A thermophysical heavy water properties package for thermal-hydraulic accident analysis codes

A thermodynamic and transport properties package for heavy water (D₂O) has been prepared. This package has been implemented in an advanced Nuclear Reactors Thermal-Hydraulic accident analysis code. Several practical problems are analyzed and a comparison between D₂O and H₂O as cooling agents is presented.     

Model for melt blockage (slug) relocation and physico-chemical interactions during core degradation under severe accident conditions

The model describing massive melt blockage (slug) relocation and physico-chemical interactions with steam and surrounding fuel rods of a bundle is developed on the base of the observations in the CORA tests. Mass exchange owing to slug oxidation and fuel rods dissolution is described by the previously developed 2D model for the molten pool oxidation. Heat fluxes in oxidising melt along with the oxidation heat effect at the melt relocation front are counterbalanced by the heat losses in the surrounding media and the fusion heat effect of the Zr claddings attacked by the melt. As a result, the slug relocation velocity is calculated from the heat flux matches at the melt propagation front (Stefan problem). A numerical module simulating the slug behaviour is developed by tight coupling of the heat and mass exchange modules. The new model demonstrates a reasonable capability to simulate the main features of the massive slug behaviour observed in the CORA-W1 test.     

A use of prototypic material for the investigation of severe accident progression

An experimental research platform using corium melts is established for the understanding of safety related important phenomena during a severe accident progression. The research platform includes TROI facility for corium water interaction experiments and VESTA facility for corium-structural material interaction experiments. A cold crucible technology is adapted and improved for a generation of 5–100 kg of corium melts at various compositions. TROI facility is used for experiments to investigate premixing and explosion behaviors during a fuel coolant interaction process. More than 70 experiments using corium at various compositions were performed to simulate steam explosion phenomena in a reactor situation. The results indicate that the conversion efficiency of steam explosion for corium is less than 1%. VESTA facility is used to investigate molten corium-structural material interaction phenomena. VESTA facility consists of two cold crucibles. One crucible is used for the melting of charged material and pouring of corium melt. The other crucible is used for the corium-structural material interaction while providing an induction heating to simulate the decay heat. The results of an experiment on the interaction between corium melt and a specimen made of Inconel performed in the VESTA facility is reported.     

Experimental investigation of in-vessel mixing phenomena in a VVER-1000 scaled test facility during unsteady asymmetric transients

In-vessel turbulent mixing phenomena affect the time and space distribution of coolant properties (e.g., boron concentration and temperature) at the core inlet which impacts consequently the neutron kinetics response. For reactor safety evaluation purposes and to characterize these phenomena it is necessary to set and validate appropriate numerical modelling tools to improve the current conservative predictions. With such purpose, an experimental campaign was carried out by OKB Gidropress, in the framework of the European Commission Project “TACIS R2.02/02 - Development of safety analysis capabilities for VVER-1000 transients involving spatial variations of coolant properties (temperature or boron concentration) at core inlet”. The experiments were conducted on a scaled facility representing the primary system of a VVER-1000 including a detailed model of the Reactor Pressure Vessel with its internals. The simulated transients involved perturbations of coolant properties distribution providing a wide validation matrix. The main achievements of the set of experiments featuring transient asymmetric pump behaviour are presented in this paper. The potential of the obtained experimental database for the validation of thermal fluid dynamics numerical simulation tools is also discussed and the role of computational fluid dynamics in supporting the experimental data analysis is highlighted.     

中国实验快堆失去厂外电后单台主泵停运的一回路瞬态特性分析

失去厂外电源是在中国实验快堆(CEFR)运行寿期内可能多次发生的预计运行事件。本文基于自主研发的系统瞬态分析程序FASYS分析了CEFR失去厂外电源后单台主泵停运的事件,并将事件过程中的关键参数与试验结果进行了对比。通过试验和模拟分析,得到了中国实验快堆失去厂外电源后单台主泵停运的一回路瞬态特性。     

离子注入法掺杂聚乙炔的初步探索

聚乙炔(CH)x是一种简单的共轭有机导电高分子材料。1975年日本白川首先用高浓度的Ti(OBu)_4-AlEt_3体系在-78℃下定向聚合得到了均匀的、具有高结晶度的顺式(CH)x薄膜。1977年白川与美国MacCDiarmid等合作,发现掺杂能改善(CH)x的导电性能。后来又发现掺入不同的杂质可使(CH)x显示n型或P型半导体的特性。聚乙炔薄膜具有物理上准一维金属膜型的一些特殊性质,因而引起物理学家的广泛兴趣。1980年美国Su、Schrieffer和Heeger提出了聚乙炔的孤立子(Soliton)导电模型,并用这种模型成功地阐述了一些实验现象。(CH)x的原料便宜,合成简单,具有广阔的应用前景。人们预期可以把聚乙炔应用于制备太阳能电池、蓄电池及塑料半导体器件等。     

Drop tests of TUK-84 for accident conditions during shipment

Tests for mechanical damage of the TUK-84 shipment assembly, used for shipping and dry storage of spent nuclear fuel, with the container drop from a height of 9 m on a rigid base and from a height of 1 m on a steel pin are described. The basic data from the measurements of the impact parameters are presented, the results of tests for defects and post-test checking are presented, and results of acceptance tests for mechanical damage are also presented. __________ Translated from Atomnaya énergiya, Vol. 100, No. 6, pp. 445–448, June, 2006.     

A comparative study of the MATRA-LMR-FB calculation with the SABRE result for the flow blockage accident in the sodium cooled fast reactor

A sub-channel flow blockage may be initiated by an ingression of damaged fuel debris or foreign obstacles into a core subassembly for the sodium cooled fast reactor (SFR) due to the compact design of the fuel arrangement. Since local coolant temperature could go up high enough to reach a safety limit by the blockage disturbance in the subassembly, the MATRA-LMR-FB code was developed to analyze such blockage effect. An effort has been undergoing to enhance its reliability.In this study, a code-to-code comparison analysis with another code, SABRE4, was performed to supplement a qualification of the MATRA-LMR-FB. The two codes were applied to the analysis of partial sub-channel blockage accidents in a subassembly of the KALIMER-150, which is a conceptual design of a sodium-cooled fast reactor with an electric output of 150 MW. The analyses were carried out not only for radially different blockage positions but also for different blockage sizes in the subassembly.In result, the two code results were generally agreed both in magnitude and trend within a range. Therefore, it was concluded that the comparison results could support complementarily the applicability of the MATRA-LMR-FB to the partial flow blockage accident in the subassembly of the SFR.     

Scaling for the ECC bypass phenomena during the LBLOCA reflood phase

As one of the advanced design features of the APR1400 (Advanced Power Reactor), a direct vessel injection (DVI) system is adopted instead of the conventional cold leg injection (CLI) system. It is known that the DVI system greatly enhances the reliability of the emergency core cooling (ECC) system. However, there is still a dispute on its performance in terms of water delivery to the reactor core during the reflood period of a large-break loss-of-coolant accident (LOCA). Thus, experimental validation is underway. In this paper, a new scaling method, using the time and velocity reduced “modified linear scaling law”, is suggested for the design of a scaled-down experimental facility to investigate the direct ECC bypass phenomena in the PWR downcomer.     

Preliminary steps toward assessing aerosol retention in the break stage of a dry steam generator during severe accident SGTR sequences

Severe accidents SGTR sequences are identified as major contributors to risk of PWRs. Their relevance lies in the potential radioactive release from reactor coolant system to the environment. Lack of knowledge on the source term attenuation capability of the steam generator has avoided its consideration in probabilistic safety studies and severe accident management guidelines. This paper describes a research program presently under way on the aerosol retention on the tubes surrounding the breach within the secondary side of the steam generator in the absence of water. Its development has been internationally framed within the EU-SGTR and the ARTIST program. Experimental activities are focused on setting up a reliable database in which the influence of gas mass flow rate, breach configuration and particle nature in the aerosol retention are properly considered. Theoretical activities are aimed at developing a predictive tool (ARISG) capable of assessing source term attenuation in the scenario with reasonable accuracy. Given the major importance of jet aerodynamics, 3D CFD analyses are being conducted to assist both test interpretation and model development.     

A simpler approach for predicting LWR vessel failure during severe accident conditions

Using closed-form solution techniques, models were developed for assessing the thermal and structural response of light water reactor (LWR) vessels and penetrations during severe accident conditions. Results from models are displayed as failure maps, generally developed in terms of non-dimensional groups, so that a broader range of reactor design parameters and severe accident conditions can be considered. In this paper, failure maps are used to compare LWR vessel response to three accident conditions. Results discussed within this paper illustrate the importance of vessel and tube geometrical parameters and material properties for predicting which vessel failure mode occurs first.     

Na~(123)I放射化学纯度测定方法的初步探讨

不同价态碘的阴离了的检测,曾有不少报道。一些作者介绍了用电泳法分离无机阴离子I~-、IO_3~-、IO_4~-。也有一些作者报道了用不同展开剂在纤维素纸上分离I~-、IO_3~-、IO_4~-。R.Vandenbosch介绍了用凝胶过滤和放射性测量分别测得I~-、IO_3~-和IO_4~-的含量。A.Moghissi介绍了用硅胶薄板层析法分离三者含量。另外,用层析法来分离不同价态的阴离子的报道也不少。     

Computational investigation of the movement of a melted mass to the bottom of the vessel during an off-design accident in a fast reactor

NGTU. Experimental and Design Office of Machines. Translated from Atomnaya énergiya, Vol. 77, No. 3, pp. 180–185, September, 1994.     

Dosimetric system for determining the emission rate of gas-aerosol radioactive impurity during a radiological accident

A dosimetric system for determining the rate at which gas-aerosol radioactive impurities are discharged during a radiological emergency is examined. It is shown that the optimal system for evaluating the rate of discharge is one that consists of flow-through and closed (hermetic) ionization chambers. The addition of a xenon spectrometer to the system makes it possible to determine the partial rate of emission of individual radionuclides and the total emission rate of all nuclides. The optimal parameters of the ionization chambers making it possible to record the emission rate of a radioactive impurity with air flow rate to 12 m/sec against background dose rate up to 200 Sv/h are found.     

Downcomer boiling phenomena during the reflood phase of a large-break LOCA for the APR1400

Downcomer boiling phenomena in a conventional pressurized water reactor has an important effect on the transient behavior of a postulated large-break LOCA (LBLOCA), because it can degrade the hydraulic head of the coolant in the downcomer and consequently affect the reflood flow rate for a core cooling. To investigate the thermal hydraulic behavior in the downcomer region, a test program for a downcomer boiling (DOBO) is being progressed for the reflood phase of a postulated LBLOCA. Test facility was designed as a one side heated rectangular channel which adopts a full-pressure, full-height, and full-size downcomer-gap approach, but with the circumferential length reduced 47.08-fold. The test was performed by dividing it into two-phases: (I) visual observation and acquisition of the global two-phase flow parameters and (II(a)) measurement of the local bubble flow parameters on the measuring planes along five elevations. In the present paper, the test results of Phase-I and a part of Phase-II(a) were introduced.     

Modeling processes in the melt localization facility for aes-2006 with vver-1200 during an unanticipated accident

The results of a computational analysis of the behavior of melt in the facility used to localize melt for VVER-1200 (AES-2006 design), which were obtained using the GEFEST-ULR code, are presented. The main parameters of the melt are determined: the time variation of the component composition, the temperature and density of the oxide and metallic components and their relative arrangement. Using the KORSAR/GP thermohydraulic code, the minimum margin to crisis of heat transfer at the outer surface of the wall of the melt localization facility is analyzed. It is confirmed that heat transfer from the outer surface of the wall with normal operation of the cooling loop is reliable.     

A simple mass and heat balance model for estimating plant conditions during the Fukushima Dai-ichi NPP accident

A simple evaluation method for the analysis of thermal-hydraulic transients in reactor pressure vessel (RPV) and primary containment vessel (PCV) is proposed to support understanding the accident behaviors of the Fukushima Dai-ichi nuclear power plant (NPP). Since most of the measurements of the plants were unavailable especially in the early stage of the accident, and the accessibility to the plants had been limited by radiation, analytical investigation for the plant was required to understand the plant conditions such as the magnitude of the damages. In order to provide easy-to-use technical tools to support the analytical investigation, we developed a simplified analysis code, named “HOTCB”, based on total mass and heat balances in a lamped parameter system. The HOTCB code has capabilities to treat two-phase fluid including water, steam, and non-condensable gas in a wide range of temperatures up to highly superheated conditions, and to consider heat structures, i.e. heat capacities and heat transfer to the fluid. The code was provided to Tokyo Electric Power Company (TEPCO) and was practically used for the analysis on the accident. This paper provides the details of the code and simulations of Unit 1 and Unit 2 reactors of Fukushima Dai-ichi nuclear power plant (NPP) as examples to show the usefulness of the code.     

Study on the passive safe technology for the prevention of air ingress during the primary-pipe rupture accident of HTGR

A primary-pipe rupture accident is one of the design-basis accidents of a high-temperature gas-cooled reactor (HTGR). When the primary-pipe rupture accident occurs, air is expected to enter the reactor core from the breach and oxidize in-core graphite structures. This study is to investigate the air ingress phenomena and to develop the passive safe technology for the prevention of air ingress and of graphite corrosion. This paper describes the method for the prevention of air ingress into the reactor during the primary-pipe rupture accident. It is found that a safe cooling rate of the reactor core exists for the prevention of air ingress. The experimental results show that the natural circulation flow of air during the accident can be controlled by the method of helium gas injection into the reactor pressure vessel.