biased overpower. Low level or LOCA signals can initiate RCIC/HPCF; ADS is used when high-pressure inventory recovery is inadequate, enabling LPFL at low pressure.

9. Transient scenarios

The scenario suite includes turbine trip, feedwater loss, condenser-vacuum loss, loss of off-site power, station blackout, three-RIP trip, all-RIP trip, small and large LOCA, MSIV closure, uncontrolled rod withdrawal, ATWS, RCIC failure, HPCF failure, feedwater-line break and service-water degradation.

10. Important limitations

No 3-D neutronics, fuel thermal limits, CPR/MCPR, LHGR, channel stability analysis, detailed turbine cycle, safety-grade containment analysis, radiological source term or severe-accident progression is calculated. Reactor water level is a simplified relative indicated coordinate rather than a calibrated Level 1/1.5/2/3/8 or top-of-active-fuel elevation. Protection thresholds, break coefficients, inventories and transient time constants are educational calibrations unless explicitly identified as public design data. The generic DC-battery endurance indicator is independent of scenario-specific coping assumptions such as the 8 h SBO RCIC-credit window. ECCS rated-flow anchors use public ABWR values, while the browser model's demand curves and initiation logic remain simplified.

Traceability & validation

Public data, calibrated assumptions and targets

Public ABWR facts are kept separate from simplified parameters introduced solely to produce a coherent browser simulation.

Model item	Value / implementation	Status and basis
Rated reactor thermal power	3,926 MWth	PUBLIC U.S. NRC ABWR design certification; UK ABWR design documentation.
Reference electrical output	~1,350 MWe	PUBLIC UK ABWR public design material.
Core	872 fuel assemblies; 205 control rods	PUBLIC UK ABWR reactor-core and waste/disposability documentation.
Reactor recirculation	10 reactor internal pumps	PUBLIC NRC and UK ABWR reactor coolant descriptions.
Rated core flow	~52,200 t/h (~14,500 kg/s)	PUBLIC UK ABWR reactor coolant specification table.
Rated steam flow	~7,640 t/h (~2,122 kg/s)	PUBLIC UK ABWR reactor coolant specification table.
Operating pressure	~7.07 MPa(g), ~7.17 MPa(a)	PUBLIC UK ABWR reactor coolant specification table.
Turbine bypass	33% of rated steam flow	PUBLIC ABWR steam and power-conversion design documentation.
ECCS architecture	RCIC; two HPCF divisions; ADS; three LPFL/RHR loops	PUBLIC NRC and UK ABWR Engineered Safety Features documentation.
ECCS flow capacities / trip thresholds	RCIC ≥182 m³/h; HPCF ~182→727 m³/h per division; LPFL/RHR ~954 m³/h per loop at public rated conditions	PUBLIC FLOW ANCHORS NRC ABWR material anchors rated flows; demand curves, thresholds and timing are educational and are not plant setpoints.
Point kinetics / feedback coefficients	6-group educational model	CALIBRATED Standard kinetics structure, not vendor core physics.
Core power map	872 schematic nodes	CONCEPTUAL Node count follows public core count; geometry/loading pattern is not represented as actual ABWR core loading.
Power / core-flow operating map	ABWR boundary concepts + educational avoid zones	PUBLIC CONCEPTS Boundary types follow UK ABWR PCSR Figure 11.5-11; exact coordinates and colored zones are schematic and are not operator limits or trip setpoints.

Steady-state validation targets

100% operation approaches 3,926 MWth, ~1,350 MWe, ~7.17 MPa(a), rated steam/feed flow and ~100% core flow.
Increasing RIP speed at power reduces void fraction and produces a positive power response unless rods compensate.
Generator output remains zero until synchronization and breaker closure.
At steady load, steam and feedwater flows converge and level approaches its setpoint.

Transient validation targets

SCRAM rapidly collapses fission power while decay heat remains.
RIP trips reduce core flow, increase voiding and initially reduce reactor power.
Loss of feedwater lowers inventory and initiates ECCS on sufficiently low level.
LOCA raises drywell pressure, isolates steam and initiates high-pressure makeup followed by depressurization/LPFL if required.

Important limitations

No validated safety-analysis or operator-training capability.
No actual core loading, proprietary algorithms or plant setpoints.
No radiological source term, dose, severe-accident or component qualification model.
Never use output for plant operation, licensing, safety decisions or engineering design.

Primary public references

U.S. Nuclear Regulatory Commission — ABWR design certification overview and Design Control Document library.

Hitachi-GE — Introduction to the UK Advanced Boiling Water Reactor.

UK ABWR Generic PCSR Chapter 11 — Reactor Core.

UK ABWR Generic PCSR Chapter 12 — Reactor Coolant and Reactivity Control Systems.

UK ABWR Generic PCSR Chapter 13 — Engineered Safety Features.