Dekui Zhan’s research while affiliated with Nuclear Power Institute of China and other places

Molten corium-concrete interaction is an important phenomenon in the late phase of severe accident, threatening the integrity of containment and might causing potential large release of radioactivity. A sensitivity and uncertainty analysis of MCCI under severe accident of a 1000 MW advanced light water reactor (ALWR) was performed with ASTEC (MEDICIS) [1], a lumped parametric integral severe accident code developed by IRSN. Several representative phenomena were screened out referenced by EURSAFE severe accident Phenomena Identification and Ranking Table (PIRT) [2], including debris bed formation, layer configuration, heat sources, and convective heat transfer correlation. Related input parameters, sensitivity coefficients, and modeling options in ASTEC code were selected, such as convective correlations, layer stratifications, initial layer compositions, and the potential ranges of these parameters were identified. A simple sampling method was used to analyze the independent effect of each parameter/model. Key parameters were chosen to evaluate the impact of sensitivity parameters to the MCCI process. A large break loss of coolant accident scenario, where in-vessel melt retention is invalid is simulated as an initial event. The results emphasize the importance of layer configuration and fission product partition. However, it should be emphasized that analysis results may be quite uncertain due to the limitation of the physical models and the adequacy or validity of the selected range of input variables.

The core of the nuclear reactor will melt during severe accidents, which may damage the integrity of the reactor pressure vessel and containment, and release radioactive materials to the environment. The third-generation pressurized water reactor is equipped with severe accident mitigation systems to prevent the high temperature (3000K) corium from melting through the reactor pressure vessel. The severe accident mitigation system mainly includes: primary depressurization system, reactor pit flooding system, containment combustible system, containment heat removal system, etc.. Benefit from these systems, large release frequency (LRF) is restricted to a low level. However, these systems cannot stop the process of core degradation. If severe accident happens, the core will melt and the whole reactor cannot be reused again, causing irreversible economic losses. In order to improve the economy and safety of nuclear power plants, this paper proposes in-vessel-injection (IVI) in the early stage of severe accident, and proves the effectiveness of this measure. According to the analysis, in-vessel-injection in the early stage of severe accident can prevent the large-scale melting of the core during severe accident, and the risk of hydrogen and source items will be greatly reduced.

In-Vessel Retention (IVR) of molten corium is an important strategy for management of severe accidents adopted by some active service generation II plus reactors and advanced generation III reactors like AP1000, Hualong one and APR1400. One of the most important factors contributing to IVR analysis can be attributed to the characteristics of heat transfer in the molten pool in the lower head of reactor pressure vessel (RPV), which directly determines the heat load imposed on the vessel wall of the RPV lower head. Besides, IVR is also an important strategy for management of severe accidents of Ocean Floating Reactor (OFR), whose flow and heat transfer characteristics are influenced by the ocean motion conditions. However, since all previous molten pool experiments were conducted under static condition, characteristics of the heat transfer in the molten pool thus obtained may be different from those under ocean motion conditions, thus not applicable to the latter. To address this issue, in this study, an experimental system was established, which includes a two-layer molten pool test facility and a swing motion platform. Based upon the system, the effects of the swing angle and swing period on the characteristics of heat transfer in the two-layer melt pool were investigated. In addition, a scaling method was proposed to select simulant materials for the two-layer corium pool. Results of the experiment show that the swing period and angle have a significant effect on the temperature field of the molten pool and the rate of heat transfer. We thus suggest that, in future studies, new heat transfer correlations, models and IVR analysis method for ocean floating reactors should be researched and developed to evaluate the IVR design of ocean floating reactors, which is critical for management of severe accidents of the ocean floating reactors.

Under hypothetical Reactor Pressure Vessel (RPV) failure accidents in Light Water Reactor (LWR), Molten Corium-Concrete Interaction (MCCI) will cause erosion of cavity concrete, possibly resulting in containment failure due to basemat penetration and overpressure. The CINA (Corium-Concrete Interaction Apparatus) experiment was conducted to investigate the MCCI of metallic and oxidic corium with siliceous concrete. Simulant melt material metallic iron and oxidic alumina generated by exothermic thermite chemical reaction was used to investigate the two-dimensional erosion of a cylindrical crucible in CINA experiment. The crucible was fabricated from siliceous concrete with an inner diameter of 300 mm and a height of 500 mm containing reinforcement (rebars). Decay heat in the melt was simulated by subsequent sustained addition of thermite. During the experiment, the melt temperature was monitored by two thermocouples and a two-color optical endurance pyrometer. Besides, the phenomenon that appeared on the melt surface was recorded by the video camera placed above the crucible and the formation process of the crust anchoring was observed. After the experiment, the crucible was split in half by a wire saw to accurately measure the ablation depth. Axial-radial ablation depths were 25 mm and 13 mm, respectively. Current findings contribute to the further understanding of MCCI mechanisms and the optimization of MCCI mitigation strategies.

Due to large uncertainty, the effectiveness of in-vessel retention (IVR) with external reactor vessel cooling (ERVC) for high-power reactors cannot be fully demonstrated during the transient process. To optimize the current IVR-ERVC strategy, the concept of IVR-ERVC with in-vessel injection (IVI) is studied. Two feasible IVI designs are proposed: 1) using the passive IVR water tank to implement simultaneous water injection in-vessel and ex-vessel and 2) using the passive severe accident dedicated in-vessel injection tanks (SADITs) to implement water injection in-vessel. The research on the feasibility and effectiveness of IVI is performed, and the corresponding negative effects are analyzed. The calculation results by using MIDAC show that the two proposed IVI concepts can greatly delay the accident progression in the core and reduce the decay heat peak of the molten pool in the lower head, thereby improving the effectiveness of the current IVR-ERVC strategy. And the negative effects are within acceptable ranges.

Investigation was conducted experimentally and numerically for ex-vessel melt behavior and concrete interaction during Molten Corium-Concrete Interaction (MCCI) in current research. The CINA (Corium-Concrete Interaction Apparatus) experiment was carried out with a total of 94 kg melt generated by exothermic thermite chemical reaction to react in a two-dimensional (2-D) cylindrical siliceous crucible with an inner diameter of 300 mm and a height of 500 mm. Meanwhile, the variation of melt temperature was monitored during the experiment. Besides, the accurate ablation depth and thickness of stratified layers were measured after the experiment. Numerical simulation was used to study the mechanism of high-temperature melt heat transfer and concrete ablation during the MCCI process. The slag film model to calculate the melt/concrete interface heat transfer coefficient and the stratified melt model were both used to simulate the above experiment. The results showed that the numerical results were in good agreement with experiment measurements by the slag film model under separated layers condition.

A Large-Break Loss of Coolant Accident (LB-LOCA) in a generic, 1000 MWe and 3-loop Chinese Pressurized Water Reactor (PWR) with the In-Vessel corium Retention (IVR) strategy was simulated by using ASTEC V2.1.1. One of the criteria of assessing the validity of IVR is that the decay heat of molten pool could be removed successfully, which means that the local heat flux will not exceed the Critical Heat Flux (CHF) of outer nucleate boiling. ASTEC determines the process of core degradation, the corium relocation and the transient heat flux to indicate whether the vessel can maintain its integrity under the LB-LOCA. From the core degradation process, for this LB-LOCA scenario, the corium with fission products slumps into the lower plenum via the downcomer, which means the sideward relocation. After the initial relocation, the mass of compositions of molten pool in lower head are almost steady except the mass of steel. The augmentation of the steel in the lower head is consistent with the melting process of the lower support plate, the fluid distribution structures and the vessel inner wall. At the azimuth from 78° to 87°, the transient heat flux is much larger compared with the other locations. The transient heat flux reaches its maximum value which is about 1.1 MW/m² at the azimuth of 84°. And the maximal ratio of the transient local heat flux to CHF is about 0.8. For the residual thickness of the lower head, the minimum is 37.35 mm at the azimuth from 78° to 87°. The results of thermal analysis conclude that the vessel can maintain its integrity during LB-LOCA scenario because the transient heat flux from the vessel wall to the external coolant do not exceed the CHF of outer nucleate boiling.

In this work, a comparative study was conducted on the tensile behaviors of two representative accident-tolerant cladding coatings, i.e., CrN coating and Cr coating, at both room temperature and 400 °C by in situ tensile testing. The surface and interfacial cracking behaviors of the coatings were experimentally tested and analyzed, with the surface crack densities predicted by a modified shear-lag model. The results showed that CrN coating exhibited brittle fracture at both room temperature and 400 °C, while the failure of Cr coating showed brittle-to-ductile transition when increasing the temperature from room temperature to 400 °C. Moreover, Cr coating exhibited better crack resistance than that of CrN coating under mechanical loading at both temperatures, due to: (i) the higher fracture toughness and ductility of Cr; (ii) better deformation compatibility between Cr coating and Zr-4 substrate. In particular, at 400 oC, Cr coating with excellent plastic deformability exhibited a protective or prohibitive effect on the crack initiation process of the Zr-4 substrate. It is indicated that the mechanical deformation properties and failure mechanisms of Cr and CrN coatings are important factors that should be considered in the selection and evaluation of accident-tolerant coatings.

In this study, a thermal-hydraulic analysis code has been developed for primary system of the integrated small modular pressurized water reactor IRIS (International Reactor Innovative and Secure). Numerical model of the primary system has been established, and characteristics of equipment and layout are considered. The sub-channel model is used for performance analysis of hot channel in the core. In modeling of the helically coiled steam generator, effects of secondary flow inside coiled tubes are taken into account. Based on the model established, the simulation code THAS (Thermal-Hydraulic Analysis code of SMRs) has been developed. To benchmark this code, steady state performance of the system has been calculated and compared with design parameters as well as results obtained with RELAP5. Furthermore, transient performance of the system is also simulated and compared with RELAP5. The comparison presents a good agreement, which confirms simulation capacity and accuracy of the THAS code.

In order to enhance the ability of severe accident mitigation for Pressurised Water Reactor (PWR), different kinds of severe accident mitigation strategies have been proposed. In-Vessel Retention (IVR) is one of the important severe accident management means by External Reactor Vessel Cooling. Reactor cavity would be submerged to cool the molten corium when a severe accident happens. The success criterion of IVR strategy is that the heat flux which transfers from the corium pool must be lower than the local critical heat flux (CHF) of the reactor pressure vessel (RPV) outside wall and the residual thickness of the RPV wall can maintain the integrity. The residual thickness of RPV is determined by the heat flux transfer from the corium pool and the cooling capability of outer wall of the RPV. There are various factors which would influence the CHF and the cooling capability of outer wall of the RPV. In order to verify the optimized design which is beneficial to the heat transfer and the natural circulation outside the actual reactor vessel, a large-scale Reactor Vessel External Cooling Test (REVECT) facility has been built. A large number of sensitivity tests were carried out, to study how these sensitivity factors affect CHF value and natural circulation. Based on the test results, the structure of the test section flow channel has an obvious effect on the CHF distribution. The flow channel optimized can effectively enhance the CHF value, especially to enhance the CHF value near the “heat focus” region of the molten pool. The water level in the reactor pit has also a great impact on the natural circulation flow. Although natural circulation can be maintained with a low water level, it will lead to a decrease of the cooling capacity. Meanwhile, some noteworthy test phenomena have been found, which are also essential for the design of the reactor pit flooding system.