Open-pool Research Reactor Research Articles

Analysis of a Hypothetical Complete Loss-of-Coolant Accident in a Typical Research Reactor

In the present study, an analysis of a hypothetical complete loss-of-coolant accident in a typical open-pool research reactor is conducted. The reactor core is assumed to be completely uncovered and exposed to the ambient air. The possibility of passively cooling the decay heat of the exposed reactor core by natural convection to air and thermal radiation until core reflooding is investigated. A three-dimensional computational fluid dynamics analysis is conducted for the uncovered core while cooled by air natural convection and thermal radiation. The reactor core is simulated as a porous zone with decay heat generation specified as a cosine-shape distribution. The reactor core decay heat acts as a driving force for the coolant flow from the cold leg to the hot leg. The thermal equilibrium porous media model is used to represent the energy equation inside the core region. This study is conducted for core uncover time (the time interval between reactor shutdown and the moment when the reactor core is drained of water) of 10E3, 10E4, 10E5, 10E6, 10E7, and 10E8 s. Contour plots of temperature, velocity, density, and pressure at different values of core uncover time are illustrated. It’s found that for core uncover times of 10E3, and 10E4 s, the maximum core temperature exceeds the cladding melting point. The core maximum temperature is well below the melting point for uncover times of 10E5, 10E6, 10E7, and 10E8 s.

Deriving formula to simulate the radiological behavior during LOCA in open pool type reactor

This study aims to model the radiological consequences that result from loss of coolant accident in open pool research reactor of 22 MW power. The loss of coolant accident results from rupture in the cooling systems of the reactor and consequently results in decreasing the water level in the reactor pool. Decreasing the water level, that represents the shielding material above the reactor core, results in increasing the dose rate in the top region in the reactor pool as well as the reactor building. LOCA normally results in shutting down the reactor due to the radioprotection system provided in the upper region of the reactor pool. A failure in shutdown systems of the reactor during LOCA is rare but expected and it represents the worst case. So, evaluating LOCA from the radiological point of view must be investigated for the safe shutdown mode as well as the failure of shutdown case. The study also includes occurring LOCA for shutdown reactor taking the decay time into consideration. ORIGEN-S module from scale system was used to estimate the delayed neutrons and gamma rays from the core for the shutdown mode. KENO-VI module from Scale system was used to estimate the prompt neutrons and gamma rays generated from the core for the failure case. Then, MAVRIC/Scale module was used to estimate the dose rate at the top point of the reactor pool for the two cases during LOCA.Studying the dose rate during LOCA aims to determine the water level that verifies the radiological safety limit at the top point of the reactor pool for the two cases. The calculations introduced a relationship between the water level that verifies the permissible radiological limit (WL) and the decay time (DT) in the range between 0 to 30 d.

Development of neutron-activated samarium-153-loaded polystyrene microspheres as a potential theranostic agent for hepatic radioembolization.

Hepatic radioembolization is an effective minimally invasive treatment for primary and metastatic liver cancers. Yttrium-90 [90Y]-labelled resin or glass beads are typically used as the radioembolic agent for this treatment; however, these are not readily available in many countries. In this study, novel samarium-153 oxide-loaded polystyrene ([153Sm]Sm2O3-PS) microspheres were developed as a potential alternative to 90Y microspheres for hepatic radioembolization. The [152Sm]Sm2O3-PS microspheres were synthesized using solid-in-oil-in-water solvent evaporation. The microspheres underwent neutron activation using a 1 MW open-pool research reactor to produce radioactive [153Sm]Sm2O3-PS microspheres via 152Sm(n,γ)153Sm reaction. Physicochemical characterization, gamma spectroscopy and in-vitro radionuclide retention efficiency were carried out to evaluate the properties and stability of the microspheres before and after neutron activation. The [153Sm]Sm2O3-PS microspheres achieved specific activity of 5.04 ± 0.52 GBq·g-1 after a 6 h neutron activation. Scanning electron microscopy and particle size analysis showed that the microspheres remained spherical with an average diameter of ~33 μm before and after neutron activation. No long half-life radionuclide and elemental impurities were found in the samples. The radionuclide retention efficiencies of the [153Sm]Sm2O3-PS microspheres at 550 h were 99.64 ± 0.07 and 98.76 ± 1.10% when tested in saline solution and human blood plasma, respectively. A neutron-activated [153Sm]Sm2O3-PS microsphere formulation was successfully developed for potential application as a theranostic agent for liver radioembolization. The microspheres achieved suitable physical properties for radioembolization and demonstrated high radionuclide retention efficiency in saline solution and human blood plasma.

Conservative analytical and experimental investigation of all pumps failure event in the primary cooling system for a 5 MWth research reactor

The 5-MWth Jordanian Nuclear Research Reactor is an open-tank and open-pool research reactor which uses narrow rectangular fuel. The reactor is cooled by demineralized water in downward forced flow and natural circulation in transient events. The reactor performed a loss of flow test in the Primary Cooling System (PCS) by turning off all PCS pumps at full-power operation. The reactor then displayed safe behavior in thermal-hydraulic, neutronics, and acceptable radiological values. Besides, the RELAP5 MOD3.3 code was used to implement the reactor in three different comprehensive models according to several safety analysis conservatisms, such as assuming a lower total core flow rate by −12% or assuming failure in a safety valve. After that, a benchmark was conducted between the experimental results, and the code calculations of implemented models showed well agreement in general thermal-hydraulic and safety analysis parameters. The code proved a safe reactor trend that used all conservative assumptions even in the most conservative implemented model. The primary safety parameters, the Maximum Temperature of Fuel Clad (MTFC) and the Minimum Critical Heat Flux Ratio (MCHFR), were 421 K and 1.82 for a short time, respectively. The resulting parameters underwent the event's acceptance criteria, where the limits of the set values are 673 K and 1.5 for MTFC and MCHFR, respectively. As a result of the study, several models illustrated differences in thermodynamic parameter effects and gave higher and additional reactor safety levels. Moreover, the test comprehensively investigated the reactor behaviors and ensured the reactor cooling system capability by free convection cooling during the event.

Modeling of the Radiological Behavior During LOCA in Open Pool Type Reactor

This study aims to model the radiological consequences that result from loss of coolant accident in open pool research reactor of 22 MW power. The loss of coolant accident results from rupture in the cooling systems of the reactor and consequently results in decreasing the water level in the reactor pool. Decreasing the water level, that represents the shielding material above the reactor core, results in increasing the dose rate in the top region in the reactor pool as well as the reactor building. LOCA normally results in shutting down the reactor due to the radioprotection system provided in the upper region of the reactor pool. A failure in shutdown systems of the reactor during LOCA is rare but expected and it represents the worst case. So, evaluating LOCA from the radiological point of view must be investigated for the safe shutdown mode as well as the failure of shutdown case. The study also includes occurring LOCA for shutdown reactor taking the decay time into consideration. ORIGEN-S module from scale system was used to estimate the delayed neutrons and gamma rays from the core for the shutdown mode. KENO-VI module from Scale system was used to estimate the prompt neutrons and gamma rays generated from the core for the failure case. Then, MAVRIC/Scale module was used to estimate the dose rate at the top point of the reactor pool for the two cases during LOCA. Studying the dose rate during LOCA aims to determine the water level that verifies the radiological safety limit at the top point of the reactor pool for the two cases. The calculations introduced a relationship between the water level that verifies the permissible radiological limit (WL) and the decay time (DT) in the range between 0 to 30 d.

On spontaneous appearance of internal waves in an open-pool-type research reactor

Variation of temperature in time and space was recorded at multiple vertical locations in the course of initiation of a heated water layer in an open-pool research reactor of the Soreq Nuclear Research Center. The pool was initially filled with warm water, and heavier cooler water was then injected at the bottom of the facility. Different modes of coolant injection were employed in two different experiments. In both cases, a finite width thermocline that separated cool water at the lower part of the pool from the warm water in its upper part was observed. The thermocline gradually moved up eventually attaining a constant raise velocity. In both experiments, the thermocline characteristics were different, but wave-trains with notable temperature fluctuations were observed within the thermocline. The characteristic frequencies of oscillations were below the Brunt–Väisälä frequencies that characterize the density gradient within the thermoclines. The finite dimensions of the tank impose conditions in which standing internal waves with the length commensurate with tank size can be expected. The oscillations were thus associated with resonant internal waves excited by disturbances introduced by the coolant flow at the lower part of the pool. In both experiments, the measured wave spectra agree with the results of linear analysis of two-layer and three-layer stratification models.

Numerical study on the uncertainty reduction for thermal power estimation of an open pool research reactor with natural convection core cooling

In this study, the calorimetric method is used to study the thermal power estimation in an open pool research reactor during power operation under natural convection core cooling. The pool temperature measuring points are used as a key factor to minimize the uncertainty of the thermal power of the research reactor. The Computational Fluid Dynamics (CFD) technique is used to analyze the natural convection flow pattern in the reactor pool. The results show how the natural convection flow pattern influences the uncertainty of the thermal power estimation in terms of the flow stabilization and how the optimal temperature measuring point can be chosen. This paper shows that an appropriate thermal power estimation with a low uncertainty can be obtained based on the temperature measurement in the lower section of the reactor pool with the proper time criteria in view of the temperature propagation.

Procedures for manufacturing an instrumented nuclear fuel element

The IEA-R1 is an open pool research reactor that operated for many years at 2 MW. The reactor uses plate type fuel elements which are formed by assembling eighteen parallel fuel plates. During the years of reactor operation at 2 MW, thermohydraulic safety margins with respect to design limits were always very high. However, more intense oxidation on some external fuel plates was observed when the reactor power was increased to 5 MW. At this new power level, the safety margins are significantly reduced due to the increase of the heat flux on the plates. In order to measure, experimentally, the fuel plate temperature under operation, an instrumented fuel element was constructed to obtain temperature experimental data at various positions of one or more fuel plates in the fuel element. The manufacturing method is characterized by keeping the original fuel element design specifications. Type K stainless sheathed thermocouples are mounted into supports pads in unrestricted positions. During the fuel element assembling, the supports pads with the thermocouples are mechanically fixed by interference between two adjacent fuel plates. The thermocouple wires are directed through the space existing at the bottom of the mounting slot where the fuel plate is fixed to the side plates. The number of thermocouples installed is not restricted and depends only on adaptations that can be made on the mounting slots of the standard fuel element side plates. This work describes the manufacturing procedures for assembling such an instrumented fuel element.