- NuScale Integral System Test (NIST-2) Facility
- Helical Coil Steam Generators
- Emergency Core Cooling System
- Fuel Design
- Control Rod Assembly
NuScale Integral System Test (NIST-2) Facility
We designed and built the state-of-the-art NuScale Integral System Test (NIST-2) facility formerly located at Oregon State University in Corvallis, Oregon. It held a one-third scale prototype that replicated the entire NuScale Power Module™ and reactor building cooling pool. It provided an electrically-heated core to bring the system up to operating temperature and pressure. Stability testing ensures that throughout the expected operating conditions, natural circulation flow is stable.
The facility, now closed, demonstrated the strong safety case and viability of the NuScale Power Module, and provided an enhanced representation of our current reactor design. NIST-2 included a data acquisition and control system as well as extensive instrumentation additions which provided the measurements necessary for safety analysis code and reactor design validation.
Helical Coil Steam Generators
Steam Generator Tube Inspection Feasibility Study
We completed a feasibility study to evaluate the use of eddy current probes to inspect our helical coil steam generator tubes. Specifically, we addressed the ability of a conventional probe system to traverse our helical coil steam generator design. Helical coils with prototypic lengths, diameters, curvatures, and inclination angles were traversed using a conventional eddy probe system, indicating acceptable insertion capability of a conventional eddy probe system.
Steam Generator Inlet Flow Restrictor Testing
We commissioned Alden Laboratory to conduct a series of Steam Generator Inlet Flow Restrictor (IFR) tests. Testing used Reynolds similitude to characterize form loss and vibration response for several candidate flow restrictor designs over a range of operating conditions, including flow velocities up to several times that which would occur in service. Test results showed excellent performance characteristics for the NPM IFR design, with no occurrence of leakage flow instability or turbulent buffeting. Post-test inspection confirmed lack of wear on the flow restrictor and inside of the steam generator tube.
SIET Separate Effects and Prototype Steam Generator Testing
We contracted with Società Informazioni Esperienze Termoidrauliche (SIET) in Piacenza, Italy to validate the steam generator design under conditions that result from natural circulation flows in the NuScale reactor coolant system. Tests, completed in early 2014, focused on the secondary side performance and consisted of an electrically-headed, highly-instrumented, full-length 3-tube bundle. A second set of tests, completed in 2015, focused on overall primary and secondary side performance and consisted of a prototypic tube bank (252-tube bundle) operated at plant primary and secondary flow conditions.
Flow Induced Vibration Tests
We have also enlisted SIET to fabricate a full-scale prototypic mock-up of our helical coil steam generator to support flow induced vibration testing of steam generator internals. Goals of this test program include determination of natural frequencies and mode shapes of steam generator tubes, determination of system damping characteristics, and determination of primary flow characteristic behavior at normal operating flow rates and above. Additionally, valuable steam generator fabrication, assembly, and material handling insights will be obtained as part of this test program.
Emergency Core Cooling System
Emergency Core Cooling System (ECCS) Valve Tests
Our SMR uses a simplified ECCS system consisting of three reactor vent valves located in the pressurizer, and two reactor recirculation valves in the downcomer. During a LOCA scenario, primary pressure decreases and the valves open when their low pressure setpoint is reached. Once open, a natural circulation flow path is established between the reactor pressure vessel and containment to provide long term core cooling.
We contracted with Target Rock to design and test these ECCS valves. Testing has been implemented in several phases, beginning with an initial proof-of-concept test to assess first-of-a-kind design features, followed by more rigorous demonstration tests that assessed functionality at operating temperatures and pressures, and lastly, by proof testing using fully prototypic valves. Proof-of-concept and demonstration testing has confirmed that the ECCS valve design functions properly, and that the low pressure setpoint, which is governed by a unique inadvertent actuation block, reliably prevents valve opening at elevated pressures. Valve proof testing will be used to demonstrate consistent and dependable valve operation in a LOCA environment.
Fuel Design
Thermal-Hydraulic and Mechanical Testing of Preliminary and Final Fuel Designs
Our SMR uses natural circulation-driven flows in the reactor coolant system to provide reliable core heat removal, both during normal plant operation and for accident conditions. To obtain critical heat flux (CHF) test data suitable to validate the fuel bundle design, we completed a major test program for the preliminary fuel design at Stern Laboratories in Ontario, Canada. Testing was conducted over a wide range of natural circulation flow rates and pressures, with both uniform and non-uniform power profiles. Results from the Stern testing have been used to characterize performance margin and to inform design optimizations and testing of the final fuel design.
Testing of the final fuel design was completed at the Framatome KATHY multifunction thermal-hydraulic test loop in Karlstein, Germany. This facility tested both steady-state and transient thermal-hydraulic behavior of the fuel assemblies.
Thermal-hydraulic fuel testing requirements extend beyond the need for CHF characterization, and the Framatome Richland Test Facility has been employed to complete lift-off, pressure drop, and hydraulic characterization of the final fuel design. Additionally, mechanical testing of the new fuel design was conducted at the Richland Test Facility to measure the physical capabilities of the design under different seismic conditions.
Control Rod Assembly
Control Rod Assembly and Drive Shaft Drop Alignment Test
Drop Alignment Tests were conducted in 2018 at the Framatome GmbH (formerly AREVA) Technical Center in Erlangen, Germany. These tests assessed control rod insertion for a variety of control rod drive shaft alignment conditions. Prototypical test hardware included a control rod assembly (CRA), control rod drive shaft structures, CRA guide tube assembly, and fuel assembly. Testing confirmed that the scram curves used in our safety analyses are conservative.
