High-Pressure Steam-Driven Jet Pump—Part I: Mathematical Modeling

2000 ◽  
Vol 123 (3) ◽  
pp. 693-700 ◽  
Author(s):  
N. Beithou ◽  
H. S. Aybar

There are several proposed advanced reactor systems, which consider the utilization of a steam-driven jet pump (SDJP) as an emergency core cooling system. The steam-driven jet pump is a device without moving parts, in which steam is used as an energy source to pump cold water from a pressure much lower than the steam pressure to a pressure higher than the steam pressure. In this study, the mathematical modeling of the SDJP has been done. An experimental analysis of the high-pressure SDJP has been reported by Cattadori et al. The results of the mathematical modeling of the SDJP have been compared with Cattadori’s experimental results. The comparisons show that the experimental and calculated pressure distributions are in good qualitative agreement. A parametric analysis of the SDJP is ongoing.

2000 ◽  
Vol 123 (3) ◽  
pp. 701-706 ◽  
Author(s):  
N. Beithou ◽  
H. S. Aybar

The steam-driven jet pump (SDJP) is a device without moving parts, in which steam is used as an energy source to pump cold water from a pressure much lower than the steam pressure to a pressure higher than the steam pressure. In the previous part of this study, the mathematical modeling of the SDJP has been done, and reported. The results of the mathematical modeling of the SDJP have been compared with Cattadori’s experimental results. The comparisons show that the experimental and calculated pressure distributions are in good qualitative agreement. For the same steam inlet pressure of 8.7 MPa, the discharge pressures in the experiment and in the simulation are 9.8 MPa and 9.54 MPa, respectively. The relative difference is two percent. It can be said that the computed discharge pressure is in good agreement with the experimental result. In the current study, a parametric analysis of the SDJP has been done in terms of four independent parameters: steam inlet pressure and temperature, supply water pressure, and temperature. The output parameters are: discharge pressure, temperature, and mass flow rate. As a result of this parametric study, the operation characteristics of the SDJP have been obtained.


2017 ◽  
Vol 370 ◽  
pp. 162-170 ◽  
Author(s):  
Luis Héctor Hernández-Gómez ◽  
Brayan Leonardo Pérez-Escobar ◽  
Juan Alfonso Beltrán-Fernández ◽  
Juan Alejandro Flores-Campos ◽  
Salatiel Pérez-Montejo ◽  
...  

In this paper, the Cumulative Usage Factor (CUF) of a High Pressure Core Cooling System (HPCS) reactor nozzle of a Boiling Water Reactor was calculated. This fatigue damage has been caused by the sudden injection of cold water into the reactor vessel through such nozzle. For this purpose, a three-dimensional analysis was carried out. Accordingly, a transient heat transfer analysis was developed. The temperature distribution was determined. With this information, the stress analysis was carried out. The safe end was restricted to move along its axial direction and the forging end was free to expand axially and radially. The resultant stress field established the magnitude of the alternative stresses. In the last step, a fatigue analysis was developed. The most critical point is the junction of the nozzle with the thermal sleeve. The fatigue performance was evaluated during a period of sixty years. It was assumed that 1.5 cycles per year will take place. The fatigue curves of ASME code section III were used. The results showed that the Cumulative Usage Factor (CUF) vary with the temperature injection, being 0.4090 when the water injected was 4.44°C and 0.3797 when the water temperature was 37.77°C. Both of them were estimated for a period of 60 years of operation. Therefore, damage is reduced as the temperature of the injected water increases. Besides, it is advisable to at least follow the recommendations of the NUREG ́s 1800 and 1801 [1, 2]. In this way, the aging of the nozzle is adequately managed.


Author(s):  
F Bakhtar ◽  
K Zidi

The paper describes the results of an experimental investigation of limiting supersaturation in high-pressure steam. It follows an earlier investigation and to avoid the uncertainties associated with leakage past sliding profiles, the test section has been redesigned and the measurements taken with fixed nozzles. Three convergent-divergent nozzles with nominal rates of expansion of 3000, 5000 and 10000 per second have been used and the inlet stagnation pressures cover the range 25–35 bar. The data consist mainly of axial pressure distributions but some droplet measurements have also been recorded.


Author(s):  
Vinesh H. Gada ◽  
Mohit P. Tandon ◽  
Jebin Elias ◽  
Andrew Splawski ◽  
Simon Lo

The Large Scale Interface (LSI) model of the Euler-Euler method in STAR-CCM+ is extended to simulate two-phase flow with phase change. This extended methodology is used to simulate direct contact condensation (DCC) of steam in a hot leg when cold water is injected by emergency core cooling system to remove the residual heat. The case corresponds an experimental study conducted at Hungarian Atomic Energy Research Institute KFKI using the PMK-2 device. Out of the several experiments reported for this scenario, the one experiment considered in this work corresponds to a case without the water hammer phenomena. It was found that the LSI model is able to capture core physics of direct contact condensation during steam-water counter-current flow in a pipe. The model could capture entrapment of steam between the interface and its subsequent rapid condensation. The role of the relaxation time-scale of the large interface drag and the turbulence damping at interface is also studied.


Author(s):  
Liu Yang ◽  
Chao Wang ◽  
Jian Zhang ◽  
Ronghua Lu ◽  
Xinhai Yu

In this study, the complete dynamic performance of the high temperature and high pressure steam pressure relief valve (HTHP PRV) from pop up to reseating was simulated by CFD software which combined with moving mesh capabilities and multiple domains. An experimental setup was established for the testing of HTHP PRV in accordance with the standard of ASME PTC 25. The dynamic performance of HTHP PRV was recorded accurately. For the transient simulation of HTHP PRV, a domain with opening boundaries connected to the outlet of PRV was proposed to avoid the direct definition of the pressure at the PRV outlet and handle the critical flow. It also can describe the surrounding flow field and help us to understand the influence of the PRV discharge on the environment better. The simulation results were verified by experimental ones. The resultant force on the disk and the lift were monitored and analyzed. A detailed contour of the compressible steam flowing through the HTHP PRV was obtained, including small scale flow features in the back pressure chamber. The effect of the adjusting sleeve on the dynamic performance of HTHP PRV was also investigated in details. The blowdown increases linearly by 0.163% with the adjusting sleeve moves by each millimeter in the direction of departing from the disk. This study sheds a light of understanding of the dynamic characteristics of HTHP PRV.


Author(s):  
Yabing Li ◽  
Lili Tong ◽  
Xuewu Cao

For advanced passive PWR, reactor coolant system (RCS) depressurization through automatic depressurization system (ADS) is an important measurement to avoid high-pressure melt ejection and direct containment heating. It allows injection from passive core cooling system and the implement of in-vessel retention. However, it has negative impact that hydrogen in the RCS can be released to the containment together with coolant, which may lead to hydrogen burning or even explosion in the containment. Therefore, this paper analyzes the RCS depressurization strategy during severe accident, and evaluates its negative impact. Severe accident sequences induced by station black out (SBO) was selected and analyzed with integral severe accident analysis code as a typical high pressure core melt accident scenario. Different depressurization strategies with ADS system were discussed based on Severe Accident Management Guideline (SAMG.) ADS valves were manually opened at a core exit temperature of 923 K with 20min delay for operator reaction. Both depressurization effect and hydrogen risk were evaluated for different strategies. Hydrogen distribution was calculated, which was used to determine the combustion mode in different compartments. Result shows all three strategies analyzed in this paper can depressurize the RCS effectively. And opening the ADS stage 1–3 valves causes rapidly increase of the hydrogen concentration in the in-containment refueling water storage tank (IRWST) compartment and may lead to hydrogen denotation. However, hydrogen can be well dispersed in the loop compartment with intentional open of ADS stage 4 valves to RCS depressurization. Therefore, suggestions are proposed for SAMG: implement RCS depressurization strategy with stage 4 ADS instead of ADS stage 1–3.


Author(s):  
Michitsugu Mori ◽  
Tadashi Narabayashi ◽  
Shuichi Ohmori ◽  
Fumitoshi Watanabe

A Steam Injector (SI) is a simple, compact, passive pump which also functions as a high-performance direct-contact compact heater. We are developing this innovative concept by applying the SI system to core injection systems in Emergency Core Cooling Systems (ECCS) to further improve the safety of nuclear power plants. Passive ECCS in nuclear power plants would be inherently very safe and would prevent serious accidents by keeping the core covered with water (Severe Accident-Free Concept). The Passive Core Coolant Injection System driven by a high-efficiency SI is one that, in an accident such as a loss of coolant accident (LOCA), attains a higher discharge pressure than the supply steam pressure used to inject water into the reactor by operating the SI using water stored in the pool as the water supply source and steam contained in the reactor as the source of pressurization energy. The passive SI equipment would replace large, rotating machines such as pumps and motors, so eliminating the possibility of such equipment failing. In this Si-driven Passive Core Coolant Injection System (SI-PCIS), redundancy will be provided to ensure that the water and steam supply valves to the SI open reliably, and when the valves open, the SI will automatically start to inject water into the core to keep the core covered with water. The SI used in SI-PCIS works for a range of steam pressure conditions, from atmosphere pressure through to high pressures, as confirmed by analytical simulations which were done based on comprehensive experimental data obtained using reduced scale SI. We did further simulations and evaluations of the core cooling and coolant injection performance of SI-PCIS in BWR using RETRAN-3D code, developed using EPRI and other utilities, for the case of small LOCA. Reactors equipped with passive safety systems — the gravity-driven core cooling/injection system (GDCS) and depressurization valves (DPV) — would inevitably end up having large LOCA, even if they are initially small LOCA, as depressurization valves are forcibly opened in order to inject coolant from the GDCS pool to the GDCS water head at up to ∼0.2MPa. On the other hand, our simulation demonstrated that SI-PCIS could prevent large LOCA occurring in reactors by having by coolant discharged into the core in the event of small LOCA or when DPV unexpectedly open.


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