Qualitative response of safety injection system for a large size break in the reactor coolant system cold leg

Safety injection system, accumulator injection system and residual heat removal system of CHASNUPP-1 were simulated using the computer code APROS. We observed the qualitative response of the simulated system during injection and re-circulation phases after LOCA. During rapid depressurization of SRC system due to leakage, these systems started coolant injection in the SRC system as per plant requirement. Different thermal-hydraulic parameters of the respective systems are presented and discussed. Results obtained are in good agreement with the reported document of the reference power plant.     

The effects of break location on PWR small break LOCA: Experimental study at the ROSA-IV LSTF

This paper discusses the effect of break location on the break flow rate and break flow quality transitions during a small-break loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR). Results from five experiments conducted at the ROSA-IV Large Scale Test Facility (LSTF) are compared for this purpose. These experiments simulated a 2-inch break at the lower plenum, upper head, pressurizer top, cold leg, and hot leg, respectively. The controlling phenomena for the break flow quality transitions in cold-leg and hot-leg break experiments are described.     

Experimental investigation on heat transfer characteristics in a safety injection nozzle component in PWR

Thermal fatigue is a potentially significant degradation mechanism in Nuclear Power Plants (NPP). For the fatigue analysis, the thermal load information about components must be determined firstly. In this paper, an experimental study was carried out to obtain local fluid temperatures and local heat transfer coefficients for the safety injection nozzle component in reactor coolant system (RCS). In this mixing tee component a hot jet issues into a cold cross-flow stream from an oblique pipe and the turbulent mixing of two fluids induces local cycling stresses on the adjacent piping wall. Experiments were performed using a special-made heat fluxmeter, which can measure the mixed fluid temperature close to the wall and the heat transfer coefficient between the fluid and the wall. Plexiglass and metallic 1/9-scale mockups were manufactured for flow visualization and heat transfer tests, respectively. All tests were conducted at range of 0–40 for the jet-to-cross-flow velocity ratio. The flow visualization test has obtained general pattern of the flow and identified sensitive zones in the component where the jet and cross-flow interact intensively to cause thermal fatigue more possibly. In the heat transfer test, heat fluxmeters were positioned in the wall at these sensitive zones. The measurement results of temperatures and heat transfer coefficients have been discussed in detail in the paper. These experimental results allow us improving the state of knowledge of the thermal load to be used in the industrial mixing tees in operating for long lifetime assessment and for the design in the basic Nuclear Power Plants.     

Study on post-blowdown of hot leg large break loss-of-coolant-accident in the view of mass and energy release analysis

This study conducted mass and energy release experiment for the hot leg large break loss-of-coolant-accident (LBLOCA) during post-blowdown with an integral test facility, Seoul National University Facility (SNUF), and its RELAP5 simulation. This facility simulated the Young Kwang Nuclear Power Plant Units 3 and 4 (YGN3&4) with volume ratio of 1:1140 based on Ishii's three level scaling. The experiments showed that safety injection (SI) water refilled the cold leg first and later the core. The SI water was vaporized in the core, which resulted in the repressurization of the reactor. This increase in pressure drove the water in the cold leg to flow up to half the height of the U tubes. However, since the water was drained back not long after, the release through the SG side broken section by evaporation was negligible. The SNUF experiment was assessed by RELAP5 simulations. Overall, the analysis of the post-blowdown phase showed that the transient of the primary pressure can be properly simulated by RELAP5 when a sufficient heat source is modeled. Consequently, the releases from reactor side broken section and steam generator side broken section were properly predicted. The pressure rise by steam generation in the core was partially well predicted. The release from the steam generator side broken section was predicted to be small except when there exists a large pressure difference between the primary system and the break boundary.     

Hydrodynamic phenomena in the downcomer during flow rate transients in the primary circuit of a PWR

Mixing phenomena observed when the flow rate in a single loop of the primary circuit is changed can influence the operation of pressurized water reactor (PWR) by inducing local gradients of boron concentration or coolant temperature. Analysis of one-dimensional Laser Doppler Anemometry (LDA) measurements during the start-up and shutdown of pump on a single loop of the ROCOM test facility has been performed. The effect of a step change and a ramped change in the flow rate on the axial and azimuthal velocities was examined. Numerical simulations were also performed for the step change in the flow rate that gave quantitative agreement with the axial velocities. Phenomenological agreement was made on the turbulent kinetic energy; however, observed values were a factor of 2.5 less than the turbulent kinetic energy derived from the measurements.     

Analysis of the OECD/NEA ROSA Project experiment simulating a PWR small break LOCA with high-power natural circulation

OECD/NEA ROSA Project experiment with the large scale test facility (LSTF) in JAEA was conducted simulating a PWR 1% cold leg small break LOCA with an assumption of high-power natural circulation due to failure of scram and total failure of high pressure injection system. The core power curve for the LSTF experiment was obtained by PWR LOCA analysis using JAEA-developed coupled three-dimensional kinetics and thermal-hydraulics code SKETCH-INS/TRAC-PF1 with detailed core model. A post-test analysis was performed against the obtained data by using JAEA-modified RELAP5/MOD3.2.1.2 code to validate the code predictability. The JAEA-modified RELAP5/MOD3.2.1.2 code was used by incorporating a break model that employs maximum bounding flow theory with a discharge coefficient of 0.61 for two-phase break flow. In the experiment, flow in hot legs became supercritical during two-phase natural circulation, causing the hot leg liquid level to be quite low. Liquid accumulation in steam generator U-tube upflow-side took place during reflux condensation mode due to high vapor velocity. The RELAP5 code predicted reasonably well the overall thermal-hydraulic phenomena observed in the experiment. The code, however, overpredicted the break flow rate especially during two-phase flow discharge period probably because of the failure in the correct simulation of the cold leg liquid level due to late decrease in the primary loop flow rate.     

An estimation of an operator's action time by using the MARS code in a small break LOCA without a HPSI for a PWR

To estimate the success criteria of an operator's action time for a probabilistic safety/risk assessment (PSA/PRA) of a nuclear power plant, the information from a safety analysis report (SAR) and/or that by using a simplified simulation code such as the MAAP code has been used in a conventional PSA. However, the information from these is often too conservative to perform a realistic PSA for a risk-informed application. To reduce the undue conservatism, the use of a best-estimate thermal hydraulic code has become an essential issue in the latest PSA and it is now recognized as a suitable tool. In the same context, the ‘ASME PRA standard’ also recommends the use of a best-estimate code to improve the quality of a PSA. In Korea, a platform to use a best-estimate thermal hydraulic code called the MARS code has been developed for the PSA of the Korea standard nuclear power plant (KSNP). This study has proposed an estimation method for an operator's action time by using the MARS platform. The typical example case is a small break loss of coolant accident without the high pressure safety injection system, which is one of the most important accident sequences in the PSA of the KSNP. Under the given accident sequence, the operator has to perform a recovery action known as a fast cooldown operation. This study focuses on two aspects regarding an operator's action; one is how they can operate it under some restrictions; the other is how much time is available to mitigate this accident sequence. To assess these aspects, this study considered: (1) the operator's action model and (2) the starting time of the operation. To show an effect due to an operator's action, three kinds of control models (the best-fitting, the conservative, and the proportional-integral) have been assessed. This study shows that the developed method and the platform are useful tools for this type of problem and they can provide a valuable insight related to an operator's actions.     

Experimental study on the direct contact condensation of steam jet in the passive safety injection tank

The direct contact condensation and subsequent thermal mixing by the injected steam jet onto a quiescent coolant inside a tank were examined experimentally to simulate the phenomena in passive safety injection systems. Specifically, the influence of the steam injection velocity was studied. Even though the total flow rate of injected steam was unchanged, the pressure inside the tank increased quickly at the larger nozzle diameter. Additionally, at a larger nozzle diameter, the thickness of the thermal mixing zone decreased because the amount of direct contact condensation decreased. For the in-depth study on the role of the nozzle size for the thermal mixing, the particle image velocimetry method was used to understand the flow field of water inside the tank. The visualization results demonstrated the formation of a flow field in the coolant due to the expansion and contraction of the steam–air mixture boundary. Furthermore, the thermal mixing zone was found to be closely related to the penetration depth. Finally, a variety of penetration models were examined and compared against the experimental observation. The correlations based on the steam condensation approach under-predicted the penetration depth, whereas the approach that considers the momentum of non-condensable gas gave the reasonable prediction capability.     

CFD studies on the phenomena around counter-current flow limitations of gas/liquid two-phase flow in a model of a PWR hot leg

In order to improve the understanding of counter-current two-phase flow and to validate new physical models, CFD simulations of a 1/3rd scale model of the hot leg of a German Konvoi pressurized water reactor (PWR) with rectangular cross section were performed. Selected counter-current flow limitation (CCFL) experiments conducted at Helmholtz-Zentrum Dresden-Rossendorf (HZDR) were calculated with ANSYS CFX using the multi-fluid Euler–Euler modelling approach. The transient calculations were carried out using a gas/liquid inhomogeneous multiphase flow model coupled with a shear stress transport (SST) turbulence model.In the simulation, the drag law was approached by a newly developed correlation of the drag coefficient (Höhne and Vallée, 2010) in the Algebraic Interfacial Area Density (AIAD) model. The model can distinguish the bubbles, droplets and the free surface using the local liquid phase volume fraction value. A comparison with the high-speed video observations shows a good qualitative agreement. The results indicate also a quantitative agreement between calculations and experimental data for the CCFL characteristics and the water level inside the hot leg channel.     

Fluid mixing and flow distribution in a primary circuit of a nuclear pressurized water reactor—Validation of CFD codes

The EU project FLOMIX-R was aimed at describing the mixing phenomena relevant for both safety analysis, particularly in steam line break and boron dilution scenarios, and mixing phenomena of interest for economical operation and the structural integrity.This report will focus on the computational fluid dynamics (CFD) code validation. Best practice guidelines (BPG) were applied in all CFD work when choosing computational grid, time step, turbulence models, modelling of internal geometry, boundary conditions, numerical schemes and convergence criteria. The strategy of code validation based on the BPG and a matrix of CFD code validation calculations have been elaborated. CFD calculations have been accomplished for selected experiments with two different CFD codes (CFX, FLUENT). The matrix of benchmark cases contains slug mixing tests simulating the start-up of the first main circulation pump which have been performed with three 1:5 scaled facilities: the Rossendorf coolant mixing model ROCOM, the Vattenfall test facility and a metal mock-up of a VVER-1000 type reactor. Before studying mixing in transients, ROCOM test cases with steady-state flow conditions were considered. Considering buoyancy driven mixing, experimental results on mixing of fluids with density differences obtained at ROCOM and the FORTUM PTS test facility were compared with calculations. Methods for a quantitative comparison between the calculated and measured mixing scalar distributions have been elaborated and applied. Based on the “best practice CFD solutions”, conclusions on the applicability of CFD for turbulent mixing problems in PWR were drawn and recommendations on CFD modelling were given. The results of the CFD calculations are mostly in-between the uncertainty bands of the experiments. Although no fully grid-independent numerical solutions could be obtained, it can be concluded about the suitability of applying CFD methods in engineering applications for turbulent mixing in nuclear reactors.     

Analysis of thermal stratification in the primary circuit of a VVER-440 reactor with the CFX code

The thermal stratification can lead an important role in the aging of the NPP piping because of the stresses caused by the temperature differences and the cyclic temperature changes. These stresses can limit the lifetime of the piping, or lead to penetrating cracks. For the stress analyses, the determination of the thermal hydraulic parameters of the stratified flow is necessary, which can be simulated by computational fluid dynamics (CFD) codes. The results of the simulation show the time development and the breaking up of the stratification and the temperature distribution of the stratified flow. The main difficulty of these CFD simulations is the uncertainty of the boundary conditions because of the unknown flow circumstances. In this paper, some results of CFX simulations are presented concerning the pressurizer surge line, and the injection pipe of the HPIS for VVER-440 type reactors.     

Gamma spectrometry in continuous monitoring of the impurity concentration in the coolant of a reactor circuit

Institute of Nuclear Energy, Academy of Sciences of the Belorussian SSR. Translated from Atomnaya Énergiya, Vol. 70, No. 3, pp. 176–177, March, 1991.     

Experimental investigations of the flow patterns in the hot leg of a pressurized water reactor

For the simulation of loss of coolant accidents in nuclear power plants, the flow patterns are predicted by using experimental results from small sized plants which usually have been achieved for fully developed flows. Experimental investigations in a large sized plant have indicated that these flow pattern maps are not fully applicable to the specific geometric properties of nuclear power plants. Therefore, we have conducted experimental investigations for cocurrent two-phase flow in the hot leg. For the experimental investigations a large sized experimental set-up has been constructed, which represents the hot leg of a pressurized water reactor at the scale of 1:1.7. To distinguish between the influence of the size of the plant and the influence of the elbow and the steam generator simulator on the flow pattern, the experimental investigations have been conducted in two steps. First, the flow in the horizontal part of the hot leg has been investigated without connecting the elbow to the plant. The flow regimes have been detected by visual observation. The experimental results are compared to those obtained for smaller pipe diameters and longer pipe lengths. Second, the 50° upwards inclined elbow and the steam generator simulator are added to the horizontal pipe and their influence on the flow patterns is investigated.     

Determination of the235U concentration in the aqueous coolant of the first circuit of a nuclear reactor

Conclusions A technique was developed for determining the²³⁵U concentration in the aqueous coolant of the first circuit of a nuclear reactor: detection limit 3·10^–12 g/cm³. Using the method in the IVV-2M reactor has shown that with this technique, an operational monitoring of the uranium concentration in the coolant and in the fluids washed from the surface of the first circuit, as well as monitoring other qqueous samples, is possible.Lavsan, which is directly irradiated in a liquid sample and electrochemically etched, can be recommended as a detector. The optimal conditions of etching 180-m-thick lavsan (after irradiation with thermal neutrons to a flux of (1–2)·10¹⁶ cm^–2) are: 30% aqueous KOH solution, a temperature of (70±0.2)°C, an electric field strength of 20 kV/cm, a frequency of 4 kHz, and an etching time of 100 min.Translated from Atomnaya Énergiya, Vol. 61, No. 5, pp. 334–338, November, 1986.