Characterization of oxide films formed on alloy 182 in simulated PWR primary water

The oxide film, formed on the alloy 182 due to the exposure to simulated Pressurized Water Reactor primary water conditions, is characterized using X-ray Photoelectron Spectroscopy, Scanning Electron Microscopy and Raman spectroscopy. As a consequence of the low redox potential for nickel, loosely bound oxide film consisting mainly of chromium oxide, is formed on the surface of the alloy. The oxide film is found to have a double-layer structure, with an inner layer rich in Cr₂O₃, outer layer rich in FeCr₂O₄. Thin Ni(OH)₂ and Fe₃O₄ clusters were observed on top of the oxide film. The morphology and thickness of the layers critically depend on the exposure times and surface treatments prior to the exposure.     

The thermal-hydraulics of a simulated PWR facility during steam generator tube rupture transients

The thermal-hydraulics of the semi-scale test facility during steam generator tube rupture transients were investigated in this paper. The test facility simulates the main features of a Westinghouse four-loop pressurized water reactor (PWR) plant.The constructed analytical model simulated both the intact and broken loops, and included the vessel (lower plenum, core, upper plenum, upper dome), the hot legs, pressurizer and the primary and secondary sides of the U-tube steam generators. The two-phase Modular Modeling System code, which was developed by the Electric Power Research Institute, and the EASY5 simulation language were used in carrying out the calculations. A control model was developed to simulate the major facility control systems and to perform the necessary control functions.Calculations were carried out during the first three hundred seconds of the event, where the automatically functioning plant protection system components were assumed to operate. The impact of reactor scram, pressurizer heater activation, main steam isolation valve closure, emergency core cooling system activation, pump trip, main feedwater termination, auxiliary feedwater injection, and atmospheric dump/safety relief valves opening/closing on the system response was calculated.The time histories of the thermal-hydraulic conditions, such as pressure and temperature, are presented for one, five and ten-tube ruptures. Comparisons with experimental data and RELAP-5 (MOD 1.5) calculations are also given.     

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.     

Influence of δ phase precipitation on the stress corrosion cracking resistance of alloy 718 in PWR primary water

The negative influence of δ phase on the intergranular stress corrosion cracking (IGSCC) resistance of alloy 718 is commonly taken for granted. In addition, δ phase formed at low temperature (about 1023 K) do not present the same characteristics than the one formed at higher temperatures (from 1173 to 1273 K). The aim of the present study is then to understand how δ phase precipitation could enhance crack initiation in alloy 718, whatever the form of δ phase is. For that purpose, several heat treatments leading to δ phase precipitation were realized on two alloy 718 heats, one sensitive to IGSCC and the second not. Specific slow strain rate tensile tests carried out on thin tensile specimens in simulated PWR primary medium at 633 K conclusively prove that δ phase has no effect on the intrinsic sensitivity to intergranular crack initiation of tested heats.     

Severe accident analysis of a 1300 MWe PWR to address the hydrogen issue

The behaviour of the potentially large quantity of hydrogen generated during a severe accident has been recognised as an issue of importance since the accident at Three Mile Island. In this article, we describe a severe accident analysis for the Neckarwestheim 2 1300 MWe PWR “Konvoi” plant, performed primarily to investigate the behaviour of hydrogen in the containment, and draw conclusions regarding the need for hydrogen control systems (igniters). The Modular Accident Analysis Program (MAAP) developed by IDCOR in the United States, and the Westinghouse COMPACT multi-compartment containment code were used. The study investigated the generation, release to containment, distribution within containment and potential combustion of hydrogen produced during two severe accident sequences. Results are summarized which show that hydrogen mixing in containment is generally good and that even without hydrogen control systems, hydrogen combustion, although possible, does not threaten containment integrity.     

压水堆核电站氧化停堆过程一回路冷却剂中58Co活度浓度分析

压水堆核电站氧化停堆过程中,一回路冷却剂中⁵⁸Co的停堆释放峰值可达上百个GBq/t,对工作人员的职业照射剂量及停堆进程都有很大影响。本文介绍了压水堆核电站氧化停堆过程,分析了对⁵⁸Co活度浓度变化有显著影响的因素,如一回路水化学、蒸汽发生器传热管材料、循环中停堆、化学和容积控制系统的净化等,同时提出了相关建议。     

Influence of oxide films on primary water stress corrosion cracking initiation of alloy 600

In the present study alloy 600 was tested in simulated pressurised water reactor (PWR) primary water, at 360 °C, under an hydrogen partial pressure of 30 kPa. These testing conditions correspond to the maximum sensitivity of alloy 600 to crack initiation. The resulting oxidised structures (corrosion scale and underlying metal) were characterised. A chromium rich oxide layer was revealed, the underlying metal being chromium depleted. In addition, analysis of the chemical composition of the metal close to the oxide scale had allowed to detect oxygen under the oxide scale and particularly in a triple grain boundary. Implication of such a finding on the crack initiation of alloy 600 is discussed. Significant diminution of the crack initiation time was observed for sample oxidised before stress corrosion tests. In view of these results, a mechanism for stress corrosion crack initiation of alloy 600 in PWR primary water was proposed.     

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.     

Effect of shot peening on primary water stress corrosion cracking of Alloy 600 steam generator tubes in an operating PWR plant

The effect of shot peening on the primary stress corrosion cracking behavior of thermally treated Alloy 600 steam generator tubes in an operating pressurized water reactor (PWR) plant was analyzed based on pulled tube examinations and in-service inspection eddy current test (ISI-ECT) data. The evaluation was focused on the shape of crack, evolution of the number of new cracks and cracked tube fraction, and variation of crack length and the corresponding eddy current amplitude before and after shot peening. The shape of the crack was changed from a half-elliptical type before shot peening, to an elliptical one with bulging after peening. It was concluded that the shot peening was not effective for retarding both crack initiation and growth for this plant.     

Study of release of nitrogen gas dissolved in water during depressurisation—application to primary circuit of PWR

The potential release of the nitrogen (N₂) gas dissolved in the water of the accumulators of the emergency core coolant system of the Loviisa Nuclear Power Plant (PWR of VVER-440 type) has been investigated. A model of the dissolution and release of N₂ gas has been implemented in the thermal-hydraulic code CATHARE for nuclear safety. In collaboration with VTT, an analytical experiment has been performed with some components of the PACTEL facility to determine, in particular, the value of the release time constant of the nitrogen gas in the depressurization conditions representative of the small and intermediate break transients postulated for the Loviisa Nuclear Power Plant. Such transients, with simplified operating procedures, were calculated using the modified CATHARE code. In comparison with the cases calculated without taking into account the release of nitrogen gas, the start of the LPIS is delayed by between 1 and 1.75 h. Applicability of the obtained results to the real safety conditions must take into account the real operating procedures used in the Loviisa Nuclear Power Plant.     

Influence of ageing heat treatment on alloy A-286 microstructure and stress corrosion cracking behaviour in PWR primary water

The influence of ageing heat treatment on alloy A-286 microstructure and stress corrosion cracking behaviour in simulated Pressurized Water Reactor (PWR) primary water has been investigated. A-286 microstructure was characterized by transmission electron microscopy for ageing heat treatments at 670 °C and 720 °C for durations ranging from 5 h to 100 h. Spherical γ′ phase with mean diameters ranging from 4.6 to 9.6 nm and densities ranging from 8.5 × 10²² m⁻³ to 2 × 10²³ m⁻³ were measured. Results suggest that both the γ′ phase mean diameter and density quickly saturate with time for ageing heat treatment at 720 °C while the γ′ mean diameter increases significantly up to 100 h for ageing heat treatment at 670 °C. Grain boundary η phase precipitates were systematically observed for ageing heat treatment at 720 °C even for short ageing periods. In contrast, no grain boundary η phase precipitates were observed for ageing heat treatments at 670 °C except after 100 h. Hardening by γ′ precipitation was well described by the dispersed barrier hardening model with a γ′ barrier strength of 0.23. Stress corrosion cracking behaviour of A-286 was investigated by means of constant elongation rate tensile tests at 1.5 × 10⁻⁷ s⁻¹ in simulated PWR primary water at 320 °C and 360 °C. In all cases, initiation was transgranular while propagation was intergranular. Grain boundary η phase precipitates were found to have no significant effect on stress corrosion cracking. In contrast, yield strength and to a lesser extent temperature were found to have significant influences on A-286 susceptibility to stress corrosion cracking.     

Zr9Ni11合金的吸氢性能测定

研究了Zr9Ni11合金室温下多次吸放氢循环中吸氢量变化及低压下吸氢动力学特征,并采用SEM和XRD研究吸氢前后合金的微观特征。结果表明:初始压力为161 kPa时,室温下Zr9Ni11合金的吸氢量较高,前六次循环平均值达0.0045 mol/g,并随循环次数的增加出现轻微下降;Zr9Ni11合金的吸氢速率虽随氢压力减小而降低,但仍可在低压时于2–3 min内达到饱和吸氢量;SEM和XRD分析表明,吸放氢后合金表面组成发生了变化,由此引起合金吸氢量变化,Zr9Ni11合金是一种优良的除氚备选材料。     

The contribution of non-wetting droplets to direct cooling of the fuel during PWR post-LOCA reflood

During the reflood of a Pressurised Water Reactor (PWR) following a loss of coolant accident, precursory cooling prior to the arrival of the rewetting front is of vital importance in limiting the rise in cladding temperature before rewet. This precursory cooling is achieved by a flow of superheated vapour, with entrained saturated drops, which evaporate into the vapour and act as a heat sink. In this paper we investigate a complementary mechanism; the direct cooling of the cladding by the drops themselves. Cladding temperatures are such that wetting by these droplets does not occur. On the contrary, droplets bounce off a vapour cushion formed during the ∼10 ms or so that they are in close proximity to the cladding. Using a combination of previous experimental correlations and recent CFD calculations, we estimate the rate of heat removal from the cladding surface as a result of the droplet impingement. Thus, we estimate the heat removed as a result of one impingement and estimate the total rate of heat removal by estimating the number droplets impinging on the cladding per unit surface area. The heat extracted by those droplets is found to be about 1/10 of the heat extracted by single-phase vapour under typical reflood conditions. Though there significant uncertainties in these estimates, it does seem that direct cooling by droplets, not generally incorporated in analyses of reflood, could actually be making a significant contribution to keeping cladding temperatures down to acceptable levels.     

高温下GH3535合金中的氢同位素扩散渗透效应分析

Flibe具有熔点低、中子性能好、沸点高等特点,是未来大型氟盐冷却高温堆的主要候选氟盐冷却剂之一,在堆芯中与中子相互作用后会导致一定量的副产物氚产生。根据哈氏N合金的成分,钍基熔盐堆核能系统(TMSR)发展了GH3535合金,作为未来大型熔盐堆的主要候选结构材料。本实验中采用GH3535合金为试样,通过使用压力差驱动原理搭建的氢同位素扩散渗透装置,试验测得了400~800℃的温度下氢气、氘气在该合金中的渗透系数、扩散系数、Sieverts常数等主要参数。实验结果表明,氢气与氘气在GH3535合金中的扩散渗透机理均属于基体扩散控制过程,扩散渗透过程中氢气、氘气的主要参数与相应温度关系均符合阿累尼乌斯公式。对于不同质量数的氢同位素原子,拟合后渗透系数和扩散系数的指前因子之比分别为1.4:1和1.2:1,扩散和渗透过程中的激活能也非常接近,符合经典扩散理论的同位素效应,可以估算得到氚在GH3535合金中扩散渗透时的主要参数大约为氢的1/3(1/2),并可能用于氚在熔盐堆中的分布计算。     

Some considerations on the estimation of the reliability of PWR vessel and primary circuit

On the basis of experience acquired during a study on the reliability of PWR vessels, this text illustrates the points requiring more thorough investigation in order to raise the confidence level of the estimation of the reliability of the vessel.We also offer evidence of the need to develop, for already existing installations, adaptive type reliability models allowing a continually updated estimate of the reliability of the plant.     

富集硼酸在压水堆一回路水化学中的应用研究

压水堆核电厂一般采用天然硼来控制反应性。在核电厂实施长循环燃料管理后,寿期初硼浓度较高,增加了水化学控制的压力。本文开展了富集硼酸(EBA)在一回路水化学中的应用可行性及其对相关水质处理系统的影响分析。研究表明一回路采用EBA有助于降低结构材料的腐蚀和堆外辐射场,提高在役核电厂的经济性。     

百万千瓦级压水堆核电厂停堆PSA研究

对百万千瓦级核电厂的停堆运行事故风险进行内部事件1级概率安全评价(PSA),并根据不同的停堆进程分别建立停堆PSA模型,分析经历LOI-RRA水位对电厂风险水平构成的影响。分析结果表明停堆工况下的电厂风险不可忽视,在冷停堆工况下经历LOI-RRA水位导致堆芯损坏频率明显增加。     

Evaluation of ASME code flaw analysis procedure using the influence function method for application to PWR primary piping

This paper discusses stress intensity factor calculations and fatigue analysis for a PWR primary coolant piping system. The influence function method is applied to evaluate ASME Code Section XI Appendix A “analysis of flaw indication” for the application to a PWR primary piping. Results of the analysis are discussed in detail.     

The impact of radiolytic yield on the calculated ECP in PWR primary coolant circuits

A code, PWR-ECP, comprising chemistry, radiolysis, and mixed potential models has been developed to calculate radiolytic species concentrations and the corrosion potential of structural components at closely spaced points around the primary coolant circuits of pressurized water reactors (PWRs). The pH(T) of the coolant is calculated at each point of the primary-loop using a chemistry model for the B(OH)₃ + LiOH system. Although the chemistry/radiolysis/mixed potential code has the ability to calculate the transient reactor response, only the reactor steady state condition (normal operation) is discussed in this paper. The radiolysis model is a modified version of the code previously developed by Macdonald and coworkers to model the radiochemistry and corrosion properties of boiling water reactor primary coolant circuits. In the present work, the PWR-ECP code is used to explore the sensitivity of the calculated electrochemical corrosion potential (ECP) to the set of radiolytic yield data adopted; in this case, one set had been developed from ambient temperature experiments and another set reported elevated temperatures data. The calculations show that the calculated ECP is sensitive to the adopted values for the radiolytic yields.