Experimental results on the coolability of a debris bed with multidimensional cooling effects

Within the reactor safety research, the removal of decay heat from a debris bed (formed from corium and residual water) is of great importance. In order to investigate experimentally the long term coolability of debris beds, the scaled test facility “DEBRIS” (Fig. 1) has been built at IKE. A large number of experiments had been carried out to investigate the coolability limits for different bed configurations ( [Rashid et al., 2008], [Groll et al., 2008] and [0055]). Analyses based on one-dimensional configurations underestimate the coolability in realistic multidimensional configurations, where lateral water access and water inflow via bottom regions are favoured. Following the experiments with top- and bottom-flooding flow conditions this paper presents experimental results of boiling and dryout tests at different system pressures based on top- and bottom-flooding via a down comer configuration.A down comer with an internal diameter of 10 mm has been installed at the centre of the debris bed. The debris bed is built up in a cylindrical crucible with an inner diameter of 125 mm. The bed of height 640 mm is composed of polydispersed particles with particle diameters 2, 3 and 6 mm. Since the long term coolability of such particle bed is limited by the availability of coolant inside the bed and not by heat transfer limitations from the particles to the coolant, the bottom inflow of water improves the coolability of the debris bed and an increase of the dryout heat flux can be observed. With increasing system pressure, the coolability limits are enhanced (increased dryout heat flux).     

Modeling of debris cooling with annular gap in the lower RPV and verification based on ALPHA experiments

For severe accident assessment in a light water reactor, heat transfer models in a narrow annular gap between the overheated core debris and the reactor pressure vessel (RPV) are important for evaluating RPV integrity and emergency procedures. Using existing data, the authors developed heat transfer models on the average critical heat flux (CHF) restricted by countercurrent flow limitation (CCFL) and local boiling heat fluxes, and showed that the average CHF depended on the steam–water flow pattern in the narrow gap and that the local heat fluxes were similar to the pool boiling curve. We evaluated the validity of heat transfer models by simple calculations for ALPHA experiments performed at Japan Atomic Energy Research Institute. Calculated results showed that heat fluxes on the crust surface were restricted mainly by thermal resistance of the crust after the crust formation, and emissivity on the crust surface did not have much effect on the heat fluxes. The calculated vessel temperature during the heat-up process and peak vessel temperature agreed well with the measurements, which confirmed the validity of the average CHF correlation. However, the vessel cooling rate was underestimated mainly due to underestimation of the gap size.     

International activities for the analysis of the TMI-2 accident with special consideration of ATHLET calculations

Several OECD countries still have great interest to analyze the TMI-2 accident. Thermal hydraulic best estimate codes and severe accident codes are used to calculate the TMI-2 analysis exercise defined by a CSNI task group. Fourteen organizations in nine OECD countries are participating in the exercise. Four thermal hydraulic best estimate codes and six severe accident codes are used. The Federal Republic of Germany (FRG) is using the thermal hydraulic code ATHLET developed in the GRS to calculate the TMI-2 analysis exercise. Lessons learned are concentrated on the assessment of ATHLET, show advantages of the two phase thermal hydraulic model used, and identify areas for further development. Results from ATHLET calculations are compared with results from other OECD-codes.     

Results of examinations of pressure vessel samples and instrument nozzles from the TMI-2 lower head

15 prism-shaped steel samples were removed from the lower head of the damaged Three Mile Island Unit 2 (TMI-2) nuclear reactor pressure vessel to assess the effects of approximately 19 tonne of molten core debris that had relocated there during the 1979 loss-of-coolant accident. Metallographic examinations of the samples revealed that inside-surface temperatures of 800–1100°C were attained during the accident, in an elliptical ‘hot spot’ with dimensions of about 1 m × 0.8 m. Tensile, creep and Charpy V-notch specimens were cut from the samples to assess the mechanical properties of the lower head material at temperatures up to the peak accident temperature. These properties were used in a margin-to-failure analysis of the lower head. Examinations of instrument nozzles removed from the lower head region assisted in defining the relocation scenario of the molten core debris and showed that the lower head was largely protected from catastrophic failure by a solidified layer below the molten core debris that acted as a partial thermal insulator.     

MELCOR/MACCS simulation of the TMI-2 severe accident and initial recovery phases, off-site fission product release and consequences

MELCOR has become the preferred code of the Swiss nuclear industry and of PSI for severe accident analysis, on account of its integrated systems-level approach and validation against experiments and more detailed codes, while MACCS is commonly used by safety authorities for independent assessment of off-site consequences, in particular health effects. The present work arises out of a programme to assess MELCOR independently using empirical data consistent with the recommendations of the OECD/CSNI validation matrix for core degradation codes. The MELCOR 1.8.5RD calculations are based on a model for phases 1 and 2 provided by the code developers but with a simplified thermal hydraulic noding in certain regions and the inclusion of a simple representation of the fission product release and transport pathways. The model has also been extended to simulate phases 3, 4, and the continuing initial period of core recovery and stabilisation. These calculations are a first attempt to demonstrate a MELCOR–MACCS capability to simulate the whole plant accident sequence beyond phase 4, including the containment response and off-site consequences arising from fission product release from the containment. Emphasis is placed on the overall accident evolution and whole plant response, rather than the detailed behaviour. Results are compared with observed and deduced data for the major accident signatures and rough estimates for exposure based on off-site monitoring. The results provide a good basis for the NPP analysis foreseen.     

Preparation,characterisation and dissolution of a CeO2 analogue for UO2 nuclear fuel

The behaviour of spent nuclear fuel under geological conditions is a major issue underpinning the safety case for final disposal. This work describes the preparation and characterisation of a non-radioactive UO₂ fuel analogue, CeO₂, to be used to investigate nuclear fuel dissolution under realistic repository conditions as part of a developing EU research programme. The densification behaviour of several cerium dioxide powders, derived from cerium oxalate, were investigated to aid the selection of a suitable powder for fabrication of fuel analogues for powder dissolution tests. CeO₂ powders prepared by calcination of cerium oxalate at 800 °C and sintering at 1700 °C gave samples with similar microstructure to UO₂ fuel and SIMFUEL. The suitability of the optimised synthesis route for dissolution was tested in a dissolution experiment conducted at 90 °C in 0.01 M HNO₃.     

Formation cross sections of nuclei with (n,2n) reactions

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Coolability of UO2 debris beds in pressurized water pools: DCC-1 and DCC-2 experiment results

The DCC-1 and DCC-2 experiments were designed to examine post-accident heat removal from reactor fuel debris using prototypic materials over a pressure range of 1 to 170 atmospheres. The purpose of these experiments is to provide dryout data for comparison with current predictive models. The experiments displayed two unexpected features. In DCC-1, the pressure dependence of the dryout flux was less than anticipated. In DCC-2, localized thermally stable dryouts were observed.     

Collection, cooling and delivery of ISOL beams

The collection of an ISOL beam in a Penning trap using implantation on a surface that is subsequently manipulated so as to become part of an end electrode of a Penning trap and reionization of the implanted material by heat has already been very productive for high-precision nuclear-mass measurements, even though it is limited to elements that are surface ionizable and the collection efficiencies are never better than about 0.1%. More recently, in 1990 a Paul trap system for electric collection of ions was installed at the ISOLDE-3 facility and collection was demonstrated for a 60 kV beam of ¹³²Xe ions. The purpose of this test setup was to determine the relationship between phase space volume of a typical trap and the collection efficiency that could be obtained in direct capture. For the modest trap used, collection efficiencies of up to 0.2% were achieved. A beam of negative bromine ions was collected by simply reversing the polarities of all voltages used. From the experience with this system it appears feasible to build a Paul trap which is about three times as large in linear dimensions as the existing one and which could be driven at up to 10 kV peak at 1 MHz using a modest rf amplifier (300 W). With moderate prebunching of the injected beam at 1 MHz, this system should achieve collection efficiencies approaching 100%. Based on these results, preliminary design work is being carried out on the collection system to be installed at the ISOLDE Booster facility. Suggestions for other uses of a Paul trap collection system for ISOL beams are presented.     

Evaluation of the weldability of heavy-section forgings with special emphasis on areas with segregation

This paper describes comparative tests performed on a component, a test coupon and a test specimen (implant) to evaluate the weldability of the material 20 MnMoNi 5 5 particularly in areas exhibiting segregation. The results show that the methods used are suitable for evaluating susceptibility to cold cracking. The test welds on the coupon, however, yielded different results with respect to the minimum preheat temperature required. Possible reasons for this are discussed.     

Proposal of cooling plant,for SPIDER and MITICA experiments

This paper presents a proposal of Cooling Plant for two new Neutral Beam experiments called MITICA and SPIDER to be realized in Padova (Italy). A large amount of Power (up to 70 MW) has to be removed from in-vessel components and auxiliary systems belonging to these two experiments. Different experimental scenarios (pulse duration ranging from few seconds up to 3600 s), requirements for operating temperature, coolant quality and voltage holding are taken into account in this conceptual design proposal.To reduce the radiological risks due to possible presence of activated corrosion products (ACP) in some water cooled components suitable design choices have been analysed.This work was carried out by considering carefully a lot of different aspects like operability, standardization of components, maintenance and repair, optimization of the installed power and the overall costs of the plant.Experiment components with similar requirements are grouped in the same primary circuits where fine temperature regulation, water quality monitoring and calorimetric measurements are the main characteristics. Each primary circuit (PC) is connected to secondary circuits which allow thermal dissipation and, in some cases, also component preheating. Secondary circuits are connected to two large basins the water of which is cooled down by active cooling rejection system such as cooling towers and air coolers. In this way the requirement for impulsive heat dissipation is fulfilled by the water basins allowing to install a less powerful active rejection system and so reducing the total costs.A large effort was done to guarantee good plant integration with the Experiment Main Hall (in which MITICA and SPIDER are located) and other technical supplies, buildings and areas.Other special requirements for stand-alone systems like Draining and Drying System, Pressure Test System and Chemical Control System are also part of this work.     

Leaching behavior of gamma-emitting fission products,calcium, and uranium from simulated MCCI debris in water

ABSTRACT

Gamma-ray radionuclides (Cs-137, Ba-140, I-131, Ru-103, and Zr-95) were produced by neutron irradiation of simulated molten core–concrete interaction (MCCI) debris, which was synthesized by the heat treatment of a mixture of UO₂ with concrete components at a relatively low temperature of 1473 K under reducing and oxidizing conditions. The major uranium solid phases were unreacted UO₂ and calcium uranate. The leaching ratio of the radionuclides in the powdered sample of the simulated MCCI debris was investigated under atmospheric conditions at 298 K in 0.1 mol/dm³ NaClO₄ after filtration of the leachate through a 0.45-µm pore membrane. The uranium molar concentration in the filtrate was affected by the oxidation state in the solids. In the present study, however, the effect of various solid phase conditions on the leaching ratio normalized to that of uranium matrix could not be clarified. It was found that the leaching ratio of various fission products (R_M) was proportional to that of Cs (R_Cs), and this trend did not depend on the oxidation state of uranium, the type of uranium complex (including a colloidal species), or the presence of Ca, Si, cement, or Zr.     

Influence of multiple jet cooling on the heat transfer and thermal stresses in DEMO divertor cooling finger

The divertor concept for DEMO fusion reactor is based on modular design cooled by multiple impinging jets. Such divertor should be able to withstand a surface heat flux of at least 10 MW/m² at an acceptable pumping power. To reduce the thermal loads the plasma-facing side of the divertor is build up of numerous small cooling fingers. Each cooling finger is cooled by an array of jets blowing through the holes on the steel cartridge.The size, number and arrangement of jets on the cartridge influences the heat transfer and pressure drop characteristics of the divertor. Five different cartridge designs are analyzed in the paper. The most critical parameters, such as structure temperature, heat removal ability, pressure drop, cooling efficiency and thermal stress loadings in the cooling finger are predicted for each cartridge design. A combined computational fluid dynamics and structural model was used to perform the necessary numerical analyses. The results have shown that the cartridge design with the best heat transfer and pressure drop characteristics is not also the most favorable choice from the point of view of minimum stress peaks.     

Homogeneity characterisation of sintered (U,Gd)O2 pellets by X-ray diffraction

(U,Gd)O₂ sintered pellets are fabricated by different methods. The homogeneity characterisation of the Gd content seems to be necessary for a production control to qualify the process and the final product obtained. In this paper, we propose an analysis of the X-ray diffraction powder patterns through the Rietveld method, in which the differences between the experimental and the calculated data proposed from a crystalline structure model are evaluated. This result allows us to determine the cell parameters, that can be correlated with the Gd concentration, and the existence of other phases with different Gd contents.     

Comparison of results of soil-structure interaction analyses based on time and frequency domain approaches with emphasis on treatment of damping

Comparison of results of soil-structure interaction analyses of the reactor building of a nuclear power plant using different analytical approaches and solution procedures is presented. The emphasis of the comparison was on the treatment of damping in these different approaches and procedures. An axisymmetric model of the reactor building was employed. The analyses were performed for the aircraft impact loadings. Two different locations were used for these loadings.The following four different sets of analyses were performed.

1. (1) Time-domain analysis using frequency-independent soil springs in conjunction with modal damping cut-off.

2. (2) Frequency-domain analysis using frequency-independent soil springs in conjunction with a complex modulus approach.

3. (3) Frequency-domain analysis using frequency-dependent soil-impedance coefficients in conjunction with a complex modulus approach.

4. (4) Frequency-domain analysis using frequency-dependent soil-impedance coefficients in conjunction with Rayleigh damping.

The frequency-independent soil springs were computed using the standard approach based on rigid base supported on an elastic layered half-space. The frequency-dependent soil impedance coefficients were computed in the form of a soil substructure matrix which included the uncoupled as well as the coupled terms. The computations were based on the use of a “flexible” base mat supported on a layered half-space. Unit dynamic loads, for each frequency, were applied to the layered half-space corresponding to each degree of freedom and the displacements were xomputed corresponding to all degrees of freedom. The compliance matrix so computed was inverted to obtain the impedance matrix for each frequency. The computations were repeated for all frequencies of interest for the aircraft impact loading.Floor response spectra were developed and compared at various floor elevations of the reactor building using the above four different sets of analyses. Conclusions were developed as a result of these comparisons.     

Influence of the specimen configuration and the insert material on fracture toughness characterisation with reconstituted specimens

Specimen reconstitution techniques offer the possibility to obtain fracture toughness measurements when only small amounts of material are available. In order to obtain extra information from charpy specimens, an electron-beam weld reconstitution method is established to obtain compact tension specimens (CT) from the broken halves of the charpy ones. Three types of reconstituted CT specimens with different weld configurations are tested in order to analyse the influence of specimen configuration on fracture toughness evaluation. The validity of the fracture toughness characterisation is analysed by comparing J-integral resistance curves (J–R curves) of specimens with insert and those of reference specimens without insert.     

Nuclear data evaluation of Mn by the EMPIRE code with emphasis on the capture cross-section

Manganese is one of the constituents of alloys for structural components of fission and fusion devices and a well-known neutron dosimeter; however, existing ENDF-B/VII.0 ⁵⁵Mn evaluation was produced by Shibata (1989). This work is an attempt to re-evaluate neutron-induced cross-sections of ⁵⁵Mn using the latest release of the EMPIRE code. Sensitivity studies on the physical and fitting parameters are presented, with special emphasis on the capture and neutron inelastic cross-sections. A calculated nuclear data file in ENDF-6 format of the neutron interaction cross-sections is produced. It extends up to 150 MeV, which is of interest for fusion and accelerator driven system applications. This evaluation is compared with the ENDF/B-VII.0 evaluation and with a selection of experimental microscopic cross-sections. The evaluation is tested using integral data: the OKTAVIAN integral experiment on a manganese shell and an FNG experiment with manganese activation foils. Benchmark results provide needed feedback for the refinement of the physics parameters.     

Prediction, analysis and solution of the flow inversion phenomenon in a typical MTR-reactor with upward core cooling

Research reactors of power greater than 20 MW are usually designed to be cooled with upward coolant flow direction inside the reactor core. This is mainly to prevent flow inversion problems following a pump coast down. However, in some designs and under certain operating conditions, flow inversion phenomenon is predicted. In the present work, the best-estimate Material Testing Reactors Thermal-Hydraulic Analysis program (MTRTHA) is used to simulate a typical MTR reactor behavior with upward cooling. The MTRTHA model consists of five interactively coupled submodels for: (a) coolant, (b) fuel plate, (c) chimney, lower plenum, suction box and cold leg, (d) flap valve and (e) natural circulation flow. The model divides the active core into a specified axial regions and the fuel plate into a specified radial zones, then a nodal calculation is performed for both average and hot channels with a chopped cosine shaped heat generation flux. The reactor simulation under loss of off-site power is performed for two cases namely: two-flap valves open and one flap-valve fails to open. The simulation is performed under a hypothetical case of loss of off-site power. Unfortunately, the flow inversion phenomenon is predicted under certain decay heat and/or pool temperature values below the design values. In most cases, the flow inversion phenomenon is accompanied by boiling which is undesirable phenomenon in this type of reactors as it could affect the fuel-clad integrity. The model results for the flow inversion phenomenon prediction are analyzed and a solution of the problem is suggested.     

Thermophysical properties of MgO, UO2, their eutectic solution and slurry of liquid-solid mixtures, concrete, sodium, stainless steel and debris beds for use in molten pool penetration of MgO substrate

We have prescribed various thermophysical and transport properties to describe various thermal states of the materials of interest such as MgO, UO₂, stainless steel, sodium, and concrete undergo during post accident heat removal (PAHR) in an ex-vessel cavity lined with MgO bricks. A number of properties, especially of molten MgO, had no experimental determination and therefore, by necessity, these were prescribed through available “best” estimates. We have also included the equivalent properties of various “composite” materials such as debris beds with a prescribed composition, solutions, and slurries to describe their participation in various thermophysical phenomena of interest in PAHR.