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.     

Experimental investigation on self-leveling behavior in debris beds

During a hypothetical core-disruptive accident in a sodium-cooled fast breeder reactor, degraded core materials can form debris beds on the core-support structure and/or in the lower inlet plenum of the reactor vessel from rapid quenching and fragmentation of core material pool. Coolant boiling may lead ultimately to leveling of the debris bed that is crucial to the relocation of molten core and heat-removal capability of the debris bed. In the present study, we elected to use depressurization boiling to simulate an axially increasing void distribution in the debris bed. Bottom-heating boiling was also chosen to confirm that characteristics of the self-leveling process do not depend on the boiling mode. Particle size (between 0.5 and 6 mm), shape (spherical and non-spherical), bed volume (between 5 and 8 l) and density (namely of alumina, zirconia, lead and stainless steel) along with boiling intensity and total volume were taken as experimental parameters to obtain the general characteristics of the self-leveling process. A series of experiments with simulant materials were conducted and analyzed in detail. The good concordance of the transient processes obtained from the different boiling methods sufficiently demonstrates that the present results obtained using the depressurization boiling method exhibit these general self-leveling characteristics. Detailed comparisons of deduced time variations of the inclination angle provides qualitative tendencies based on the experimental parameters considered influential to self-leveling behavior. The rationale behind the definition introduced for equivalent power density is also presented.     

The effect of lateral flooding on the coolability of irregular core debris beds

The coolability of ex-vessel core debris is an important issue in the severe accident management strategy of, e.g. the Nordic boiling water reactors. In a core melt accident, the molten core material is expected to discharge into the containment and form a porous debris bed on the pedestal floor of a flooded lower drywell. The debris bed generates decay heat which must be removed by boiling in order to stabilize the debris bed and to prevent local dryout and possible re-melting of the material. The STYX test facility which consists of a cylindrical bed of irregular alumina particles has been used to investigate the effect of lateral coolant inflow on the dryout heat flux of the particle bed. The lateral flow was achieved by downcomers attached on the sides of the test rig. The downcomers provide coolant into the lower region of the bed by natural circulation. Both homogenous and stratified bed configurations have been examined. It was observed that the dryout heat flux is increased by 22-25% for the homogenous test bed compared to the case with no lateral flooding. For the stratified configuration with a fine particle layer on top of the bed, no significant increase in the dryout heat flux was observed. The experiments have been analyzed by using the MEWA-2D code. Models which include explicit consideration of gas-liquid friction were used in the calculations in order to realistically capture the lateral flow configuration.     

Analytical solution for the transient thermal response of volumetrically heated fixed porous nuclear debris beds

An analytical solution is presented for the problem of the transient distribution of fluid and solid temperatures in a volumetrically heated fixed porous nuclear debris bed through which a single-phase fluid, either subcooled liquid or superheated vapor, is flowing. The one-dimensional, time-dependent Modified Dispersion-Concentric Model (D-C Model) is used in the analysis of this problem. The method of solution is based on transformations that reduce the equations and then the application of Green's function in a straightforward manner. The mathematical solution characteristics for the transient fluid and solid temperature distributions are analyzed for subcooled liquid flow and for superheated vapor flow. Also, the transient fluid and solid phase temperature distributions for infinite time are compared with steady-state experimental data.     

Transient cooling rate of partially dry fuel elements

Scientific Production Association of the I. I. Polzunov Central Boiler and Turbine Design Institute. Science and Technology Center for Nuclear Radiation Safety of the State Atomic Inspection Agency of the Russian Federation St. Petersburg State Technical University. Translated from Atomnaya Énergiya, Vol. 75, No. 4, pp. 276-281, October, 1993.     

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).     

Transient natural convection cooling of a vertical circular cylinder

A numerical solution is presented for transient natural convection cooling of a finite vertical circular cylinder standing on a semi-infinite horizontal base. The method used is the quasi-finite element approach introduced by Mansour et al. in the determination of the steady state temperature field in metal cutting. The method is able to solve heat transfer problems in a more economical way (memory and computing time requirements) than the classical finite element method and the finite difference techniques.     

Concept of a core cooling system and experiments performed

One of the problems which must be solved in severe accidents is the melt-concrete interaction which does occur when the core debris penetrates the lower pressure vessel head and contacts the basement. To prevent these accident consequences, a core catcher concept is proposed to be integrated into a new pressurized-water reactor design. The core catcher achieves coolability by spreading and fragmentation of the ex-vessel core melt based on the process of water inlet from the bottom.In order to justify the dominant process during flooding of the melt from the bottom, prototypic experiments with thermite melts in laboratory scale have been carried out. In these experiments flooding and early coolability of the melt is demonstrated. To obtain more detailed information on the important process of water penetration into the melt, a simulant experiment has been conducted using a transparent plastic melt with the typical viscosity behaviour of an oxidic corium melt and a temperature allowing evaporation of water. In every experiment the melt is flooded, and complete freezing in the form of a porous layer occurs within a few minutes only.     

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.     

Transient analytical temperature distributions in cylindrical packed beds volumetrically heated by radiogenic decay energy

Analytical solutions are presented for the problem of the transient distribution of fluid and solid phase temperatures in a packed, porous, cylindrical particle bed with constant thermophysical properties. The packed particle bed is volumetrically heated by radiogenic decay energy from fission products. Flowing through the particle bed by forced convection is a single-phase fluid, either subcooled liquid or superheated vapor. The dynamic response of the packed bed is for low Reynolds numbers. In this case the transient will develop through the packed bed slowly enough for interphase heat transfer to keep the fluid and solid phase temperatures from having large differences. The two-dimensional, time-dependent Modified Dispersion-Concentric Model (D-C model) is used in the analysis of this problem. The D-C model energy equations are solved using Green's function. The mathematical solution characteristics for the transient fluid and solid phase temperature distributions are presented for three different volumetric heat generation terms: two-dimensional, time-dependent; simplified two-dimensional, time-dependent; and two-dimensional, time-independent. Using the two time-dependent volumetric heat generation terms, a comparison is presented for the transient fluid and solid phase temperatures and the radioactive decay heat power coming from the fission products in the particles.     

Formation, characterisation and cooling of debris: Scenario discussion with emphasis on TMI-2

Regarding safety improvements for existing nuclear power plants, the TMI-2 accident is interesting because of the present commercial dominance of light water reactors (LWR). This accident demonstrated that the nuclear safety philosophy evolved over the years has to cover accident sequences involving massive core melt progression in order to develop reliable mitigation strategies for both, existing and advanced reactors. Although the TMI-2 core was reflooded, the results also appear applicable to the general melt progression phenomenology of most unrecovered (unreflooded) blocked core accident scenarios. Nevertheless, a large range in the initial conditions of core melt progression provides significant uncertainties in assessing the integrity of the lower head, the containment in severe reactor accidents, and the consequences of recovery actions in accident management, as well as core reflooding in particular. The probability of success of reflooding as an accident management strategy – in-vessel reflooding to terminate the accident and ex-vessel flooding to prevent reactor vessel melt-through – has to be assessed and discussed in detail.     

Commissioning of the first KSTAR neutral beam injection system and beam experiments

The neutral beam injection (NBI) system was designed to provide plasma heating and current drive for high performance and long pulse operation of the Korean Superconducting Tokamak Advanced Research (KSTAR) device using two co-current beam injection systems. Each neutral beam injection system was designed to inject three beams using three ion sources and each ion source has been designed to deliver more than 2.0 MW of deuterium neutral beam power for the 100-keV beam energy. Consequently, the final goal of the KSTAR NBI system aims to inject more than 12 MW of deuterium beam power with the two NBI for the long pulse operation of the KSTAR. As an initial step toward the long pulse (～300 s) KSTAR NBI system development, the first neutral beam injection system equipped with one ion source was constructed for the KSTAR 2010 campaign and successfully commissioned. During the KSTAR 2010 campaign, a MW-deuterium neutral beam was successfully injected to the KSTAR plasma with maximum beam energy of 90 keV and the L-H transition was observed with neutral beam heating. In recent 2011 campaign, the beam power of 1.5 MW is injected with the beam energy of 95 keV. With the beam injection, the ion and electron temperatures increased significantly, and increase of the toroidal rotation speed of the plasma was observed as well. This paper describes the design, construction, commissioning results of the first NBI system leading the successful heating experiments carried in the KSTAR 2010 and 2011 campaign and the trial of 300-s long pulse beam extraction.     

Thermal mixing processes and RPV wall loads for HPI-emergency core cooling experiments in the HDR-pressure vessel

This paper reports about experimental and analytical results of a first series of three thermal mixing experiments at HDR with high-pressure cold water injection (20°C) of a complete 3-D, large scale, thick-walled PV at 11 MPa. This experimental setup leads to a localized, stripe-like asymmetric cooldown of downcomer and vessel wall for the conditions examined. With respect to this asymmetric thermal loading, a first unique data set of wall temperatures and surface strains has been generated as decision basis for code validations and future fracture mechanic oriented HDR experiments. The paper summarizes the experimental results of the Preliminary Test Phase of HDR TEMB (thermal mixing experiments) consisting of the three experiments T32.15, T32.17 and T32.18.Major findings with respect to fluid mixing behavior, the decrease of fluid/wall temperature in the HPI-nozzle/cold leg region, the cold leg nozzle and along the downcomer are reported. Also, transient axial and azimuthal strains and deduced stresses at the inside RPV surface are reported in- and outside the plume. In addition, comparisons between measured data and blind pretest predictions by best-estimate codes COMMIX-1B and SOLA-PTS as well as engineering models REMIX, VOLMIX and JETMIX are presented and discussed. Measured strains and stresses are compared with VISA predictions at different axial positions.     

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.     

The influence of subcooling on dryout inception in sodium-saturated fuel particle beds with top cooling and adiabatic bottom

In the frame of a one-dimensional theory for the heat transfer behaviour of a top-cooled sodium-saturated particle bed it is possible to describe the onset of packed and/or channeled dryout as a function of the subcooling of the overlaying sodium plenum. The onset of bed channeling is modeled with the aid of capillary pressure. A critical value for the subcooling is derived for the change from packed to channeled dryout for decreasing subcooling. This model is applied for the interpretation of the results of the D4-inpile particle bed experiment performed at SANDIA Laboratories.     

Analysis of loss-of-coolant and loss-of-flow accidents in the first wall cooling system of NET/ITER

This paper presents the thermal-hydraulic analysis of potential accidents in the first wall cooling system of the Next European Torus or the International Thermonuclear Experimental Reactor. Three ex-vessel loss-of-coolant accidents, two in-vessel loss-of-coolant accidents, and three loss-of-flow accidents have been analyzed using the thermal-hydraulic system analysis code RELAP5/MOD3. The analyses deal with the transient thermal-hydraulic behavior inside the cooling systems and the temperature development inside the nuclear components during these accidents. The analysis of the different accident scenarios has been performed without operation of emergency cooling systems. The results of the analyses indicate that a loss of forced coolant flow through the first wall rapidly causes dryout in the first wall cooling pipes. Following dryout, melting in the first wall starts within about 130 s in case of ongoing plasma burning. In case of large break LOCAs and ongoing plasma burning, melting in the first wall starts about 90 s after accident initiation.     

Neutron-physical calculations and experiments on a reactor in the first nuclear power plant built in the world

Methods for performing neutron-physical calculations of a reactor and the experiments performed during physical startup and subsequent operation are described. Starting with the simplest one-dimensional calculations during the reactor design process, the operation of the reactor is completed with three-dimensional calculations taking account of the heterogeneous structure of the core.Among the experiments investigations of the effect of water entering the graphite masonry during an accident with the fuel assemblies on the reactivity should be singled out.The method of partial fuel reloading, making it possible to decrease fuel-assembly burnout, has been implemented for the first time for the AM reactor.__________Translated from Atomnaya Énergiya, Vol. 98, No. 1, pp. 13–18, January, 2005.     

Prediction calculations and experiments for the first criticality of the 10 MW High Temperature Gas-cooled Reactor-Test Module

The 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) is a pebble bed experimental reactor built by the Institute of Nuclear Energy Technology (INET), Tsinghua University. This paper introduces the first critical prediction calculations and the experiments for the HTR-10. The German VSOP neutronics code is used for the prediction calculations of the first loading. The characteristics of pebble-bed high temperature gas-cooled reactors are taken into account, including the double heterogeneity of the fuel element, the buckling feedback of the spectrum calculation, the effect of the mixture of fuel elements and graphite balls, and the correction of the diffusion coefficients in the upper cavity based on transport theory. Also considered are the effects of impurities in the fuel elements, in the graphite balls and in the reflector graphite on the reactivity. The number of fuel elements and graphite balls in the initial core is predicted to provide reference for the first criticality experiment. The critical experiment adopts a method of extrapolating to approach criticality. The first criticality was attained on December 1, 2000. The first criticality experiment shows that the predicted critical number of the fuel elements and graphite balls is in close agreement with the experimental results. Their relative error is less than 1.0%, implying the physical predictions and the results of the criticality experiment are much beyond expectations.