Ex-vessel core melt stabilization research (ECOSTAR)

In Chaps. 7 and 9 the results of the risk studies for LWRs and the facts and consequences of the severe nuclear accidents of Chernobyl and Fukushima were assessed. The severe nuclear accidents of Chernobyl and Fukushima resulted in contamination of large areas and in the evacuation of the population of large areas.It is concluded that LWRs in Europe are built and operated in densely populated areas with cities of several 100,000 inhabitants. It is impossible to evacuate cities of that size in due time and their contamination is beyond anybody’s imagination. Therefore, the KHE safety concept was proposed around 1990 to build future LWRs such that the main accident path of the risk studies leading to the high number of deaths and to the large contamination of land should be eliminated by the safety design concept of the plant. The probabilistic safety studies leading to low frequencies of core melt in the range of 10−6 per year should be retained. However, the present risk concept should not be applicable any more to LWR plants built in future in Europe. It was concluded that the consequences of the most severe accidents in risk studies should be managed by the inner and outer containment of the reactor plant. A severe accident research program of the Research Center Karlsruhe, Germany, lasting for about two decades showed that it is possible to fulfill the requirements of the KHE concept. The KHE concept was essentially accepted by the German and French Safety Commissions and also incorporated in the German Atomic Law in 1994.The main findings of the Karlsruhe research program for severe accidents were as follows: A large scale steam explosion in the reactor pressure vessel of a PWR would not jeopardize the mechanical integrity of the reactor pressure vessel. Consequently a steam explosion followed by failure of the integrity of the reactor pressure vessel and of the containment as assumed in WASH-1400 and the German Risk study can be considered impossible. A large scale hydrogen detonation after core melt in the spherical containment of a KWU PWR-1300 reactor plant cannot jeopardize the integrity of the steel containment. Nevertheless hydrogen recombiners in the containment are useful for preventing or mitigating some accident sequences. Core melt down after a break of a pipe of the residual heat removal system in the annulus of the containment system or core melt down after an uncontrolled large scale steam generator tube break can be avoided by proper design. Containment failure after core melt down under high primary coolant pressure, as assumed in WASH-1400 and the German risk study, can be avoided by manual or automatic opening of pressure relief valves (ADS systems of Chap. 3) or—as a limiting case—reinforced anchorage of the pressure vessel cover. Core melting through the bottom part of the reactor pressure vessel can be counteracted by flooding the reactor pressure vessel on the outside with water (severe accident measure). Similar results can be obtained by installing a molten core spreading and cooling device (core catcher). Appropriate core catcher designs were developed after research. A rising steam pressure in the inner containment after a core melt can be avoided by water spray systems. A mechanically intact double containment, where the inner containment is either a steel containment or a concrete containment with an inner steel liner, having a leak rate of 0.3–1 % per day after a core melt accident and where the leaking radioactive gases and aerosols are passed from the annulus through aerosol filters to a stack can fulfill the requirements of the KHE safety concept. In this case the contamination by radioactive fission products is essentially limited to the site of the reactor plant. No evacuation of the population is necessary. In summary all most severe accident consequences found in the WASH-1400 or the German risk study can be either controlled or eliminated or managed in future LWRs by appropriate design measures. Examples for such designs are EPR and KERENA (SWR-1000). Concluding remarks compare the KHE safety concept with the safety concept of presently operating reactors and more recent designs for LWRs.

Controlling and cooling core melts outside the pressure vessel

A study on the minimum steam cooling pressure of EOP for BWR power plants

Ablation and thermal stress analysis of RPV vessel under heating by core melt

In-vessel Melt Retention (IVR) system is a typical design feature of the the generation III nuclear power plant. In case of severe accident, enough water is injected into the reactor pit to cool the outside of the Reactor Pressure Vessel (RPV) to maintain the integrity of the reactor pressure vessel, so that the core melt is retained in RPV. The design concept of IVR is very meaningful for maintaining the integrity of RPV and reducing the risk of large radioactivity release, but on the other hand, it may also introduce the possibility of inadvertent reactor pit flooding during normal operation, which may threaten the integrity of RPV. At home and abroad, the research on IVR mainly focuses on whether its safety function can be realized, but less attention is paid to its potential negative impact. This paper takes the China generation III reactor with the IVR design feature for the study, carry out the possible inadvertent reactor pit flooding path analysis and frequency analysis. Based on the risk study, put forward the concerns to prevent inadvertent reactor pit flooding, which will provide guide for the subsequent IVR design and optimization.

Reactor pressure vessels (RPV) are components with the highest importance for the reactor safety and operation as they contain practically whole inventory of fission material but they are damaged/aged during their operation by an intensive reactor radiation. Surveillance specimen programs are the best method for monitoring changes in mechanical properties of reactor pressure vessel materials if they are designed and operated in such a way that they are located in conditions close to those of the vessels. Reactor Codes and standards usually included requirements and conditions for such programs to assure proper vessel monitoring. WWER (Water-Water-Energetic Reactors) reactor pressure vessels are designed according to former Russian Codes and rules with somewhat different requirements using different materials comparing e.g. with ASME Code. Two principal types of WWER reactors were designed, manufactured and are operated in several European countries (and also in China, Iran): WWER-440 and WWER-1000. Their surveillance programs were designed in quite different way, with some modifications due to the time, country of manufacturing and experience gained from their operation. The paper gives a critical comparison of these programs in both types of reactors with requirements of both Russian and ASME/ASTM Codes and Standards. Finally, information about creation of the Integral Surveillance Program for WWER-1000 type reactor pressure vessels covering vessels from several countries is described.

Integrity assessment of reactor pressure vessels (RPVs) can be performed either by deterministic fracture mechanics (DFM) or/and by probabilistic fracture mechanics (PFM) analyses. In European countries and Switzerland, only DFM analyses are required. However, in order to establish the probabilistic approach in Switzerland, the advantages and shortcomings of the PFM are investigated in the frame of a national research project. Both, the results from DFM and PFM depend strongly on the previous calculated thermal-hydraulic boundary conditions. Therefore, complete integrity analyses involving several integrated numerical codes and methods were performed for a reference pressurized water reactor (PWR) RPV subjected to pressurized thermal shock (PTS) loads. System analyses were performed with the numerical codes RELAP5 and TRACE, whereas for structural and fracture mechanics calculations, the FAVOR and ABAQUS codes were applied. Additional computational fluid dynamics analyses were carried out with ANSYS/FLUENT, and the plume cooling effect was alternatively considered with GRS-MIX. The results from the different analyses tools are compared, to judge the expected overall uncertainty and reliability of PTS safety assessments. It is shown that the scatter band of the stress intensities for a fixed crack configuration is rather significant, meaning that corresponding safety margins should be foreseen. The conditional probabilities of crack initiation and RPV failure might also differ, depending on the considered random parameters and applied rules.

Characterization of PWR vessel steel tearing under severe accident condition temperatures

During postulated severe accidents in a light water reactor (LWR), the core melt (corium) may relocate to the lower head and fail the reactor pressure vessel (RPV). The corium is expected to undergo fuel coolant interactions (FCI) if the reactor cavity is flooded with water. Both FCI energetics and resulting debris bed coolability are of paramount importance to reactor safety, since the ex-vessel corium poses a threat to the containment integrity if steam explosion occurs or the debris bed is uncoolable, leading to release of radioactive fission products to the environment. The present study is intended to quantify the characteristics of a debris bed resulting from FCI, which are crucial to debris bed coolability. Different from the previous studies with only oxidic materials, various materials, including metallic ones of Sn, Sn-Bi and Zn as well as oxidic one of Bi2O3-WO3, were employed as the simulants of corium (mixture of UO2/ZrO2/Zr/Fe) in the present study to investigate the effects of melt materials, melt superheat and coolant subcooling on debris bed formation in a water pool. High-speed photography was applied to visualize melt jet breakup, droplets fragmentation, as well as fragments sedimentation on the pool floor. Other obtained data are debris bed shape (profile) and porosity, as well as morphology and size distribution of debris particles. The comparative results of various tests provided insights toward filling the knowledge gap on debris bed characteristics under different melt materials and compositions.

For mitigation of severe accident with core melting, core-catcher has been developed to catch and cool molten core. The core-catcher developed by Toshiba for Advanced Boiling Water Reactor was designed to be installed under Reactor Pressure Vessel (RPV) and catch molten core in basin with thermal-resistant material. It also has structure including risers and downcomers to generate natural circulation flow of cooling water. On the other hand, there is not enough space to install it in the existing plants due to the height of the inclined cooling channels. Then, we have been developing flat core-catcher with flat cooling channels for the existing plants to reduce total height of the structure. Finned channels will be adopted to increase heat transfer rate by increasing heat transfer area. However, the thermal-hydraulics characteristics of such core-catcher has not been clarified due to the specific configuration, that is, the horizontal rectangular finned channel with the heated surface from upper side. This present study investigated the natural circulation characteristics and heat transfer behavior in the horizontal rectangular finned channel by experiment. Pressure drop, natural circulation flow and temperature were measured by changing heat fluxes. The flow was visualized to obtain flow pattern in the finned channel by a high-speed camera. The maximum value of test range of heat flux was 250 kW/m2, which is the value when the total amount of the molten core would be dropped to the core-catcher. The partial simulated test section of finned channel with 3.5 m length was heated from upper side by heaters to simulate the heat flux from the molten core. This length is the same as the inner diameter of RPV pedestal of the existing plants. The natural circulation mass flux increased as the heat flux increased and then two-phase flow pressure loss also increased. Consequently, circulation flow turned to decrease. As a result of the test, when the heat flux was 50 kW/m2, the circulation mass flux got to the maximum value, 230 kg/m2s. Under all the conditions except for the maximum heat flux of 250 kW/m2, the fin surface temperature was around the saturated temperature. At maximum heat flux 250 kW/m2, the temperature got 540 K. However, the structural soundness was maintained because it is lower than melting temperature of the fin. It can be concluded that the flat and high-thermal-conductivity core-catcher has enough cooling performance to catch and stabilize the molten core.

The gamma radiation emitted by a nuclear reactor core might contain information about the reactor state. This information may be used in a monitoring system for severe accidents. The Technische Universität Dresden and the Zittau/Görlitz University of Applied Sciences are currently carrying out feasibility studies for the development of such a system in a collaborative effort. As one part of such feasibility studies we performed Monte Carlo simulations on a simplified model of a generic pressurized water reactor. For a set of states which represent scenarios of a coolant level decrease and core melt, the gamma radiation distribution outside the reactor pressure vessel has been computed. The results are presented in this paper. They indicate that different coolant levels yield different gamma radiation distributions, and that an accumulation of corium inside the lower head is detectable from the outside.

Three major severe accidents with core meltdown/core disruption occurred at Three Mile Island (USA) in 1979, Chernobyl (Ukraine) in 1986 and Fukushima (Japan) in 2011.

In hypothetical severe accidents of Nordic boiling water reactors (BWRs), the core melt (corium) has the potential to relocate to the lower head, leading to the failure of the reactor pressure vessel (RPV). Consequently, the corium, a molten mixture of either UO2-ZrO2 or Fe-Zr, is anticipated to undergo fuel coolant interactions (FCI) in a deep-water pool within the reactor cavity which is supposed to be flooded during severe accidents. The characteristics of the resulting debris bed play an important role in corium coolability that is paramount to accident stabilization and termination. This study is dedicated to characterizing the debris bed resulting from the FCI using Fe-Sn as the simulant of Fe-Zr. Compared with other simulants in our previous studies, the Fe-Sn alloy has a a higher melting point of 1130 °C and the same element of Fe as in prototypical Fe-Zr. The Fe-Sn is melted in a clay graphite crucible with glass coating layer to minimize contamination of molten materials by graphite. Argon containing 1.26 % hydrogen is employed as cover gas to avoid the oxidation of Fe-Sn at high temperature. The process of debris bed formation, including melt jet breakup, debris fragmentation, and fragments’ sedimentation on the pool bottom, is meticulously recorded using high-speed cameras. The porosity of the debris bed is determined in the post-test analyses using a three-dimensional laser scanner and a graduated water tank. The laser scanner data are also used to reconstruct the geometric appearance of the final debris bed. The debris particulates are sieved to obtain their size distribution. The chemical interaction between water and melt is examined by the SEM analysis. The experimental results show that the debris bed of Fe-Sn has significant differences from those of low melting point simulants (Sn-Bi, Sn, Zn) in our previous studies.

During severe accidents in a light water reactor, the core melt (corium) may relocate to the lower head and fail the reactor pressure vessel (RPV). The corium will be ejected to the reactor cavity upon the RPV failure and undergo melt coolant interactions (FCI) if the cavity is flooded with water. The FCI process does not only determines the characteristics of the resulting debris bed which are important to coolability, but also induces a steam explosion risk which may threaten containment integrity. The present study is concerned with characterization of debris bed formed from FCI of metal-rich corium failing into a deep water pool in the reactor cavity. Low melting-point metals Tin and Tin–Bismuth (20 kg) were employed as the simulant materials of metal-rich corium melt. Ten tests were carried out on the DEFOR-M test facility at KTH to investigate the effects of various parameters on debris bed formation, such as melt superheat, coolant subcooling, material. The melt jet fragmentation and fragments movement in the water pool as well as debris deposition on the pool floor were recorded by high-speed cameras. Melt sensors and weight sensors were installed to detect the period of melt jet discharge and the mass of forming debris bed. The porosity of debris bed was obtained through the debris bed volume measured by a three-dimensional laser scanner and the pore volume measured by water absorption. The final configuration of debris bed was also reconstructed through the laser scanner data, and the debris particles were sieved for their size distribution. The experimental results revealed the FCI phenomena and debris bed characteristics including configuration and porosity of debris bed as well as morphology and size distribution of debris particles under different melt superheats, coolant subcooling, materials.

Reactor core degradation and in-vessel and ex-vessel corium behavior have been major research topics for the last three decades to which Institut de Radioprotection et de Sûreté Nucléaire (IRSN) strongly contributed by the coordination of or the contribution to large research programs and through the development and validation of the severe accident (SA) ASTEC code. In recent years, the balance of research efforts has trended toward analyses of pros and cons and assessments of mitigation measures. The outcomes of risk significance analysis [including fuel-coolant interaction (FCI), hydrogen combustion, and molten core–concrete interaction (MCCI) risks] performed in France and corium behavior research are described. The focus these days is on (1) in-vessel melt retention (IVMR) strategies for future reactor concepts and the need to establish the reliability of such strategies when implemented in existing reactors and (2) in-containment corium cooling for existing reactors.This paper summarizes the main achievements and remaining issues related to understanding and modeling of (1) reflooding of a degraded core where, despite substantial knowledge gained through research programs, additional efforts are required to establish the efficiency of such a measure and the associated risks for largely degraded cores; (2) corium behavior in the reactor pressure vessel (RPV) lower head where, despite the Organisation for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA) MASCA program results, efforts remain necessary to predict RPV thermal loadings resulting from corium layer evolution and RPV resilience with and without IVMR measures (internal and/or external cooling); (3) FCI for which, despite the OECD/NEA SERENA program results, the knowledge is not sufficient to assess with confidence the induced risk of containment failure; and (4) MCCI, where the knowledge on corium cooling in the containment by top and/or bottom water flooding is insufficient to formulate conclusions regarding the efficiency of such measures. Of particular interest for top flooding are the water ingress and corium eruption processes. Specifically for top flooding, respective impacts of water ingress and corium eruption processes remain to be quantified in reactor conditions.In support of these activities, substantial efforts are also being conducted at IRSN to constantly improve and validate nuclear material property databases that are key tools for corium behavior analysis.This paper describes ongoing and future research programs performed at IRSN or internationally with IRSN coordination or participation to tackle the remaining issues and summarizes expected progress in modeling for SA codes, in risk analysis and in SA management.