Preliminary computational and experimental design studies of the ISHTAR thermostatic rig for the high-temperature reactors materials irradiation

The Irradiation System for High-Temperature Reactors (ISHTAR) thermostatic rig will be used to irradiate advanced core material samples in conditions corresponding to those prevailing in the high-temperature reactors (HTRs): these conditions include a stable temperature extending up to 1000°C in the helium atmosphere. Computational and experimental studies concerning the design have been conducted, proving the possibility of these conditions’ fulfillment inside the rig while maintaining the safety limits for MARIA research reactor. The outcome is the thermostatic rig design that will be implemented in the MARIA reactor. Appropriate irradiation temperature will be achieved by a combination of electric heating with the control system, gamma heating, and a helium insulation gap with precisely designed thickness. The ISHTAR rig will be placed inside the vertical irradiation channel, which is located in the reactor pool. The device is being developed from scratch at the Nuclear Facilities Operation Department of the National Centre for Nuclear Research as a part of the GOSPOSTRATEG programme.

Reactivity induced transient analysis when the occurrence of leakage in the dry irradiation channels of the Dalat Nuclear Research Reactor

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.

Experimental evaluation of natural convection in the IPR-R1 Triga research reactor at 264 kW and 105 kW

In order to study the safety aspects connected with the permanent increase of the maximum steady state power of the IPR-R1 Triga Reactor of the Nuclear Technology Development Center (CDTN), experimental measurements were done with the reactor operating at power levels of 265 kW and 105 kW, with the pool forced cooling system turned off. A number of parameters were measured in real-time such as fuel and water temperatures, radiation levels, reactivity, and influence of cooling system. Information on all aspects of reactor operation was displayed on the Data Acquisition System (DAS) shown the IPR-R1 online performance. The DAS was developed to monitor and record all operational parameters. Information displayed on the monitor was recorded on hard disk in a historical database. This paper summarizes the behavior of some operational parameters, and in particular, the evolution of the temperature in the fuel element centerline positioned in the core hottest location. The natural circulation test was performed to confirm the cooling capability of the natural convection in the IPR-R1 reactor. It was confirmed that the IPR-R1 has capability of long-term core cooling by natural circulation operating at 265 kW. The measured maximum fuel temperature of about 300 oC was lower than the operating limit of 550 oC. It has been proven that without cooling in the primary the gamma dose rate above reactor pool at high power levels was rather high.

Experimental study of ISHTAR thermostatic irradiation device for the MARIA research reactor

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.

Experimental Study on Natural Convection Heat Transfer Outside Tube Bundle in Space Under Low Temperature Difference

The Passive Residual Heat Removal System (PRHRS) is very important for the safety of the heating reactor after shutdown. PRHRS is a natural circulation system driven by density difference, therefore, the heat transfer performance of the Passive Residual Heat Removal Heat Exchanger (PRHR HX) has a great impact to the heat transfer efficiency of PRHRS. However, the most research object of PRHR HX is the C-shape heat exchanger at present, which located in In-containment Refueling Water Storage Tank (IRWST). This heat exchanger is mainly used for the PRHRS of nuclear power plants. In the swimming pool-type low-temperature heating reactor (SPLTHR), the PRHR HX is placed in the reactor pool, which the pressure and temperature of the reactor pool are relatively low, and the outside heat transfer mode of tube bundle is mainly natural convection heat transfer. In this study, a miniaturized single-phase pool water cooling system was built to investigate the natural convective heat transfer coefficient of the heat exchanger under the large space and low temperature conditions. The experimental data had been compared with several correlations. The results show that the predicted value of Yang correlation is the closest to the experimental data, which the maximum deviation is about 11%.

RELAP5-3D Three-Dimensional Analysis Based on PHÉNIX Dissymmetric Transient Test

Before the final shutdown of the PHÉNIX fast reactor, the CEA carried out a final set of experimental tests to gather data and additional knowledge on relevant sodium fast reactors (SFR) operation and safety aspects. One of these experiments conducted was the dissymmetrical configuration test, which was selected as benchmark transient within the H2020 SESAME project. ENEA and Sapienza University of Rome are participating in the benchmark using the RELAP5-3D© code. The thermal hydraulic analysis focuses on adequate core cooling prediction in accidental scenario. With the goal of investigating asymmetric thermal hydraulic behavior inside of the reactor pool, two different nodalization approaches have been applied for the RELAP5-3D model, which adopt the same geometrical scheme for the primary flow path, with the exception of the hot and cold pools and the core bypass. The first scheme has been developed using vertical parallel pipes with cross junctions for the hot and cold pools and an equivalent pipe to reproduce the core bypass. The second model includes a multidimensional (MULTID) component, which simulates the pools and provides a detailed nodalization of the core bypass. This study aims at assessing whether the two modeling approaches are equally capable to predict the asymmetrical temperature evolution over the test, caused by the azimuthal asymmetry of the boundary conditions. Blind calculation results are presented and discussed. The paper will be a first step toward the RELAP5-3D code assessment against the experimental results collected as part of the PHÉNIX dissymmetric test.

Analysis of cable effect on dynamic motion of an underwater vehicle for welding of reaction pool

Remotely operated vehicle is a reliable tool in an emergency rescue and routing inspection of a reactor pool. In practice, a cable has been considered as an important part of a vehicle system (i.e. winch, vehicle, and cable) to evaluate and predict the mechanical characteristics, and this article presents a study on a dynamic mechanism modeling of a cable partially in water and air based on an omnidirectional motion, and a numerical simulation is employed. In this work, we programmed the model governed by a partial differential equation set, at the discrete time node which was transformed into an ordinary differential equation set regarded as an initial value problem. The dynamic mechanical characteristics of the lower endpoint (i.e. a connection point between vehicle and cable) and the upper endpoint (i.e. a connection point between cable and winch) were, respectively, quantified with acceleration and a compounded motion including a uniform and rolling motion. A dynamic-state mechanism test was carried out to verify an authenticity of the three-dimensional mechanical model and numerical solution in a circulating tank. The results demonstrated that the presented method was used to evaluate the dynamic mechanism, and held a potential to improve a vehicle design and control strategy.