Design and modeling of the hydraulic control rod steering mechanism for the passive cooling system inside the core. Hydraulic CRDM design for previous studies and PINC; (a) Conventional IRIS-based CRDM, (b) Hydraulic CRDM for PINC.
Research background and motivation
However, an in-ship hydraulic CRDM has a smaller or negligible permeable area, a compact CRDM geometry, and a short control rod structure compared to the general CRDMs in nuclear power plants.
Review on hydraulic control rod drive systems
Introduction to passive in-core cooling system (PINCs)
Research Objectives and Scope
Bottom up CRDM BWR, Germany Prototype Test Test of elevation control and free rod fall behavior. Characteristics of the flow channel and the flow hole for height change Define the effect of the delay time for height control Wang (1993).
DESIGN AND MODELING OF THE HYBRID CONTROL ROD
Driving mechanism and its theoretical models
A ball bearing device is considered to linearly control the length of the hybrid control rod, so the friction force in the interface area is one of the main driving forces. The gravitational force for each hybrid control rod was calculated as a function of length taking into account the weight of the working fluid.
Experimental setup and procedures
- Heat transfer test
- Length control of the hybrid control rod
Experimental results
- Heat transfer performance of the single hybrid control rod
- Change of the pressure and operation condition
In the tests, the control parameters were the length of the hybrid control rod and the starting pressure. From the experimental results, the theoretical model predicted the pressure of the hybrid control rod well.
Validation of the theoretical model
The general trend line of the heat transfer coefficients is the same as that of general heat pipes. In previous studies, the trend line of the hybrid control bar is similar to the effect of the fill ratio 2.25-2.27. To analyze the pressure change of the hybrid control rod, pressure control test was performed.
In the test, the length of the hybrid control stick is controlled from the minimum to the maximum length without additional gases. To confirm the balance of the forces of the system, the forces in the equilibrium state were measured and a comparative study was carried out. in the case of a fill ratio of 100%, the pressure change of the hybrid joystick increased slightly. Evaporator and adiabat Diameter 10.9 mm. Instruments Connection pipe volume 1124.0 mm3. a) Hybrid joystick design.
Fig.2.9 Results of the pressure change of the hybrid control rod according to the control length.
DESIGN AND MODELING OF THE HYDRAULIC CONTROL ROD DRIVE
Driving mechanism and its theoretical models
- Design of the hydraulic CRDM with flow control system
- Theoretical model for conventional hydraulic CRDM
The steady flow rate and the flow behavior of hydraulic cylinder were defined as the function of the inlet flow rate and flow resistance coefficient. The continuity equation and momentum equation were driven using the geometry of the flow control loop. The driving mechanism of the hydraulic CRDM is highly related to the flow-driven forces at the moving cylinder.
The hydraulic force characteristics are defined as a stepped/patterned geometry within the cylinder. In SBWR, hydraulic cylinder design is just defined as a balance between weight and flow force against mass flow rate. But the CAREM reactor does not consider the force balance equation and the transient behavior of the hydraulic system.
Each hydraulic CRDM concept has different parameters to analyze the behavior of the hydraulic cylinder.
Experimental setup and procedures
- Step control test
- Rod drop test
The drop height was 600 mm and the tests were performed five times for each weight. During the drop test, a high-speed camera was used to measure the position of the CRDM.
Experimental results
- Conventional hydraulic CRDM (CRDM-C)
- Hydraulic CRDM of PINCs (CRDM-P)
The pressure and flow characteristics of the hydraulic CRDM during the single step down are shown in Fig. The test results of the bar fall time according to the weight of the hydraulic CRDM are shown in Fig. The experimental test was performed at the maximum height (600 mm) of the hydraulic CRDM in the test facility.
The main reason for these delayed results was the properties of the water and the weight of the hydraulic CRDM. Mass flow test results and pressure difference between the inlet and outlet streams. The step control prediction was very sensitive to operating conditions and cylinder geometry.
In the case of the hydraulic CRDM-P, the pitch (lift of the hydraulic CRDM) is easy to predict based on the mass flow rate.
Modified theoretical model
The height of the hydraulic CRDM is determined by the steady state mass flow rate. During surge control, the mass flow rate and pressure of the hydraulic CRDM increased dramatically. In addition, a hydraulic CRDM needs a position indicator to measure the location of the control rod.
The mass flow rate was increased or decreased to control the height of the hybrid control rod. In contrast, the hydraulic CRDM has additional weight due to the shape of the cylinder. Based on the force balance equation, the hybrid control rod hydraulic CRDM modeling was theoretically developed.
Experimental study of the pressure discharge process for the hydraulic control rod drive system stepped cylinder.
Validation of the theoretical models
- Validation of the CRDM-C
- Validation of the CRDM-P
APPLICABLE DESIGN OF THE NUCLEAR POWER PLANT
Design of the PINCs
Based on the functional requirements for safe shutdown of the nuclear power plant on an SBO, the reactor vessel, hydraulic CRDM and water storage pool are modified. During normal operation of the reactor, the hybrid control rod assembly (full strength; 708 rods) used for reactor shutdown is held in the upper plenum of the reactor vessel. As the first step in the reactor shutdown operation, the inlet flow of the hydraulic CRDM decreased by shutting down the control pump.
In a previous study, the detailed design information was presented, and the optimal results of the heat transfer and thermal resistance were determined as 18.20 kW and 0.015 k/W, respectively 4.8, 4.9. When the stroke length of the reactor core is 4 m, the required height of the moving part of the hydraulic CRDM is about 8 m. During steady state, full strength assemblies cannot be placed in the reactor core, and are therefore positioned on top of the reactor core (upper plenum). strength compositions are shifted to control the reactivity of reactor power.
Decomposition The heat from the nuclear fuel is transported by the phase change heat transfer of the working fluid between the inner cladding and the outer cladding.
Applicable heat removal design and its feasibility test
- Experimental setup and procedures
- Experimental results
In general, rod steering systems have control steps for systematically controlling the position of the control rod assemblies. As one of the main components of PINCs, the hydraulic control rod steering mechanism (CRDM) has been designed and modeled. Due to the high pressure and temperature conditions of nuclear power plants, the full-scale design of the hydraulic CRDM was carried out.
In nuclear power plants, hydraulic CRDM will be a good candidate due to good passivity and elimination of rod ejection accidents. Design and testing of a reactor hydraulic control rod drive for a nuclear power plant. Study of braking mechanism of new hydraulic control stick brake for CRHDS based on numerical simulation and experiment.
Annals of Nuclear Energy Study on the deceleration mechanism of the new control rod hydraulic decelerator for the CRHDS based on numerical simulation and.
Design of the hydraulic CRDM with sensitivity of the operation temperature
Prediction of heat removal performance in nuclear power plant
Based on the proposed design, the heat removal performance of PINCs was predicted by the safety analysis using the multidimensional analysis of reactor safety code (MARS-KS) in APR1400. In previous studies, the optimal heat removal of single hybrid control rod was determined based on the figure of merit of thermosyphon heat pipe 4.18. During the station blackout, the shutdown of the reactor coolant pump and turbine assumed to be 1.0 sec. after reactor shutdown signal.
The failure of the high pressure, low pressure safety injection and turbine driven auxiliary water system was assumed. Based on the calculated results, the timing of fuel melting and hydrogen production was delayed. The core cover time and the hydrogen production time are delayed by 35 minutes and 50 minutes, respectively, as shown in fig.
From the calculated results, it is possible to break down critical phenomena such as the time of maximum coating temperature and core detection.
Failure/degradation of PINCs in nuclear power plant
The critical failure and degradation of PINCs are 1) Hybrid control rod failure and 2) Hydraulic CRDM (CRDM-P) failure. In case of failure of the hybrid control rod, the heat removal capacity is reduced and the driving force of the hydraulic CRDM is changed according to the number of failed hybrid control rods. The degraded heat capacities of the hybrid control rod (Failure rate: 0%, 25% and 50%) were calculated using the MARS-KS code.
To mitigate the effect of hybrid control rod failure, inspection of the PINCs during the refueling period is required. If the hybrid control rods failed, the driving force of the hybrid control rod was reduced according to the failure ratio. The pressure difference and mass flow rate will increase as the failure ratio increases, due to the decrease in the driving force of the hybrid control rod (up force).
The operating range is determined by the range between with and without hybrid control rod.
CONCLUSIONS AND RECOMMENDATIONS
Modeling of the hydraulic control rod drive mechanism
Application of the hydraulic CRDM
In addition, the hydraulic CRDM model can be used in general nuclear power plants for reactor power control (without hybrid control rod). Therefore, the load-following mode of the nuclear power plant is a key issue for sharing the role of electricity generation with renewable energy sources. In order to become suitable for operation in the load control method, a high-precision control method and an improved safety function are required.
The proposed hydraulic CRDM could be a good candidate for post-load operation due to its high accuracy and passivity compared to the previous hydraulic CRDM concept.
Recommendations
Experimental study on the operational and cooling performance of the APR + Passive Auxiliary Feedwater System. Hydraulic control rod drive mechanism concept for passive nuclear cooling system (PINCs) in fully passive advanced nuclear power plant. Experimental study on the heat recovery characteristics of a new type of flat micro heat pipe array heat exchanger using nanofluid.
A study on sodium heat pipe characteristics in a passive molten salt reactor residual heat removal system. Batheja P, Meier WJ, Rau PJ, Design and testing of an internal hydraulic reactor control drive for a nuclear power plant. An in-vessel control rod drive mechanism using magnetic force latching for a very small reactor.
The passive cooling system study helped with a separate heat pipe for removing decay heat in spent fuel pool.
