Open-pool Type Research Reactor Research Articles

Plant-scale experiments of an air inflow accident under sub-atmospheric pressure by pipe break in an open-pool type research reactor

In an open-pool type research reactor with a downward forced flow in the core, pipes can be under sub-atmospheric pressure because of the large pressure drop at the reactor core in the atmospheric pool. Sub-atmospheric pressure can result in air inflow into the pipe from the pressure difference between the atmosphere and the inside of the pipe, which in a postulated pipe break scenario can lead to the breakdown of the cooling pump. In this study, a plant-scale experiment was conducted to study air inflow in large piping systems by considering the actual operational conditions of an advanced research reactor. The air inflow rate was measured, and the entrained air was visualized to investigate the behavior of air inflow and flow regime depending on the pipe break size. In addition, the developed drift-flux model for a large vertical pipe with a diameter of 600 mm was compared with other correlations. The flow regime transition in a large vertical pipe under downward flow was also studied using the newly developed drift-flux model. Consequently, the characteristics of two-phase flow in a large vertical pipe were found to differ from those in small vertical pipes where liquid recirculation was not dominant.

Experimental and analytical natural circulation analysis of the Jordanian 5 MW research reactor using plate type fuel by RELAP5 MOD3.3

The 5-MWth Jordanian research reactor is an open-pool type research reactor using plate-type fuel. It is cooled by the upward natural flow at low power-operation. Natural circulation cooling is important in the safety analysis because it is the last scenario of all accidents. Therefore, studying natural circulation is important in determining the safety limits, such as the maximum allowed temperature and the minimum critical heat flux ratio of fuel. In this paper, the reactor operated at 50KW by the natural circulation to investigate its neutronics and thermal–hydraulic characteristics. Then, the results were benchmarked with RELAP5 MOD 3.3 by implementing models with and without considering the effect of heat loss through the Upper Guide Structure. As a result of this study, different models show minor differences in thermodynamic parameters and an additional safety margin. Eventually, benchmarking the results and the calculations of RELAP5 led to a reasonable agreement in both simulated models.

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

An open-pool type research reactor is designed and operated considering the accessibility around the pool top area to enhance the reactor utilization. The reactor structure assembly is placed at the bottom of the pool and filled with water as a primary coolant for the core cooling and radiation shielding.Most radioactive materials are generated from the fuel assemblies in the reactor core and circulated with the primary coolant. If the primary coolant goes up to the pool surface, the radiation level increases around the working area near the top of the pool. Hence, the hot water layer is designed and formed at the upper part of the pool to suppress the rising of the primary coolant to the pool surface. The temperature gradient is established from the hot water layer to the primary coolant. As this temperature gradient suppresses the circulation of the primary coolant at the upper region of the pool, the radioactive primary coolant rising up directly to the pool surface is minimized. Water mixing between these layers is reduced because the hot water layer is formed above the primary coolant with a higher temperature. The radiation level above the pool surface area is maintained as low as reasonably achievable since the radioactive materials in the primary coolant are trapped under the hot water layer. The key to maintaining the stable hot water layer and keeping the radiation level low on the pool surface is to have a stable flow of the primary coolant.In the research reactor with a downward core flow, the primary coolant is dumped into the reactor pool and goes to the reactor core through the flow guide structure. Flow fields of the primary coolant at the lower region of the reactor pool are largely affected by the dumped primary coolant. Simple, circular, and duct type discharge headers are designed to control the flow fields and make the primary coolant flow stable in the reactor pool.In this research, flow fields of the primary coolant and hot water layer are numerically simulated in the reactor pool. The heat transfer rate, temperature, and velocity fields are taken into consideration to determine the formation of the stable hot water layer and primary coolant flow. The bulk Richardson number is used to evaluate the stability of the flow field. A duct type discharge header is finally chosen to dump the primary coolant into the reactor pool. The bulk Richardson number should be higher than 2.7 and the temperature of the hot water layer should be 1 °C higher than the temperature of the primary coolant to maintain the stability of the stratified thermal layer.

개방수조형 저출력 연구로의 환경적 열 손실 평가에 대한 분석적 접근

개방수조형 저출력 연구로의 환경적 열 손실 평가에 대한 분석적 접근

개방 수조형 연구용 원자로의 조사체 냉각 계통 개념 설계

개방 수조형 연구용 원자로의 조사체 냉각 계통 개념 설계

Simulation and analysis of the Loss of Flow Accident (LOFA) scenarios for an open pool type research reactor by using the RELAP5/MOD3.2 code

Abstract In this paper, a 5 MW open pool type research reactor is simulated and analyzed to check and pass a complete and practical safety assessment against various LOFA scenarios. Generally, LOFA may occur in a research reactor when the reactor primary pump or the safety flapper valve fail. The simulation includes the study of a downward core cooling system circulated by the gravity driven force, the performance and failure effects of the safety flapper valve, the flow reversal mode and transition, the natural convection mode, different possible types of LOFA, and impacts of the surrounding empty boxes. First of all, the code nodalization is successfully benchmarked against available experimental data of the reactor operating parameters. Then, all possible DBAs in this field of study are simulated and discussed in detail. Results are completely satisfactory to simulate and analyze a pool-type research reactor in response to LOFA using the RELAP5 code. Furthermore, transients including the natural convection mode and even flow reversal mode are following naturally without any oscillation or source of errors. Finally, TRR is completely safe against the DBA type of LOFA scenarios.

개방수조형 연구용 원자로에서 일차냉각계통의 설계

고온 고압조건을 가진 발전용 상업로와는 달리 저온 저압조건에서 작동되는 연구용 원자로는 일반적으로 핵반응이 발생하는 노심이 개방된 수조수에 담겨 있는 개방수조형 방식으로 설계된다. 이 수조수의 냉각 및 정화를 위해 일차냉각계통을 포함한 다양한 유체계통이 필요하다. 특히 일차냉각계통은 정상출력운전 중 노심에서 발생되는 열을 일차냉각계통 펌프와 열교환기를 이용하여 제거하는 역할을 한다. 일차냉각계통 펌프가 정지되면 일차냉각계통 펌프의 관성바퀴에 의한 관성서행 유량을 통해 노심냉각을 유지하고, 안전잔열제거계통 펌프를 정지하면 플랩밸브를 통한 수조수의 자연순환에 의해 노심붕괴열이 제거된다. 본 논문에서는 개방수조형 연구용 원자로의 노심 냉각을 위한 일차냉각계통에 대해서 설계요건, 계통설명, 계통 운전(기동, 출력운전, 원자로 정지), 설계평가, 검사 및 시험의 항목으로 나누어 자세히 기술하였다.

Modeling and Simulation for Coolant Activity and Dose Rate Estimation in a Pool-Type Nuclear Research Reactor

Coolant activity in the primary coolant system and reactor pool in an open pool-type research reactor is very important in view of operational and radiological safety considerations. As pool water acts as the shielding medium to minimize the radiation dose at the top of the reactor pool, an estimation of the activities of radionuclides in pool water is essential to establish the safety of the operating personnel and researchers working at the reactor pool top. A system is provided to create a hot water layer (HWL) at the top of the pool by supplying water at a temperature more than that of the pool water so that the dose rate at the reactor pool top can be minimized. This HWL system helps in breaking the natural convection current of reactor pool water by maintaining a higher temperature at this layer so that high-density pool water below this layer cannot replace this low-density HWL. Therefore, pool water that is comparatively more radioactive will not be able to cross this HWL by convection. Hence, diffusion will be the only mechanism by which radioactivity can reach the pool top. So eliminating the convection current keeps the activity at this topmost layer of the pool at a minimum value. The estimation of the activity of the radioactive nuclides is required to assess the radiation field at different locations in the primary coolant loop for designing proper shielding requirements of the system. The radionuclides of interest are the activation products of aluminum (24Na, 27Mg, 28Al) and 41Ar. In this technical note, a transient code is presented for estimating the activity of radioactive nuclides in the coolant loop and reactor pool of a nuclear research reactor. The reduced activity level at the pool top is estimated considering the presence of the HWL at the top of the pool. It is observed that purification flow plays a major role on the activity level of radioactive nuclides in reactor pool water. The variation of the activity dose rate at the reactor pool top with purification flow is also discussed.

Method development and validation for simultaneous determination of IEA-R1 reactor’s pool water uranium and silicon content by ICP OES

IPEN’s research reactor, IEA-R1, an open pool type research reactor moderated and cooled by light water. High quality water is a key factor in preventing the corrosion of the spent fuel stored in the pool. Leaching of radionuclides from the corroded fuel cladding may be prevented by an efficient water treatment and purification system. However, as a safety management policy, IPEN has adopted a water chemistry control which periodically monitors the levels of uranium (U) and silicon (Si) in the pool’s reactor, since IEA-R1 employs U3Si2-Al dispersion fuel. An analytical method was developed and validated for the determination of uranium and silicon by ICP OES. This work describes the validation process, in a context of quality assurance, including the parameters selectivity, linearity, quantification limit, precision and recovery.

A multi-physics analysis for the actuation of the SSS in opal reactor

OPAL is a 20 MWth multi-purpose open-pool type Research Reactor located at Lucas Heights, Australia. It was designed, built and commissioned by INVAP between 2000 and 2006 and it has been operated by the Australia Nuclear Science and Technology Organization (ANSTO) showing a very good overall performance. On November 2016, OPAL reached 10 years of continuous operation, becoming one of the most reliable and available in its kind worldwide, with an unbeaten record of being fully operational 307 days a year. One of the enhanced safety features present in this state-of-art reactor is the availability of an independent, diverse and redundant Second Shutdown System (SSS), which consists in the drainage of the heavy water reflector contained in the Reflector Vessel. As far as high quality experimental data is available from reactor commissioning and operation stages and even from early component design validation stages, several models both regarding neutronic and thermo-hydraulic approaches have been developed during recent years using advanced calculations tools and the novel capabilities to couple them. These advanced models were developed in order to assess the capability of such codes to simulate and predict complex behaviours and develop highly detail analysis. In this framework, INVAP developed a three-dimensional CFD model that represents the detailed hydraulic behaviour of the Second Shutdown System for an actuation scenario, where the heavy water drainage 3D temporal profiles inside the Reflector Vessel can be obtained. This model was validated, comparing the computational results with experimental measurements performed in a real-size physical model built by INVAP during early OPAL design engineering stages. Furthermore, detailed 3D Serpent Monte Carlo models are also available, which have been already validated with experimental data from reactor commissioning and operating cycles. In the present work the neutronic and thermohydraulic models, available for OPAL reactor, are coupled by means of a shared unstructured mesh geometry definition of relevant zones inside the Reflector Vessel. Several scenarios, both regarding coupled and uncoupled neutronic & thermohydraulic behavior, are presented and analyzed, showing the capabilities to develop and manage advanced modelling that allows to predict multi-physics variables observed when an in-depth performance analysis of a Research Reactor like OPAL is carried out.

Calibration of digital wide-range neutron power measurement channel for open-pool type research reactor

Abstract As the modernization of the nuclear instrumentation system progresses, research reactors have adopted digital wide-range neutron power measurement (DWRNPM) systems. These systems typically monitor the neutron flux across a range of over 10 decades. Because neutron detectors only measure the local neutron flux at their position, the local neutron flux must be converted to total reactor power through calibration, which involves mapping the local neutron flux level to a reference reactor power. Conventionally, the neutron power range is divided into smaller subranges because the neutron detector signal characteristics and the reference reactor power estimation methods are different for each subrange. Therefore, many factors should be considered when preparing the calibration procedure for DWRNPM channels. The main purpose of this work is to serve as a reference for performing the calibration of DWRNPM systems in research reactors. This work provides a comprehensive overview of the calibration of DWRNPM channels by describing the configuration of the DWRNPM system and by summarizing the theories of operation and the reference power estimation methods with their associated calibration procedure. The calibration procedure was actually performed during the commissioning of an open-pool type research reactor, and the results and experience are documented herein.

Uncertainty assessment for loss of flow accident of a 5 MW pool-type research reactor

Best estimate plus uncertainty (BEPU) is a promising approach to the safety analysis of nuclear reactors, and the uncertainty calculation is a very important concern about it. BEPU ensures realistic safety margins and secures higher reactor effectiveness by adopting best-estimate codes and realistic input data with uncertainties, whereas the previous conservative analysis generates excessive conservatism by considering each input parameter separately. A loss of flow accident (LOFA) of a 5MW open-pool type research reactor was selected as a sample problem for a BEPU uncertainty assessment. We selected the failures of all primary cooling system (PCS) pumps, which would cause the abrupt reduction of flow and the reversal of core flow. The significant contributors to the reactor safety were identified and then input sets were sampled. For the uncertainty evaluation, 124 calculations were performed. This is the number of code runs required for a 95%/95% level with the 3rd order Wilk’s formula. The MOSAIQUE software developed by Korean Atomic Energy Research Institute (KAERI) was used for automated sampling of the uncertainty parameters, a global uncertainty calculation, and post processing of the results. The critical heat flux ratio (CHFR) and the fuel centerline temperature (FCT) were calculated at the 95%/95% level and were compared with those from conservative analyses. In addition, the impact of each design variables on the safety parameters was estimated by sensitivity analysis.

An Innovative Passive Residual Heat Removal System of an Open-Pool Type Research Reactor with Pump Flywheel and Gravity Core Cooling Tank

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.

ANALYSIS OF AN EXTREME LOSS OF COOLANT IN THE IPR-R1 TRIGA REACTOR USING A RELAP5 MODEL

The RELAP5/MOD3.3 code has been applied for thermal hydraulic analysis of power reactors as well as nuclear research reactors with good predictions. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA have been validated for steady state and transient situations. The reactor is located at the Nuclear Technology Development Centre (CDTN), Brazil. It is a 250 kW, light water moderated and cooled, graphite-reflected, open pool type research reactor. In this work, an extreme transient case of loss of coolant accident (LOCA) has been simulated. For this type of analysis, the automatic scram of the reactor was not considered because the main aim was to verify the evolution of the fuel elements heating in the absence of coolant. The temperature evolutions are presented as well as an analysis about the temperature safety limits.

Uncertainty analysis of criticality safety for the plate type fuel assembly storage rack

To evaluate the criticality safety of the fresh and the spent fuel storage racks in an open pool type research reactor designed by KAERI, the upper subcriticality limit (USL) analysis was carried out. First, the bias and its uncertainty of MCNP code system with ENDF/B-VII library were evaluated using the calculation results of the 183 benchmark experiments. The criticality calculations for the fuel storage rack are carried out under a normal state, an increased water temperature, a fuel assembly drop, and an eccentric insertion which can affect the criticality. Considering biases and uncertainties for the MCNP code system, abnormal conditions, and the manufacturing tolerance of the cell tube thickness, the USL value that can guarantee sufficient subcriticality is determined. It was found that the criticality of the fresh and the spent fuel storage racks currently designed satisfy the USL condition. Additionally, it was concluded that the pitch size of a fresh fuel storage rack can be reduced for efficient space availability, and even under a worst case in which the fresh storage rack is surrounded by a lower water density and the smallest pitch size satisfies the USL conditions.

5MW 연구용 원자로의 1차 냉각 계통 유동 특성

5㎿, open pool type research reactor, is commonly used to education and experimental purpose. It is necessary to prepare a standardization of system designs for considering a demand. HANARO has prepared the standardization of 5MW research reactor system designs based on the design, installation, commissioning and operating experiences of HANARO. For maintaining an open pool type reactor safety, a primary cooling system (after below, PCS) should remove the heat generated by the reactor under a reactor normal operation condition and a reactor shutdown condition. For removing the heat generated by the reactor, the PCS should maintain a required coolant flow rate. For a verification of the required flow rate, a flow network analysis of the PCS was carried under a normal operating condition. Based on the flow network analysis result, this paper describes the PCS flow characteristics of a 5MW open pool type research reactor. Through the result, it was confirmed that the PCS met design requirements including design flow rate without cavitation.

Thermal hydraulic analysis of the IPR-R1 TRIGA research reactor using a RELAP5 model

The RELAP5 code is widely used for thermal hydraulic studies of commercial nuclear power plants. Current investigations and code adaptations have demonstrated that the RELAP5 code can be also applied for thermal hydraulic analysis of nuclear research reactors with good predictions. Therefore, as a contribution to the assessment of RELAP5/MOD3.3 for research reactors analysis, this work presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor at 50 kilowatts (kW) of power operation. The reactor is located in the Nuclear Technology Development Center (CDTN), Brazil. It is a 250 kW, light water moderated and cooled, graphite-reflected, open pool type research reactor. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of the RELAP5 model validation. The RELAP5 results were also compared with calculated data from the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The results obtained have shown that the RELAP5 model for the IPR-R1 TRIGA reproduces the actual steady-state reactor behavior in good agreement with the available data.

Assessment of RELAP5 model for the University of Massachusetts Lowell Research Reactor

RELAP5 is a system code developed at the Idaho National Environmental and Engineering Laboratory for thermal hydraulic analysis of nuclear reactors. The code RELAP5 is widely used for safety analysis studies of commercial nuclear power plants. However, recent released version of RELAP5/3.2 and over present significant capabilities for analysis of nuclear reactor research systems. As a contribution to the assessment of RELAP5/3.3 for research reactor safety analysis, experimental data from the University of Massachusetts Lowell Research Reactor - UMLRR are used. The UMLRR is a 1 MW light water moderated and cooled, graphite-reflected, open-pool type research reactor. This paper presents the development and the validation of a UMLRR-RELAP model using experimental data. For this purpose, a series of experiments were performed for benchmarking RELAP5 calculations for research reactor systems. As a result of this study, the UMLRR nodalization is shown to be representative of the experimental data reactor behavior.

Analytical study on thermal-hydraulic behavior of transient from forced circulation to natural circulation in JRR-3.

Transient thermal-hydraulic behaviors of the JRR-3 which is an open-pool type research reactor has been analyzed with the THYDE-P1 code. The focal point is the thermal-hydraulic behaviors related to the core flow reversal during the transition from forced circulation downflow to natural circulation upflow. In the case of a loss-of-coolant accident (LOCA), for example, the core flow reversal is expected to occur just after the water pool isolation from the primary cooling loop with a leak. The core flow reversal should cause a sudden increase in fuel temperature and a steep decrease in the departure-from-nucleate-boiling ratio (DNBR) and the phenomenon is, therefore, very important especially for safety design and evaluation of research reactors. Major purposes of the present work are to clarify physical phenomena during the transient and to identify important parameters affecting the peak fuel temperature and the minimum DNBR. The results calculated with THYDE-P1 assuming the sequences of events of the loss-of-offsite power and LOCA help us to understand the phenomena both qualitatively and quantitatively, with respect to the safety design and evaluation.