Computational Analysis of Downcomer Boiling Phenomena Using a Component Thermal Hydraulic Analysis Code CUPID

A component scale thermal hydraulic analysis code, CUPID (Component Unstructured Program for Interfacial Dynamics), is being developed for the analyses of components of a nuclear reactor, such as reactor vessel, steam generator, containment, etc. It adopts three-dimensional, transient, two-phase and three-field model, and includes various physical models and correlations of the interfacial mass, momentum and energy transfer for the closure relations of the two-fluid model. In the present paper, the two-phase models were assessed against the DOBO (DOwncomer BOiling) experiment, which was constructed to simulate the downcomer boiling phenomenon. It may happen in the downcomer of a nuclear reactor vessel during the reflood phase of a postulated loss of coolant accident. The stored energy release from the reactor vessel to the liquid inside the downcomer causes the boiling on the wall, and it can reduce the hydraulic head of the accumulated water, which is the driving force of water reflooding to the core. This phenomenon has been considered as a crucial safety issue of an advanced power reactor because it is concerned with the core cooling capability of the safety injection system. In this paper, the physical models and correlations that were incorporated into the CUPID code were introduced and the validation results against the experiment were reported. The benchmark calculation results concluded that the CUPID code can appropriately predict the boiling phenomena under a low pressure and low flow rate condition with modification of the bubble size correlation.

Characteristic Optimization of the Double-Tube OTSG

The OTSG (Once-Through Steam Generator) is usually used in the integral nuclear power equipment which requires smaller size and better effect of heat transfer. The OTSG with double-side heat transfer component is presented in this paper. The heat transfer component is composed of straight tube outside and helix tube inside. In the both sides of the helix tube, the flow is spirally, therefore, the heat transfer is enhanced. The smaller the pitch, the stronger the spirally flow, the effect of heat transfer is better, but the flow resistance is raised. Especially the increased flow resistance in the secondary side brings a great influence to the pump. The heat transfer region of the secondary fluid are divided into three regions: sub-cooled region, boiling region, and superheated region, the effects of heat transfer induced by the spirally flow vary in different regions. Thus, there is an optimization problem which is to find an optimization pitch of the inner helix tube with the best effect of heat transfer and the minimum flow resistance. Based on analyzing the effects of the pitch on heat transfer enhancement and flow resistance, the pitch is optimized by the constrained nonlinear optimization method.

An Experimental Study of Bubble Nucleation Frequency in Subcooled Flow Boiling

In this paper, a simultaneous visualization and measurement study have been carried out to investigate bubble nucleation frequency of water in micro-channel at various heat fluxes and mass fluxes. A single micro-channel with an identical rectangular cross-section having a hydraulic of 137 μm and a heating length of 30 mm was used in this experiment. It is shown that the frequency of bubble nucleation increased drastically with the increase of heat flux and was also strongly dependent on the mass flux. A dimensionless frequency of bubble nucleation was correlated in terms of the Boiling number. The predictions of bubble nucleation frequency in the microchannel are found in good agreement with experimental data with a MAE of 10.4%.

Thermal Hydraulic Analysis of Severe Accident in PFBR

During a core disruptive accident (CDA), the amount of primary sodium that can be released to Reactor Containment Building (RCB) in Prototype Fast Breeder Reactor (PFBR) is estimated to be 350 kg/s, by a transient fluid dynamic calculation. The pressure and temperature evolutions inside RCB, due to consequent sodium fire have been estimated by a constant burning rate model, accounting for heat absorption by RCB wall, assuming RCB isolation based on area gamma monitors. The maximum pressure developed is 7000 Pa. In case RCB isolation is delayed, then the final pressure inside RCB reduces below atmospheric pressure due to cooling of RCB air. The negative pressure that can be developed is estimated by dynamic thermal hydraulic modeling of RCB air / wall to be −3500 Pa. These investigations were useful to arrive at the RCB design pressure. Following CDA, RCB is isolated for 40 days. During this period, the heat added to RCB is dissipated to atmosphere only by natural convection. Considering all the possible routes of heat addition to RCB, evolution of RCB wall temperature has been predicted using HEATING5 code. It is established that the maximum temperature in RCB wall is less than the permissible value.

Advanced Dryout Correlation for BWR Fuel

Each BWR fuel design requires a method to predict its dryout performance in order to be licensed. Presently, the assessment of dry-out risk is based on empirical correlations, which sometimes results in inaccurate or non-physical predictions in certain portions of operational space. This poses a number of limitations as plant operators seek to extract additional value from the fuel through more aggressive operation strategies. A new form of BWR dryout correlation is developed. Accuracy of predictions outside of experimental data range is increased by employing a non-linear correlation form and the transformation to axial power profile, which is based on physical considerations. Proper qualitative behavior is assured by the correlation form itself rather than values of regression coefficients.

Validation of CATHARE V2.5 Thermal-Hydraulic Code Against Full-Scale PERSEO Tests for Decay Heat Removal in LWRs

The PERSEO experimental program was performed in the framework of a domestic research program on innovative safety systems with the purpose to increase the reliability of passive decay heat removal systems implementing in-pool heat exchangers. The conceived system was tested at SIET laboratories by modifying the existing PANTHERS IC-PCC facility utilized in the past for testing a full scale module of the GE-SBWR in-pool heat exchanger. Integral tests and stability tests were conducted to verify the operating principles, the steadiness and the effectiveness of the system. Two of the more representative tests have been analyzed with CATHARE V2.5 for code validation purposes. The paper deals with the comparison of code results against experimental data. The capabilities and the limits of the code in simulating such kind of tests are highlighted. An improvement in the modeling of the large water reserve pool is suggested trying to reduce the discrepancies observed between code results and test measurements.

Statistical Uncertainty Analyses of Void Fraction Predictions Using Two Different Sampling Strategies: Latin Hypercube and Simple Random Sampling

In recent years, a more realistic safety analysis of nuclear reactors has been based on best estimate (BE) computer codes. Because their predictions are unavoidably affected by conceptual, aleatory and experimental sources of uncertainty, an uncertainty analysis is needed if useful conclusions are to be obtained from BE codes. In this paper, statistical uncertainty analyses of cross-sectional averaged void fraction calculations using the POLCA-T system code, and based on the BWR Full-Size Fine-Mesh Bundle Test (BFBT) benchmark are presented by means of two different sampling strategies: Latin Hypercube (LHS) and Simple Random Sampling (SRS). LHS has the property of densely stratifying across the range of each input probability distribution, allowing a much better coverage of the input uncertainties than SRS. The aim here is to compare both uncertainty analyses on the BWR assembly void axial profile prediction in steady-state, and on the transient void fraction prediction at a certain axial level coming from a simulated re-circulation pump trip scenario. It is shown that the replicated void fraction mean (either in steady-state or transient conditions) has less variability when using LHS than SRS for the same number of calculations (i.e. same input space sample size) even if the resulting void fraction axial profiles are non-monotonic. It is also shown that the void fraction uncertainty limits achieved with SRS by running 458 calculations (sample size required to cover 95% of 8 uncertain input parameters with a 95% confidence), result in the same uncertainty limits achieved by LHS with only 100 calculations. These are thus clear indications on the advantages of using LHS.

Research on the Influence of Ocean Condition on the Oscillation in Parallel Double-Channel

In this paper, the flow oscillation in the parallel multichannel system has been studied under ocean condition with RELAP5/MC. A double-channel boiling system model is built using the code RELAP5/MC and it is proved to be capable of simulating the instable phenomena. The influences of rolling, heaving and inclination on the flow oscillation have been analyzed. The result shows that the effect of lengthways rolling on the oscillation in parallel channels can be ignored. The influence of athwartships rolling and heaving reveals resonance effect, that is, when the rolling cycle is close to the oscillation cycle and half cycle respectively, the oscillation would happen earlier, when the difference between the cycles is large, the impact of rolling can also be ignored. The influence of inclination on the oscillation in parallel channels is feeble.

Non-Linear Time Series Analysis of the Infuence of Rolling Motion on Natural Circulation System

Two-phase natural circulation flow instability under rolling motion condition was studied experimentally and theoretically. Experimental data were analyzed with nonlinear time series analysis methods. The embedding dimension, correlation dimension and K2 entropy were determined based on phase space reconstruction theory and G-P method. The maximal Lyapunov exponent was calculated according to the methods of small data sets. The nonlinear features of the two phase flow instability under rolling motion were analyzed with the results of geometric invariants coupling with the experimental data. The results indicated that rolling motion strengthened the nonlinear characteristics of two phase flow instability. Some typical nonlinear phenomena such as period-doubling bifurcations and chaotic oscillations were found in different cases.

Thermal-Hydraulic R&Ds for the APR+ Developments in Korea

This paper briefly introduces recent progress in thermal-hydraulic R&Ds, which is mainly being performed at KAERI, for the APR+ (Advanced Power Reactor plus) development. The main R&D items for the APR+ reactor are associated directly with recent efforts to introduce new safety concepts in the APR+ standard design developments, which is currently in progress in the Republic of Korea. The R&D activities reported here mainly cover the thermal-hydraulic and severe accident areas and are being performed in experimental and/or analytical ways. They include: (1) advancement and optimization of safety injection system, (2) incorporation of passive safety features, such as advanced Fluidic Device (FD+) and passive auxiliary feedwater system (PAFS), and (3) incorporation of severe accident mitigation features.