Principal Research Results

Development of BWR Regional Stability Experimental Facility SIRIUS-F,
which Simulates Thermalhydraulics-Neutronics Coupling in Reactor Core,and Stability Evaluation of ABWR
Background
In BWRs, the reactor power is suppressed when the amount of steam bubbles (voids) increases in the core (void-reactivityfeedback). The feedback may cause self-sustained oscillations depending on the core design and operating conditions. There are corewide instability (in-phase oscillation) and regional instability (out-of-phase or first azimuthal mode oscillation). Current BWRs are designed and operated to prevent those instabilities from occurring.
Stability estimation methods were reassessed according to the experiences such as core-wide instability observed at La-Salle
2 plant in the United States and regional instability, which occurred at Caorso plant in Italy. Currently, BWRs are designed and operated on the basis of stability analyses by the developed stability estimation methods. These analyses are validated with instability events
observed at power plants and experimental data from out-of-pile thermal-hydraulic loops, which ignore void-reactivity feedback. It isof importance to build an out-of-pile experimental facility which simulates void-reactivity feedback to yield wide range of experimental database to validate analysis code.

Objectives
This study is intended to develop an experimental facility to simulate channel, core-wide and regional stabilities, and to evaluate stabilities of the ABWR for validation of design stability analysis code, ODYSY for a wide range including conditions for thelicensing.

Principal Results(1) Development of Stability Estimation Method
The SIRIUS-F facility was designed and constructed for highly accurate simulation of channel, core-wide and regional instabilities of an ABWR. A real-time simulation is performed for modal-point kinetics of reactor neutronics and fuel-rod conduction on
the basis of measured void fractions in a reactor core section of the thermal-hydraulic facility (Fig-1). Two tie plates and seven spacers
were installed in the facility at the same location of a commercially employed 9 ×9A fuel. The pressure profile of thermal-hydraulicfacility along the flow direction agrees with that of the 9 ×9A fuel (Fig-2).
A noise analysis method was performed to calculate stability indexes (decay ratios and resonance frequencies) from dominant poles of transfer function on the basis of the AR method by applying time series of a core inlet flow rate. The online analysis allows us to obtain stability indexes directly in any operating conditions without assuming hypothetical conditions.
(2) Thermal-hydraulic stability including the licensing conditions
Channel stability experiments were conducted for a wide range of operating conditions including maximum power points along the minimum pump speed line and the natural circulation line. The decay ratio and the resonance frequency of the experiments
are in good agreement with analytical results by ODYSY (Fig-3). The SIRIUS-F facility is the first out-of-pile facility to estimate thedecay ratio as an index of the stability for code validation, while stability boundary is only obtained as an index obtained in the otherfacility.
(3) Thermalhydraulic-neutronic coupling stability including the licensing conditions
Core-wide and regional stability experiments were conducted with artificial void reactivity feedback including maximum
power points along the minimum pump speed line and the natural circulation line (Fig-4). The regional stability was observed, whichis the least stable mode. There is a distinct peak of the cross correlation function with a negative value observed at delay time of 0s.
Therefore, regional mode was exited rather than core-wide mode.
The decay ratio and the resonance frequency of the experiments are in good agreement with analytical results by ODYSY(Fig-4).
In conclusions, although available stability data from the operating reactors are limited, the SIRIUS-F facility is the first outof-pile facility which enables us to evaluate stability of BWR for wide range of operating conditions accurately and systematically tovalidate design stability codes.
Main Researcher: Masahiro Furuya,
Research Scientist, Nuclear Power Generation Technology Sector, Nuclear Technology Research Laboratory
Reference
M. Furuya and F. Inada, “BWR Regional Stability Estimation using SIRIUS Facility with Noise Analysis Technology,” Proc. TenthInternational Meeting on Nuclear Reactor Thermal-Hydraulics, A503 (CD-ROM Publication), Seoul, Korea, October 5-9,2003.

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5. Nuclear - Improvement of economics and reliability of LWR power generation

Pressure
Release Line

4.5
Simulated Fuel Channel (left)Elevation from Core Inlet, H (m)4.0

Separator

Condenser

Orifice Flow Meter

Simulated Fuel Channel (right)

TurbineFlow Meter
Simulated Fuel Channel (left)

Bypass Channel

VacuumPump

Fine-Tuning
Subcooler
Recirculation
Pump
Main
Subcooler

Simulated Fuel Channel (right)
Design Specification of ABWR

3.5
3.0
2.5
2.0
1.5
1.0
0.5
0.0

Gate Valve
Lower Plenum

–0.5
6.85

NitrogenGas Tank

6.86

6.87

6.88

Pressure, P (MPa)Preheater

Fig.2 Comparison of Pressure Profile

Fig.1 Thermal-hydraulic Loop of SIRIUS-F Facility

Resonance Frequency, f (Hz)

0.7
42% Flow Rate
0.60.5

Maximum Power Point along
Minimum Pump Speed Line

0.40.31.00.8SIRIUS–F Experimental Data
Design Analysis Code (Channel Stability)

Stable0.60.4
Maximum Power Point along
Minimum Pump Speed Line
0.2

0.060

650.7
42% Flow Rate

7075800.5
Core–Wide Stability
0.4Core Thermal Power (%)

Maximum Power Point along
Minimum Pump Speed Line

0.3
Unstable

42% Flow Rate
1.00.8SIRIUS–F Experimental Data
Design Analysis Code (Resional Stability)
Design Analysis Code (Core–Wide Stability)

Stable0.60.4
Regional Stability
0.20.06085

Regional Stability

0.6

Unstable

42% Flow Rate

Core–Wide Stability and Regional
Stability Decay Ratios, DR (–)

Channel Stability Decay Ratio, DR (–) Resonance Frequency, f (Hz)13m-high SIRIUS-F facility is designed to simulate boiling Pressure profile is sensitive for stability estimation.
two-phase flow accurately in reactor core of ABWR.
Pressure profile of the SIRIUS-F facility is in good
agreement with that of ABWR.

Core–Wide Stability
Maximum Power Point along Minimum Pump Speed Line
6570758085

Core Thermal Power (%)

Fig.3 Channel Stability Results

Fig.4 Core-Wide and Regional Stability Results

Experiments were conducted with fix power level.
The results of design analysis code agree with thoseof experimental results.

Experiments were conducted with artificial neutronic
feedback. The results of regional stability by the codeagree with those of experimental results.

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