The Reactor Pressure Vessel (RPV) inlet nozzles and downcomer wall in Pressurized Water Reactors (PWR) may suffer serious thermal shock caused by cold water from reactor Safety Injection System (SIS) in some unexpected accident scenarios. It implies the formation of great temperature gradient on the inlet nozzles and RPV wall, leading to the localized stresses and propagation of possible flaws that appeared in the material. In this paper, the multiscale thermal hydraulic analysis was performed for Chashma Nuclear Power Plant (NPP) under the inadvertent SIS operation scenario. The primary loop and SIS were modeled using one-dimensional method, while the three-dimensional models of reactor cold leg, RPV inlet nozzles, and downcomer were established. Then, the inadvertent Safety Injection System operation scenario was simulated using RELAP5 code, providing the boundary conditions for three-dimensional Computational Fluid Dynamics (CFD) analysis. The fluid and solid coupling heat transfer simulation method was employed. Results show that the maximum temperature difference was about 80 K in the most conservative condition and the RPV inlet nozzle region was the most critical region during the accident. This work could provide in-depth understanding on the effect of cold coolant injection along the main pipes and RPV wall during the accident scenario.
The Pressurized Thermal Shock (PTS) phenomenon study for Pressurized Water Reactor (PWR) has been a hot research topic for several decades. PTS is identified as one of the most important Nuclear Reactor Safety (NRS) issues since the primary loop boundary is one of the barriers against fission product release. In the event of Loss of Coolant Accident (LOCA) or some unexpected accident scenarios, the cold injection water from Safety Injection System (SIS) or Emergency Core Cooling (ECC) system flows to the cold leg and Reactor Pressure Vessel (RPV) downcomer, leading to the material rapidly cooling and great thermal loads on the reactor main pipes and RPV under pressurized conditions. The thermal stress is induced during the process and it would threaten the RPV structure integrity [
Generally, the PTS thermal hydraulic study is involved in the complex system operation and local detailed three-dimensional analysis. Therefore, the combination of different thermal hydraulic analysis scales is necessary in order to achieve more reliable results. The one-dimensional system analysis is used to study system performance during the transient scenarios. However, it could not give detailed three-dimensional temperature distributions of key components. The Computational Fluid Dynamics (CFD) study could achieve the detailed three-dimensional fluid-solid coupling heat transfer features, but currently it could not realize a long time scale transient calculation for a complex structure, such as reactor, due to the computational resources limit. Therefore, the multiscale method was adopted for the PTS thermal hydraulic study under the inadvertent Safety Injection System operation scenario of typical PWR in this work.
The nuclear power plant one-dimensional safety analysis is widely implemented using system codes, such as RELAP5, TRACE, and TRAC, which has been accepted by scholars around the world. In recent years, one of the main applications of CFD method for NRS is the PTS study, which is regarded as high priority in the Phenomena Identification and Ranking Table (PIRT) [
As stated above, some literatures were focused on the nuclear power plant transient safety analysis, while some were concentrated on the specific coolant mixing phenomenon study in a limited time scale. Very rare research was given consideration to the multiscale thermal hydraulic study for the reactor long time simulation. In this paper, the Chashma Nuclear Power Plant (NPP) was taken as the research object. The Reactor Coolant System (RCS) and the SIS were modeled using the RELAP5 code firstly [
The procedure of multiscale thermal hydraulic study for Chashma NPP is shown as Figure
The procedure diagram of multiscale thermal hydraulic study.
The transient behavior of Chashma NPP was studied under the inadvertent Safety Injection System operation scenario. The temperatures of primary loop and SIS and flow rates of primary loop and SIS achieved through one-dimensional study were handled using MATLAB software as the inputs for the boundary and initial conditions of CFD simulation. Then the detailed thermal hydraulic characteristics of the primary loop cold leg inlet nozzles and RPV downcomer were obtained.
The Chashma NPP is a two-loop type PWR and the RCS was divided into several important parts according to the basic principles and processes of primary coolant system during system modeling, including the reactor core, pressurizer and pressure control system, steam generators, main pumps, and the corresponding pipes and valves. The diagram of RCS and RELAP5 nodalization is shown as Figure
The main parameters of Chashma NPP.
Parameters | Value |
---|---|
Reactor thermal power | 998.6 MW |
Operation pressure | 15.2 MPa |
Arrange type of fuel assembly | 15 × 15 |
SG pressure (full power) | 5.4 MPa |
Reactor type | 2 Loops PWR |
Reactor outlet temperature | 315.5°C |
Reactor inlet temperature | 288.5°C |
Thermal designed flow rate | 3227.896 kg/s |
Steam flow rate (each SG) | 280.56 kg/s |
The Chashma NPP RCS and RELAP5 nodalization diagram.
Schematic diagram
Nodalization
There are four high-pressure-safety-injection pumps in high pressure SIS, two of which compose a group whose inlets are connected to the refueling water tank through check valves and power-operation-isolation valves. All the inlets in the same group are connected to the outlets of the shutdown-margin pumps through power-operation-isolation valves. The outlets are connected to a header pipe which is divided into four branch pipes. The four pipes are connected to the cold legs or hot legs of RCS.
Under the condition of safety injection, the high-pressure-safety-injection pumps pump cold water from the refueling water tank to the reactor core. During the recirculation phase, the high-pressure-safety-injection pumps work as supplement of low pressure safety injection pumps, to pump the water in the containment sump back to the reactor core. The RELAP5 nodalization diagram of Chashma NPP SIS is shown as Figure
The RELAP5 nodalization of Safety Injection System.
Generally, The RELAP5 code is a general transient analysis program for thermal hydraulic system and its application scope includes coolant loss accidents, running transient state, power supply lose, flow loss, and subcooling transient state. The RELAP5 code has been widely accepted for the system safety analysis of nuclear power plants. For a special work, the steady-state simulation could be used to verify the feasibility of the built RELAP model and the transient scenarios are simulated based on the steady state with the “Restart function”; in this paper, the steady-state simulation of Chashma NPP was performed firstly to verify the RELAP5 model established in this work. The whole system calculation reached steady state at about 3460 s. The comparisons between the calculation results and the rated full-power operation parameters obtained from the Chashma NPP design report are shown in Table
Steady calculation results and comparison with design value.
Parameters | RELAP5 value | Design value | Error |
---|---|---|---|
Primary loop pressure (MPa) | 15.26 | 15.2 | 0.39% |
Reactor outlet temperature (°C) | 316.1 | 315.5 | 0.19% |
Reactor inlet temperature (°C) | 289.0 | 288.5 | 0.17% |
Primary flow rate (kg/s) | 3230.25 | 3227.896 | 0.07% |
SG mass flow rate (kg/s) | 283.62 | 280.56 | 1.09% |
SG pressure (MPa) | 5.544 | 5.54 | 0.07% |
Upper plenum flow rate (kg/s) | 31.65057 | 32.2788 | 1.94% |
In this section, one-dimensional safety analysis of inadvertent Safety Injection System operation was simulated. The variations of SIS and primary loop coolant flow rates and temperatures with time were studied. In the scenario of inadvertent Safety Injection System operation, due to the unexpected action of the Safety Injection System valves, the cold water in SIS was injected into the reactor primary loop and the thermal shock damage would happen. According to the operating experience of nuclear power plant, this type accident has a relatively high frequency.
According to the logic of inadvertent Safety Injection System operation accident, the Safety Injection System was acted by mistake at 100 s. Then the reactor shuts down and the main pumps run out. The steam generator secondary water supply switched from the main water supply system to the auxiliary feed-water system after the trip signal. The important thermal hydraulics characteristics results are shown in Figures
Variations of SIS flow rate and coolant temperature versus time.
Variations of primary loop pressure and flow rate versus time.
The variations of flow rates and coolant temperatures in Safety Injection System and primary loop during the accident scenario were shown in the figures. The reactor shuts down immediately receiving shut-down signal after inadvertent Safety Injection System operation. The cold water from Safety Injection System began to be injected into primary loop, leading to rising of the primary loop pressure, making the safety injection flow rate decrease. The maximum injection flow rate was about 13 kg/s. The coolant temperature declined from the beginning of this scenario. In this case, the thermal shocks could be loaded on the reactor cold leg inlet nozzles and RPV downcomer and it would bring great threat to the reactor safe operation.
In order to achieve the detailed three-dimensional temperature distributions of RPV inlet nozzles and downcomer wall, the CFD work was performed and the conjugate heat transfer simulation method was adopted to handle the coupling phenomenon between fluid region and solid region.
The detailed three-dimensional models of Chashma NPP primary loop cold pipes and the RPV downcomer were established, as shown in Figure
The main parameters of RPV.
Parameters | Unit | Value |
---|---|---|
Designed pressure | Mpa | 17.16 |
Designed temperature | °C | 350 |
Shell material | / | SA-508C1 |
Total height | mm | 10705 |
Barrel outer diameter | mm |
|
Shell inner diameter | mm |
|
Shell wall thickness | mm | 175 (without reactor weld layer 4 mm) |
Total volume | m3 | 74.47 |
Junction inner diameter | mm |
|
Junction outer diameter | mm |
|
The whole geometry of main pipes and RPV.
The hexahedral mesh with refined meshes near the wall was generated using ICEM software for the cold legs, nozzles, and downcomer fluid region, as shown in Figure
Grid of the main pipes and upper section of downcomer.
For the solid region, the mesh of pressure vessel was divided into two parts. The tetrahedral mesh was employed in the upper part, while hexahedral mesh was used in the lower part. Since the mass conservation, momentum conservation, and turbulence equations were not solved in the solid region, no large number of grids were required in this part. The detailed mesh condition is shown in Figures
The grid of RPV.
Grid of the core barrel.
For the three-dimensional flow and heat transfer study, the continuity equation, momentum conservation equation, and energy conservation equation were established. The continuity equation is Momentum conservation equation is Energy conservation equation is
The energy conservation equation in solid region is
There were several turbulent models in the ANSYS-CFX, and the well-known single-phase turbulence models are usually used to model turbulence of the liquid phase in Eulerian-Eulerian multiphase simulations. Based on the applicability of different turbulent models, the
The conjugate heat transfer method was adopted in order to simulate the fluid-structure interaction process. The solid conduction and convective heat transfer in pressure vessel are taken into consideration. The governing equations were established for each physical region during boundary coupling implementation. The temperature continuation and heat flux continuation boundary conditions were employed in this study. Temperature continuation is Heat flux continuation is
The initial accident scenario was divided into three parts due to the parameter variation principle. The time intervals were 102 s–123 s, 123 s–150 s, and 150 s–186 s, respectively. Then, in each time interval, the CFD simulation boundary and initial conditions were input using fitted polynomials based on the system analysis results. Figures
Fitted boundary conditions from 102 s to 123 s.
Inlet flow rate
Inlet temperature
Pressure
Injection flow rate
Fitted boundary conditions from 123 s to 150 s.
Inlet flow rate
Inlet temperature
Pressure
Injection flow rate
Fitted boundary conditions from 150 s to 186 s.
Inlet flow rate
Inlet temperature
Pressure
Injection flow rate
Based on the above boundary conditions, mathematical models, the detailed solid region temperature distribution characteristics of RPV wall were studied in this section. Figures
Temperature distribution of RPV wall at 102 s.
Temperature distribution of RPV wall at 103 s.
Figure
Temperature distribution of RPV wall at 110 s.
Temperature distribution of the RPV wall at 150 s.
Temperature distribution of the RPV wall at 185 s.
Figure
Temperature distribution of the T-junction in the first phase.
102 s
104 s
As could be seen from Figure
Temperature distribution of the T-junction in the second phase.
109 s
111 s
123 s
In this work, the multiscale method was implemented for the Chashma NPP thermal hydraulic study under the inadvertent Safety Injection System operation scenario. A fine model of primary loop and Safety Injection System was established using system analysis code RELAP5 and then the transient scenario was simulated. The transient variations of important thermal hydraulic parameters were achieved and investigated. The injected coolant mass flow rate increased sharply and then dropped gradually due to the increase of primary loop pressure. The maximum injection flow rate from Safety Injection System was about 13 kg/s during this scenario.
The detailed primary loop inlet nozzles and RPV geometries were established using three-dimensional method. The one-dimensional safety analysis results were fitted and imported to the CFD code as the initial and boundary conditions. Based on this model, the variations of RPV wall temperature distribution with time were achieved under the inadvertent Safety Injection System operation scenario for Chashma NPP. The maximum temperature difference was about 80 K and the most critical region was marked. The detailed three-dimensional thermal hydraulic simulation results provide the necessary input conditions for three-dimensional stress analysis for PTS study.
The multiscale simulation method could provide an effective and precise approach for the PWR PTS thermal hydraulic study. It is of great significance and meaningful for the reactor key equipment structure assessment and lifetime fatigue research.
The authors declare that they have no conflicts of interest.
This research has been supported by National Natural Science Foundation of China (no. 11705139).