Development of the Method to Evaluate the Concentration of Containment Gas Considering the Radiolysis of Water
1Toshiba Energy Systems & Solutions Corporation; 2Hitachi-GE Nuclear Energy, Ltd.
In BWRs, a containment is inerted with nitrogen during the normal operation; even if a severe accident happens, in which a large amount of hydrogen is generated by metal-water reaction, hydrogen combustion will not occur immediately in the containment due to the lack of oxygen. However, if no countermeasure to control the composition of containment gas is conducted, oxygen concentration will gradually increase by the radiolysis of water so that the risk of hydrogen combustion will also increase. Hence, it is important to evaluate the generation of the hydrogen and oxygen by the radiolysis of water and transportation of the each containment gas species.
However, most of the major severe accident code widely used by utilities do not model the radiolysis of water. It means that it is difficult to estimate the distribution of hydrogen and oxygen generated by the radiolysis of water along the severe accident; we cannot correctly anticipate the time when the concentration of oxygen reaches the flammable limit in the containment.
From the above background, the method that can evaluate the concentration of containment gases during severe accident considering the radiolysis of water, which is affected by the distribution of radioactivity source, that is, fission product and corium, related to the thermal hydraulic condition of containment, has been developed by the superposition of the containment gas distribution calculated by the severe accident code and radiolysis gas distribution calculated by this model itself. Using this method, severe accident analysis for BWR Mark-I type of containment was carried out and its appropriateness was confirmed. As the result, it was confirmed that the developed method is appropriate to evaluate the containment gas distribution and the concentration of each containment gas considering radiolysis of water at the severe accident condition.
Comparison of decontamination factors from coupled and uncoupled aerosol removal models in rising bubble region under pool scrubbing
Korea Advanced Institute of Science and Technology
After the Fukushima accident in 2011, extensive research has been conducted to prevent severe accidents. For instance, containment filtered venting systems CFVS have been introduced to reduce the amount of radioactive material spreading into the environment. These systems make use of the pool scrubbing method, according to which radioactive materials are removed by injecting a gas containing such materials into the water. Several codes have been developed to evaluate pool scrubbing effect, which calculates the decontamination factor to predict the removal of radioactive material by the pool. Nowadays, three representative pool scrubbing codes exist, namely BUSCA, SPARC, and SUPRA. These pool scrubbing codes are based on several phenomena: Brownian diffusion, gravitational sedimentation, and inertial impaction. Although these codes perform calculations based on Fuchs' theory in the bubble rise region, SUPRA assumes that each aerosol removal mechanism occurs independently. On the other hand, BUSCA and SPARC describe the pool scrubbing phenomenon with all removal mechanisms coupled. As for calculations based upon uncoupled removal mechanisms, the areas of bubble deposition are not considered, while the other calculations describe these areas using the net velocity. This study aims at investigating the best computational approach to physically described the pool scrubbing phenomenon.
Development of Korean Filtered Containment Venting System
FNC Technology Co., Ltd.
A Filtered Containment Venting is one of the options to prevent the containment over- pressurization during the severe accident. The Korean Filtered Containment Venting System (K-FCVS) based on the wet scrubber was developed with the key filtration components such as a self-priming scrubber nozzle submerged in water pool, cyclone for droplet separation, metal fiber filter as well as molecular sieve. The scrubbing performance of K-FCVS has been verified by the component tests, integral tests and independent third party tests against the aerosol, elemental iodine and organic iodide under the various operating conditions. The integral tests have been performed in the full height integral test facility named ARIEL (Aerosol Removal and Iodine Elimination) installed at KAERI (Korea Atomic Energy Research Institute) and independent third party tests were conducted by PSI (Paul Scherrer Institut) in Switzerland using a full height test mockup of K-FCVS connected to VEFITA facility. Its integral scrubbing performances as Decontamination Factor (DF) shows over 1,000 for submicron aerosol, over 100 for elemental iodine and over 50 for organic iodide.
An Impact Analysis of MCCI Uncertainty Factors in Level-2 PSA for PWR Large Dry Containment
Korea Atomic Energy Research Institute
Ex-vessel debris coolability (EDC) is a major issue threatening containment integrity during core melt accidents, inducing molten core concrete interaction (MCCI) in the reactor cavity and then a large release of fission products to the environment. Nuclear safety law, revised in 2015, specified the safety goal of “Cs-137 release”, for which safety research on EDC and MCCI possibility is in active progress in Korea. As a severe accident management strategy, pre-flooding of coolant into the reactor cavity of large dry containment has been adopted for both ex-vessel corium cooling and stabilization in most operating Korean PWR NPPs. If the molten corium is not coolable, MCCI phenomena occurs in the reactor cavity, which can be a grave threat to containment due to over-pressurization and/or basemat melt-through (BMT). To find out impact factors on MCCI (and finally on BMT), uncertain parameters are selected based on the success criteria of EDC applied in the existing Level-2 PSA (Probabilistic Safety Assessment) for Korean PWRs. Then, using sensitivity analyses performed from a perspective of BMT frequency determination, the important but uncertain factors are shown for both PSA and SAM (Severe Accident Management) application on EDC/MCCI.
In conclusion, from the design standpoint, it is essential under severe accident conditions both to be able to always supply water to the cavity and to make the corium layer in the cavity spread evenly. Therefore, in terms of the phenomenological study, it is important to provide PSA branch information such as (1) the accumulated shape associated with the thickness of the corium layer in the reactor cavity, (2) the condition in which the corium is not cooled despite the wet or flooded cavity, and (3) the initial conditions (within 3 days or in the long term) for BMT occurrence under situations imposed by (2) and (3).
A Simplified Core Catcher Model for the Containment Code AC2/COCOSYS
In reactor designs of Generation 3+ plants like EPR or WWER-1200 the concept of an ex-vessel core catcher plays an important role for the individual severe accident mitigation strategy. The computer code AC2/COCOSYS is designed to simulate the conditions in light water reactor containments during all relevant stages - from normal operation up to severe accidents. A versatile core catcher model in COCOSYS is developed with the main objective to include core melt retention and cooling in a generic core catcher within thermal-hydraulic containment simulations. Focus is on the characterization of the corium in the core catcher during the simulation time, including the processes during melting of sacrificial material and during the long-term cooling phase utilizing a passive or active cooling circuit. The core catcher model is based on the existing molten corium / concrete interaction (MCCI) model in COCOSYS with additions for melt confinement within cooled structures and it is flexible enough to accommodate various designs like e. g. the EPR or the WWER.
The current state of the model is outlined. As a first example, the calculation of the water-cooled basemat test OECD WCB-1 is discussed. The specific objective of the WCB-1 test conducted in the frame of the OECD MCCI project was to provide prototypic data on the evolution and stabilization of a core melt in a generic water-cooled core catcher design. The most relevant experimental data of WCB-1 are adequately captured in the COCOSYS simulation when using empirical data for heat transfers, including the MCCI phase. These empirical heat transfer data agree well with other findings from recent analytical work on the interpretation of MCCI tests.