Bottom Flooding Research Articles

Application of thermal hydraulic and severe accident code SOCRAT/V2 to bottom water reflood experiment PARAMETER-SF4

Abstract The PARAMETER-SF4 test conditions simulated a severe LOCA (loss of coolant accident) NPP (nuclear power plant) sequence in which the overheated core up to 1700–2300 K would be reflooded from the bottom on ECCS (emergency core cooling system) recovery. The test was successfully conducted at the NPO “LUTCH”, Podolsk, Russia, on July 21, 2009, and was the fourth of four experiments in the series PARAMETER-SF. The PARAMETER facility of NPO “LUTCH”, Podolsk, is designed for studies of VVER fuel assembly behavior under conditions simulating design basis, beyond design basis and severe accidents. After the maximum cladding temperature of about 1900 K had been reached in the bundle during PARAMETER-SF4 test, bottom flooding was initiated. The thermal hydraulic and SFD (severe fuel damage) best-estimate computer modeling code SOCRAT/V2 was used for the calculation of PARAMETER-SF4 experiment. The important feature of PARAMETER-SF4 was the air ingress phase during which air was supplied to the working section of experimental installation. It is known that zirconium oxidation in the air proceeds in a different way in comparison to oxidation in the steam. Thermal hydraulics in PARAMETER-SF4 experiment played a very important role and its adequate modeling is important for the thermal analysis. The results obtained by SOCRAT/V2 were compared with experimental data concerning different aspects of the air ingress phase and thermal hydraulic behavior during the reflood. The temperature experimental data were found to be in a good agreement with calculated results. It is indicative of the adequacy of modeling the complicated thermal hydraulic behavior in the PARAMETER-SF4 test.

Validation and application of the WABE code: Investigations of constitutive laws and 2D effects on debris coolability

The WABE-2D model aims at the problem of coolability of degraded core material during a severe accident in a light water reactor (LWR) and describes the transient boil-off and quenching behavior of debris beds. It is being developed in the frame of the KESS code system, which is considered to describe the processes of core heatup, melting, degradation and relocation to the lower plenum as well as the subsequent behavior. The models developed in this frame are being integrated in the German system code ATHLET-CD. An emphasis of the present contribution lies on multidimensional aspects of the cooling behavior. From multidimensional features a significant improvement of overall coolability is expected compared to what is concluded based on classical one-dimensional analyses. Such analyses – also mainly oriented at top cooling conditions – additionally miss the expected importance of interfacial drag which should support coolability in co-current flow situations due to bottom flooding. The latter situation plays a role in the multidimensional behavior expected under realistic conditions. Thus, a further emphasis in the present contribution lies on the constitutive drag laws and their effects in such configurations. Calculations comparing top and bottom flooding and the influence of interfacial friction are presented. An explanation for effects observed in related experiments at Forschungszentrum Karlsruhe is provided based on this influence. The significant increase of dryout heat flux with water inflow from below, driven by a lateral water column, is reproduced and understood. Enhanced cooling due to this and in general by lateral inflow is also demonstrated for reactor scenarios, considering particulate debris in the lower head of the reactor pressure vessel (RPV) of a LWR or in a deep water pool in the reactor cavity of a boiling water reactor (BWR). Cooling by steam flow through local dry zones can establish under lateral water supply to regions below and yield a further extension of coolability. Quenching of hot material is also analyzed. Finally, cases with loss of coolability, dry zone formation and melting are considered, especially in the perspective to analyze melt pool formation in the lower head of the RPV and the history of thermal interaction with the lower head wall. The latter will determine failure possibilities and modes of the RPV.

Ex-vessel core melt stabilization research (ECOSTAR)

The project on ex-vessel core melt stabilization research (ECOSTAR) started in January 2000 to be concluded by end of 2003. The project is performed by 14 partner institutions from five European countries and involves a large number of experiments with low- and high-temperature simulant melts and real corium at different scales. Model development and scaling analysis allows application of the research results to existing and to future LWRs in the area of reactor design and accident mitigation. The project is oriented toward the analysis and mitigation of severe accident sequences that could occur in the ex-vessel phase of a postulated core melt accident. The issues are: (1) the release of melt form the pressure vessel, (2) the transfer and spreading of the melt on the basement, (3) the analysis of the physical–chemical processes that are important for corium behavior especially during concrete erosion with onset of solidification, and (4) stabilization of the melt by cooling through direct water contact. The results achieved so far resolve a number of important issues: the amount of melt that could be transferred at RPV failure from the RPV into the containment can be substantially reduced by lowering the residual pressure in the primary circuit. It is found that melt dispersion also strongly depends on the location of the RPV failure, and that lateral failure results in substantially less melt dispersion. During melt release, the impinging melt jet could erode parts of the upper basement surface. Jet experiments and a derived heat transfer relation allow estimation of its contribution to concrete erosion. Spreading of the corium melt on the available basement surface is an important process, which defines the initial conditions for concrete attack or for the efficiency of cooling in case of water contact, respectively. Validation of the spreading codes based on a large-scale benchmark experiment is underway and will allow determination of the initial conditions, for which a corium melt can be assumed to spread homogeneously over the available surface. Experiments with UO 2-based corium melts highlight the role of phase segregation during onset of melt solidification and during concrete erosion. To cool the spread corium melt, the efficacy of top flooding and bottom flooding is investigated in small-scale and in large-scale experiments, supported by model developments. Project assessment is continuing to apply the results to present and future reactors.

Effects of dispersed flow interfacial area/shear modeling and node size in the core on TRAC-BFl rod surface temperature (rst) histories of separate-effect and system calculations

As part of the task of assessing and qualify the frozen version of PennState University TRAC-BF1/v2001.2 as a BWR LOCA code, in this work, we shall study the extent to which changes in physical models in the code can affect the predicted rod surface temperature histories during separate-effect tests and system calculations. In particular, we shall concentrate our attention to the reflooding phase and after first demonstrating by two examples the necessity of using the code with the PSI updates, we shall implement in the code a dispersed flow interfacial shear model which is consistent with the corresponding interfacial heat transfer model, i.e. it utilizes the interfacial area per unit volume for both closure relations and not only for the interfacial heat transfer. For this, we shall implement in the code the dispersed flow interfacial shear model of RELAP5/MOD3; furthermore, we shall assume that the average droplet diameter is a function of the distance from the bottom quench front. Having done this, we shall analyze one FLECHT bottom flooding test with both code versions and show that the differences between measured and predicted PCTs are for both cases approximately the same. Finally, we shall study the effect of the number of nodes (8, 24 and 48) in the core by analyzing a hypothetical LB-LOCA in a BWR with limited ECC availability both with the frozen and the modified versions of the code and we shall show that when the frozen version of the code is used, the predicted peak clad temperatures (PCTs) are 120 K lower if 48 rather than when 8 nodes are used in the active core.

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 studies on rewetting of hot vertical annular channel

Studies on the rewetting behaviour of hot vertical annular channels are of interest in the context of emergency core cooling in nuclear reactors following LOCA. Experimental studies were carried out to study the rewetting behaviour of a hot vertical annular channel, with hot inner tube, for bottom flooding and top flow rewetting conditions. The length of the inner tube of the test section was 3030 mm for bottom flooding rewetting experiments and 2630 mm for top flow rewetting experiments. The tube was made of stainless steel. Experiments were conducted for water flow rates in the annulus upto 7 lpm (11.7×10 −5 m 3 s −1). The initial surface temperature of the inner tube was varied from 200 to 500°C. The experimental studies show that for a given initial surface temperature of the tube, the rewetting velocity increases with an increase in flow rate of water and it decreases with an increase in the initial surface temperature for a given water flow rate. For a given water flow rate and initial surface temperature, the rewetting velocity is higher in the case of rewetting under bottom flooding conditions as compared to that in the case of rewetting under top flow conditions. These conclusions agree with the conclusions reported in the earlier literature. Using the experimental data of the present work, correlations for bottom flooding and top flow rewetting velocities are developed.

Simulation of mould cooling in injection moulding: R.T. Fox, D. Lee, (Rensselaer Polytechnic Inst, Troy, New York, USA)

The PARAMETER-SF4 test conditions simulated a severe LOCA (loss of coolant accident) NPP (nuclear power plant) sequence in which the overheated core up to 1700–2300 K would be reflooded from the bottom on ECCS (emergency core cooling system) recovery. The test was successfully conducted at the NPO “LUTCH”, Podolsk, Russia, on July 21, 2009, and was the fourth of four experiments in the series PARAMETER-SF.The PARAMETER facility of NPO “LUTCH”, Podolsk, is designed for studies of VVER fuel assembly behavior under conditions simulating design basis, beyond design basis and severe accidents.After the maximum cladding temperature of about 1900 K had been reached in the bundle during PARAMETER-SF4 test, bottom flooding was initiated.The thermal hydraulic and SFD (severe fuel damage) best-estimate computer modeling code SOCRAT/V2 was used for the calculation of PARAMETER-SF4 experiment.The important feature of PARAMETER-SF4 was the air ingress phase during which air was supplied to the working section of experimental installation. It is known that zirconium oxidation in the air proceeds in a different way in comparison to oxidation in the steam.Thermal hydraulics in PARAMETER-SF4 experiment played a very important role and its adequate modeling is important for the thermal analysis. The results obtained by SOCRAT/V2 were compared with experimental data concerning different aspects of the air ingress phase and thermal hydraulic behavior during the reflood. The temperature experimental data were found to be in a good agreement with calculated results. It is indicative of the adequacy of modeling the complicated thermal hydraulic behavior in the PARAMETER-SF4 test.

An experimental investigation of multidimensional quenching of a simulated core debris bed

Quenching by bottom flooding of a particulate bed simulating core debris has been experimentally studied in this work. In the experiments 6.4 mm diameter apriori oxidized stainless steel particles were placed in a basket up to a height of 28 cm from the bottom. These particles were topped with a 27 cm deep layer of 4.8 mm diameter particles. An annular passage 1.45 cm in width existed between the basket containing the particles and the test chamber. Fresh Zircaloy was distributed uniformly in the central portion of the particulate bed. The initial average temperature of the particles varied from 889 to 979°C. During flooding, though the system pressure varied, the pressure difference across the particulate layer was held constant. The experiments were carried out to determine the effect on the quenching process of the magnitude of the pressure difference across the particulate bed, the amount of Zircaloy, and the initial temperature distribution. From experiments it is found that increased pressure drop across the bed increases the multidimensional effects and reduces the quenching time. The presence of Zircaloy leads to increased local temperatures and delays complete quenching of the particulate layer. The initial temperature distribution can affect the shape of the quench front, the hydrogen production rate and the total time to quench.

Inverted annular flow boiling

The convective heat-transfer coefficients of inverted annular flow film boiling in the region immediately downstream of the rewetting front in the bottom flooding mode were calculated and were found to agree very well with the experimental values. Also, the results of the analysis reconfirmed that having more accurate heat-transfer coefficients in a confined space immediately downstream of the rewetting front substantially improves the accuracy of the estimate for rewetting velocities.

Validation of a bottom flooding model

To validate a thermal-hydraulic model, which predicts the wall temperature history and the quench front propagation during the bottom reflooding of tubular channels, a comparison is made between the predicted results and the experimental data performed on a facility with a 2 m long, 0.01 m inner diameter and a 0.001 m thick AISI 304 vertical tubular test section. The initial wall temperature ranges from 600 to 700°C, the flooding mass velocity ranges from 127 to 254 kg/m 2s, the inlet water temperature ranges from 20 to 80°C, the pressure is 0.1 MPa. New void fraction, vapour enthalpy, critical Weber number and heat flux correlations in the inverted annular flow region are presented. The actual quality and the heat transfer in the dispersed flow film boiling are evaluated on the basis of Mastanaiah and Ganic model. The strong effect of the inverted annular film boiling heat transfer on the quench mechanism is shown. The agreement between the predictions and the data is fairly good especially when the quench front propagation rate change is not strong, but also in the other cases the trend is correct and the results are encouraging.

Effect of coolant vapor quality on rewetting phenomena

The effect of the coolant vapor quality on the rewetting process in bottom flooding has been experimentally studied. The flow visualization showed that when the coolant vapor quality was greater than about 1%, a completely different flow pattern was noted from that observed when the coolant had no vapor quality. For a given test condition, an increasing vapor quality was accompanied by an increasing rewetting velocity. The results from the experiment are discussed in terms of the deduced heat transfer coefficient distribution and rewetting temperature.

Parametric effects on the onset of flooding in flat-plate geometries

Flooding of countercurrent stratified steam-water flow in flat-plate geometries is investigated. Both local and global effects of condensation on the onset of flooding are discussed, based on the results of top and bottom flooding tests. The influence of the channel depth and inclination angle on the flooding velocity is also explored. An envelope model, which considers flooding to be the limiting condition for countercurrent flow, is presented and compared with extensive data at various inclinations. Quite good agreement is shown at the nearly horizontal and moderately steep inclinations, but the onset of flooding in nearly vertical flow is expedited by significant droplet entrainment, and thus takes place earlier than the theory predicts.

Influence of Partial Blockage of a BWR Bundle on Heat Transfer, Cladding Temperature, and Quenching During Bottom Flooding or Top Spraying Under Simulated LOCA Conditions

In a test facility with two parallel boiling water reactor fuel assemblies, experiments were carried out with top spray and bottom flooding, simulating loss-of-coolant accident (LOCA) conditions. The flow area restriction, caused by the ballooning of fuel rod cladding within one of the bundles, was provided by blockage plates, which had reductions of 37% in one case and in a second series 70% of the flow area. Test parameters were system pressure (1, 5, and 10 bars), spray (0.68 and 1.02 m/sup 3//h) and flooding rates (1.5,2, and 3.3 cm/s), power input (520 and 614 kW), and the initial cladding temperature (600 and 800/sup 0/C at midplane) of the heaters. The test results showed no significant variations from those without blockage, except in the blocked region. An enhancement of heat transfer was observed in a close region downstream from the blockage in cases such as bottom flooding and top spray tests. The results will serve the purpose of code verification for reactor LOCA analysis.

Analysis of Precursory Cooling in Quenching Phenomena

It has been noted that precursory cooling plays an important role in quenching phenomena. In this work, a new model is presented by assuming that the heat transfer in precursory cooling is mainly due to film boiling, which persists in a finite length. Then the quench velocity and temperature profile are obtained based on the three-region model for one- dimensional axial heat conduction in a heated tube. We applied our model to several existing experimental results and obtained a correlation to predict the effective length of precursory cooling region. It turns out that the correlation takes an identical form for both falling-film rewetting and bottom flooding.

Analysis of precursory cooling in quenching phenomena.

It has been noted that precursory cooling plays an important role in quenching phenomena. In this work, a new model is presented by assuming that the heat transfer in precursory cooling is mainly due to film boiling, which persists in a finite length. Then the quench velocity and temperature profile are obtained based on the three-region model for one- dimensional axial heat conduction in a heated tube. We applied our model to several existing experimental results and obtained a correlation to predict the effective length of precursory cooling region. It turns out that the correlation takes an identical form for both falling-film rewetting and bottom flooding.

A correlation of rewetting temperature

An empirical correlation of “apparent rewetting temperature” using dimensional analysis was attempted for vertical circular channels under bottom flooding conditions. The variables used in the analysis were k, ϱ, c p, σ, D, T C, G, Z, T g, and P. The effect of each variable on “rewetting temperature” is critically discussed. From the comparison correlated values fall within about ±10% of the experimental values.

Correlation of Quench Phenomena for Bottom Flooding during Loss-of-Coolant Accidents

Correlation of Quench Phenomena for Bottom Flooding during Loss-of-Coolant Accidents

Correlation of Quench Phenomena for Bottom Flooding during Loss-of-Coolant Accidents

The three quench modes based on the observation with an outside-heated quartz tube experiment, i.e. (1) liquid column type, (2) dryout type and (3) droplet-rewetting type are discussed quantitatively to provide correlations for a reflood analysis code. For the quench velocity of liquid column type and dryout type, using the data of PWR-FLECHT experiments, a correlation is obtained. This correlation is compared with experimental data and found that it can predict the inverse quench velocity within ±20% accuracy under the following conditions: pressure 1.0–4.1 kg/cm2-a, quench temperature 370~510°C, local subcooling 0~30°C, flow rate 5~25 cm/sec. For the quench temperature of droplet-rewetting type, a correlation is derived and the prediction from it gives the qualitative agreement with few available data.

Methods and correlations for the prediction of quenching rates on hot surfaces

A substantial quantity of experimental data on rewetting, much of which has not been previously reported, is analysed using the results of calculations of the two-dimensional conduction processes taking place in the walls of tubes, which have been used to simulate the cladding of nuclear fuel elements. Correlations giving the quenching heat-transfer coefficient and sputtering temperature are proposed as a result of the analysis. These correlations may be combined with the previously reported conduction analysis to predict rewetting rates under a wide range of conditions. The new data include falling film rewetting rates measured for a range of system pressures (1–15 bars), initial wall temperatures (200–650°C), coolant mass flowrates (3–50 g sec −1) and subcoolings (0–90°C). Measurements have also been made of rewetting rates by bottom flooding of both saturated and subcooled water at atmospheric pressure.

The quenching of irradiated fuel pins

This paper presents the results of falling film rewetting experiments on two types of irradiated fuel pin and on a complementary range of tubes and heaters. Preliminary measurements of rewetting rate in bottom flooding are presented and a brief comparison is made between falling film and bottom flooding rewetting, including the effects of surface finish. It is demonstrated that fuel pin rewetting behaviour is broadly consistent with that for tubes and heaters and that there are no large systematic effects of irradiation. Surface finish is shown to have an important influence on the falling film rewetting rate and could have influenced the fuel pin behaviour. The preliminary findings for bottom flooding are that there are smaller differences between various heaters than for falling film, and surface finish does not significantly change rewetting behaviour. It is shown that subcooling local to the quench front is an important parameter in rewetting, and can be used as a basis for correlating data. This is in agreement with recent ideas.