Volume 83, Issue 3

Recent nation-wide measures and actions taken in Korea after March 11, 2011 for enhancing nuclear safety and reducing public concerns are discussed by considering the lessons learned from the Fukushima Daiichi accidents as well as the amended National Safety Act, recently occurred historic earthquakes, and advanced light water reactor developments in Korea. In addition, recent R&D activities in the area of enhancing nuclear safety are introduced by focusing on the prevention and mitigation of severe accidents and the reduction of their environmental impact.

Small modular reactors (SMRs) are one interesting option for new builds in almost all countries worldwide continuing to use nuclear energy for commercial power generation. For asserting of legitimate nuclear safety and/or security interests German authorities require in this context own and independent expertise for the safety assessments. The Gesellschaft für Anlagen und Reaktorsicherheit (GRS) mbH, which is the main technical support organization in nuclear safety for the German federal government, has among others performed a study on Safety and International development of Small Modular Reactors (GRS-376) to identify essential issues of SMR safety as well as needs for adaption, improvement and validation of nuclear evidence tools developed and applied by GRS. For the light-water-cooled SMRs, which have (at least according to GRS) the best prospects for realization, selected results of this study will be presented.

Integral test plays essential role to assess the design of the emergency cooling system of nuclear reactors. Different from full height integral test facilities, reduced height integral test facilities have new problems on the pressure drop scaling. This paper mainly focuses on scaling of pressure drop across the core as it is the major pressure drop in primary loop. The analysis of pressure drop across the core has been divided into three terms and each term has been discussed separately based on two conditions: the normal operation condition and natural circulation condition. After that, the total pressure drop ratios under these two conditions have been discussed.

The overall scaled-down separate-effects IRWST&PRHR HX&ADS test facility has been built to simulate the thermal hydraulic behavior of the passive residual heat removal system under accident conditions. The measured 3-D temperature and velocity results show obvious thermal stratification in IRWST. The Richardson number ( Ri ) could be used to predict the occurrence of possible thermal stratification, and the Stratification number ( Str ) to evaluate the thermal stratification extent. Both dimensionless parameters provide criteria for thermal stratification and experimental references for the design and operation of engineering equipment.

Based on the phase change model developed by lattice Boltzmann method (LBM), an improved source term which is responsible for phase change is proposed. The computation of the latent heat is taken into consideration to avoid the error and instability caused by the equation of state (EOS). Periodic bubble growth and departure from a heated surface during the boiling process is investigated based on the newly developed lattice Boltzmann model. Effects of gravity, contact angel and superheat on bubble departure diameter and release period under constant wall temperature conditions are illustrated. The three-phase contact line movements of the vapor bubble and the wall heat flux are also described. The results reflect the dynamic features and the influencing mechanism of bubble in boiling period, and this research lays the foundation of further studies on thermal-hydraulics and safety analysis by LBM.

Plate fuel assembly is widely used in reactor cores due to benefits of the narrow rectangular channels. Compared with rod assembly, the heat transfer area and power density can be increased in narrow rectangular channel, where little researches on post-CHF thermal-hydraulic characteristics have been done. Once the loss of coolant accident happens, post-CHF thermal-hydraulic characteristics are important for accident mitigation. This paper mainly evaluates the corresponding reflood models in RELAP5/MOD3 code. Combined with the post-CHF thermal-hydraulic characteristics in the narrow channel, interfacial friction model and post-CHF heat transfer models are modified. According to the comparison results with Saxena reflood experiments, conclusions can be obtained that the original RELAP5/MOD3 code overestimates quench front velocity and heat transfer in the film boiling region. The results from the code with new models agree better with the experimental data.

Constrained by limited experimental data, development of CHF correlation for PWR fuel assemblies under transient and accidental conditions at low pressure levels (2–10 MPa) is a typical statistical problem with small sample amounts, but simultaneously has requirements of high prediction accuracy. In this study, stepwise regression method was used to develop a new CHF correlation for application in PWR under low pressure conditions. First, several essential thermal-hydraulic parameters which might influence CHF were selected based on consensus characteristics of DNB phenomenon. With stepwise regression, the form and coefficients of the proposed CHF correlation were optimized in a dynamic manner. Compared to currently available CHF correlations, represented by the Westinghouse W-3 correlation, the CHF correlation obtained by stepwise regression has a much simpler form and matches also well with the experimental data.

Post dryout heat transfer and rewetting are important processes determining the level and the time duration of high temperature phase and the integrity of the fuel cladding. In spite of extensive studies in the past decades, reliable prediction methods are still missing due to the complexity of processes involved, which consist mainly of interaction between solid wall, main gas flow and droplets. In the present study, a phenomenological model is proposed considering the three individual heat transfer processes between the three parts. Main new features of the present model compared to the models available in the open literature are the mechanistic modeling of (a) droplet concentration and droplet size, (b) turbulent fluctuation velocity of droplets and its critical value, (c) evaporation rate of droplets arriving the wall. Comparison of the new model with selected experimental data shows at least qualitatively good agreement. The experimental behavior of wall temperature can be well explained. According to the new model the Leidenfrost effect results in the hysteresis behavior of wall temperature.

Safety is an essential topic in the development process of nuclear power plant. Several Generation III and III+ reactor designs contain passive safety system to control accident without external power. An example is the Emergency Condenser (EC) of the KERENA reactor design. The EC removes heat from the Reactor Pressure Vessel in the case of design accidents. The experimental facility COSMEA at Helmhotz Zentrum Dresden Rossendorf (HZDR) was set up to investigate the flow morphology and heat transfer structure of condensation inside a slightly inclined tube. In this paper, the condensation process in COSMEA was simulated with the thermal hydraulic system codes ATHLET. The performance of the ATHLET heat transfer models were identified. The simulation results were compared against the experiments. The heat flux, condensation rate and temperature of cooling water during the condensation was analyzed.

During accident scenarios the effective cooling of spent fuel directly affects the safety of nuclear power plants. Two experiments were performed in a full-height facility to study the thermal-hydraulic behavior in spent fuel pool. In spent fuel pool boiling experiment, the heat transfer characteristics are related to the flow patterns. However, the flow pattern in narrow and long channel is different from the traditional flow pattern. In the semi-dry of heated rod, wall temperature oscillation occurs for a long time. In the spent fuel pray experiment, the liquid film thickness varies randomly with time and space. As the spray flow density increase, the maximum wall temperature decrease gradually with a certain linear characteristic.

The spent fuel pool cooling system in a nuclear power plant, which is comprised mainly by the cooling pumps and heat exchangers, ensures the safety of the spent fuel assemblies and the integrity of the fuel rods during the period of storage. With the development of the passive cooling technique, a spray cooling system for the spent fuels based on the gravity was designed to further enhance the safety of the spent fuel pool in case of accident conditions. This paper presents an experimental investigation of the validity of the spray-cooling system using two types of tight rod bundles, namely a 5 × 5 heated rod bundle and a 17 × 17 isothermal rod bundle. Results shows that the rod bundle heated with a lower power can be effectively cooled only by air without any spray water. With the increase of the heated power, the rod surface temperature increases gradually and the spray cooling has to be implemented to maintain the wall temperature at a certain level. The effect of flow rate on wall temperature was investigated. For the isothermal rod bundle, main interests were focused on the distribution of the spray water after it flowed along the rods.

An empirical correlation for the estimation of the droplet entrainment from boiling water pools is under development. The correlation is validated on the basis of experiments, taken from test facilities of different scale, which investigate the effect of several boundary conditions on the droplet entrainment. The calculated values for the droplet entrainment are in good agreement with the experimental data, indicating that relevant influences are considered in the developed correlation with an appropriate quantity. Nevertheless, the application of the correlation on scenarios with flashing pools and non-boiling pools with gas flows through the pool surface have to be investigated in the ongoing validation process. Furthermore, other experiments are needed, which consider pools containing surface-active substances to depict the influence of surface tension on the droplet entrainment.

Sodium-cooled fast reactor (SFR) is one of the most promising reactors among the six Gen-IV reactor systems due to its significant advantages in close fuel cycle, comprehensive technology foundations and operation experiences. China is designing and constructing a demonstration SFR, in which, to ensure the reactor passive safety, a direct reactor auxiliary cooling system (DRACS) with the inter-wrapper flow is proposed as the decay heat removal system. Xi'an Jiaotong University is in charge of the DRACS analysis code development. The physical models in the DRACS are extracted and the numerical models for each component are established. The Gear method and the SIMPLE method are adopted as the primary solution algorithm. The code is developed and validated by EBR-II and PHENIX benchmarks, the results of which indicate that the code can predict experimental results very well.

Because of the large variation of thermal–physical properties near pseudo-critical point, heat transfer of supercritical water shows abnormal behavior. Thus, an accurate prediction of heat transfer between the cladding and fluid plays a very important role. It is necessary to investigate the reliably of heat transfer models in the vicinity of the pseudo-critical point. In the frame of evaluating the present HT models and developing new reliable prediction models, databank of heat transfer of supercritical water flowing in tube were established by Karlsruhe Institute of Technology (KIT) with more than 35,000 experimental data. A thorough analysis and assessment was carried out, to give an insight into the characters of the database. Experimental data from different sources are compared to figure out the deviation between different sources and to present the reliability of the database.

Blankets are key elements of a future fusion power reactor, as they breed the fusion fuel tritium, extract the heat from the reactor for power generation and contribute to the nuclear shielding of the plasma confining magnetic field coils. On the way to the engineering implementation of fusion, in particular the blanket design approach has changed substantially. Novel blanket designs require, already from the beginning, incorporating close coupling of plasma physics with engineering physics to develop robust solutions coping with thermal, mechanical and also electrodynamic loads – not only during the stationary operating phase, but also during transients. Simultaneously, nuclear licensing feasibility as well as component failure safety must be part of the design approach. This article describes advanced blanket design approaches undertaken in the past years by the example of the helium cooled pebble bed blanket (HCPB), aiming at an efficient blanket engineering design, starting from the development of modular integral reactor analysis tools, via design analysis and engineering validation of fabrication and interface performance, towards safety analysis on the reactor level.

The helium cooled solid breeder (HCSB) blanket, which is one of the blanket candidates for Chinese Fusion Engineering Test Reactor (CFETR), has been developed in University of Science and Technology of China (USTC). And the blanket cooling system (HCS) is also designed for HCSB. The steady-state results show that the blanket system has a good relevance to the design parameters. The transient analysis of possible accidents should proceed to evaluate its safety performance. The accident analysis in this paper adopts the ITER-like conditions and the ex-vessel loss of coolant accident (LOCA) case is investigated with RELAP5 code. The results show that the ex-vessel LOCA with detection couldn't cause serious consequence, however the ex-vessel LOCA without detection has potential to melt the blanket that we should take measures to avoid it.

Within the framework of EUROfusion research activities, the Karlsruhe Institute of Technology (KIT) is developing the Helium Cooled Pebble Bed (HCPB) breeding blanket (BB) for European demonstration fusion power plant (DEMO), with the support of BME and Wigner-RCP (Hungary) and CIEMAT (Spain). During the development of the BB concept in the past years, the in-box Loss Of Coolant Accident (LOCA) has been identified as a main design driver for BB. During this event, a leakage of pressurized coolant within the blanket box happens (in-box LOCA), pressurizing the blanket module and endangering its structural integrity. This accident has not been investigated in a detailed approach yet. In this paper, a more precise methodology is introduced taking into account the temperature evolution of the structure for assessing the structural integrity of BB under in-box LOCA and has been applied to the HCPB BB. The methodology and results are presented and critically discussed.