Lawrence E. Hochreiter

A model for droplet entrainment in heated annular flow

Abstract The ability to accurately predict droplet entrainment in annular two-phase flow is required to effectively calculate the interfacial mass, momentum, and energy transfer, which characterizes nuclear reactor safety, system design, analysis, and performance. Most annular flow entrainment models in the open literature are formulated in terms of dimensionless groups, which do not directly account for interfacial instabilities. However, many researchers agree that there is a clear presence of interfacial instability phenomena having a direct impact on droplet entrainment. The present study proposes a model for droplet entrainment, based on the underlying physics of droplet entrainment from upward co-current annular film flow that is characteristic to light water reactor safety analysis. The model is developed based on a force balance and stability analysis that can be implemented into a transient three-field (continuous liquid, droplet, and vapor) two-phase heat transfer and fluid flow systems analysis computer code.

Simulated AP600 Response to Small-Break Loss-of-Coolant-Accident and Non-Loss-of-Coolant-Accident Events: Analysis of SPES-2 Integral Test Results

As part of the AP600 design certification program, a series of component separate effects tests and two integral systems tests of the nuclear steam supply system were performed. These tests were designed to provide data necessary to validate Westinghouse safety analysis codes for AP600 applications. In addition, the tests have provided the opportunity to investigate the thermal-hydraulic phenomena that are expected to be important in AP600 transients. One series of integral systems tests was undertaken on the SPES-2 facility in Italy, a full-height, full-pressure, 1/395th-power and -volume scale simulation of the AP600 nuclear steam supply system and passive safety features. A series of thirteen design-basis events were simulated at SPES-2 to obtain data for verification and validation of the computer models used for the safety analysis of the AP600. The modeled initiating events included a series ofsmall-break loss-of-coolant accidents (SBLOCAs), single steam generator tube ruptures, and a main steam-line break. The results of the analyses of the SPES-2 test data, performed to investigate the performance of the safety-related systems are reported. These analyses were also designed to demonstrate, through mass and energy inventory calculations, mass and energy balances, and event timing analyses, the applicability of the SPES-2 tests for computer model verification and validation. The key thermal-hydraulic phenomena simulated in the SPES-2 tests and the performance and interactions of the passive safety-related systems that can be investigated through the SPES-2 facility are emphasized. The latter includes the impact of accumulator nitrogen and nonsafety- related system actuation on the passive safety-related system performance. It is concluded that the key thermal-hydraulic phenomena that characterize the SBLOCA and non-LOCA transients have been successfully simulated in the SPES-2 facility, and the test results can be used to validate the AP600 safety analysis computer codes. The SPES-2 tests demonstrate that the AP600 passive safety-related systems successfully combine to provide a continuous removal of core decay heat. The SPES-2 tests also showed no adverse interactions between the passive safety-related system components or with the nonsafety-related systems. In particular, it was found that the effect of non-condensable nitrogen on passive safety-related system performance was negligible.

Application of code scaling applicability and uncertainty methodology to the large break loss of coolant

Abstract In the late 1980s, after completion of an extensive research program, the United States Nuclear Regulatory Commission (USNRC) amended its regulations (10CFR50.46) to allow the use of realistic physical models to analyze the loss of coolant accident (LOCA) in a light water reactors. Prior to this time, the evaluation of this accident was subject to a prescriptive set of rules (Appendix K of the regulations) requiring conservative models and assumptions to be applied simultaneously, leading to very pessimistic estimates of the impact of this accident on the reactor core. The rule change therefore promised to provide significant benefits to owners of power reactors, allowing them to increase output. In response to the rule change, a method called Code Scaling, Applicability and Uncertainty (CSAU) was developed to apply realistic methods, while properly taking into account data uncertainty, uncertainty in physical modeling and plant variability. The method was claimed to be structured, traceable, and practical, but was met with some criticism when first demonstrated. In 1996, the USNRC approved a methodology, based on CSAU, developed by a group led by Westinghouse. The lessons learned in this application of CSAU will be summarized. Some of the issues raised concerning the validity and completeness of the CSAU methodology will also be discussed.

Application of the (W)under-barCOBRA/TRAC best-estimate methodology to the AP600 large-break LOCA analysis

The AP600 is a simplified advanced pressurized water reactor (PWR) design incorporating passive safety systems that perform the same function as the active emergency core cooling systems (ECCSs) on the current reactors. In order to verify the effectiveness of the AP600 design features for mitigation of a postulated large-break loss-of-coolant accident (LOCA), the recently United States Nuclear Regulatory Commission (USNRC)-approved best-estimate LOCA methodology (BELOCA) was applied to perform the AP600 standard safety analysis report large-break LOCA analysis. The applicability of the (W) under bar WCOBRA/TRAC code to model the AP600 unique features was validated against cylindrical core test facility (CCTF) and upper plenum test facility (UPTF) downcomer injection tests, the blowdown and reflood cooling heat transfer uncertainties were re-assessed for the AP600 large-break LOCA. conditions and a conservative minimum film boiling temperature was applied as a bounded parameter for blowdown cooling. The BELOCA methodology was simplified to quantify the code uncertainties due to local and global models, as well as the statistical approximation methods, with the other uncertainties being bounded by limiting assumptions on the initial and boundary conditions. The final 95th percentile peak cladding temperature (PCT95%) was 1186 K, which meets the 10CFR50.46 criteria with a considerable margin. It is therefore concluded that the AP600 design is effective in mitigation of a postulated large-break LOCA

A moving subgrid model for simulation of reflood heat transfer

Abstract In the quench front and froth region the thermal-hydraulic parameters experience a sharp axial variation. The heat transfer regime changes from single-phase liquid, to nucleate boiling, to transition boiling and finally to film boiling in a small axial distance. One of the major limitations of all the current best-estimate codes is that a relatively coarse mesh is used to solve the complex fluid flow and heat transfer problem in proximity of the quench front during reflood. The use of a fine axial mesh for the entire core becomes prohibitive because of the large computational costs involved. Moreover, as the mesh size decreases, the standard numerical methods based on a semi-implicit scheme, tend to become unstable. A subgrid model was developed to resolve the complex thermal-hydraulic problem at the quench front and froth region. This model is a Fine Hydraulic Moving Grid (FHMG) that overlies a coarse Eulerian mesh in the proximity of the quench front and froth region. The fine mesh moves in the core and follows the quench front as it advances in the core while the rods cool and quench. The FHMG software package was developed and implemented into the COBRA-TF computer code. This paper presents the model and discusses preliminary results obtained with the COBRA-TF/FHMG computer code.

Critical Heat Flux During Reflood Transients in Small-Hydraulic-Diameter Geometries

Abstract Critical heat flux (CHF) at a low-flow condition in a small-hydraulic-diameter duct is an important phenomenon for a Materials Test Reactor/Advanced Test Reactor (MTR/ATR) design under a number of accident conditions, including reflood transients. Current CHF models in the literature, such as the Mishima/Nishihara and Oh/Englert CHF models, are based on macroscopic system parameters and not local thermal-hydraulic conditions. These macroscopic parameter–based models cannot be readily used for analysis in transient best-estimate thermal-hydraulic codes. The present work focuses on developing a low-flow-rate CHF correlation, based on local conditions, that is amenable to implementation into a best-estimate transient thermal-hydraulic code for a small-hydraulic-diameter duct. The model development proceeds with a means of correlating CHF data to local conditions parameters and then applying a correction factor to the resulting correlation, subsequently permitting accurate predictions over a range of pressures. An evaluation of the proposed local conditions-based CHF model is conducted by predicting independent sets of CHF experimental results over a range of flow rate, pressure, and subcooling conditions. Conclusions on the viability of the proposed CHF model and suggestions for future efforts in improving the reflood heat transfer CHF models for small-hydraulic-diameter ducts are provided with an evaluation of the model results.

Effect of chemical cleaning on steam generator tube performance

Boiling experiments were performed on new, chemically cleaned, and fouled steam generator tubes to determine the heat transfer performance of each. It was found that the heat transfer performance of the fouled tube was the best, followed by the chemically cleaned tube. The performance of the new tube was the worst. Scanning electron microscope (SEM) photographs of the boiling surfaces were taken to identify differences in surface characteristics. Results revealed the presence of significant amounts of porous deposits on the surface of the fouled tube that provided ample nucleation sites for boiling. Chemical cleaning removed most of the deposits such that the boiling performance of the cleaned surface was degraded. The new tube was very smooth and there were relatively fewer nucleation sites as evidenced in the SEM photographs. Available correlations were used to predict the heat flux for a given wall superheat and were compared with the experimental data.

Reflood heat transfer correlation

An empirical reflood heat transfer correlation has been developed from the FLECHT reflood data for different axial power shapes and variable flooding rate conditions. This correlation consists of a separate quench correlation and a heat transfer coefficient correlation. The reflood correlation predicts both the quench front location and the heat transfer coefficient above the quench front. The reflood heat transfer correlation prediction is in good agreement with both the cosine and the skewed axial power shape FLECHT reflooding data as well as other rod bundle reflood data.

Modeling of droplet entrainment phenomena at a quench front

Abstract The phenomena of droplet entrainment at a quench front is of practical importance as a clear understanding of the underlying mechanisms required to effectively calculate the interfacial mass, momentum, and energy transfer, which characterizes nuclear reactor safety, system design, analysis, and performance. The present study proposes a model for droplet entrainment at a quench front that is based on the best-understood physics related to the Lagrangian quenching phenomenon characteristic to light water reactor (LWR) safety analysis. The model is based on a film boundary layer and stability analysis that attempts to match the characteristic time and length scales of the entrainment phenomenon. This model has been developed such that direct implementation can be made into any two-phase flow simulation code with a three-field (continuous liquid, droplet, and vapor) flow model. Comparisons with integrated transient test data independent of those used for model development have been performed to verify the applicability of the proposed model for the prediction of the entrainment rate of liquid droplets at a quench front under typical reflood conditions envisioned in LWRs.

Optimal Gap Size for Downward Facing Boiling and Steam Venting in a Hemispherical Annular Channel

Abstract A theoretical model has been developed to predict the behavior of a buoyancy-driven upward co-current two-phase flow in an annular channel with uniform gap size that forms between a hemispherical vessel and its surrounding structure. The vessel is fully submerged in water and is heated from within, leading to downward facing boiling on its outer surface. The problem under consideration is relevant to the so-called in-vessel retention (IVR) of core melt, which is a key severe accident management strategy for some advanced pressurized water reactors (APWRs). One available means for IVR is the method of external reactor vessel cooling by flooding of the reactor cavity with water during a severe accident. Design features of most APWRs have the provision for substantial water accumulation in the reactor cavity during numerous postulated accident sequences. With water covering the lower external surfaces of the reactor pressure vessel, significant energy (i.e., decay heat) could be removed from the core melt through the vessel wall by downward facing boiling on the vessel’s outer surface. As boiling of water takes place on the vessel outer surface, the vapor generated on the surface would flow upward through the annular channel under the influence of gravity. The vapor motions would entrain liquid water, thus resulting in a buoyancy-driven upward co-current two-phase flow in the channel. While the flow is induced entirely by the boiling process, the rate of boiling, in turn, can be significantly affected by the resulting two-phase flow. As long as the heat flux from the core melt to the vessel wall does not exceed the critical heat flux limit for downward facing boiling, nucleate boiling is the prevailing regime and the vessel wall can be maintained at relatively low temperatures to prevent failure of the lower head. With this scenario in mind, the problem is formulated by considering the conservation of mass, momentum, and energy in the two-phase mixture, along with the use of available information on two-phase frictional drop and void fraction. The resulting governing system is solved numerically to predict the total mass flow rate that would be induced in the channel by the boiling process. Based on the numerical results, the optimal gap size that would maximize the steam venting rate and the rate of downward facing boiling over a range of wall heat fluxes is determined. The effects of system pressure and liquid level in the reactor cavity on the induced mass flow rate have also been identified.