Reactor Containment Research Articles

Characterization and mapping of the neutron fields around Bruce Power’s 177Lu isotope production system

Bruce Power operates a first-of-its-kind isotope production system (IPS) that enables continuous production of 177Lu within Canada Deuterium Uranium (CANDU) commercial power reactors. Located on the reactivity mechanisms deck of Unit 7, just outside of reactor containment but in close proximity to the primary heat transport (PHT) pumps, this facility offers unique advantages for 177Lu production. However, employees working in this area encounter a radiation hazard which consists primarily of photoneutrons. These originate from the base of the PHT pumps and are only present when the reactor is operating. This study evaluates neutron exposure at Bruce Power’s IPS by using a nested neutron spectrometer (NNS) to determine the neutron energy spectra and absolute dosimetric quantities such as the ambient dose equivalent, H*(10). The results from the NNS are then compared to surveys performed by a portable neutron rem meter (Model NP-2 by Nuclear Research Corporation), routinely used by Bruce Power staff for workplace monitoring. While the Model NP-2 generally showed consistent results across locations, a 50% dose correction factor was identified when operators were harvesting 177Lu from the IPS. This finding highlights an opportunity to reduce the neutron dose that is assigned to operators when producing 177Lu.

Asymmetric Two-Step Thermosiphon for Containment Cooling of Small Modular Reactors

A thermosiphon is a heat transfer device that utilizes the phase change of a liquid and has a single closed-loop shape in a gravity-dominant field. This can be expressed as a single-step thermosiphon because boiling and condensation occur once per cycle. In contrast, the multistep thermosiphon, introduced for the first time in the field of thermal engineering in this study, is a new heat transfer mechanism in which boiling and condensation occur several times per cycle in a single loop with multiple channels. The new mechanism has a superior heat transfer rate compared to the existing single-step thermosiphon, and the operating pressure of the loop can be lowered. However, as the heat transfer rate increases, the circulation flow in the channel tends to pulsate. This thermohydraulic characteristic was confirmed through theoretical and computational analyses of a two-step thermosiphon. In this study, an improved concept of an asymmetric two-step thermosiphon was developed that can be applied to heat exchanger design by eliminating pulsating flow while maintaining the advantages of a two-step thermosiphon. The newly proposed heat transfer mechanism, termed the multistep thermosiphon, can be effectively used in the design of heat exchangers in industrial fields. In particular, if the asymmetric two-step thermosiphon is applied to the design of small nuclear reactor containments currently being developed in several countries, there are several advantages associated with the reduction of the containment volume and design pressure.

Review of the research on the scrubbing of fission products in liquid metal pool

In recent times, the risk assessment and design optimization of advanced fourth-generation nuclear reactors have garnered significant global scholarly attention. Specifically, researchers worldwide have been focusing on the comprehensive analysis of the scrubbing process of radioactive fission product aerosols within the reactor containment pool under severe accident scenarios. This study delves into the historical progression of investigating the pool scrubbing phenomenon, which dates back to the mid-20th century. International efforts have encompassed various aspects, including assessing the migration behavior of radioactive nuclides within molten metal environments, conducting sensitivity analyses to gauge the impact of diverse variables on pool scrubbing, formulating models for analyzing the pool scrubbing process, and conducting relevant experimental validations. Drawing upon an extensive body of literature, this article presents a synthesized overview of research into the pool scrubbing process of fission product aerosols, focusing on three principal dimensions: model elucidation, experimental investigations, and computational simulations. The insights provided herein serve as a consolidated reference point for forthcoming endeavors in this domain of research.

Dust detection and cleanliness assessment based on S-YOLOv5s for NPP reactor containment wall-climbing cleaning robot

NPP reactor containment dust can easily turn into radioactive dust, endangering staff health and the environment. However, the nuclear reactor containment wall-climbing cleaning robot cleans blindly without the ability to clean the dust in a timely and thoroughly. In this paper, ShuffleNetV2-YOLOv5s (S-YOLOv5s) model is proposed to solve the problem of wall-climbing robots unable to detect different categories of dust in time. The use of ShuffleNetV2 in the backbone of the network not only ensures a large number of characterized channels and a large network capacity, but also reduces the complexity of the model; SIoU is chosen for the loss function to improve the model accuracy. Then, planar cleaning index (PCI) is proposed by combining the results of S-YOLOv5s to evaluate whether the wall-climbing cleaning robot cleans thoroughly. Compared to other methods, PCI considers amount and area occupation of different classes of dust. The dust data set is collected to train the designed model, and the model size is reduced to 14 % of the original model, and the FPS is 7.313 higher than the original model. Especially when compared with the state-of-the-art lightweight methods, our model has smaller model size and higher recognition speed. Experimental results have shown that our dust detection and cleanliness assessment method can be used on a wall-climbing cleaning robot to clean dust in time and thoroughly.

Numerical analysis of sodium-argon two-phase flow in narrow channels under severe accident conditions in pool type SFR

A two-phase flow model for sodium-argon mixture flow through the reactor vessel head penetrations under the energetic core disruptive accident in a sodium fast reactor (SFR) is presented. The model evaluates the total sodium mass in sodium-argon mixture released to the reactor containment building (RCB) through penetrations. The driving pressure at the penetration entrance is given as input. The two-phase frictional and local pressure drops, inertial and piezo-metric heads are accounted for in the model. Depending on the argon volume fraction, either a homogenous or separated flow approach is adopted to evaluate the mixture flow. Reactor-scale analysis show that the presence of argon in vessel head penetrations can significantly mitigate the total sodium release mass during the slug impact phase. Also, the two-phase frictional pressure drop plays a mitigating role and needs to be included in the model formulation to avoid overestimation of the sodium release mass. Fluid-structure interaction effects such as vessel head bending and hold-down bolt elongation have negligible influence on sodium release mass as long as the deformations are within the elastic limit. The model is conceived to be part of an integrated code system to evaluate the sodium fire and the maximum RCB pressure under severe accident conditions in a pool-type SFR.

Analytical model development for investigation on hydrogen management studies in the containment of a nuclear reactor

Containment is the ultimate physical barrier to prevent the spread of active air-borne fission products as well as radiation shielding in normal operation as well as accident condition and is designed to enclose whole reactor systems. The design of containment system for Fleet mode of 700 MWe Indian Pressurized Heavy Water Reactors (IPHWRs) incorporates a Primary Containment (PC) of pre-stressed concrete enveloped by a Secondary Containment (SC) of reinforced concrete. For the specification of the design pressure of the containment system, analysis is performed for various postulated accidents such as Loss of Coolant Accident (LOCA) and Main Steam Line Break Accident (MLSB). Further Studies on hydrogen generation and distribution for various accident scenarios in reactor containment of Indian Pressurized Heavy Water Reactors (IPHWR) indicate that during severe accident with core damage, the possibility of global deflagration in absence of mitigating features is high and may challenge the integrity of containment. Therefore, it is essential to incorporate mitigating provisions inside the containment to prevent the hydrogen reaching local detonation and global deflagration during such a scenario.In order to arrive at the design basis of containment system and simulation of large numbers of accident sequence a computer code is required, where mitigating features such as Passive Auto-Catalytic Recombiners, Containment Spray System & Containment Filtered Venting System can be modeled. A system thermal hydraulic computer code has been developed for containment response calculation during normal operation as well as accident conditions including, severe accident to include the new models for mitigating features. This article presents the details of various models of computer code PACSR-SI-2.0, its validation with data observed for hydrogen distribution & removal in Recombiner Test Facility,Battelle Spray Test for containment system & Spray test at IIT Mumbai.

Application of analytical wall functions to CFD analysis of condensation flow

The wall function (WF) enables analyzing condensation flow in a nuclear reactor containment vessel with reasonable computational costs. However, conventional wall treatments rely on the logarithmic laws for velocity, temperature, and concentration, limiting applicability. In this paper, we applied the analytical wall function approach to the condensation flow analysis of steam/air mixtures. This approach features the analytical integration of transport equations considering the buoyancy, the material property change, and the convective terms. We conducted CFD analysis with the analytical wall function models for the forced, mixed, and natural convection and confirmed good prediction, especially when the log law does not hold.

Numerical investigation on seismic performance of reinforced rib-double steel plate concrete combination shear wall

Double steel plate concrete composite shear wall (SCSW) has been widely utilized in nuclear power plants and high-rise structures, and its shear connectors have a substantial impact on the seismic performance of SCSW. Therefore, in this study, the mechanical properties of SCSW with angle stiffening ribs as shear connections were parametrically examined for the reactor containment structure of nuclear power plants. The axial compression ratio of the SCSW, the spacing of the angle stiffening rib arrangement and the thickness of the angle stiffening rib steel plate were selected as the study parameters. Four finite element models were constructed by using the finite element program named ABAQUS to verify the experimental results of our team, and 13 finite element models were established to investigate the selected three parameters. Thus, the shear capacity, deformation capacity, ductility and energy dissipation capacity of SCSW were determined. The research results show that: compared with studs, using stiffened ribs as shear connectors can significantly enhance the mechanical properties of SCSW; When the axial compression ratio is 0.3–0.4, the seismic performance of SCSW can be maximized; with the lowering of stiffener gap, the shear bearing capacity is greatly enhanced, and when the gap is lowered to a specific distance, the shear bearing capacity has no major affect; in addition, increasing the thickness of stiffeners can significantly increase the shear capacity, ductility and energy dissipation capacity of SCSW. With the rise in the thickness of angle stiffening ribs, the improvement rate of each mechanical property index slows down. Finally, the shear bearing capacity calculation formula of SCSW with angle stiffening ribs as shear connectors is derived. The average error between the theoretical calculation formula and the finite element calculation results is 8% demonstrating that the theoretical formula is reliable. This study can provide reference for the design of SCSW.

A study on transport and plugging of sodium aerosol in leak paths of concrete blocks

In the event of a severe accident in Sodium Cooled Fast Reactors (SFR), the sodium combustion aerosols along with fission product aerosols would migrate to the environment through leak paths of the Reactor Containment Building (RCB) concrete wall under positive pressure. Understanding the characteristics of sodium aerosol transport through concrete leak paths is important as it governs the environmental source term. In this context, experiments are conducted to study the influence of various parameters like pressure, initial mass concentration, leak path diameter, humidity etc., on the transport and deposition of sodium aerosols in straight leak paths of concrete. The leak paths in concrete specimens are prepared by casting and the diameter of the leak path is measured using thermography technique. Aerosol transport experiments are conducted to measure the transported and plugged aerosol mass in the leak paths and corresponding plugging times. The values of differential pressure, aerosol concentration and relative humidity taken for the study are in the ranges 10–15 kPa, 0.65–3.04 g/m3 and 30–90% respectively. These observations are numerically simulated using 1-Dimensional transport equation. The simulated values are compared with the experimental results and reasonable agreement among them is observed. From the safety assessment view of reactor, the approach presented here is conservative as it is with straight leak paths.

Investigating the effect of spray system on the containment condition of VVER1000/V446 NPP during LBLOCA and TLOFW accidents

The containment system serves as the last barrier against the release of fission products. However, over pressurization or hydrogen detonation can compromise its effectiveness, as evidenced by the Fukushima nuclear power plant (NPP) disaster. This study describes the efficacy of the spray system in reducingcontainment pressure andtemperature, as well as,itsimpacton hydrogen mitigation during large break loss of coolant accident (LBLOCA) and total loss of feed water (TLOFW) accident in the VVER-1000/V446NPP. The evaluation is conductedusing theMELCOR 1.8.6code.The results of the LBLOCA analysis demonstrate that the spray system significantly reduces the pressure and temperature of the containment. It also enhances hydrogen mitigation by increasing the amount of recombined hydrogen by around 4% and reducing the onset time of hydrogen recombination by 90,000 s. In the TLOFW accident scenario, the spray system is effective in reducing the containment pressure from 0.41 MPa to 0.27 MPa. The investigation of hydrogen behavior in the reactor containment atmosphere reveals that the simultaneous operation of the spray and the hydrogen recombiner systems is the most effective strategy for hydrogen mitigation. However, activating the spray system without the passive autocatalytic recombiners (PARs) function can lead to an increase in the hydrogen mole fraction in all containment compartments, thereby increasing the likelihood of hydrogen combustion.

Comprehensive Analysis of CdS Ultrathin Films Modified by the Substrate Position inside the Reactor Container Using the CBD Technique.

The cadmium sulfide (CdS) n-type semiconductor is one of the most used as a window layer in thin-film solar cells, such as CdTe, CIS, CIGS, and CZTS. Optoelectronic properties are the most important characteristics for window materials. CdS thin films obtained using the chemical bath deposition technique (CBD) have been reported; however, large amounts of precursor solutions are used, which generate considerable amounts of toxic waste. The aim of this work is to reduce the amount of precursor solutions used for CdS growth; for this, it is necessary to consider an efficient position for the substrate inside the reactor container and at the same time allow obtaining CdS thin films with adequate physical properties to be applied in the photovoltaic solar cell development. CdS thin films were deposited on soda-lime/SnO2:F substrates (FTO) using the CBD technique; the substrates were placed in three different arrangements [rack system, step system (up), and step system (down)]. CdS samples with areas of 4 cm2 and a thickness of 27-48 nm were obtained; the X-ray diffraction patterns show CdS thin films with different polycrystalline structures. The morphological measurements reveal different surface formations depending on the substrate position, and resistivity values of around 105 Ω*cm were measured. UV-vis spectra show transmittance values of around 45-94% in the visible region with band gap energy values of around 2.1-2.36 eV. The best physical properties of CdS thin films and an efficient CBD process were obtained when the FTO substrates were located near the bottom of the reactor container with the FTO side down, leading to an optimal configuration that allows reducing the amount of precursor solutions and in this way reduces the toxic waste generated. These results are important in the photovoltaic technology process and environmental impact.

Effect of morphological properties on the particle size distribution measurements and modelling of porous and nonspherical aerosol behaviour

Estimating the time evolution of the suspended aerosol concentration in nuclear reactor containment is critical in assessing the environmental source term during severe accident conditions. Morphological properties of porous and non-spherical aerosol particles are necessary to convert the measured equivalent size distributions to actual size distributions and modelling the aerosol processes. Towards this, the dynamics of liquid PAO (Poly Alpha Olefin) and metal aerosol (non-radioactive fission product equivalent SrO2; and structural steel) are studied individually in a 1 m3 closed chamber. The effect of initial particle size distributions (PSDs) based on different equivalent diameters and the morphological properties on realistic modeling of experimental results is analysed. Electrical Low-Pressure Impactor (ELPI) and Scanning Mobility Particle Sizer (SMPS) are used to measure concentration decay and size growth. The aerosol dynamics model is validated with PAO aerosol data and applied to metal aerosol. Modelled dynamics using both the aerodynamic and effective density (ρe) corrected mobility equivalent PSDs of ELPI (APSD and MEPSD) agree with the measured values within 10% deviation for PAO aerosol, which are spherical with ρe (0.82 g cm−3) slightly differed from 1 g cm−3. Modelled dynamics using APSD significantly deviate around 77.5–137% for the porous and non-spherical/fractal metal aerosols, irrespective of the morphological properties explicitly used to model the aerosol processes with marginal effect. As ρe of SrO2 and steel aerosols (3.16 and 3.34 g cm−3) are >1 g cm−3, aerodynamic distribution underestimates the coagulation phenomenon, the primary aerosol dynamics process to a larger extent. The mobility equivalent size distributions of ELPI model the aerosol dynamics within 6–23% deviation, as supported by SMPS measurements. This study can be extended to ascertain the input parameters required for improved estimations of sodium combustion aerosol dynamics, which has added complexity of hygroscopic growth and time-varying chemical speciation.

Use of supporting software tool for decision-making during low-probability severe accident management at nuclear power plants

In the project NARSIS – New Approach to Reactor Safety ImprovementS – possible advances in safety assessment of nuclear power plants (NPPs) were considered, which also included possible improvements in the field of management of low probability accident scenarios. As a part of it, a supporting software tool for making decisions under severe accident management was developed.
 The mentioned tool, named Severa, is a prototype demonstration-level decision supporting system, aimed for the use by the technical support center (TSC) while managing a severe accident, or for the training purposes. Severa interprets, stores and monitors key physical measurements during accident sequence progression. It assesses the current state of physical barriers: core, reactor coolant system, reactor pressure vessel and containment. The tool gives predictions regarding accident progression in the case that no action is taken by the TSC. It provides a list of possible recovery strategies and courses of action. The applicability and feasibility of possible action courses in the given situation are addressed. For each action course, Severa assesses consequences in terms of probability of the containment failure and estimated time window for failure. At the end, Severa evaluates and ranks the feasible actions, providing recommendations for the TSC. The verification and validation of Severa has been performed in the project and is also described in this paper. Although largely simplified in its current state, Severa successfully demonstrated its potential for supporting accident management and pointed toward the next steps needed with regard to further advancements in this field.

Spatio-temporal evolution of sodium aerosol and temperature distribution during sodium pool fire in a confined environment

In case of a Core Disruptive Accident in sodium-cooled fast reactor, sodium will be ejected from the reactor vessel into the Reactor Containment Building (RCB) and results in sodium fire. Liquid sodium combustion generates aerosol, which gets dispersed and remains suspended in RCB based on geometry of the building and degree of turbulence. The Environment Source Term (EST) primarily depends on the dynamics of sodium aerosols inside the containment. Hence, it is essential to carry out a safety analysis of SFR using computational tools and models shall be validated with experimental data, particularly aerosol dynamics inside the containment along with thermal hydraulics. Besides, all codes assume uniform aerosol concentration in large containment buildings and the evolution of suspended concentration was predicted. However, in actual case a non-uniform concentration profile exists, which leads to stratification of aerosol concentration inside the containment. Towards assessment of aerosol dynamics in realistic scenario, spatial–temporal characteristics of sodium aerosol and temperature distribution were studied by conducting sodium pool fire experiments in a large containment.The sodium fire experiment was conducted by using ∼2.0 kg of sodium in a pool area of ∼0.25 m2. The measured maximum pool and gas temperatures are 685 °C and 126 °C respectively, and the temperature gradient at the wall boundary layer was 7.5 °C/cm. The estimated average sodium burning rate and aerosol release rate are 17.3 kg/m2hr and 0.24 g/s, respectively. The spatial variation of aerosol characteristics was measured for both horizontal and vertical planes. The measured maximum aerosol concentration in the horizontal plane ranges from 2.76 to 5.32, 2.16–3.30 and 2.09–4.21 g/m3 for the bottom, middle and top elevations respectively. For the vertical plane, aerosol concentration varied from 2.16 to 3.65 g/m3 for central positions and 2.53–5.32 g/m3 for off-center positions. The trend of average suspended aerosol concentration at each elevation showed that the concentration was higher at the top elevation than at the middle and bottom for up to 15 min i.e., during aerosol generation period and after that concentration was found to be higher at the bottom elevation for the next 25 min. Based on the measurements, it is observed that the aerosol distribution is not well mixed within the experimental period. A significant difference in aerosol mass concentration was observed between each elevation, between estimation and observations based on a zero-dimensional model. The evolutions of aerosol median size by theoretical simulation capture the overall trend reasonably.

Importance, influence and limits of CFD radiation modeling for containment atmosphere simulations

In the appearance of steam, the CFD modeling of thermal-hydraulic phenomena in reactor containments during various phases of an accident necessarily requires consideration of radiative heat transfer. These radiation phenomena involve (i) energy transfer within the gas mixture and (ii) between the gas and the surrounding structures which have a much higher thermal inertia. Preliminary calculations performed for the present experiments in the OECD/NEA HYMERES-2 project with the CFD code containmentFoam using a Monte Carlo thermal radiation solver with non-gray gas modeling of steam absorption, showed that radiant heat transfer is important even with a very low amount of steam (≈ 2%). Therefore, the test matrix was tailored to the two opposite extremes: Either gas compositions with low steam content, where radiative heat transfer can be neglected, or gas mixtures containing larger amounts of steam, so that radiative heat transfer plays a dominant role. For the two selected experiments within the H2P2 series and the corresponding CFD calculations, a vessel with a diameter of 4 m and a height of 8 m was preconditioned with different mixtures of air/steam at room and elevated temperatures. This was followed by the build-up of a stable helium stratification in the upper part of the vessel. Helium was then injected from the top at a higher mass flow rate, which (a) compresses the gas atmosphere, (b) resembles piston compression and (c) leads to a height-dependent and transient increase in gas temperature below this helium stratification, influenced by thermal radiation effects — or their absence. These experiments and the associated CFD calculations were developed to isolate the thermal radiation phenomena as much as possible from convective and diffusive effects. They are the first of their kind to be conducted in a large-scale thermal-hydraulic facility. It is demonstrated (a) that neglecting thermal radiation for the high steam content case, the temperature rise is significantly over-predicted by a factor of two compared to the application of the suggested Monte Carlo thermal radiation solver with non-gray gas modeling, (b) the computational effort considering CFD and non-gray gas radiation is feasible with the suggested tailored Monte Carlo solver and (c) that the suggested radiation model outperforms the widespread used P1 radiation model.

Reevaluation of failure criteria location and novel improvement of 1/4 PCCV high fidelity simulation model under material uncertainty quantifications

Reactor containment buildings serve as the last barrier to prevent radioactive leakage due to accidents and their safety is crucial in overpressurization conditions. Thus, the Regulatory Guide (RG) 1.216 has mentioned the global strain as one of failure criteria in the free-field for cylindrical prestressed concrete containment vessels (PCCV) subject to internal pressure. However, there is a limit that RG 1.216 shows the free-field without the specific locations of failure criteria and also the global strain corresponding to only azimuth 135° has been mentioned in NUREG/CR-6685, regardless of the elevations of the structure. Therefore, in order to reevaluate the failure criteria of the 1:4 scaled PCCV, the high fidelity simulation model based on the experimental test was significantly validated in this study, and it was interesting to find that the experimental and numerical result was very close to each other. In addition, for the consideration of the material uncertainties, the Latin hypercube method was used as a statistical approach. Consequently, it was revealed that the radial displacements of various azimuth area such as 120°, 135°, 150°, 180° and 210° at elevations 4680 mm and 6,200 mm can represent as the global deformation at the free-field, obtained from the statistical approach.

TRACE code simulation of the interaction between reactor coolant system and containment building with passive heat removal system

TRACE code simulation of the interaction between reactor coolant system and containment building with passive heat removal system

Film condensation with high heat fluxes and scaled experiments using pure steam for reactor containment cooling

Condensation tests were performed using a newly developed test facility for scaling the passive containment cooling system (PCCS) to a small modular reactor (SMR). The PCCS of the SMR plays a pivotal role in ensuring greater safety, reliability, and compactness than what is afforded by traditional reactors. Therefore, a well-designed PCCS is essential to SMRs. However, previous studies and test data were unsuitable for scaling, due to high variation in the test geometry and operating conditions. This study intends to close this research gap by using a novel designed scaled test facility consisting of vertical condensing test sections featuring 1-, 2-, and 4-inch-diameter condensing tubes with annular water cooling, and by applying superheated and saturated steam with different steam mass flow ranges of 5–25 g/s. The primary test data, including axial temperatures, mass flow rates, and pressures, were used in conjunction with a standard data reduction method to estimate critical parameters such as heat fluxes, heat transfer coefficients, and condensation rates. These scaled test data would support improving empirical correlations and validating condensation models to identify scaling distortion for SMR PCCSs.

Improvement and validation of aerosol models for natural deposition mechanism in reactor containment

Nuclear safety is the lifeline for the development and application of nuclear energy. In severe accidents of pressurized water reactor (PWR), aerosols, as the main carrier of fission products, are suspended in the containment vessel, posing a potential threat of radioactive contamination caused by leakage into the environment. The gas-phase aerosols suspended in the containment will settle onto the wall or sump water through the natural deposition mechanism, thereby reducing atmospheric radioactivity. Aiming at the low accuracy of the aerosol model in the ISAA code, this paper improves the natural deposition model of aerosol in the containment. The aerosol dynamic shape factor was introduced to correct the natural deposition rate of non-spherical aerosols. Moreover, the gravity, Brownian diffusion, thermophoresis and diffusiophoresis deposition models were improved. In addition, ABCOVE, AHMED and LACE experiments were selected to validate and evaluate the improved ISAA code. According to the calculation results, the improved model can more accurately simulate the peak aerosol mass and respond to the influence of the containment pressure and temperature on the natural deposition rate of aerosols. At the same time, it can significantly improve the calculation accuracy of the residual mass of aerosols in the containment. The performance of improved ISAA can meet the requirements for analyzing the natural deposition behavior of aerosol in containment of advanced PWRs in severe accident. In the future, further optimization will be made to address the problems found in the current aerosol model.

Experimental assessment of hydrogen removal rate for passive auto catalytic recombiners

Generation and accumulation of hydrogen in containment atmosphere could pose a potential threat to the integrity of containment as hydrogen can form flammable mixture with air in the containment during postulated severe accident scenario involving multiple failures. Studies on hydrogen generation and distribution for various accident scenarios in reactor containment of Indian pressurized heavy water reactors (PHWRs) indicate that during severe accident with core damage, the possibility of global deflagration in absence of mitigating features is high and may challenge the integrity of containment. Therefore, it is essential to incorporate mitigating provisions inside the containment to prevent the hydrogen reaching deflagration/ detonation limits during such a scenario.Out of several options available for hydrogen mitigation, Passive Auto-catalytic Recombiner (PAR) is one of the most effective and recommended (IAEA -TECDOC-1661(2011)) technology for the removal of hydrogen inside the containment. The testing and evaluation of the performance of PAR for different conditions envisaged in the containment (such as pressure, temperature, steam concentration and poison) is a vital requirement for the deployment of these devices in the nuclear power plants (NPPs). The behavior of catalytic recombiners in terms of hydrogen removal rate with varying hydrogen concentrations is required to be evaluated essentially to work out the numbers and location of recombiners inside the containment. Hence, for testing of catalytic recombiners, an instrumented test facility simulating different reactor containment environment under controlled conditions, has been commissioned. Recombiner Test Facility (RTF) is a test rig for testing of full scale prototypes of catalytic recombiners.Results of Hydrogen Removal Rates (HRR) for varying hydrogen concentration, method for the calculation of HRR and its validation, model equation developed for Corrected Hydrogen Removal Rates (CHRR) usable in numerical simulations are presented in this article. With the proposed method for CHRR, it was established that the Recombiner performance can be evaluated.