Reactor Containment Research Articles

Long-term decay heat removal in a submerged SMR

Following the Fukushima-Daiichi nuclear accident in March 2011, innovations are needed to improve the reliability of new generation nuclear power plants toward scenarios where electrical power and ultimate heat sink are lost. This paper describes an integral design and basic passive safety strategy of a Small Modular Reactor (SMR) submerged in the sea or in an artificial lake, then performing a preliminary analysis of the long-term decay heat removal. The analysis considers a pressurized reactor placed in a horizontal cylindrical hull, which is surrounded by the external water. The simulated system is based on the Flexblue concept, developed by French company DCNS (now Naval Group). The object of the investigation is the natural circulation in the submerged containment, which is the key component for the long-term cooling. Following a rupture in the primary circuit, decay heat must be removed according to a fully passive safety strategy for an indefinitely long period. The purpose of this work is to study the effectiveness of a sump natural circulation flow to cool the fuel rods, up to several days after the scram. Decay heat generates steam in the core, which is released in the containment and condensed on the metal surface, transferring the heat to the exterior. Relap5-Mod3.3 has been employed to simulate the accident scenario. Results show the consistency of the safety principles and stimulate experimental investigations. However, the sensitivity analysis identifies the nodalization of the reactor containment as a modeling and numerical issue, deserving further analyses.

Numerical Investigation of Natural Convective Condensation with Noncondensable Gases in the Reactor Containment after Severe Accidents

The heat and mass transfer processes of natural convective condensation with noncondensable gases are very important for the passive containment cooling system of water cooled reactors. Numerical simulation of natural convective condensation with noncondensable gases was realized in the Fluent software by adding condensation models. The scaled AP600 containment condensation experiment was simulated to verify the numerical method. It was shown that the developed method can predict natural convective condensation with noncondensable gases well. The velocity, species, and density fields in the scaled AP600 containment were presented. The heat transfer rate distribution and the influences of the mass fraction of air on heat transfer rate were also analyzed. It is found that the driving force of natural convective condensation with noncondensable gases is mainly caused by the mass fraction difference but not temperature difference. The natural convective condensation with noncondensable gases in AP1000 containment was then simulated. The temperature, species, velocity, and heat flux distributions were obtained and analyzed. The upper head of the containment contributes to 35.1% of the total heat transfer rate, while its area only takes 25.4% of the total condensation area of the containment. The influences of the mass fraction of low molecular weight noncondensable gas (hydrogen) on the natural convective condensation were also discussed based on the detailed species, density, and velocity fields. The results show that addition of hydrogen (production of zirconium-water reaction after severe accident) will weaken the intensity of natural convection and the heat and mass transfer processes significantly. When hydrogen contributes to 50% mole fraction of the noncondensable gases, the heat transfer coefficient will be reduced to 45%.

Investigation of Mechanical Properties and Ductile-Brittle Transition Behaviors of SA738Gr.B Steel Used as Reactor Containment

The low temperature tensile properties, Charpy-V notch impact performance and fracture toughness of SA738Gr.B steel plate for domestic CAP1400 containment vessel were tested. On this basis, the reference temperature T0 of the master curve method was obtained. The fracture toughness distribution of the steel in the whole ductile-brittle transition zone is predicted and its applicability is verified by the theoretical basis of the master curve method. The results show that the reference temperature of SA738Gr.B steel master curve method is-123.6 °C. The master curve method is appropriate for SA738Gr.B steel with domestic nuclear containment vessel.

Aluminum corrosion in reactor containment environment following a loss of coolant accident (LOCA): High-temperature flow loop tests

A high-temperature flow loop system was designed to study the effects of buffers agents, temperature and flow conditions on the release of aluminum corrosion products in post-LOCA scenarios. Inductively coupled plasma-Optical Emission Spectrometry (ICP-OES) and X-ray photoelectron spectroscopy techniques are used to characterize the aluminum release in solution and the oxide layers formed on the surface. Aluminum release in trisodium phosphate solution is significantly less than in sodium tetraborate solution under same operating conditions. Flow accelerated aluminum release is evident, and power law correlations between the aluminum release rate and mass transfer coefficient are developed, allowing transfer of data between different flow systems.

The corrosion of aluminium alloy and release of intermetallic particles in nuclear reactor emergency core coolant: Implications for clogging of sump strainers

Abstract Clogging of sump strainers that filter the recirculation water in containment after a loss-of-coolant accident (LOCA) seriously impedes the continued cooling of nuclear reactor cores. In experiments examining the corrosion of aluminium alloy 6061, a common material in containment equipment, in borated solutions simulating the water chemistry of sump water after a LOCA, we found that Fe-bearing intermetallic particles, which were initially buried in the Al matrix, were progressively exposed as corrosion continued. Their cathodic nature vis-a-vis the Al matrix provoked continuous trenching around them until they were finally released into the test solution. Such particles released from Al alloy components in a reactor containment after a LOCA will be transported to the sump entrance with the recirculation flow and trapped by the debris bed that typically forms on the strainer surface, potentially aggravating strainer clogging. These Fe-bearing intermetallic particles, many of which had a rod or thin strip-like geometry, were identified to be mainly the cubic phase αc-Al(Fe,Mn)Si with an average size of about 2.15 μm; 11.5 g of particles with a volume of about 3.2 cm3 would be released with the dissolution of every 1 kg 6061 aluminium alloy.

Challenges and Safety in Erection and Commissioning of 280/85 Tons Single Failure Proof EOT Crane at PFBR

The Electric overhead traveling (EOT) crane installed in Prototype Fast Breeder Reactor (PFBR), Reactor Containment Building (RCB) is the largest capacity EOT crane installed in any nuclear power plants in India with some unique features of its own. It has been built with a single failure proof concept which is one of the prime requirements of a nuclear facility as per the standards of NUREG/IS. It is a double bridge girder type; the girder used in this crane is a single structure having a length of 34.8m without any joints as designed by IGCAR/BHAVINI (first of its kind). The span of the EOT crane is 33.4m and installed at a height of 42m from the ground level with the lifting height of 42.5m. This crane was manufactured by an Indian company M/s. UNIQUE CRANE, NASIK. Two hooks (280 tons / 85 tons) are connected in the same trolley sharing the Long Travel (LT). Load testing of the complete crane was done at factory before transporting it to PFBR site. The components of the crane were assembled in PFBR site. The erection methodology adopted was distinct w. r. t the site condition, as the components which were handled were very heavy, available space for installation was restricted and the head room was also very less. This erection has been completed successfully and very safely without any notable incidents. This activity is a biggest achievement / milestone in the nuclear industry in India. Later the EOT crane was commissioned, load tested with 350 tons load (negligible deflection) and being operated successfully without any incidents so far. This was achieved by accurate erection procedure, stringent Quality Assurance Plan (QAP), Job Hazard Analysis (JHA) and Safe Operation Procedure (SOP). Many challenges were faced during manufacturing, testing, erection and load testing at site. This paper deliberates those challenges.

Numerical method to simulate detonative combustion of hydrogen-air mixture in a containment

During an accident accompanied by core damage in a water-cooled nuclear reactor, a large amount of hydrogen is generated by fuel clad oxidation and released into the reactor containment. Hydrogen mitigation in the containment during a severe accident is important because the integrity of the structures important to safety could be threatened by an explosive combustion of the hydrogen. Even when equipped with a hydrogen mitigation system, if a highly concentrated hydrogen mixture cloud is developed in a compartment of the containment, the compartment integrity with a pressure load from a local detonation must be evaluated. In this study, a reliable numerical method to simulate the detonation of a hydrogen-air mixture was developed for the evaluation of a pressure load from a detonation. The numerical method has the capability to capture shocks robustly as well as resolve combustion phenomena stably. The Euler equations of mass, momentum, energy and gas species are employed as the governing equations of detonation. A central-upwind scheme was adopted as a shock capturing scheme and a reduced chemical kinetic model was applied for combustion. The numerical method was validated by solving experimental cases and applied for understanding detonation shock behaviour in a generic containment.

CADOR “Core with Adding DOppleR effect” concept application to sodium fast reactors

Generation-IV sodium fast reactors (SFR) will only become acceptable and accepted if they can safely prevent or accommodate reactivity insertion accidents that could lead to the release of large quantities of mechanical energy, in excess of the reactor containment's capacity. The CADOR approach based on reinforced Doppler reactivity feedback is shown to be an attractive means of effectively preventing such reactivity insertion accidents. The accrued Doppler feedback is achieved by combining two effects: (i) introducing a neutron moderator material in the core so as to soften the neutron spectrum; and (ii) lowering the fuel temperature in nominal conditions so as to increase the margin to fuel melting. This study shows that, by applying this CADOR approach to a Generation-IV oxide-fuelled SFR, the resulting core can be made inherently resistant to reactivity insertion accidents, while also having increased resistance to loss-of-coolant accidents. These preliminary results have to be confirmed and completed to meet multiple safety objectives. In particular, some margin gains have to be found to guarantee against the risk of sodium boiling during unprotected loss of supply power accidents. The main drawback of the CADOR concept is a drastically reduced core power density compared to conventional designs. This has a large impact on core size and other parameters.

Feasibility Study of Beta Detector for Small Leak Detection inside the Reactor Containment

This work was supported by the Korea Institute of EnergyTechnology Evaluation and Planning (KETEP) grant fundedby the Korea government (MOTIE) (20181510102340, Development of a real-time detection system for unidentified RCSleakage less than 0.5 gpm).

Gamma and neutron dosimetry of Tehran Research Reactor containment during and after LOCA accident

The radiation effects of loss of coolant accident on the reactor operators are critical issues. In the present study, computational codes were used to investigate gamma and neutron dose rates inside the 5 MW pool-type Tehran Research Reactor containment. The carried out calculations showed, when the remaining coolant over the core is decreasing less than 1 m, the containment should be evacuated. In addition, 24 h after LOCA, the containment gamma dose rates drop to one tenth of their initial values immediately after LOCA.

Experimental investigation of gas transport induced by containment spray activation during a scaled severe accident scenario

Abstract We present the results of an experimental study on the activation of sprays representative of those used in a generic light water reactor containment for severe accident mitigation. Depending on the initial conditions, the spray nozzle type and the spray water flow rates, the spray may cause higher hydrogen concentration during depressurization due to steam condensation, or it may erode the hydrogen stratification by enhanced mixing. To investigate these phenomena, the tests were performed in the PANDA facility at Paul Scherrer Institut, Switzerland in the frame of the ERCOSAM-SAMARA project. The experiments followed a test sequence based on a postulated severe accident scenario scaled from a nuclear power plant design. The scaled scenario consisted of the formation of a stratified helium-rich (substitute for hydrogen) atmosphere in one of the PANDA vessels followed by the activation of a spray. Temporal evolution and spatial distribution of the gas temperature and the gas concentrations at selected positions were measured using thermocouples and mass spectrometers. Four tests were performed with two different spray nozzle types, two spray flow rates and two initial vessel wall temperatures of 105 °C and 135 °C, which created condensing and non-condensing environments, respectively. The different initial conditions lead to different density stratification. The effect of these different initial and boundary conditions on the transport and mixing of gases in the vessels due to the activation of the spray is revealed by these tests. The spray causes a fast breakup of the stratification and a complete mixing in the vessel in which it was injected. The concentration measurements in the adjacent connected vessel showed the presence of helium at the bottom of the vessel suggesting the possibility of hydrogen transport to the lower parts of the containment in a three gas mixture environment.

Numerical simulation of injection characteristics, hydrodynamics and absorption of iodine vapour in a venturi scrubber operating in self-priming mode

Abstract Post Fukushima, many utilities are planning to install the Filtered Containment Venting System (FCVS) to protect the nuclear reactor containment under severe accident scenarios. FCVS aims at depressurizing the reactor containment and filtering the radioactive gas leaving the containment boundary under such scenarios. A key component of FCVS is the venturi scrubber. The hydrodynamics of the fluid flow inside the FCVS is complex involving multi-phase and multi-component systems with droplets formed by entrainment of high velocity gas stream with liquid jet. The size of droplets and their trajectories are critical with regard to decontamination of iodine which combines with the droplets inside the venturi. Thus, understanding the injection characteristics, hydrodynamics and absorption of iodine vapour with the droplets in the venturi scrubber is very important. There are very limited studies on this aspect. In view of this, a CFD simulation study has been performed in this work for understanding of venturi scrubber operation and performance in self-priming mode. The interaction of the continuous gas phase and droplets is assumed to occur through drag force. The gas flow rate is varied from 1700 to 3400 lpm for two submergence heights of 3 m and 4 m. The liquid loading is estimated from pressure drop at the nozzles observed in the experiments. The retention of iodine in venturi scrubber is calculated assuming instantaneous mass transfer at gas droplet interface. The mass transfer coefficient is evaluated from Steinberg and Treybal correlation. Our results indicate that the liquid loading to the scrubber increases at lower flow rates of 1700 lpm gas flow and further decreases with the increase in gas flow. The two phase pressure profile predicted is close to measured values for each case. The pressure decreases in the converging section; decreases sharply in the throat section due to liquid injection as droplets and shows negligible recovery in the diverging section of the venturi scrubber. The retention of iodine in the venturi scrubber increases with the increase in gas flow rate in the venturi scrubber. This indicates that the participation of venturi scrubber in the iodine retention increases as the air flow rate increases.

Computational studies on passive containment isolation system of AHWR

The Advanced Heavy Water Reactor (AHWR) incorporates several passive safety systems. Passive containment isolation system (PCIS) is one of such passive system, which is designed to prevent release of radioactivity to environment in the event of accidents such as Loss of Coolant Accident (LOCA) or Main Stream Line Break. PCIS isolates the reactor containment from atmosphere by making a water seal in ‘U’ shaped ventilation duct using the differential pressure between containment and atmosphere. This paper deals with Computational Fluid Dynamics (CFD) simulation of PCIS using Volume of Fluid (VOF) approach using OpenFOAM CFD code. CFD model is validated using Passive Containment Isolation Experimental Facility and dam break problem. CFD simulations are performed for PCIS of AHWR for 2%, 20%, 100% and 200% break size LOCA and estimated the time of isolation and amount of water transfer to ‘U’ duct. CFD result shows that PCIS of AHWR performed its intended work for all LOCA break sizes.

Исследование эффективности противоаварийных систем снижения давления в защитной оболочке реакторных установок атомных объектов морской техники

Исследование эффективности противоаварийных систем снижения давления в защитной оболочке реакторных установок атомных объектов морской техники

Risk and regulatory considerations for small modular reactor emergency planning zones based on passive decontamination potential

It has been argued that risk and performance-based approaches to licensing would be appropriate for Small Modular Reactors (SMRs) because their risk profiles differ from large-scale reactors. This is based on several factors including their limited electrical capacity of 300 MW, the below grade reactor vessel, and passive safety features. One design feature that can significantly reduce accident severity is the larger lateral surface area-to-volume (A/V) ratio of SMRs. Following a nuclear accident, this larger A/V ratio can increase the removal of radioactive particles due to natural phenomena compared to large light water reactors (LWRs). To quantify the improvements in safety, this work estimates the airborne radioactivity within containment and environmental dose exposure in a post-accident scenario for an advanced Generation III+ LWR (AP1000), a representative Generation II LWR (Surry), and an SMR. On average, the AP1000, Surry, and SMR produces 139, 153, and 104 curies/ft3 (182, 200, and 136 terabecquerels/m3) 75 min after a Loss-of-coolant-accident (LOCA). Using Monte Carlo simulations, the SMR produces less radioactivity per volume in containment than the AP1000 and Surry 84% and 96% of the time, respectively. On average, the AP1000, Surry, and SMR produces 84, 106, and 7 thousand curies/MWth (3.1, 3.9, and 2.5 petabecquerels/MWth) 75 min after a LOCA. The larger A/V ratio of the SMR plays a substantial role in reducing the radioactivity. While it is expected that the SMR would have a lower levels of radioactivity compared to the AP1000 and Surry, the SMR produces less radioactivity after normalizing by thermal reactor power and containment volume. With respect to environmental dose exposure, the US Environmental Protection Agency 1–5 rem (0.01–0.05 sieverts) Protective Action Guide (PAG) limits for whole body exposure is not exceeded at the 10-mile (16.1-km) EPZ using the mean estimates for the AP1000 and Surry. The iPWR does not exceed the 1 rem (0.01 sieverts) lower PAG limit for whole body exposure at the 5-mile (8-km) EPZ using the mean value. These findings can be used in conjunction with the improved analytical methods, found in the SOARCA study, to provide accurate and realistic estimates for exposure. This will help create a pathway to develop a regulatory basis for technology-neutral, risk-based approach to EPZs for iPWRs.

An evaluation of the range-gated-imaging technology under dense aerosol environments

Range-gated-imaging system, which can be used to eliminate backscatter in strong scattering environments, is based on two high speed technologies. It uses high power, ultra-short pulse laser as the light source. And it opens the optical gate of an ICCD camera with a micro-channel-plate image intensifier in a very short time while the laser pulses reflected by the object is coming back to the ICCD camera. Using this range-gated-imaging technology, the effect of scattered light can be reduced and a clear image is obtained. In this paper, the test results of the range-gated-imaging system under dense aerosol environments, which simulates environments in the reactor containment building when the severe accident of the nuclear power plant occurred, are described. To evaluate the observation performance of the range-gated-imaging system under such dense fog environment, we made a test facility. Fog particles are sprayed into the test facility until fog concentration is reached to the postulated concentration level of the severe accident of the nuclear power plant. At such dense fog concentration conditions, we compared and evaluated the observation performances of the range-gated-imaging system and the CCD camera.

Contribution of production and loss terms of fission products on in-containment activity under severe accident condition for VVER-1000

The purpose of this paper is to study the source term behavior after severe accidents by using a semi-kinetic model for simulation and calculation of in-containment activity. The reactor containment specification and the safety features of the containment under different accident conditions play a great role in evaluating the in-containment activity. Assuming in-vessel and instantaneous release of radioactivity into the containment, the behavior of in-containment isotopic activity is studied for noble gasses (Kr and Xe) and the more volatile elements of iodine, cesium, and aerosols such as Te, Rb and Sr as illustrative examples of source term release under LOCA conditions. The results of the activity removal mechanisms indicates that the impact of volumetric leakage rate for noble gasses is important during the accident, while the influence of deposition on the containment surfaces for cesium, mainly iodine isotopes and aerosol has the largest contribution in removal of activity during evolution of the accident.

Prevention of hydrogen accumulation inside the vacuum vessel pressure suppression system of the ITER facility by means of passive auto-catalytic recombiners

Hydrogen safety is a relevant topic for both nuclear fission and fusion power plants. Hydrogen generated in the course of a severe accident may endanger the integrity of safety barriers and may result in radioactive releases. In the case of the ITER fusion facility, accident scenarios with water ingress consider the release of hydrogen into the suppression tank (ST) of the vacuum vessel pressure suppression system (VVPSS). Under the assumption of additional air ingress, the formation of flammable gas mixtures may lead to explosions and safety component failure.The installation of passive auto-catalytic recombiners (PARs) inside the ST, which are presently used as safety devices inside the containments of nuclear fission reactors, is one option under consideration to mitigate such a scenario. PARs convert hydrogen into water vapor by means of passive mechanisms and have been qualified for operation under the conditions of a nuclear power plant accident since the 1990s.In order to support on-going hydrogen safety considerations, simulations of accident scenarios using the CFD code ANSYS-CFX are foreseen. In this context, the in-house code REKO-DIREKT is coupled to CFX to simulate PAR operation. However, the operational boundary conditions for hydrogen recombination (e.g. temperature, pressure, gas mixture) of a fusion reactor scenario differ significantly from those of a fission reactor. In order to enhance the code towards realistic PAR operation, a series of experiments has been performed in the REKO-4 facility with specific focus on ITER conditions. These specifically include operation under sub-atmospheric pressure (0.2–1.0 bar), gas compositions ranging from lean to rich H2/O2 mixtures, and superposed flow conditions.The paper gives an overview of the experimental program, presents results achieved and gives an outlook on the modeling approach towards accident scenario simulation.

Analysis of a Large-Scale PWR In-Vessel Retention Device

This study assessed the effectiveness of in-vessel retention (IVR) in terminating the progression of an accident sequence initiated by a station blackout and large loss-of-coolant accident in a pressurized water reactor with thermal power of approximately 5000 MW. In the IVR design, external reactor vessel cooling is established by flooding of the reactor cavity. A water channel is introduced into the outer wall of the reactor vessel, and an insulated layered structure is added around the vessel. The amount of heat removed from the corium pool in the vessel lower plenum is limited by the critical heat flux (CHF) at the outer surface of the vessel wall. An integrated assessment was conducted in three steps. First, the responses of the reactor coolant system and containment were simulated using MELCOR. The predicted transient heat load at the vessel wall was then fed into RELAP5-3D, where the flow of natural, buoyancy-driven convection within the IVR water channel was simulated. Finally, the main thermal-hydraulic parameters in the IVR channel were substituted into the ULPU, SULTAN, SBLB, and MELCOR CHF correlations, and the effectiveness of IVR was assessed. The MELCOR simulation demonstrated that the heat load at the vessel wall of the lower plenum is dependent on the configuration of the debris. The heat flux to the vessel wall reached a maximum at 483 min, at an inclination angle of approximately 68 deg. The peak heat flux moved from a small inclination angle to a larger angle as the accident progressed. Both MELCOR and RELAP5-3D calculations predicted a gradual buildup of natural convection flow within the IVR channel following the application of a heat load to the vessel wall. The MELCOR code significantly overpredicts the mass flow of natural convection flow. Both codes predicted that the flow would experience large-amplitude fluctuations as the water in the IVR flow channel reached saturation. These fluctuations were attributed to instability induced by two-phase flow. If the inlet temperature can be kept sufficiently low to obviate boiling in the IVR channel, RELAP5-3D predicts that the channel flow will approach an approximately steady state. The selected CHF correlations predicted significantly different CHFs. The MELCOR correlation, which is a correlation based on pool boiling, produced the most conservative predictions, and the CHFs predicted by SBLB had the highest value. The minimum margin was found between 55 and 75 deg in all correlations. With the exception of the MELCOR correlation, the CHF ratio predicted by the other three correlations is greater than 1.2.

Radiation dose level inside the control room of open pool type reactor during emergency situation

The objective of this paper is to estimate radiation dose level in the control room due to air contamination in the containment of open pool type reactor during emergency situation. A postulated core degradation accident, due to fuel element blockage, causes fission products to release to the reactor containment and then, the ventilation system would convert automatically from the normal situation to the emergency situation to purify the contaminated air by forcing it to pass through a group of filters. The study computed the dose rate in the control room, adjacent to the containment, during the emergency situation. The dose rate level in the control room depends on the degree of the core degradation and so, the maximum dose rate is corresponding to the complete degradation of the core and the minimum dose rate is corresponding to the degradation of one fuel plate. The dose rate in the control room was calculated using MCNP5 code and it ranged from 1.12 to 2.03E-3 Sv/h. The results also show that the dose rate level in the control room would continue having values more than the permissible limit for a long time and so, a suggestion of venting the radioactive isotopes from the containment into the environment was studied to decrease the dose rate level in the control room. The suggestion was possible because the maximum dose, resulting from the contaminated air venting, would fortunately locate in a free-inhabitant area.