Numerical analysis of pressure load in a PWR cavity in an ex-vessel steam explosion

Ex-vessel steam explosion may happen as a result of melting core falling into the reactor cavity after failure of the reactor vessel and interaction with the coolant in the cavity pool. It can cause the formation of shock waves and production of missiles that may endanger surrounding structures. Ex-vessel steam explosion ener- getics is affected strongly by three dimensional （3D） structure geometry and initial conditions. Ex-vessel steam explosions in a typical pressurized water reactor cavity are analyzed with the code MC3D, which is developed for simulating fuel-coolant interactions. The reactor cavity with a venting tunnel is modeled based on 3D cylin- drical coordinate. A study was performed with parameters of the location of molten drop release, break size, melting temperature, cavity water subcooling, triggering time and explosion position, so as to establish parame- ters＇ influence on the fuel-coolant interaction behavior, to determine the most challenging cases and to estimate the expected pressure loadings on the cavity walls. The most dangerous case shows the pressure loading is above the capacity of a typical reactor cavity wall.     

熔融物液柱碎化模型对先进压水堆堆外蒸汽爆炸计算的影响分析

在堆外蒸汽爆炸计算中,液柱碎化模型影响着熔融物液滴生成速率、液滴直径、液滴分布、液滴凝固和气泡比例等粗混合参数和现象,从而影响了蒸汽爆炸的冲击载荷。本文基于MC3D V3.8程序,采用不同的液柱碎化模型（CONST模型和KHF模型）对先进压水堆堆外蒸汽爆炸进行计算分析,探讨了CONST和KHF模型对蒸汽爆炸计算的影响。结果表明,两种模型计算的粗混合状态类似;在熔融物触底时刻,爆炸性准则几乎相同,此时触发爆炸得到的冲击载荷差别很小,表明该时刻触发爆炸时不同液柱碎化模型对爆炸冲击计算的影响较小;在本文所定义的工况下,先进压水堆堆坑墙体承受的最高压力约为20 MPa,最大冲量小于0.2 MPa•s。     

Corrosion and deposition on the secondary circuit of steam generators

This study conducts a critical review on the studies of material corrosion and deposition on the secondary circuit of a pressurized water reactor, especially on the steam generators (SGs). Available knowledge has shown that the structural materials in the environment of the secondary circuit are susceptible to flow-accelerated corrosion and deposition-induced degradation. The deposition of the non-volatile impurities, especially the corrosion products, on the SG surfaces can be a primary cause of material degradations, including stress corrosion cracking. The review will analyze the fouling mechanisms and behaviors, the source of impurities, corrosion mechanisms, and the factors that affect the deposition and corrosion behaviors.     

Effect of water injection on hydrogen generation during severe accident in PWR

Effect of water injection on hydrogen generation during severe accident in a 1000 MWe pressurized water reactor was studied.The analyses were carried out with different water injection rates at different core damage stages.The core can be quenched and accident progression can be terminated by water injection at the time before cohesive core debris is formed at lower core region.Hydrogen generation rate decreases with water injection into the core at the peak core temperature of 1700 K,because the core is quenched and reflooded quickly.The water injection at the peak core temperature of 1900 K,the hydrogen generation rate increases at low injection rates of the water,as the core is quenched slowly and the core remains in uncovered condition at high temperatures for a longer time than the situation of high injection rate.At peak core temperature of 2100-2300 K,the Hydrogen generation rate increases by water injection because of the steam serving to the high temperature steam-starved core.Hydrogen generation rate increases significantly after water injection into the core at peak core temperature of 2500 K because of the steam serving to the relocating Zr-U-O mixture.Almost no hydrogen generation can be seen in base case after formation of the molten pool at the lower core region.However,hydrogen is generated if water is injected into the molten pool,because steam serves to the crust supporting the molten pool.Reactor coolant system (RCS) depressurization by opening power operated relief valves has important effect on hydrogen generation.Special attention should be paid to hydrogen generation enhancement caused by RCS depressurization.     

高温气冷堆蒸汽发生器舱室混凝土温度场分析

采用ABAQUS6.7有限元分析软件,对高温气冷堆蒸汽发生器舱室混凝土在正常工况和设备冷却水系统停止供水事故工况下的温度场进行了计算。结果表明,在正常工况下,蒸汽发生器舱室混凝土的最高温度低于规定的限值;在设备冷却水系统停止对屏蔽冷却水系统供水事故工况下,7天内混凝土最高温度低于100℃,屏蔽冷却水系统能够保证对蒸汽发生器舱室的冷却。     

Investigation on two-phase critical flow for loss-of-coolant accident of pressurized water reactor

The previous investigations were mainly conducted under the condition of low pressure,however,the steam-water specific volume and the interphase evaporation rate in high pressure are much different from those in low pressure,Therefore,the new experimental and theoretical investigation are performed in Xi‘an Jiaotong University.The investigation results could be directly applied to the analysis of loss-of -coolant accident for pressurized water reacor.The system transition characteristics of cold leg and hot leg break loss-of -coolant tests are described for convective circulation test loop.Two types of loss-of-coolant accident are identified for :hot leg” break,while three types for “cold leg”break and the effect parameters on the break geometries.Tests indicate that the mass flow rate with convergent-divergent nozzle reaches the maximum value among the different break sections at the same inlet fluid condition because the fluid separation does not occur.A wall surface cavity nucleation model is developed for prediction of the critical mass flow rate with water flowing in convergentdivergent nozzles.     

Effects of molten material temperatures and coolant temperatures on vapor explosion

An observable experiment facility for low-temperature molten materials to be dropped into water was set up in this study to investigate the mechanism of the vapor explosion. The effect of the fuel and coolant interaction (FCI) on the vapor explosion during the severe accidents of a fission nuclear reactor has been studied. The experiment results showed that the molten material temperature has an important effect on the vapor explosion behavior and pressure. The increase of the coolant temperature would decrease the pressure of the vapor explosion.     

Theoretical investigation on the steady-state natural circulation characteristics of a new type of pressurized water reactor

1 Introduction With respect to the inherent safety of nuclear re- actors, application of passive systems/components including natural circulation phenomena as the main mechanism is intended to simplify the safety-related systems and to improve their reliability, to reduce the effect of human errors and equipment failures, and to provide more time to enable the operators to prevent or mitigate serious accidents. Natural circulation is the main mode of heat removal for removing decay heat from t…     

压水堆核电厂流出物中14C的排放限值探讨

研究分析了压水堆核电厂中¹⁴C的产生途径与排放量,调研了美国和欧洲运行压水堆核电厂气态流出物和液态流出物中¹⁴C的排放水平,分析了我国国家标准《核动力厂环境辐射防护规定》(GB 6249—2011)对美国和欧洲运行压水堆核电厂流出物排放¹⁴C的包络性,同时分析了多堆厂址、AP1000和EPR等新堆型电厂的运行需求对目前标准规定的¹⁴C排放限值管理带来的挑战,提出了¹⁴C的减排和资源化利用建议。     

Visualization on the behavior of inert gas jets impinging on a single glass tube submerged in liquid sodium

In order to accurately model sodium–water reaction jets in steam generators of fast breeder reactors, knowledge of size distributions or mean diameters of liquid sodium droplets entrained into the reaction jets is prerequisite. In the present study, argon-gas jet behaviors, without chemical reaction, injected into liquid sodium were successfully visualized using an endoscope and a glass tube, and the size distributions and mean diameters of liquid sodium droplets entrained into the gas jet were also obtained in the bubbling regime. Most of the liquid sodium droplets were observed to be intermittently produced in the vicinity of a gas nozzle in the present study. The droplet size distributions of entrained sodium droplets were found to agree well with the Nukiyama–Tanasawa distribution function when the arithmetic mean diameter was used. The Sauter mean diameters obtained in the present study were also found to be well correlated with an empirical equation proposed by Epstein et al. The present study shows that the existing knowledge, which is based on the results of water experiments, is suitable in terms of accuracy in practice.     

Effect of Pu-rich agglomerates in MOX fuel on reactivity analysis of light water reactor MOX core physics experiments

The effect of Pu-rich agglomerates in U-Pu mixed oxide (MOX) on the reactivity analysis of light water reactor MOX core physics experiments was studied with the continuous-energy Monte Carlo calculation code MVP II. First, the following three different models were compared in the analysis of a representative unit cell of a MOX core tested at the KRITZ reactor: a Lattice model where Pu-rich agglomerates were assumed to exist in a fixed pitch, a statistical geometry (STG) model of MVP II, and a Random model where the random distribution of Pu-rich agglomerates was directly modeled. Since the three models gave comparable results, the STG model was used in parametric calculations to systematically understand the reactivity effect depending on the characteristics of Pu-rich agglomerates. In addition, the selected unit cells composing the MOX cores and one representing MOX core tested at the EOLE criticality facility were analyzed with the measured characteristics of Pu-rich agglomerates in MOX fuel. Consequently, the reactivity differences between the calculations assuming the homogeneous Pu distributions and those considering Pu-rich agglomerates were less than 0.0005 Δk/k/k', indicating that the effect of Pu-rich agglomerates was small on the reactivity analysis of the MOX cores tested in the EOLE facility.     

Effect of hydrocarbons on the efficiency of catalytic reactor of detritiation system in an event of fire

Detritiation system of a nuclear fusion plant is mandatory to be designed and qualified taking carefully into consideration all the possible extraordinary situations in addition to that in a normal condition. We focused on the change in the efficiency of tritium oxidation of a catalytic reactor in an event of fire where the air accompanied with hydrocarbons, water vapor, and tritium is fed into a catalytic reactor at the same time. Our test results on the effect of these gases on the efficiency of tritium oxidation of the catalytic reactor indicated; (1) tritiated hydrocarbon produces significantly by reaction between tritium and hydrocarbons in a catalytic reactor; (2) there is little possibility of degradation in the detritiation performance because the tritiated hydrocarbons produced in the catalyst reactor are combusted; (3) there is no possibility of uncontrollable rise in the temperature of the catalytic reactor by heat of reactions; and (4) saturated water vapor could temporarily poison the catalyst and degrades the detritiation performance. Our investigation indicated a saturated water vapor condition without hydrocarbons would be the dominant scenario to determine the amount of catalyst for the design of catalytic reactor of the detritiation system.     

Effect of corrosion potential on the corrosion fatigue crack growth behaviour of low-alloy steels in high-temperature water

The low-frequency corrosion fatigue (CF) crack growth behaviour of different low-alloy reactor pressure vessel steels was characterized under simulated boiling water reactor conditions by cyclic fatigue tests with pre-cracked fracture mechanics specimens. The experiments were performed in the temperature range of 240-288 °C with different loading parameters at different electrochemical corrosion potentials (ECPs). Modern high-temperature water loops, on-line crack growth monitoring (DCPD) and fractographical analysis by SEM were used to quantify the cracking response. In this paper the effect of ECP on the CF crack growth behaviour is discussed and compared with the crack growth model of General Electric (GE). The ECP mainly affected the transition from fast (‘high-sulphur’) to slow (‘low-sulphur’) CF crack growth, which appeared as critical frequencies ν_crit = f(ΔK, R, ECP) and ΔK-thresholds ΔK_EAC = f(ν, R, ECP) in the cycle-based form and as a critical air fatigue crack growth rate da/dt_Air,crit in the time-domain form. The critical crack growth rates, frequencies, and ΔK_EAC-thresholds were shifted to lower values with increasing ECP. The CF crack growth rates of all materials were conservatively covered by the ‘high-sulphur’ CF line of the GE-model for all investigated temperatures and frequencies. Under most system conditions, the model seems to reasonably well predict the experimentally observed parameter trends. Only under highly oxidizing conditions (ECP ? 0 mV_SHE) and slow strain rates/low loading frequencies the GE-model does not conservatively cover the experimentally gathered crack growth rate data. Based on the GE-model and the observed cracking behaviour a simple time-domain superposition-model could be used to develop improved reference CF crack growth curves for codes.     

Effect of spring water on the radon concentration in the air at Masutomi spa in Yamanashi Prefecture,Japan

The concentrations of ²²²Rn existing in air have been studied by using a convenient and highly sensitive Pico-rad detector system at Masutomi spa in Yamanashi Prefecture, Japan. The measurements in air were carried out indoors and outdoors during the winter of 2000 and the summers of 1999 and 2005. The concentrations of ²²²Rn in spring water in this region were measured by the liquid scintillation method. The concentrations of natural radionuclides contained in soils surrounding spa areas were also examined by means of the γ-ray energy spectrometry technique using a Ge diode detector to investigate the correlation between the radionuclides contents and ²²²Rn concentrations in air at each point of interest. The atmospheric ²²²Rn concentrations in these areas were high, ranging from 5 Bq/m³ to 2676 Bq/m³. The radon concentration at each hotel was high in the order of the bath room, the dressing room, the lobby, and the outdoor area near the hotel, with averages and standard deviations of the concentration of 441 ± 79 Bq/m³, 351 ± 283 Bq/m³, 121 ± 5 Bq/m³, and 23 ± 1 Bq/m³, respectively. The source of ²²²Rn in the air in the bath room is more likely to be the spring water than the soil. The spring water plays carries the radon to the atmosphere. Our measurements indicated that the ²²²Rn concentration in the air was affected by the ²²²Rn concentration in spring water rather than that in soil.     

Quantities of I-131 and Cs-137 in accumulated water in the basements of reactor buildings in process of core cooling at Fukushima Daiichi nuclear power plants accident and its influence on late phase source terms

During the process of core cooling at Fukushima Daiichi nuclear power plants accident, large amount of contaminated water was accumulated in the basements of the reactor buildings at Units 1–4. The present study estimated the quantities of I-131 and Cs-137 in the water as of late March based on the press-opened data. The estimated ratios of I-131 and Cs-137 quantities to the core inventories are 0.51%, 0.85% at Unit 1, 74%, 38% at Unit 2 and 26%, 18% at Unit 3, respectively. According to the Henry's law, certain fraction of iodine in water could be released to atmosphere due to gas–liquid partition and contribute to increase in the release to environment. A lot of evaluations for I-131 release have been performed so far by the MELCOR calculation or the SPEEDI reverse estimation. The SPEEDI reverse predicted significant release until 26 March, while no prediction in MELCOR after 17 March. The present study showed that iodine release from accumulated water may explain the release between 17 and 26 March. This strongly suggests a need for improvement of current MELCOR approach which treats the release only from containment breaks for several days after the core melt.