Analysis of Flow Blockage of a Single Fuel Assembly in the JRR-3 20MW Research Reactor

The case of flow blockage of a single fuel assembly in the JRR-3 20MW open-pool-type research reactor is investigated without taking into account the effect of the power regulation system. The coolant system and multi-channel reactor core are modeled in detail using thermal hydraulic system analysis code RELAP5/MOD3.4. MDNBR (Minimum Departure From Nucleate Boiling Ratio) and the maximum fuel central temperature are investigated to assess the integrity of fuels. The fuel plates in blocked assembly are not damaged until the blockage ratio exceeds 70%. In addition, the mitigative effect of the assumed 18 MW lower power emergency shutdown operation on the accident is also discussed qualitatively. Results indicate that although the assumed lower power emergency shutdown operation cannot avoid the most severe operating condition, it can obviously mitigate the consequences of the accident. The reactor eventually remains in the long-term safe state when natural circulation is established.

Numerical Study of Deteriorated Convection Heat Transfer of Supercritical Fluid Flowing Through Vertical Mini Tube at Relatively Low Reynolds Numbers

Numerical investigations are performed on the convection heat transfer of supercritical pressure fluid flowing through vertical mini tube with inner diameter of 0.27 mm and inlet Reynolds number of 1900 under various heat fluxes conditions using low Reynolds number k-ε turbulence models due to LB (Lam and Bremhorst), LS (Launder and Sharma) and V2F (v2-f). The predictions are compared with the corresponding experimentally measured values. The prediction ability of various low Reynolds number k-ε turbulence models under deteriorated heat transfer conditions induced by combinations of buoyancy and flow acceleration effects are evaluated. Results show that all the three models give fairly good predictions of local wall temperature variations in conditions with relatively high inlet Reynolds number. For cases with relatively low inlet Reynolds number, V2F model is able to capture the general trends of deteriorated heat transfer when the heat flux is relatively low. However, the LS and V2F models exaggerate the flow acceleration effect when the heat flux increases, while the LB model produces qualitative predictions, but further improvements are still needed for quantitative prediction. Based on the detailed flow and heat transfer information generated by simulation, a better understanding of the mechanism of heat transfer deterioration is obtained. Results show that the redistribution of flow field induced by the buoyancy and flow acceleration effects are main factors leading to the heat transfer deterioration.

The Experimental Investigation of Steam Condensation With Non-Condensable Gas Under Natural Convection

The non-condensable gas is used to keep the pressure stable in the steam-gas pressurizer. The processes of heat and mass transfer during steam condensation in the presence of non-condensable gas play an important role and the thermal hydraulic characteristics in the pressurizer is particularly complicated due to the non-condensable gas. The effects of non-condensable gas on the process of heat and mass transfer during steam condensation were experimental investigated. A steam condensation experimental system under high pressure and natural convection was built and nitrogen was chosen in the experiments. The steam and nitrogen were considered in thermal equilibrium and shared the same temperature in the vessel under natural convection. In the experiments, the factors, for instance, pressure, mass fraction of nitrogen, subcooling of wall and the distribution of nitrogen in the steam, had been taken into account. The rate of heat transfer of steam condensation on the vertical wall with nitrogen was obtained and the heat transfer coefficients were also calculated. The characteristics curve of heat and mass transfer during steam condensation with non-condensable gas under high pressure were obtained and an empirical correlation was introduced to calculated to heat transfer coefficient of steam condensation with nitrogen which the calculation results showed great agreement with the experimental data.

Experimental Investigation on Dependence of Decontamination Factor on Aerosol Number Concentration in Pool Scrubbing Under Normal Temperature and Pressure

During a severe accident of a nuclear reactor, radioactive aerosols may be released from degraded nuclear fuels. Pool scrubbing is one of the efficient filters with a high aerosol removal efficiency, in other words a high decontamination factor (DF). Because of its high performance, many pool scrubbing experiments have been performed and several pool scrubbing models have been proposed. In the existing pool scrubbing experiments, an experimental condition of aerosol number concentration was seldom taken into account. It is probably because DF is assumed to be independent of aerosol number concentration, at least, in the concentration where aerosol coagulation is limited. The existing pool scrubbing models also follow this assumption. In order to verify this assumption, we performed a pool scrubbing experiment with different aerosol number concentrations under the same boundary conditions. The test section is a transparent polycarbonate pipe with an inner diameter of 0.2 m. 0.5 μm SiO2 particles were used as aerosols. As a result, DF was increasing as decreasing the aerosol number concentration. In order to ensure a reliability of this result, three validation tests were performed with meticulous care. According to the results of these validation tests, it was indicated that DF dependence on the aerosol concentration was not because of our experimental system error including measurement instruments but a real phenomenon of the pool scrubbing.

A Simulation of Small Break Loss of Coolant Accident in Nuclear Heating Reactor Based on RELAP5

Nuclear heating reactor is integrated designed without main pump and safety injection system. The loss of coolant accidents are mainly in the form of small break LOCA. As no safety injection system is designed for coolant makeup, the water volume in the reactor vessel is critical since it determines whether the reactor will be submerged during the whole scenario. Therefore, the study on coolant loss in this pool system is indispensable. The RELAP5 code has been developed for best-estimate transient simulation of light water reactor coolant systems during postulated accidents. The long term effect in nuclear heating reactor is important. In this paper we investigated the influential factors on SBLOCA scenario and found the long term residual heat removal capacity is decisive in determining the loss of coolant. The residual heat removal capacity should be greater than 2% of reactor thermal power if ensuring the core submerged in the long run.

Research on the Physical Modelling for the Subchannel Analysis of PWR Core

To obtain a better subchannel analysis tool for PWR core, an investigation on the physical modelling was done via CFD analysis in this paper. A detailed discussion on the TH status was carried out to consider the suitability of the physical modelling in the subchannel analysis. The counteracted effect on the balance of the mass, momentum, and energy was found. Its influence on the physical mechanism was analyzed for the TH status in PWR core. Then it was found that the subchannel analysis based on each flow channel is very necessary for the accuracy of the 3D engineering TH analysis for the whole PWR core.

Fourier-Bessel Series Model for the Stefan Problem With Internal Heat Generation in Cylindrical Coordinates

A nonlinear, first-order ordinary differential equation that involves Fourier-Bessel series terms has been derived to model the time-dependent motion of the solid-liquid interface during melting and solidification of a material with constant internal heat generation in cylindrical coordinates. The model is valid for all Stefan numbers. One of the primary applications of this problem is for a nuclear fuel rod during meltdown. The numerical solutions to this differential equation are compared to the solutions of a previously derived model that was based on the quasi-steady approximation, which is valid only for Stefan numbers less than one. The model presented in this paper contains exponentially decaying terms in the form of Fourier-Bessel series for the temperature gradients in both the solid and liquid phases. The agreement between the two models is excellent in the low Stefan number regime. For higher Stefan numbers, where the quasi-steady model is not accurate, the new model differs from the approximate model since it incorporates the time-dependent terms for small times, and as the system approaches steady-state, the curves converge. At higher Stefan numbers, the system approaches steady-state faster than for lower Stefan numbers. During the transient process for both melting and solidification, the temperature profiles become parabolic.

Transient Boiling and Cross Flow in 5×5 Rod Bundle With Rapid Heating

Rapid thermal elevation in nuclear reactor is an important factor for nuclear safety. It is indispensable to develop a three-dimensional nuclear thermal transient analysis code and confirm its validity in order to accurately evaluate the effectiveness of the running nuclear safety measures when heating power of reactor core rapidly rises. However, the heat transfer characteristics such as reactivity feedback characteristics due to moderator density and the technical knowledge explaining the uncertainty are insufficient. In particular, the cross propagation behavior of vapor bubble (void) in cross section of fuel assembly is not grasped. This study evaluates the cross propagation void behavior in a simulated fuel assembly at time of rapid heat generation with a thermal hydraulic test loop including a 5 × 5 rod bundle having the heat generation profile in the flow cross sectional direction. In this paper, the branching heat output condition of transient cross propagation was investigated from visualization of high speed video camera and void fraction measurement by wire mesh sensor with the inlet flow rate 0.3m/s and the inlet coolant temperature 40°C, which are based on the transient safety analysis condition. In addition, we applied the particle imaging velocimetry (PIV) technique to measure liquid-phase velocity profile of the coolant in the transient cross flow and experimentally clarified the relationship with the cross flow.

Numerical Investigation of Heat Transfer Characteristics of Debris Bed After Severe Accident of SFR Based on 1-D Heat Conduction Model

Nuclear reactor severe accidents can lead to the release of a large amount of radioactive material and cause immense disaster to the environment. Since the Fukushima nuclear accident in Japan, the severe accident research has drawn worldwide attention. Based on the one-dimensional heat conduction model, a DEBRIS-HT program for analyzing the heat transfer characteristics of a debris bed after a severe accident of a sodium-cooled fast reactor was developed. The basic idea of the DEBRIS-HT program is to simplify the complex energy transfer process in the debris bed to a simple one-dimensional heat transfer problem by solving the equivalent thermal conductivity in different situations. In this paper, the DEBRIS-HT program code is prepared by using the existing model and compared with the experimental results. The results show that the DEBRIS-HT program can correctly predict the heat transfer process in the fragment bed. In addition, the heat transfer characteristics analysis program is also used to model the core catcher of the China fast reactor. Firstly, the dryout heat flux when all of molten core dropped on the core catcher was calculated, which was compared with the result of Lipinski’s zero dimensional model, and the error between two values is only 11.2%. Then, the temperature distribution was calculated with the heat power of 15MW.

Experimental Study on Two-Phase Natural Circulation Flow Instability in Rod Bundle Channel Under Low Pressure Condition

An experimental investigation was conducted in a natural circulation (NC) loop to study the characteristics of two-phase flow instability under low pressure condition. A 3 × 3 rod bundle channel was used as the test section. The effects of heating power, inlet subcooling degree and system pressure on the two-phase NC flow instability types and stable boundaries were studied. The experimental results show that three typical flow conditions can occur in rod bundle channel under NC condition, which are single-phase NC flow, subcooled boiling NC flow oscillation and density wave oscillations (DWO). The oscillation amplitude and period of DWO can be enlarged by increasing the heat flux. Increasing the inlet subcooling degree can increase the marginal heating power of flow instability in NC system. The occurrence of DWO can be suppressed by increasing the system pressure. The flow instability boundary presented by the subcooling number and phase change number was also obtained in present work.