Experimental Demonstration of Decay Heat Removal by Submerged Feeders in a Full-Scale Test Facility of a Natural Circulation Boiling Water Reactor

2019 ◽  
Vol 5 (4) ◽  
Author(s):  
A. K. Nayak ◽  
Mukesh Kumar ◽  
Sumit V. Prasad ◽  
V. Jain ◽  
D. K. Chandraker
Keyword(s):  
Natural Circulation ◽  
Heat Removal ◽  
Test Facility ◽  
Full Scale Test ◽  
Water Pool ◽  
Water Reactor ◽  
Decay Heat ◽  
The Core ◽  
Lower Elevation ◽  
Scale Test

Removal of decay heat with nonavailability of active systems is a safety issue especially during station blackout (SBO) in a light water reactor. Passive systems are being incorporated in the new designs of nuclear reactors for this purpose. Some of the advanced reactors such as Indian advanced heavy water reactor (AHWR) have dedicated isolation condensers (ICs) which are submerged in large water pool called gravity driven water pool (GDWP). These ICs remove decay heat from the core by natural circulation cooling and dissipate it to the GDWP by natural convection. There is a concern that cracks may develop in the GDWP if a large seismic event similar to Fukushima type occurs. In that case, the pool water is lost and it can threaten the core coolability because of loss of heat sink. In AHWR, the cracks in the water pool leads to the relocation of the water of the pool to the reactor cavity. Feeders of AHWR are positioned in the reactor cavity. Thus, the water relocated in the cavity, will eventually submerge the feeders and these submerged feeders have the potential to remove the decay heat of the core. However, the feeders are located at a lower elevation as compared to the core, and hence, there is concern on the heat removal capability by the submerged feeders by natural convection. To understand this aspect and to establish the core coolability under the above-mentioned conditions, experiments were performed in a full-scale test facility of AHWR. Experiments showed that the decay heat can be safely removed in natural circulation mode of cooling with heat sink located at lower elevation than the heat source.

Study on Cooling Process in a Reactor Vessel of Sodium-Cooled Fast Reactor Under Severe Accident: Velocity Measurement Experiments Simulating Operation of Decay Heat Removal Systems

2020 ◽  
Author(s):  
Mitsuyo Tsuji ◽  
Kosuke Aizawa ◽  
Jun Kobayashi ◽  
Akikazu Kurihara ◽  
Yasuhiro Miyake
Keyword(s):  
Natural Circulation ◽  
Heat Removal ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Decay Heat ◽  
The Core ◽  
Severe Accidents ◽  
Piv Measurement ◽  
Image Velocimetry ◽  
Experimental Apparatus

Abstract In Sodium-cooled Fast Reactors (SFRs), it is important to optimize the design and operate decay heat removal systems for safety enhancement against severe accidents which could lead to core melting. It is necessary to remove the decay heat from the molten fuel which relocated in the reactor vessel after the severe accident. Thus, the water experiments using a 1/10 scale experimental apparatus (PHEASANT) simulating the reactor vessel of SFR were conducted to investigate the natural circulation phenomena in a reactor vessel. In this paper, the natural circulation flow field in the reactor vessel was measured by the Particle Image Velocimetry (PIV) method. The PIV measurement was carried out under the operation of the dipped-type direct heat exchanger (DHX) installed in the upper plenum when 20% of the core fuel fell to the lower plenum and accumulated on the core catcher. From the results of PIV measurement, it was quantitatively confirmed that the upward flow occurred at the center region of the lower and the upper plenums. In addition, the downward flows were confirmed near the reactor vessel wall in the upper plenum and through outermost layer of the simulated core in the lower plenum. Moreover, the relationship between the temperature field and the velocity field was investigated in order to understand the natural circulation phenomenon in the reactor vessel. From the above results, it was confirmed that the natural circulation cooling path was established under the dipped-type DHX operation.

Study on the Reliability Improvement of a Mechanical Seal for the Primary Loop Recirculation Pumps (PLR Pump)

2013 ◽  
Author(s):  
Naoto Hikida ◽  
Hiroyuki Torigoe ◽  
Kikuo Nishiyama ◽  
Kenichi Sugiyama
Keyword(s):  
Test Facility ◽  
Mechanical Seal ◽  
Full Scale Test ◽  
Shaft Seal ◽  
Primary Loop ◽  
Water Reactor ◽  
Reliability Improvement ◽  
Operation Pattern ◽  
Ceramic Property ◽  
Scale Test

Primary Loop Recirculation pump (PLR pump) for Boiling Water Reactor (BWR) uses a mechanical seal as a shaft seal device, which the seal material is made of a ceramic. Recently, a new ceramic suitable as a seal material has been developed. We will explain about the introduction of the material to a mechanical seal for the PLR pump. First, we checked the material characteristics of the ceramic. Next, we carried out the abrasion acceleration test at the mechanical seal test facility for PLR pump, and performed an analysis of seal face deformation for the applied ceramic. Then, we performed a full scale test that simulated the operation pattern of the PLR pump. As a result, we confirmed that the new ceramic property suits for the use as a mechanical seal for the PLR pump.

scholarly journals CFD Analysis of the Passive Decay Heat Removal System of an LBE-Cooled Fast Reactor

2018 ◽  
Vol 2018 ◽  
pp. 1-11
Author(s):  
Jiarun Mao ◽  
Lei Song ◽  
Yuhao Liu ◽  
Jiming Lin ◽  
Shanfang Huang ◽  
...  
Keyword(s):  
Heat Exchange ◽  
Fast Reactor ◽  
Natural Circulation ◽  
Heat Removal ◽  
Engineering Project ◽  
Decay Heat ◽  
The Core ◽  
Main Vessel ◽  
Heat Removal System ◽  
Removal System

This paper presents capacity of the passive decay heat removal system (DHRS) operated under the natural circulation conditions to remove decay heat inside the main vessel of the Lead-bismuth eutectic cooled Fast Reactor (LFR). The motivation of this research is to improve the inherent safety of the LFR based on the China Accelerator Driven System (ADS) engineering project. Usually the plant is damaged due to the failure of the main pumps and the main heat exchangers under the Station Blackout (SBO). To prevent this accident, we proposed the DHRS based on the diathermic oil cooling for the LFR. The behavior of the DHRS and the plant was simulated using the CFD code STAR CCM+ using LFR with DHRS. The purpose of this analysis is to evaluate the heat exchange capacity of the DHRS and is to provide the reference for structural improvement and experimental design. The results show that the stable natural circulations are established in both the main vessel and the DHRS. During the decay process, the heat exchange power is above the core decay heat power. In addition, in-core decay heat and heat storage inside the main vessel are efficiently removed. All the thermal-hydraulics parameters are within a safe range. Moreover, the highest temperature occurs at the upper surface of the core. A swirl occurs at the corner of the lateral core surface and some improvements should be considered. And the natural circulation driving force can be further increased by reducing the loop resistance or increasing the natural circulation height based on the present design scenario to enhance the heat exchange effect.

Generic Experimental Investigations of Thermohydraulic Instabilities With Void Fraction Measurement at Natural Circulation Test Facility Geneva

2014 ◽  
Author(s):  
Tim Cloppenborg ◽  
Christoph Schuster ◽  
Antonio Hurtado
Keyword(s):  
Void Fraction ◽  
Power Plants ◽  
Natural Circulation ◽  
Saturation Temperature ◽  
Heat Removal ◽  
Open Loop ◽  
Test Facility ◽  
Experimental Investigations ◽  
Decay Heat ◽  
System Pressure

Passive systems like natural circulation (NC) loops can offer reliable and cost efficient alternatives to common active systems for decay heat removal in nuclear power plants. During the transition between stable single and stable two phase flows, instabilities e. g. flashing and geysering may occur in the riser due to low system pressure and saturation temperature conditions. These instabilities may cause severe stress to the system components. This paper presented some results of the study on the decay heat removal system based on natural circulation, performed on the open loop NC test facility GENEVA, built at TU Dresden in 2013. 16 probes were used to determine void fraction along the riser on nine different levels in high time and spatial resolution, and stability maps was created for riser with inner diameters of 20 mm and 38 mm and up to 85 kW evaporator power.

Performance Assessment of Sodium to Air Finned Heat Exchanger for FBR

2006 ◽  
Author(s):  
V. Vinod ◽  
V. A. Suresh Kumar ◽  
I. B. Noushad ◽  
T. R. Ellappan ◽  
K. K. Rajan ◽  
...  
Keyword(s):  
Heat Exchanger ◽  
Cross Flow ◽  
Heat Removal ◽  
High Elevation ◽  
Test Facility ◽  
Finned Tubes ◽  
Decay Heat ◽  
The Core ◽  
Flow Heat ◽  
Air Flows

In pool type Fast Breeder Reactors (FBR) a passive Safety Grade Decay Heat Removal (SGDHR) system removes decay heat produced in the core when normal heat removal path through steam water system is not available. This is essential to maintain the core temperatures within limits. A Decay Heat Exchanger (DHX) picks the heat from the pool and transfers the heat to atmosphere through sodium to Air Heat Exchanger (AHX) situated at high elevation. Due to the temperature differences existent in the system density differences are generated causing a buoyant convective heat transfer. The system is completely passive as primary sodium, secondary sodium and air flows under natural convection. DHX is a sodium to sodium counter flow heat exchanger with primary sodium on shell side and secondary sodium on tube side. AHX is a cross flow heat exchanger with sodium on tube side and air flows in cross flow across the finned tubes. Capacity of a single loop of SGDHR is 8MW. Four such loops are available for the decay heat removal. It has been seen that the decay heat removal to a large extent depends on the AHX performance. AHX tested have shown reduced heat removal capacity much as 30 to 40%, essentially due to the bypassing of the finned tubes by the air. It was felt that a geometrically similar AHX be tested in sodium. Towards this a 2MW Sodium to air heat exchanger (AHX) was tested in the Steam Generator Test Facility (SGTF) constructed at Indira Gandhi Center for Atomic Research (IGCAR), Kalpakkam. The casing arrangement of the AHX was designed to minimise bypassing of air.

Experimental Demonstration of Safety During Extended Station Blackout in an Integral Test Loop of a Natural Circulation Boiling Water Reactor

2019 ◽  
Vol 5 (4) ◽  
Author(s):  
A. K. Nayak ◽  
Mukesh Kumar ◽  
A. K. Vishnoi ◽  
Vikas Jain ◽  
D. K. Chandraker
Keyword(s):  
Safety Concern ◽  
Stress Test ◽  
Heat Removal ◽  
Test Facility ◽  
Integral Test ◽  
Water Reactor ◽  
Decay Heat ◽  
Safety Margins ◽  
Test Loop ◽  
Station Blackout

Decay heat removal for prolonged period of station blackout (SBO) is a safety concern of the nuclear reactors. Aftermath of Fukushima, safety evaluation (performance under severe conditions: stress test) of the reactors was carried out worldwide. It includes establishment of grace period of the reactors. Similar exercises for advanced heavy water reactor (AHWR) were also performed and the design of AHWR was established for its robustness against such events. Decay heat removal during extended SBO is such a condition to be qualified. In this regard, experiments in the integral test loop (ITL), a full scale test facility of AHWR, were conducted for continuous 7 days of extended SBO. Experiment was started with 6.8 MPa as the initial reactor pressure and decay heat removal was demonstrated for 7 days of SBO by passive means. It is observed that the pressure falls down to 1 MPa in 3 h. The design of AHWR was evaluated from safety critical aspects during such an event experimentally. During this event, the clad surface temperature was found to be well within safe limits of operations. As a result of this experiment, it can be concluded that the design of AHWR is capable to remove decay heat for 7 days of SBO with sufficient safety margins.

scholarly journals Effect of Coolant Inventories and Parallel Loop Interconnections on the Natural Circulation in Various Heat Transport Systems of a Nuclear Power Plant during Station Blackout

2008 ◽  
Vol 2008 ◽  
pp. 1-11 ◽  
Author(s):  
Avinash J. Gaikwad ◽  
P. K. Vijayan ◽  
Sharad Bhartya ◽  
Kannan Iyer ◽  
Rajesh Kumar ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Heat Transport ◽  
Nuclear Power ◽  
Natural Circulation ◽  
Heat Removal ◽  
Transport Systems ◽  
Critical Examination ◽  
Decay Heat ◽  
Station Blackout

Provision of passive means to reactor core decay heat removal enhances the nuclear power plant (NPP) safety and availability. In the earlier Indian pressurised heavy water reactors (IPHWRs), like the 220 MWe and the 540 MWe, crash cooldown from the steam generators (SGs) is resorted to mitigate consequences of station blackout (SBO). In the 700 MWe PHWR currently being designed an additional passive decay heat removal (PDHR) system is also incorporated to condense the steam generated in the boilers during a SBO. The sustainability of natural circulation in the various heat transport systems (i.e., primary heat transport (PHT), SGs, and PDHRs) under station blackout depends on the corresponding system's coolant inventories and the coolant circuit configurations (i.e., parallel paths and interconnections). On the primary side, the interconnection between the two primary loops plays an important role to sustain the natural circulation heat removal. On the secondary side, the steam lines interconnections and the initial inventory in the SGs prior to cooldown, that is, hooking up of the PDHRs are very important. This paper attempts to open up discussions on the concept and the core issues associated with passive systems which can provide continued heat sink during such accident scenarios. The discussions would include the criteria for design, and performance of such concepts already implemented and proposes schemes to be implemented in the proposed 700 MWe IPHWR. The designer feedbacks generated, and critical examination of performance analysis results for the added passive system to the existing generation II & III reactors will help ascertaining that these safety systems/inventories in fact perform in sustaining decay heat removal and augmenting safety.

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