Study on local hydrogen behaviors in a subcompartment of the NPP containment

Hydrogen control in the case of severe accidents has been required by nuclear regulations to ensure the integrity of containment after TMI accidents. Up to now, many experiments have been conducted to estimate the distribution of hydrogen during accidents in nuclear power plants. In this article, we proposed a computer code named HYCA3D developed to calculate the local hydrogen distribution with three-dimensional time-dependent governing equations, which can simulate the transport of multiple species. Also, local hydrogen behavior has been experimentally investigated in a cylindrical multi-subcompartment mixing chamber, measuring the local concentration in various conditions. Hydrogen is simulated by helium in the experiments. The proposed code was verified with these experimental results, followed by pre-tests with EPRI/HEDL standard problems. The calculation results show good agreement with the experimental data.     

Deformable-volume-based thermal-hydraulic model for a tanks-in-series water-gas system in a vertically stratified flow. Part I: Under adiabatic conditions

Gases such as nitrogen can be used as a working medium for a pressurizing system like a gas pressurizer. A gas-pressurizing system can consist of several tanks in series where one of the tanks contains a water-gas interface. In general, existing system codes such as RELAP5/MOD3, MARS3.1, RETRAN, and TASS are based on a fixed-grid system and specialize in analyzing a water-steam system with a small portion of non-condensable gas. With these fixed-volume-based system codes, some special models may be needed in order to assess the water-gas interface velocity and track the water level. Furthermore, in the equilibrium model of a fixed-grid system code, the gas temperature in a water-gas coexistence region cannot be distinguished from the water temperature. In this study, we propose a deformable-volume-based thermal-hydraulic model consisting of five equations of the mass, momentum, enthalpy, volume, and pressure for a tanks-in-series water-gas system in a vertically stratified flow operating under adiabatic conditions. In particular, the relative velocity of the fluid at the moving boundary of the flow path, which is an important contributor to the momentum-flux term, is expressed in terms of the absolute velocity of the fluid at the neighboring fixed-grid point. With the proposed model, the fluid velocity of the water-gas interface and the water level can be directly obtained from the state variables, and the gas temperature in the water-gas coexistence region can be calculated separately from the water temperature. This characteristic feature suggests that the proposed model is simpler and more accurate than the models used in the existing system codes for a high-pressure water-gas system with a clear phase separation. The proposed model was verified and validated by comparisons with the results of a semi-analytical model and the TASS code under adiabatic conditions, respectively.     

Experimental study on hydrogen behavior at a subcompartment in the containment building

Hydrogen control in the case of severe accidents has been required by nuclear regulations to ensure the integrity of the nuclear containment building after Three Miles Island (TMI) accidents. Up to now, many experiments have been conducted to estimate the distribution of hydrogen during accidents in nuclear power plants. In this study, local hydrogen behavior has been experimentally investigated in a cylindrical multi-subcompartment mixing chamber of the SNU (Seoul National University) hydrogen mixing facility, measuring the local concentration in various conditions and mixture injection locations. Hydrogen is simulated by helium in the experiments. Results showed remarkably different local behavior of helium in experiments of several conditions, and the local analysis for hydrogen concentration rather than the lumped compartment analysis, used widely in most plants, would be important to ensure the equipment survivability or to determine the positions of ignitors.     

Effect of uncertainties in best-estimate thermal hydraulic analysis on core damage frequency for PSA

Generally, thermal hydraulic (TH) analyses have been performed as part of a probabilistic safety assessment (PSA) to construct event trees and to evaluate success criteria. Even though an accident scenario in an event tree for PSA is exceedingly dependent on many uncertainty parameters, TH analysis in PSA, up to now, has been performed without considering the uncertainties for the important parameters. In the present study, TH analysis was carried out using the MARS code to simulate the large break loss of coolant accident (LBLOCA) which is one of the event sequences of level 1 PSA in an optimized power reactor 1000 MWe (OPR1000). First, the phenomena identification and ranking table (PIRT) for LBLOCA were established, and the candidate parameters were set-up. Once the input file for the MARS code was made with consideration of the uncertainties of the candidate parameters, and a parameter assessment was carried out with the MARS code to rank the candidate parameters according to the effect on peak cladding temperature (PCT). For the five highest-ranking parameters resulting from parameter assessment, the probability density function (PDF) of PCT was derived by the response surface method (RSM), and comparative Monte Carlo calculations were also performed to assess the accuracy of the RSM. As a result, it was shown that by considering the uncertainties of the TH analysis, the accident sequence, which had filed in the PSA result in the established PSA results, had a possibility of succeeding, and thus, be able to modify the core damage frequency (CDF).     

Numerical treatment of pebble contact in the flow and heat transfer analysis of a pebble bed reactor core

This paper studies the numerical treatment of the inter-pebble regions in the modeling of a packed bed geometry for the computational fluid dynamics (CFD) analysis of a pebble bed reactor (PBR) core, where the pebbles are physically in contact with each other. In some studies, the inter-pebble regions have been approximated with gaps, in consideration of the problems on mesh quality or economy of the CFD calculation. To examine such a methodology, a sensitivity analysis for the gap size was conducted with two spherical pebbles, where the inter-pebble region was modeled by means of two kinds of inter-pebble gap and two kinds of direct contact. The cases of direct contact showed numerous differences in the results of the flow regime around the pebbles as well as in the wake, compared to the cases of the inter-pebble gap. No large differences were found between the two cases of direct contact. Based on the result of the sensitivity analysis, the two cases of inter-pebble modeling, i.e., the 1-mm gap and area-contact, were applied to the PBR simulation. It was concluded that the flow regimes and their relevant flow-induced local heat transfer were significantly dependent on the modeling of the inter-pebble region.     

Drift-flux model in a sub-channel of rod bundle geometry

In view of the practical interest of the drift-flux model for two-phase flow analysis, the distribution parameter and drift velocity constitutive equations have been obtained for subcooled boiling flow in a sub-channel of rod bundle geometry. The constitutive equation of the distribution parameter for subcooled boiling flow in a sub-channel is obtained from the bubble-layer thickness model. In this derivation an existing constitutive equation for subcooled boiling flow in a round pipe is modified by taking account of the difference in the flow channel geometry between the sub-channel and round pipe. The constitutive equation of the drift velocity is proposed based on an existing correlation and considering the rod wall and sub-channel geometry effects. The prediction accuracy of the newly developed correlations has been checked against experimental data in a 3 × 3 rod bundle sub-channel, obtaining better predicting errors than the existing correlations most used in literature.     

RELAP5/MOD3.3 analysis of coolant depletion tests after safety injection failure during a large-break loss-of-coolant accident

This study is concerned with development of a coupled calculation methodology with which to continually and consistently analyze progression of an accident from the design-basis phase via core uncovery to core melting and relocation. Experiments were performed to investigate the core coolant inventory depletion after safety injection failure during a large-break loss-of-coolant accident in a cold leg utilizing the Seoul National University Facility (SNUF). The SNUF is an integral test loop scaled down to 1/6.4 in length and 1/178 in area from the Advanced Power Reactor 1400 MWe (APR1400). The SNUF tests are simulated with the RELAP5/MOD3.3 code. The test results revealed that the core coolant inventory decreased five times faster during the sweepout in the downcomer than after termination of the sweepout. The sweepout was observed to take place on top of spillover from the downcomer region to expedite the depletion of the core coolant inventory. The calculation results of RELAP5/MOD3.3 deviated from the experimental data in terms of entrainment from the surface of core coolant, condensation and sweepout in the downcomer. Thereby, the core coolant level was computed to decrease faster than the measured from the experiment due to the overestimated spillover by the evaporation of the entrained droplets by the uncovered heaters. Notwithstanding the occasional disparities, the code prediction is in reasonable agreement with the overall behavior of the tests.     

Optimum RCS depressurization strategy for effective severe accident management of station black out accident

Plant specific severe accident management guidelines (SAMG) for operating plants are developed and implemented in Korea as was required by government policy on severe accident. Korea Institute of Nuclear Safety (KINS) has recently reviewed feasibility of the developed SAMG for Ulchin unit 1 plant. Among the strategies referred in SAMG, we have intensively analyzed the reactor coolant system (RCS) depressurization strategy during station black out (SBO) accident scenario, which has a high probability of occurrence according to Ulchin unit 1 Probabilistic Safety Analysis (PSA). In depressurization strategy of the current SAMG, operators need to depressurize rapidly RCS pressure below 2.75 MPa using pressurizer (PZR) pilot operated safety relief valves (POSRVs) for high pressure accident like SBO. The rapid depressurization is effective in allowing the water of safety injection tank (SIT) to be injected into the core, but an excessive discharge of the SIT water is not desirable for an economical use of SIT inventory. Lack of SIT water accelerates the core damage in case the failed electric power do not recover in due to time. The SIT inventory economy means here that we should not waste the water inventory of SIT and use it in the most efficient way to cool the core. In case we do not use it in an economical way, the SIT might be depleted too rapidly, thus leaving an insufficient reservoir for post-depressurization cooling. The quantification of this SIT inventory economy for plant specific situation is of interest to develop an optimum depressurization strategy. In this study we have analyzed an effectiveness of current depressurization strategy for SBO accident with the severe accident analysis code MELCOR 1.8.5 which has been used for regulatory purpose in KINS. The entry time of severe accident management, a grace time gained by the current strategy, and the economy of the discharge mass flow rate for Ulchin plant were evaluated. Moreover, through a simple energy balance equation we could find an optimum strategy for RCS depressurization. The proposed strategy is based on finding an optimum discharge rate for an efficient use of the SIT inventory and it allows us to handle an SBO accident with higher confidence. The proposed strategy is yet a theoretical one, but possibilities of how to incorporate this strategy into engineered safety features are also discussed.     

Condensation heat transfer characteristic in the presence of noncondensable gas on natural convection at high pressure

The steam-gas pressurizer in integrated small reactors experiences very complicated thermal-hydraulic phenomena. Especially, the condensation heat transfer with noncondensable gas under natural convection is an important factor to evaluate the pressurizer behavior. However, few studies have investigated the condensation in the presence of noncondensable gas at high pressure. In this study, therefore, a theoretical model is proposed to estimate the condensation heat transfer at high pressure using the heat and mass transfer analogy. For the high pressure effect, the steam and nitrogen gas tables are used directly to determine the density of the gas mixture and the heat and mass transfer analogy based on mass approach is applied instead of that based on the ideal gas law. A comparison of the results from the proposed model with experimental data obtained from Seoul National University indicates that the condensation heat transfer coefficients increase with increasing system pressure and with decreasing mass fraction of the nitrogen gas. The proposed model is also compared with other conventional correlations proposed in the literature. The proposed model demonstrates the capability to predict the condensation heat transfer coefficients at high pressure better than any other correlation. Finally, the condensate rate is compared to verify the application of the heat and mass transfer analogy at high pressure. The comparison results confirm that the heat and mass transfer analogy can be applied to evaluate the condensation heat and mass transfer at high pressure.