Two-dimensional interaction of oxidic corium with concretes: The VULCANO VB test series

Three two-dimensional Molten Core–Concrete Interaction tests have been conducted in the VULCANO facility with prototypic oxidic corium. The major finding is that for the two tests with silica-rich concrete, the ablation was anisotropic while it was isotropic for limestone-rich concrete. The cause of this behaviour is not yet well understood.Post Test Examinations have indicated that for the silica-rich concrete, the corium melt mixed specifically with mortar, while, for limestone-rich concretes, the analysed samples were in accordance with a corium–concrete mixing. The experimental results are described and compared to numerical codes. Separate Effect Tests with Artificial Concretes and prototypic corium are proposed to understand the phenomena governing the ablation geometry.     

High pressure corium melt quenching tests in FARO

First-of-a-kind experimental data on the quenching of large masses of corium melt of realistic composition when poured into pressurised water at reactor scale depths are presented and discussed. The tests involved 18 and 44 kg of a molten mixture 80 w% UO₂-20 w% ZrO₂, which were delivered by gravity through a nozzle of diameter 0.1 m to 1 m depth nearly saturated water at 5.0 MPa. The objective was to gain early information on the melt/water quench process previous to tests that will involve larger masses of melt (1.50 kg of mixtures UO₂---ZrO₂---Zr). Particularly, pressures and temperatures were measured both in the gas phase and in the water. The results show that significant quenching occurred during the melt fall stage with 30% to 42% of the melt energy transferred to the water. About two-thirds of the melt broke up into particles of mean size of the order of 4.0 mm. The remaining one-third collected still molten in the debris catcher but did not produce any damage to the bottom plate. The maximum downward heat flux was 0.8 MW m². The maximum vessel overpressurisation, i.e. 1.8 MPa, was recorded with 44 kg of melt poured into 255 kg of water and a gas phase volume of 0.875 m³. No steam explosions occurred.     

Results of recent KROTOS FCI tests: alumina versus corium melts

Recent results from KROTOS fuel-coolant interaction experiments are discussed. Five tests with alumina were performed under highly subcooled conditions, all of these tests resulted in spontaneous steam explosions. Additionally, four tests were performed at low subcooling to confirm, on one hand, the suppression of spontaneous steam explosions under such conditions and, on the other hand, that such a system is still triggerable using an external initiator. The other test parameters in these alumina tests included the melt superheat and the initial pressure. All the tests in the investigated superheat range (150–750 K) produced a steam explosion and no evidence of the explosion suppression by the elevated initial pressure (in the limited range of 0.1–0.375 MPa) was observed in the alumina tests. The corium test series include a test with 3 kg of melt under both subcooled and near saturated conditions at ambient pressure. Two additional tests were performed with subcooled water; one test was performed at an elevated pressure of 0.2 MPa with 2.4 kg of melt and another test with 5.1 kg of melt at ambient pressure. None of these tests with corium produced a propagating energetic steam explosion. However, propagating low energy (about twice the energy of the trigger pulse) events were observed. All corium tests produced significantly higher water level swells during the mixing phase than the corresponding alumina tests. Present experimental evidence suggests that the water depletion in the mixing zone suppresses energetic steam explosions with corium melts at ambient pressure and in the present pour geometry. Processes that could produce such a difference in void generation are discussed.     

Ex-vessel releases: Aerosol source terms in reactor accidents

An ex-vessel aerosol and fission-product source term may arise from various events occurring in the containment building of a nuclear reactor. The research into the source terms associated with three of these events is reviewed. These source terms are from steam explosions, pressurized melt ejection, and melt/concrete interaction.The least is known about the steam explosion source term. Analyses indicate that its magnitude is likely lower than that assumed in the Reactor Safety Study (WASH-1400), but no conclusive experimental data are as yet available.The aerosol and fission-product source term from pressurized ejection of melt is an issue only recently addressed. Experimental evidence has allowed estimates to be made of the magnitude of this source term.The source term from melt/concrete interaction has been long recognized and has the largest data base. Experimental programs have addressed this source term for several years. A mechanistic model of material release has been developed and is discussed.     

Fukushima Daiichi fuel debris retrieval: results of aerosol characterization during laser cutting of non-radioactive corium simulants

ABSTRACT

The production and dispersion of contaminated aerosols during the laser cutting of corium can potentially provide useful insights into the dispersion of contamination during the evacuation of damaged reactors during decommissioning. Quantitative assessments of contamination dispersion are fundamental to the development of a safety case for the decommissioning of the Fukushima Daiichi plant. This collaborative work between IRSN, ONET Technologies and CEA, managed by the Mitsubishi Research Institute on behalf of the Japanese Ministry of Economy, Trade and Industry, presents the characterization of aerosols generated during laser cutting of corium simulants both in air and under water.

The objective is to obtain quantitative data for risk assessment related to the contamination released and disseminated when implementing this technique, over the next few years, in the process of decommissioning the damaged reactors. This paper presents a part of the results stemming from this project, focused on the characterization of aerosols produced during laser cutting of two representative corium simulants in air and underwater conditions. The experimental configuration also enabled investigation of the production of other material residues such as particle dross and water purity on the particulate composition of the aerosols. Ultimately, the radioisotope concentration distribution in the aerosols are transposed to radioactivity in order to assess the risk to radiation workers during decommissioning.     

Loss-of-primary-flow-without-scram tests: Pretest predictions and preliminary results

A series of tests in the Experimental Breeder Reactor No. 2 (EBR-II) has been concluded that investigated the effects of a complete loss of primary flow without scram. The development and preliminary study of these events is first discussed, including the test limits and controlling parameters. The results of two of the tests, SHRT 39 and 45, are examined in detail, although a compact summary of all the tests is included. The success in meeting the objectives of the test program served to verify that natural processes will shut down the reactor and maintain adequate cooling without control rod or operator intervention. The good comparison between predicted and measured results confirms that such events can be analyzed without elaborate codes if the basic processes are understood. Furthermore, recent studies suggest that the EBR-II results are characteristic of new innovative LMR designs being pursued in the U.S. that incorporate metallic driver fuel.     

Influence of corium oxidation on fission product release from molten pool

Qualitative and quantitative determination of the release of low-volatile fission products and core materials from molten oxidic corium was investigated in the EVAN project under the auspices of ISTC. The experiments carried out in a cold crucible with induction heating and RASPLAV test facility are described. The results are discussed in terms of reactor application; in particular, pool configuration, melt oxidation kinetics, critical influence of melt surface temperature and oxidation index on the fission product release rate, aerosol particle composition and size distribution. The relevance of measured high release of Sr from the molten pool for the reactor application is highlighted. Comparisons of the experimental data with those from the COLIMA CA-U3 test and the VERCORS tests, as well as with predictions from IVTANTHERMO and GEMINI/NUCLEA codes are made. Recommendations for further investigations are proposed following the major observations and discussions.     

UPTF test results: First three separate effect tests

The first separate effect tests were run in the Upper Plenum Test Facility — a 1:1 representation of a PWR primary system. These tests were focusing the simultaneous steam up- and water down flow phenomena at the upper tie plate, the fluid-fluid mixing in the cold leg and downcomer and the countercurrent flow conditions of steam and saturated water in a PWR-hot leg.     

In-vessel retention of corium at the Loviisa plant

In-vessel retention of corium has been approved to be part of the severe accident management strategy for IVO's Loviisa plant. The approach selected takes advantage of the unique features of the plant such as a low power density, a reactor pressure vessel (RPV) without penetrations at the bottom, and ice-condenser containment which ensures a flooded cavity in all risk significant sequences. The thermal analyses, which are supported by an experimental program, demonstrate that, in Loviisa, the molten corium on the lower head of the RPV is externally coolable with wide margins. This paper summarizes the approach, the thermal analyses and the plant modifications being implemented.     

FCI experiments in the corium/water system

The KROTOS fuel coolant interaction (FCI) tests are aimed at providing benchmark data to examine the effect of fuel/coolant initial conditions and mixing on explosion energetics. Experiments, fundamental in nature, are performed in well-controlled geometries and are complementary to the FARO large scale tests. Recently, a test series was performed using 3 kg of prototypical corium (80 w/o UO₂, 20 w/o ZrO₂) which was poured into a water column of ≤1.25 m in height (95 and 200 mm in diameter) under 0.1 MPa ambient pressure. Four tests were performed in the test section of 95 mm in diameter (ID) with different subcooling levels (10–80 K) and with and without an external trigger. Additionally, one test has been performed with a test section of 200 mm in diameter (ID) and with an external trigger. No spontaneous or triggered energetic FCIs (steam explosions) were observed in these corium tests. This is in sharp contrast with the steam explosions observed in the previously reported alumina (Al₂O₃) test series which had the same initial conditions of ambient pressure and subcooling. The post-test analysis of the corium experiments indicated that strong vaporisation at the melt/water contact led to a partial expulsion of the melt from the test section into the pressure vessel. In order to avoid this and to obtain a good penetration and premixing of the corium melt, an additional test was performed with a larger diameter test section. In all the corium tests an efficient quenching process (0.8–1.0 MW kg-melt⁻¹) with total fuel fragmentation (mass mean diameter 1.4–2.5 mm) was observed. Results from alumina tests under the same initial conditions are also given to highlight the differences in behaviour between corium and alumina melts during the melt/water mixing.     

Considerations on ex-vessel corium behavior: Scenarios, MCCI and coolability

The aim of this editorial article is to provide some structuring of the subject addressed in this part of the issue, which, in the view of the present authors, has not that clearly been reached by the specific contributions. Ex-vessel corium behavior is a wide field. Therefore, classification of the goals of investigations within a safety philosophy is especially required to get not lost in detailed aspects. The overall goals here are the coolability and retention options under ex-vessel conditions. Based on scenario considerations generally addressing risks and options, general principles of cooling and retention devices are outlined. Since the concrete erosion by melt yields a major risk and has to be considered in the concepts and devices, and also since several contributions to this part are dealing with specific aspects of this molten core-concrete interaction (MCCI), a large part of this editorial paper concerns the status of knowledge and modeling and the lines of research in this area. The status and the perspectives of codes is especially addressed by own contributions of one of the authors with the GRS code WEX and with MEDICIS, both codes included in the European integral code ASTEC. Finally, the coolability questions are discussed with respect to the different concepts in general and those addressed specifically in the present contributions. In some considerations, gas production by erosion plays a role to produce porous or particulate debris and thus to enhance coolability. Water injection from bottom is a more direct and probably more effective measure to reach this, specifically designed in the COMET core catcher concept. Specific contributions in this part deal with this concept, which is most closely related to the general subject of the present issue.     

Ex-vessel fuel-coolant interaction experiment in the DISCO facility in the LACOMECO project

In the frame of the LACOMECO (large scale experiments on core degradation, melt retention and containment behavior) project of the 7th European Framework Program, a test in the DISCO (dispersion of corium) facility was performed in order to analyze the phenomena which occur during an ex-vessel fuel–coolant interaction (FCI). The test is focused on the premixing phase of the FCI with no trigger used for explosion phase. The objectives of the test were to provide data concerning the dispersion of water and melt out of the pit, characterization of the debris and pressurization of the reactor compartments for scenarios, where the melt is ejected from the reactor pressure vessel (RPV) under pressure. The experiment was performed for a reactor pit geometry close to a French 900 MWe reactor configuration at a scale of 1:10. The corium melt was simulated by a melt of iron–alumina with a temperature of 2400 K. A containment pressure increase of 0.04 MPa was measured, the total pressure reached about 0.24 MPa. No spontaneous steam explosion was observed. About 16% of the initial melt (11.62 kg) remained in the RPV vessel, 60% remained in the cavity mainly as a compact crust. The fraction of the melt transported out of the pit was about 24%.     

Analysis of corium recovery concepts by the EUROCORE group

The objective of the EUROCORE (European Group for Analysis of Corium Recovery Concepts) Concerted Action is to obtain a clear view of the state-of-the-art for melt stabilisation as considered in accident management schemes and to better identify Research and Development (R&D) needs. Five different melt stabilisation concepts have been discussed: in-vessel retention with external cooling, core-concrete interaction with top cooling, ex-vessel spreading with top flooding, water injection by bottom flooding, and crucible concept with sacrificial material. For each concept, main unresolved problems are discussed in this paper and recommended R&D actions are outlined. The project started on 1 March 2000 and ended on 28 February 2002.     

Experimental evaluation of the water ingression mechanism for corium cooling

Experiments were performed to assess the significance of water ingression cooling in the quenching of molten corium. Water ingression is a mechanism by which water penetrates into cracks and pores of solidified corium to enhance cooling that would otherwise be severely limited by the low thermal conductivity of the material. Quench tests were conducted with 2100 °C melts weighing 75 kg composed of UO₂, ZrO₂ and chemical constituents of concrete. The amount of concrete in the melts was varied between 4% and 23%. The melts were quenched with an overlying water layer; three tests were conducted at a system pressure of 1 bar and four tests at 4 bar. The measured cooling rates were found to decrease with increasing concrete content and, contrary to expectations, are essentially independent of system pressure. For the lower concrete content melts, cooling rates exceeded the conduction-limited rate with the difference being attributed to the water ingression mechanism. Measurements of the permeability of the corium “ingots” produced by the quench tests were used to obtain a second, independent set of dryout heat flux data, which exhibits the same trend as the quench test data. The data was used to validate an existing dryout heat flux model based on corium permeability associated with thermally induced cracking. The model uses the thermal and mechanical properties of the corium and coolant, and it reproduces the very particular data trend found for the dryout heat flux as a function of concrete content. The model predicts that water ingression cooling would be most effective for concrete-free corium mixtures such as in-vessel type melts. For such a melt the model predicts a dryout heat flux of 400 kW/m² at a pressure of 1 bar. The results of this study provide an experimental basis for a water ingression model that can be incorporated into computer codes used to assess accident management strategies.     

Determination of leaching factors and effective diffusion coefficients of radionuclides from the results of long tests

radon Scientific-Industrial Amalgamation, Moscow. Translated from Atomnaya Énergiya, Vol. 70, No. 3, pp. 196–197, March, 1991.     

The results of 5% small-break LOCA tests and natural circulation tests at the ROSA-IV LSTF

Five 5% small-break loss-of-coolant accident (SBLOCA) experiments and two natural circulation experiments were conducted at the ROSA-IV Large Scale Test Facility (LSTF). The liquid holdup in the upflow side of steam generator (SG) U-tubes temporarily depressed the core collapsed liquid level below the bottom of core during the loop seal clearing in the cold-leg break SBLOCA tests. This phenomena was affected by the core power and core bypass but was affected little by the actuation of the high pressure injection system. Overall thermal-hydraulic phenomena in a loop seal line break test was similar to that of cold-leg break tests, however, the liquid holdup phenomena played a little role. In a hot-leg break test a temporary but rapid depression of the core liquid level was observed immediately after the initiation of accumulator injection which caused condensation and depressurization in the cold leg. The change of natural circulation flow rate with the decrease of primary system mass inventory was qualitatively the same as observed in Semiscale, LOBI and PKL. The SG effective overall heat transfer coefficient below the secondary-side collapsed liquid level was weakly dependent on the secondary side liquid level and the core power. The measured minimum heat transfer coefficient was 1.7 kW/m²K for the full secondary side mass inventory.     

An evaluation of a direct cooling method for the ex-vessel corium stabilization

An evaluation of the ex-vessel core catcher system of a sample advanced light water reactor was presented. The core catcher was designed to cool down the molten corium through a combined injection of water and gas from the bottom of the molten corium, which could be effective in the reduction of rapid steam generation. By using the MELCOR code, a scenario analysis was performed for a representative severe accident scenario of the ALWR, that is, the 6-in. large break loss of coolant accident without safe injection. The spreading characteristics of ejected corium at vessel breach were asymptotically evaluated on the core catcher horizontal surface. The composition of the molten corium, the decay power level, and the sacrificial concrete ablation depth with time were obtained by a sacrificial concrete ablation analysis. The corium cooling history in the core catcher during the coolant injection was evaluated to calculate the temporal steam generation rate by considering an energy conservation equation. These were used as the major inputs for the temporal calculations of containment pressure which was performed by using the GASFLOW code. Several cases with change of water and gas injection rates were calculated. It was confirmed that the bottom water/gas injection system was an effective corium cooling method in the ex-vessel core catcher to suppress the quick release of steam.     

Corrosion of vessel steel during its interaction with molten corium: Part 1: Experimental

This paper is concerned with corrosion of a cooled vessel steel structure interacting with molten corium in air and neutral (nitrogen) atmospheres during an in-vessel retention scenario. The data on corrosion kinetics at different temperatures on the heated steel surface, heat flux densities and oxygen potential in the system are presented. The post-test physico-chemical and metallographic analyses of melt samples and the corium–specimen ingot have clarified certain mechanisms of steel corrosion taking place during the in-vessel melt interaction.     

Early results of gate valve flow interruption blowdown tests

This paper presents the preliminary results of the USNRC/INEL** high-energy BWR line break flow interruption testing. Two representative nuclear valve assemblies were cycled under design basis vector water cleanup pipe break conditions to provide input for the technical basis for resolving the Nuclear Regulatory Commission's Generic Issue 87. The effects of the blowdown hydraulic loadings on valve operability, especially valve closure stem force, were studied. The blowdown tests showed that, given enough thrust, typical gate valves will close against the high flow resulting from a line break. The tests also showed that proper operator sizing depends on the correct identification of values for the sizing equation. Evidence exists that values used in the past may not be conservative for all valve applications. The tests showed that improper operator lock ring installation following test or maintenance can invalidate in-situ test results and prevent the valve from performing its design function.     

Water boiling on the corium melt surface under VVER severe accident conditions

Experimental results are presented on the interaction of corium melt with water supplied on its surface. The tests were conducted in the ‘Rasplav-2’ experimental facility. Corium melt was generated by induction melting in the cold crucible. The following data were obtained: heat transfer at boiling water-melt surface interaction, gas and aerosol release, post-interaction solidified corium structure. The corium melt charge had the following composition, mass%: 60% UO_2+x–16% ZrO₂–15% Fe₂O₃–6% Cr₂O₃–3% Ni₂O₃. The melt surface temperature ranged within 1920–1970 K.