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.     

FCI experiments in the aluminum oxide/water system

The KROTOS facility at JRC Ispra was recently used to study experimentally melt-coolant premixing and steam explosion phenomena in Al₂O₃/water mixtures with approximately 1.5 kg melt at 2300–2400 °C. In the five tests performed the main parameter was the water subcooling, 10, 40 and 80 X, respectively. In the nearly saturated system, steam explosions could be externally triggered, which resulted in high (supercritical) explosion pressures in the test tube: KROTOS 26, 28. Without triggering, melt penetration in water and melt agglomeration on the bottom plate of the test tube could be observed, which gave rise to strong steaming during the melt cooling-down process: KROTOS 27. In the two tests KROTOS 29, 30, performed with 80 K subcooled water, self-triggered steam explosions occurred with pressures of more than 100 MPa. Post-test analysis of the debris revealed that 85% of the interacting fuel mass fragmented in particles of sizes smaller than 250 μm. An energy conversion ratio of 1.25% was estimated from vessel pressurization data taking into account the energy content in the fuel mass which fragmented to particle diameters of less than 250 μm. The test section was damaged in the test KROTOS 30.     

Characteristics of corium debris bed generated in large-scale fuel-coolant interaction experiments

The formation of corium debris as the result of fuel-coolant interaction (energetic or not) has been studied experimentally in the FARO and KROTOS facilities operated at JRC-Ispra between 1991 and 1999. Experiments were performed with 3–177 kg of UO₂–ZrO₂ and UO₂–ZrO₂–Zr melts, quenched in water at depth between 1 and 2 m, and pressure between 0.1 and 5.0 MPa. The effect of various parameters such as melt composition, system pressure, water depth and subcooling on the quenching processes, debris characteristics and thermal load on bottom head were investigated, thus, giving a large palette of data for realistic reactor situations.Available data related to debris coolability aspects in particular are:

• Geometrical configuration of the collected debris.

• Partition between loose and agglomerated (“cake”) debris.

• Particle size distribution with and without energetic interaction.

These data are synthesised in the present contribution.     

Multi-phase flow aspects of fuel-coolant interactions in reactor safety research

The energetic FCI has long been recognized as an industrial hazard, and more recently has been considered as a possible hazard during a severe accident in a nuclear power plant. The focus of this paper is on the latter application with specific emphasis on in-vessel and ex-vessel situations in which molten fuel may come into contact with the water coolant. Our focus is twofold; first, to explain the rationale for current research into FCIs and second, to discuss the important multi-phase flow issues that arise from such investigations, particularly experimental. After the many years of research on energetic FCIs there still appears to be three areas where the FCI is important to consider: (1) fuel melt quenching in a water pool, (2) adding water to a degraded core, and (3) FCI energetics. Under current agreements these areas are being actively investigated by researchers in the European community as well as the United States. Such experiments with international cooperation are briefly discussed (e.g., FARO, KROTOS and MACE). In such experiments difficulties arise in measuring the appropriate quantities to characterize the FCI phenomena due to the high transient nature of processes involved. We discuss the important multi-phase flow topics in which further basic research may be needed to aid in FCI model validation of FCI related physical processes, and how these subjects relate to the FCI.     

多维熔融物与冷却剂相互作用分析程序COSMETRIC的开发与验证

基于MCBA-SIMPLE算法开发了自主化的多维熔融物与冷却剂相互作用分析程序COSMETRIC。为验证该程序,针对熔融物与冷却剂相互作用实验KROTOS的典型工况进行了模拟计算。通过与KROTOS37实验结果对比,验证了程序模拟高温熔融物与冷却剂混合过程中熔融物液柱碎化、熔融物液滴迁移以及冷却剂蒸发的能力;通过与KROTOS21实验结果对比,验证了程序对蒸汽爆炸压力脉冲峰值及传播速度预测的合理性。在此基础上,对KROTOS21爆炸工况计算的初始空泡份额、熔滴水力学碎化无量纲时间和熔融物碎片初始直径等参数进行了敏感性分析,评估了这些参数对最终压力脉冲的影响。敏感性分析结果发现,较大的初始空泡份额会抑制压力峰值和传播速度;增大熔融物碎片初始直径和水力学碎化无量纲时间,会提升压力波传播速度,降低压力峰值。     

Insight into steam explosions with corium melts in KROTOS

This paper describes two KROTOS tests in which 4 kg of corium melt was injected into a 1-m deep subcooled water pool. These tests were performed in a test vessel that allowed direct visual observations of melt injection and mixing conditions. Visual observations showed that the corium jet penetrated deep into the water while maintaining its shape. The tests did not produce spontaneous explosions. However, the second test with an external trigger produced an explosion with a relatively low efficiency of 0.15%. This is different from the behaviour of alumina melts, which exhibit one order of magnitude higher explosion efficiencies. This paper describes possible mechanisms that could have contributed to the reduced efficiency with corium melt.     

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.     

Transient cooling of shallow debris beds in the first two FARO-LWR experiments

The investigation of the early behaviour of reactor core debris beds in high pressure water and the heating of the bottom plate on which the debris rests are important aspects of the FARO-LWR out-of-pile tests. In the first two tests 18 and 44 kg of molten U0₂-ZrO₂ fell through 1 m of water and led to the fragmentation of of the melt in both tests and to the formation of shallow debris beds of 3 and 7 cm in height. The unfragmented melt ended up as frozen cakes on the bottom plate. The latter got heated rather little in comparison with the heating of the overlying water by the debris bed. From global energy balances of these tests, heat fluxes from the transient debris beds to the water could be evaluated. They turned out to be between 7.8 and 10.5 MW m⁻² during the first 10 s and about 3.9 MW m-⁻² from 10 to 24 s. In particular the first values are considerably higher than dryout heat fluxes measured in experiments with steady state debris beds. This implies that the shallow and initially very hot FARO debris beds must have cooled down rapidly and that they did not dry out significantly. Probably some fluidization occurred in the early life of these beds.     

Development of a multidimensional model for the premixing phase of a fuel-coolant interaction

A transient, three-dimensional, four-field model is under development to deal with the premixing of large amounts of molten corium falling down in the lower plenum of a PWR. This paper presents the main features of the code, written in as mechanistic a manner as possible, in order to provide best-estimate results. Calculations with a two-dimensional, three-field version of MC3D of two FARO tests are presented.     

Pre- and post-test calculations to natural circulation experiments at the integral test facility ISB-VVER using the thermalhydraulic code ATHLET

In 1995 at the integral test facility ISB-VVER in Elektrogorsk near Moscow natural circulation experiments were performed, which were scientifically accompanied by the Forschungszentrum Rossendorf. These experiments were the first of this kind at a test facility, which models VVER-1000 thermalhydraulics. Using the code ATHLET which is being developed by ‘Gesellschaft für Anlagen und Reaktorsicherheit’, pre- and post-test calculations were done to determine the thermalhydraulic events to be expected and to define and tune the boundary conditions of the test. The conditions found for natural circulation instabilities and cold leg loop seal clearing could be confirmed by the tests. Besides the thermalhydraulic standard measuring system, the facility was equipped with needle shaped conductivity probes for measuring the local void fractions.     

Hydrogen distribution tests under severe accident conditions at the large-scale HDR-facility

The E11 tests series covered a combination of important issues dominating the physical phenomena controlling the hydrogen distribution mechanisms, namely: large-scale, multi-compartment, geometry with large-sized dome volume, high gas release rates, multiple steam and gas injection phases at different axial positions and examinations of the efficiencies of mitigative system features including the impact of external sprays at the top of the dome.The test series consisted of a total of eight different experiments covering all aspects of the H₂-distribution and potential mitigation features.A total of 700 sensors were applied during these experiments.The paper outlines experimental and computational results of tests E11.2 and E11.4 which were chosen for two computational PHDR-Benchmark Exercises in the context of blind posttest predictions with broad international participation applying the majority of known computer codes. In addition test E11.2 was selected as an open post-test, OECD International Standard Problem No. 29 (H. Karwat, Distribution of hydrogen within the HDR-containment under severe accident conditions — Task specifications (July 1990)) which is presently in progress.     

Molten corium oxidation model

In order to model oxidation of Zr–O and U–Zr–O melts, post-test appearance of refrozen oxidised melts in the CORA and QUENCH bundle tests performed at the Research Centre Karlsruhe (FZK) are analysed. Furthermore, data from new separate effect tests on ZrO₂ crucible dissolution by molten Zry, specially designed for investigation of long-term behaviour during the melt oxidation stage, are taken into consideration. On this base, a new model on oxidation of molten Zr–O and U–Zr–O mixtures in steam was developed, which allows interpretation of melt oxidation and hydrogen production observed in various bundle tests. The complete formulation of the analytical model, development of the numerical model and its validation against the crucible tests are presented.     

Experimental and calculation results of the integral reflood test QUENCH-14 with M5® cladding tubes

The QUENCH-14 experiment investigated the effect of M5® cladding material on bundle oxidation and core reflood, in comparison with tests QUENCH-06 (ISP-45) that used standard Zircaloy-4 and QUENCH-12 that used VVER E110-claddings. The PWR bundle configuration of QUENCH-14 with a single unheated rod, 20 heated rods, and four corner rods was otherwise identical to QUENCH-06. The test was conducted in principle with the same protocol as QUENCH-06, so that the effects of the change of cladding material could be observed more easily. Pre-test calculations were performed by the Paul Scherrer Institut (Switzerland) using the SCDAPSIM, SCDAP/RELAP5 and MELCOR codes. Follow-on post-test analyses were performed using SCDAP/RELAP5 and MELCOR as part of an ongoing programme of model validation and code assessment. Alternative oxidation correlations were used to examine the possible influence of the M5® cladding material on hydrogen generation, in comparison with Zircaloy-4.     

On the diffusion coefficient calculation in two-step light water reactor core analysis

This paper presents consistent and rigorous accuracy assessments of various methods for calculating the diffusion coefficients in a two-step reactor core analysis of light water reactors (LWRs). The diffusion coefficients are significantly affected by the transport correction and critical spectrum calculations. There are various methods for the transport corrections (inflow/outflow/hybrid corrections) and critical spectrum calculations (B₁/P₁/CASMO-4E methods) so that it is necessary to decide the best combination to achieve a high accuracy in the transport/diffusion two-step analysis. Numerical tests are performed step-by-step to search for the best combination of the methods by comparing each other the transport one-step results, transport/diffusion two-step results, and Monte Carlo results. Numerical test results with a large and a small LWR core show that the combination of inflow transport correction and CASMO-4E critical spectrum calculation is most accurate than the other combinations in terms of eigenvalues and assembly power distributions.     

Mechanical design assessment approaches of actual spent fuel and HLW transport package designs

Abstract

In recent years, BAM Federal Institute for Materials Research and Testing finalised the competent authority assessment of the mechanical and thermal package design in several German approval procedures of new spent fuel and high level waste package designs. The combination of computational methods and experimental investigations in conjunction with materials and cask components testing is the most common approach to mechanical safety assessment. The methodology in the field of safety analysis, including associated assessment criteria and procedures, has evolved rapidly over the last years. The design safety analysis must be based on a clear and comprehensive safety evaluation concept, including defined assessment criteria and constructional safety goals. In general, for new package designs, the implementation of experimental package drop tests in the approval process should be obligatory. Additionally, pre- and post-test calculations as well as components or material testing could be important. The extent to which drop tests are necessary depends on the individual package construction, the materials used and identified safety margins in the design.     

Development of Assessment Methods For Transport and Storage Containers with a Higher Content of Recycled Metal

Abstract

The mechanical behaviour of transport and storage containers made of ductile cast iron melted with a higher content of recycled metal from decommissioning and dismantling of nuclear installations is investigated. Using drop tests with cubic container-like models, the influence of different real targets on the stresses in the cask body and the fracture behaviour is examined. A foundation for a test stand is suggested, which is simple to manufacture and which greatly improves the reproducibility of the test results. Dynamic fracture mechanics analyses of artificial crack-like defects in the test objects were performed by means of finite-element calculations to uncover safety margins. Numerous test results have shown that containers for final disposal can be built from a ductile cast iron with a fracture toughness of more than 50 per cent less than the lower bound value for the current licensed material. The limits of application of the material are also determined by the opportunities for safety assessment.     

Drop tests of TUK-84 for accident conditions during shipment

Tests for mechanical damage of the TUK-84 shipment assembly, used for shipping and dry storage of spent nuclear fuel, with the container drop from a height of 9 m on a rigid base and from a height of 1 m on a steel pin are described. The basic data from the measurements of the impact parameters are presented, the results of tests for defects and post-test checking are presented, and results of acceptance tests for mechanical damage are also presented. __________ Translated from Atomnaya énergiya, Vol. 100, No. 6, pp. 445–448, June, 2006.     

Verification of FEMAXI-7 code by using irradiation test in Halden reactor for He-pressurization effect on FGR of BWR fuels under power transient

Power ramp test for He-pressurization effect on fission gas release (FGR) of about 42GWd/tUO₂ boiling water reactor (BWR) fuel rods was analyzed by the fuel performance code FEMAXI-7. The experimental data were obtained with the two rods, which were base irradiated in the Halden reactor for 12 years (IFA-409), then subjected to the power ramp tests (IFA-535) to investigate the He-pressurization effect. The FEMAXI-7 calculations were performed by inputting rod specifications and experimental conditions in both the baseand test irradiations. The results showed that the calculations reasonably followed the trends of measured cladding elongation and FGR during the power ramp test, depending on the pellet temperature and fission gas atoms diffusion rate. Based on the calculated results, the reason that no apparent He-pressurization effect was observed in the experiment was considered to be caused by insufficient gas communication during strong pellet–clad mechanical interaction (PCMI) and enhanced gap thermal conductance by the solid–solid contact due to gap closure.     

Iodine release behavior from iodine-immobilized cement solid under geological disposal conditions

The iodine release behavior from the iodine-immobilized cement was investigated. From the results of immersion tests using ion-exchanged water (IEW) and calculations, the solubility equilibrium model could describe the iodine release behavior. To assess the performance of cement in an actual environment, it is important to confirm that the solubility equilibrium model is applicable to the geological disposal conditions. From immersion tests using simulated ground water, the release of iodine from the cement occurred in a shorter period of time than in the case of IEW, and reactions with CO₃²⁻ and Cl⁻, which were contained in the simulated ground water, had an influence on the iodine release behavior. As a result of calculations using the solubility equilibrium model, the liquid-solid ratio at which the iodine was completely released was mostly in agreement with the results of immersion tests. The results show that this model can be applied in a wide range of environments.