scholarly journals Discharge header design inside a reactor pool for flow stability in a research reactor

2020 ◽  
Vol 52 (10) ◽  
pp. 2204-2220
Author(s):  
Hyungi Yoon ◽  
Yongseok Choi ◽  
Kyoungwoo Seo ◽  
Seonghoon Kim
2006 ◽  
Author(s):  
Ping-Rey Jang ◽  
Rangaswami Arunkumar ◽  
Teresa Leone ◽  
Zhiling Long ◽  
Melissa A. Mott ◽  
...  

2021 ◽  
Vol 253 ◽  
pp. 04001
Author(s):  
Maciej Lipka ◽  
Anna Talarowska ◽  
Grzegorz Wojtania ◽  
Marek Migdal

Materials and core components for the next generation power reactors technologies require testing that can be performed in existing research reactors. Such experiments employ devices dedicated to reflect the relevant thermal and neutron parameters simulating conditions present in, for example, but not limited to, HTGR reactors. A novel thermostatic irradiation device named ISHTAR (Irradiation System for High-Temperature Reactors) has been designed and constructed in the MARIA research reactor. Its mission is to enable irradiation of samples in controlled, homogeneous temperature field reaching 1000°C and inert gas atmosphere. The high temperature is achieved by a combination of electric and gamma heating, together with carefully designed thermal insulation. Additionally, samples holder made of graphite with high thermal conductivity enables the temperature homogenization in all directions. Device will be placed inside the Beryllium matrix of MARIA core and cooled with forced circulation of water from the reactor pool loop. This paper presents the outcome of experiments conducted with the rig prototype in external hydraulic mock-up of the MARIA reactor irradiation channel. The results have proved that the desired conditions for irradiation of the samples were achieved and their comparison against computational data has shown the adequacy of the design process. Finally, the loss of flow scenario was tested in protected and unprotected conditions (meaning with and without the safety system based on temperature feedback), proving the operational safety of the ISHTAR design. Experimental results will be used in the future to validate the numerical models (two and three dimensional) of the irradiation rig, providing an improved understanding of free convection and radiation phenomena modeling.


Author(s):  
Young-Chul Park

During an open-pool-type research reactor operation, it is necessary to access the pool top area for un/loading irradiation test pieces by a required irradiation period. However, when the reactor pool top radiation level exceeds the limit of radiation level by the rising of reactor chimney water contaminated by radioactivity due to a natural convection of the pool water, access the reactor pool top area is denied due to the high radiation level. In the case of HANARO, a hot-water layer (HWL, hereinafter) is maintained below a depth of 1.2 m from the top of the reactor pool in order to reduce the radiation level of the reactor pool top area. After a normal operation of the HWL, the pool top radiation level is safely maintained below the limit of the pool top radiation level. For studying more the characteristics of the HWL under a reactor coolant downward flow condition, The HWL heat loss is calculated based on the HANARO HWL calculation model. The HWL heat loss characteristics were reviewed by variations of the HWL temperature, reactor core coolant flow direction, and reactor power. It was confirmed through the results that the HWL heat loss under a reactor coolant downward flow condition was increased by about 20% to 60% over that under a reactor coolant upward flow condition, as per the HWL temperature variation. It was the reason that the HWL bottom convection heat loss was increased by the higher flow rate under a reactor coolant downward flow condition than that under a reactor coolant in an upward flow condition.


2013 ◽  
Vol 255 ◽  
pp. 28-37 ◽  
Author(s):  
Soon Ho Kang ◽  
Ho Seon Ahn ◽  
Ji Min Kim ◽  
Hyeong Min Joo ◽  
Kwon-Yeong Lee ◽  
...  

2005 ◽  
Vol 20 (2) ◽  
pp. 3-9 ◽  
Author(s):  
Ahmed Khedr ◽  
Francesco d’Auria

The present work deals with the anal y sis of RELAP5 results obtained from the evaluation study of the total loss of flow transient with the deficiency of the heat removal system in a research reactor using two different nodalizations. It focuses on the effect of nodalization on the thermal-hydraulic evaluation of the re search reactor. The analysis of RELAP5 results has shown that nodalization has a big effect on the predicted scenario of the postulated transient. There fore, great care should be taken during the nodalization of the reactor, especially when the avail able experimental or measured data are insufficient for making a complete qualification of the nodalization. Our analysis also shows that the research reactor pool simulation has a great effect on the evaluation of natural circulation flow and on other thermal-hydraulic parameters during the loss of flow transient. For example, the on set time of core boiling changes from less than 2000 s to 15000 s, starting from the beginning of the transient. This occurs if the pool is simulated by two vertical volumes in stead of one vertical volume.


Author(s):  
Ping-Rey Jang ◽  
Rangaswami Arunkumar ◽  
Jeffrey S. Lindner ◽  
Zhiling Long ◽  
Melissa A. Mott ◽  
...  

The Oak Ridge Research Reactor (ORRR) was operated as an isotope production and irradiation facility from March 1958 until March 1987. The US Department of Energy permanently shut down and removed the fuel from the ORRR in 1987. The water level must be maintained in the ORRR pool as shielding for radioactive components still located in the pool. The U.S. Department of Energy’s Office of Environmental Management (DOE EM) needs to decontaminate and demolish the ORRR as part of the Oak Ridge cleanup program. In February 2004, increased pit corrosion was noted in the pool’s 6-mm (1/4")-thick aluminum liner in the section nearest where the radioactive components are stored. If pit corrosion has significantly penetrated the aluminum liner, then DOE EM must accelerate its decontaminating and decommissioning (D&D) efforts or look for alternatives for shielding the irradiated components. The goal of Mississippi State University’s Institute for Clean Energy Technology (ICET) was to provide a determination of the extent and depth of corrosion and to conduct thermodynamic modeling to determine how further corrosion can be inhibited. Results from the work will facilitate ORNL in making reliable disposition decisions. ICET’s inspection approach was to quantitatively estimate the amount of corrosion by using Fouriertransform profilometry (FTP). FTP is a non-contact 3-D shape measurement technique. By projecting a fringe pattern onto a target surface and observing its deformation due to surface irregularities from a different view angle, the system is capable of determining the height (depth) distribution of the target surface, thus reproducing the profile of the target accurately. ICET has previously demonstrated that its FTP system can quantitatively estimate the volume and depth of removed and residual material to high accuracy. The results of our successful initial deployment of a submergible FTP system into the ORRR pool are reported here as are initial thermodynamic modeling results.


2021 ◽  
Vol 8 (4) ◽  
pp. 1-9
Author(s):  
Duc Tu Dau ◽  
Minh Tuan Nguyen ◽  
Vinh Vinh Le ◽  
Ton Nghiem Huynh ◽  
Cuong Nguyen Kien ◽  
...  

The leakage from the reactor pool back into the dry irradiation channels due to corrosion or mechanics based reason is a postulated event that could occur under operating conditions of the Dalat nuclear research reactor (DNRR), especially the channel 7-1 which has been installed more than 30 years. When it occurs, the air space in these channels will be occupied by the water, subsequently a water column will appear in fuel region. The appearance of water column considerably enhances medium of neutron moderation for its surrounding fuel assemblies. As a result, a positive reactivity is inserted in the core and this event is classified as an insertion of excess reactivity. This event needs to be addressed by analysis and assessment from safety point of view and the results of analysis are also important for updating the reactor operating procedures. This paper presents assumptions, computer models and the results of analysis for such event in the DNRR by using MCNP5 code (code for neutronics analysis) and EUREKA-2/RR code (code for transient analysis). The calculation results include value of reactivity insertion, change in power of reactor, as well as surface temperature of the hottest fuel assembly. This research contributes to updating the reactor operating procedure.


2015 ◽  
Vol 2015 ◽  
pp. 1-7 ◽  
Author(s):  
Kwon-Yeong Lee ◽  
Hyun-Gi Yoon

In an open-pool type research reactor, the primary cooling system can be designed to have a downward flow inside the core during normal operation because of the plate type fuel geometry. There is a flow inversion inside the core from the downward flow by the inertia force of the primary coolant to the upward flow by the natural circulation when the pump is turned off. To delay the flow inversion time, an innovative passive system with pump flywheel and GCCT is developed to remove the residual heat. Before the primary cooling pump starts up, the water level of the GCCT is the same as that of the reactor pool. During the primary cooling pump operation, the water in the GCCT is moved into the reactor pool because of the pump suction head. After the pump stops, the potential head generates a downward flow inside the core by moving the water from the reactor pool to the GCCT and removes the residual heat. When the water levels of the two pools are the same again, the core flow has an inversion of the flow direction, and natural circulation is developed through the flap valves.


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