Fluid-structure interaction analysis for pressurizer surge line subjected to thermal stratification

Serious mechanical damages such as cracks and plastic deformations due to excessive thermal stress caused by thermal stratification have been experienced in several nuclear power plants. In particular, the thermal stratification in the pressurizer surge line has been addressed as one of the significant safety and technical issues. In this study, a detailed unsteady computational fluid dynamics (CFD) analysis involving conjugate heat transfer analysis is performed to obtain the transient temperature distributions in the wall of the pressurizer surge line subjected to stratified internal flows either during out-surge or in-surge operation. The thermal loads from CFD calculations are transferred to the structural analysis code which is employed for the thermal stress analysis to investigate the response characteristics, and the fatigue analysis is ultimately performed. In addition, the thermal stress and fatigue analysis results obtained by applying the realistic temperature distributions from CFD calculations are compared with those by assuming the simplified temperature distributions to identify some requirements for a realistic and conservative thermal stress analysis from a safety point of view.     

Structural evaluation of a piping system subjected to thermal stratification

Piping systems in nuclear power plants are often designed for pressure, mechanical loads originating from deadweight and seismic events and operating thermal transients using the equations in the ASME Boiler and Pressure Vessel Code, Section III. In the last few decades a number of failures in piping have occurred due to thermal stratification caused by the mixing of hot and cold fluids under certain low flow conditions. Such stratified temperature fluid profiles give rise to circumferential metal temperature gradients through the pipe leading to high stresses causing fatigue damage. A simplified method has been developed in this work to estimate the stresses caused by the circumferential temperature distribution from thermal stratification. It has been proposed that the equation for the peak stress in the ASME Section III piping code include an additional term for thermal stratification.     

Thermal fatigue damage evaluation of a PWR NPP steam generator injection nozzle model subjected to thermal stratification phenomenon

Thermal stratification phenomenon with the same thermodynamic steam generator (SG) injection nozzle parameters was simulated. After 41 experiments, the experimental section was dismantled; cut and specimens were made of its material. Other specimens were made of the preserved pipe material. By comparing their fatigue tests results, the pipe material damage was evaluated. The water temperature layers and also the outside pipe wall temperatures were measured at the same level. Strains outside the pipe in 7 positions were measured. The experimental section develops thermal stratified flows, stresses and strains caused enlargement of material grain size and reduction in fatigue life.     

Stress analysis of a 900 MW pressurizer surge line including stratification effects

Following temperature monitoring programmes performed on 900 MW pressurized water reactor pressurizer surge lines, it has been reported that those lines are stratified in steady state, owing to their geometry. The highest temperature difference occurs during reactor heat-up and cool-down, reaching 110°C. Obviously, this phenomenon was not considered in nuclear steam supply system (NSSS) design transients and stress reports.Based on Electricité de France and FRAMATOME experiences, such as temperature measurements on site and mock-up, and thermal hydraulic computations, NSSS transients are updated. Stratification conditions are defined in different cross-sections of the line, using pressurizer temperature, hot leg temperature and flow rate, through the Froude number. A complete stress analysis of surge lines is performed including the updated transients and bending moment increase due to stratification. First of all different sensibility studies are carried out in order to simplify assumptions.Using a two-dimensional-one-dimensional method developed by FRAMATOME, the usage factor is then computed in different cross-sections, distinguishing upper and lower parts. In the presence of stratification, the surge line is subjected to thermal stresses following thermal shocks and to bending moment variation. These two load types are studied vs. time in order to reduce conservatism present in usual analyses.     

Structural response function approach for evaluation of thermal striping phenomena

A rational analysis method for thermal stress induced by fluid temperature fluctuation is developed, by utilizing frequency response characteristics of structures. High frequency components of temperature fluctuation are attenuated during a transfer process from fluid to structures. Low frequency components hardly induce thermal stress, since temperature homogenization in structures. Based on investigations of the frequency response mechanism, a frequency response function of structures was derived, which can predict stress amplitudes on structural surfaces from fluid temperature amplitudes and frequencies. It is formulated by multiplication of the effective heat transfer and the effective thermal stress functions. The frequency response function was applied to fatigue analysis of nuclear components, and clarified relation of fatigue damage to thermal hydraulic and structural design parameters.     

PWR annulus thermal hydraulics important to analysis of pressurized thermal shock

Analysis of plume mixing in the annulus of a pressurized water reactor (PWR) are presented. The plume mixing analysis is based on a simple two-dimensional model that accounts for the surrounding flow and confinement. A correlation for entrainment is presented and comparison with experiment is made.Mixed convection resulting from downflow between parallel heated plates is studied experimentally. The experimental system used to obtain the data is described with the scaling rationale for choosing the working fluid. Heat transfer results are presented in terms of a Nusselt number and a correlation is given. Results show an enhancement in heat transfer with increasing Gr_Dh/Re_Dh² due to an increase in turbulence intensity associated with the buoyant wall layer. The correlation obtained for Gr_Dh/Re_Dh² < 2.29 was found to be Nu_Dh/Nu_Dh,0 = 1 + 2.93 (Gr_Dh/Re_Dh²)^0.54, where where Nu_Dh,0 is given by the Dittus-Boelter correlation. Use of this correlation for Gr_Dh/Re_Dh² > 2.29 is not recommended due to an observed flow bifurcation in this neighborhood.     

Prediction method for thermal stratification in a reactor vessel

The prediction method for thermal stratification phenomena in a fast breeder reactor is described. The focus of attention is placed on the applicability of water test results to predict thermal stratification phenomena in a real plant. The basic feature of thermal stratification was examined in a cylindrical plenum, using water and sodium as test fluids. The similitude relationship between a small-scale test and a real plant is discussed in order to understand the experimental results. The scale-model experiments for LMFBRs (liquid metal-cooled fast breeder reactors) were also performed to see the effects of a reactor configuration and reactor-trip operation condition. Then the magnitudes of the temperature gradient and the ascending speed of stratified interface in the hot plenum of LMFBRs were predicted, based on the results of the water scale-model.     

Investigation of turbulence modelling in thermal stratification analysis

In-vessel thermal stratification analysis was carried out using a multidimensional thermohydraulic analysis code, in which a higher-order finite difference scheme was applied to the convection terms. Discussions centred on the buoyancy modelling in the vicinity of the stratification interface through comparisons between experiment and calculation.Computational results were obtained from the following three turbulence models: (i) the k−ε model with a constant turbulent Prandtl number Pr_t, (ii) the k−ε model with the turbulent Prandtl number being dependent on the local Richardson number Ri, and (iii) the algebraic stress model. Numerical analysis of the stratification phenomena using the higher-order scheme showed that, in general, the modelling of the buoyancy terms appearing in the turbulence transport equations was the most important key to successful results. When the k−ε model was used, it was pointed out that a dependence on the local Richardson number must be carefully included in the turbulent Prandtl number. In this case, however, the range of applicability was limited to the phenomena observed in the water system in general because the model was constructed and calibrated for water experiments. Overall it was found that the calculated stratification interface rise agreed well with experimental results in water and sodium insofar as the algebraic stress model was utilized. As a conclusion, in predicting the behaviour of the thermal stratification phenomena in liquid metal cooled reactors, the coupled use of the higher-order difference scheme and the algebraic stress model was most appropriate and recommended.     

Analysis of PWR SBO sequences with RCP passive thermal shutdown seals

The operation of recently implanted low-leakage seals after Fukushima has altered the analysis of classical PWR Station Blackout (SBO) sequences , as Seal Loss of Coolant Accident (SLOCA) is no longer one of the dominant factors in the accident progression . An analysis of different management strategies in non-SLOCA sequences has been performed by means of the Integrated Safety Assessment (ISA) methodology using the SCAIS-MAAP model of a 3-Loop PWR Westinghouse design. Through the use of the Damage Domain concept(i.e. the region of the uncertain crew actuation times or physical parameters space where each damage limit is exceeded for each sequence), the times for reaching different damage limits are obtained. Results evidence the positive impact of low-leakage seals, which greatly increase the margin to core uncoveryand reduce core damage frequency. Results also allow concluding that an SBO is dominated, namely by the Auxiliary Feed-Water (AFW) mass flow(turning blind AFW management into an essential procedure), SLOCA (in case the new low-leakage seals fail or they are not present), an excessive AFW mass flow (leading to Turbine-Driven Pump failure) and the DC failure time (losing control valves and the instrumentation).     

Piping specific analysis of stresses due to thermal stratification

Within the framework of the ‘Component Specific Analysis of Mechanical Behaviour’, the nuclear codes and standards, e.g. KTA Standard KTA 3201.2 on Components of the Reactor Coolant Pressure Boundary of Light Water Reactors (Design and Analysis) allow a simplified stress and fatigue analysis for piping by applying the stress index method. The equations supplementary to present KTA Standard given in this context can be used for the consideration of the thermal loads arising from piston flow but not of those from thermal stratification. Thermal stratification occurs e.g. in the surge line of pressurised water reactors during start-up and shut-down processes. This phenomenon is explained and its effects are described. The equations supplementary to the present KTA Standard are derived which enable the consideration of the load case ‘thermal stratification’ in the simplified stress and fatigue analysis. Their applicability is demonstrated by a numeric example.     

Three-dimensional transport coefficient model and prediction-correction numerical method for thermal margin analysis of PWR cores

Combustion Engineering Inc. designs its modern PWR reactor cores using open-core thermal-hydraulic methods where the mass, momentum and energy equations are solved in three dimensions (one axial and two lateral directions). The resultant fluid properties are used to compute the minimum Departure from Nucleate Boiling Ratio (DNBR) which utlimately sets the power capability of the core. The on-line digital monitoring and protection systems require a small fast-running algorithm of the design code. This paper presents two techniques used in the development of the on-line DNB algorithmFirst, a three-dimensional transport coefficient model is introduced to radially group the flow subchannel into channels for the thermal-hydraulic fluid properties calculation. Conservation equations of mass, momentum and energy for these channels are derived using transport coefficients to modify the calculation of the radial transport of enthalpy and momentum.Second, a simplified, non-iterative numerical method, called the prediction-correction method, is applied together with the transport coefficient model to reduce the computer execution time in the determination of fluid properties.Comparison of the algorithm and the design thermal-hydraulic code shows agreement to within 0.65% equivalent power at a 95/95 confidence/probability level for all normal operating conditions of the PWR core. This algorithm accuracy is achieved with 1/800th of the computer processing time of its parent design code.     

Analytical assessment of an inclad defect in a PWR vessel belt line

The French approach to the assessment of the integrity of PWR vessels requires, in particular, that existence of large margins with respect to fast fracture shall be demonstrated for all kinds of defects which can be produced during manufacturing, taken with envelope sizes. The case of defects in the cladding with one tip against the base metal interface raises several difficult problems, mainly on the effect of residual stresses in the cladding, and on the choice of a relevant criterion for the risk of initiating cleavage cracking in the base material. The behaviour of a typical defect has been computed with elastic-plastic analyses and the criteria of the local approach of fracture: the effect of residual stresses is negligible and the margins with respect to fast fracture are much larger than those indicated previously by LEFM computations with plasticity corrections. The values obtained ensure that the integrity of the vessel would not be affected if such defects had occurred during manufacturing.     

The study on a statistical methodology for PWR fuel rod internal pressure evaluation

The most limiting design criteria for high Burnup PWR fuel are known to be rod internal pressure and cladding oxidation. Some fuel vendors have been increasing the design margin of rod internal pressure by increasing fuel rod plenum volume or optimizing fuel pellet grain size. In this study, a sophisticated statistical methodology that employs the response surface method and Monte Carlo simulation has been proposed to increase the design margin of rod internal pressure and subsequently a simplified statistical methodology has been developed to simplify the sophisticated statistical methodology. The simplified statistical methodology utilizes the system moment method combined with a deterministic approach for calculating a maximum variance of rod internal pressure. This simplified statistical methodology may be more efficient in the reload core fuel rod performance analyses than the sophisticated statistical methodology since the former eliminates numerous calculations needed for the evaluation of power history-dependent variances. It is found that this simplified methodology also generates more conservative rod internal pressure than the typical statistical methodology.     

压水堆的信号证实

本文介绍信号证实技术的基本原理及其在压水堆上的应用。     

Fatigue design of nuclear class 1 piping considering thermal stratification

In ASME B&PV Code, Section III, Subsection NB-3600, thermal stratification is not taken into account to determine the peak stress intensity range for fatigue design of nuclear class 1 piping. Therefore, the effects of several parameters such as boundary layer thickness, temperature difference, stratification length, wall thickness, inner diameter and material properties on peak temperature and peak stress intensity due to non-linear temperature distribution of thermal stratification in a pipe cross-section are studied through the numerical parametric study. The results of the parametric study are closely examined and consolidated to introduce an additional term into the equation of ASME so that the modified equation can be used to determine the peak stress intensity range due to all loads including thermal stratification.     

Consequence evaluation for pressurized thermal shock

In order to estimate the risk associated with Pressurized Thermal Shock (PTS), a sample calculation of the core melt frequency and offsite consequences has been performed for Oconee Unit 1, a Babcock and Wilcox pressurized water reactor located in the United States. Core melt frequency was derived from through-wall-crack frequency estimates based on thermal-hydraulic and fracture mechanics analyses performed by Oak Ridge National Laboratory and Pacific Northwest Laboratory. The mode and timing of containment response was estimated from previous risk studies for Oconee Unit 3 and other plants with large dry containments.The core melt frequency was calculated to be 6 × 10⁻⁶ per reactor year for operation at the PTS screening criterion. Operation of redundant and independent containment heat removal systems results in low probability of containment failure. The risk dominant scenario involves overpressure failure of containment due to failure of containment heat removal. Prompt containment failure was assigned a very low probability (10⁻⁴), and hydrogen burn failure was not considered.The central estimate of annual risk was 5 × 10⁻⁷ early fatalities, 2 × 10⁻⁴ latent cancer fatalities and 0.7 person-rem. These values are minimal compared with other severe accident scenarios.Uncertainties and sensitivies to important parameters are discussed. The response of other types of plants is briefly described.     

先进压水堆核电厂主管道结构完整性的研究

以某先进压水堆核电厂主管道为例,对核安全一级管道的结构完整性进行分析评价,并对根据规范设计的管道设计裕量进行了分析。管道结构完整性评价内容包括依据规范对管道强度进行评价、采用解析法求解管道温度场进行热棘轮评价、采用简化雨流法对管道进行疲劳寿命评价。计算结果表明,主管道最小壁厚减少至55 mm能够满足标准规范要求,但安全裕度较小,其中主管道支管位置的疲劳和热棘轮评价结果裕量最小。     

Strategy evaluation for fire spray system on Advanced Passive PWR Severe Accident Management Guideline

The fire spray system (FSS) of the Advanced Passive PWR, as a part of the fire protection system, can provide a non-safety related containment spraying function for severe accident mitigation which is included in the Severe Accident Management Guidelines (SAMG) of the Advanced Passive PWR when dealing with severe accidents. The effectiveness of the FSS is investigated on three effects for severe accident mitigation which are controlling the containment condition, washing out fission product and injecting into the containment through three representative severe accident scenarios analysis with integral accident analysis code since there is no sufficient data support, besides the negative impact is also discussed. Results show that the FSS can be effective for controlling the containment condition, washing out fission product and injecting into the containment, however the effect is limited due to system limitation: the FSS can only cool the containment atmosphere for a short term; the flow rate of FSS cannot fulfill the success criteria given in the PRA report of the Advanced Passive PWR. Meanwhile, the hydrogen concentration and the containment water level should be the long-term monitored because actuating the FSS may cause hydrogen risk in the containment and containment flooding. Despite its limitation and negative impact, the FSS can be effective as an alternative severe accident mitigation measurement for postponing the process of accidents for safety system recovery.     

The effective thermal conductivity of crud and heat transfer from crud-coated PWR fuel

Water-filled crud on the surface of PWR fuel could offer resistance to the flow of heat, which might be expected to cause higher clad temperatures, and probably more fuel failures, than are actually observed. However, there is some evidence from post-irradiation inspection that the crud is penetrated by pores large enough to permit vapour formation, and it is believed these provide a mechanism for ‘wick boiling’ to occur, which modifies, and indeed can under some circumstances actually improve, heat transfer. This phenomenon is investigated using a two-dimensional coupled multi-physics model, accounting for the flow of water, heat and dissolved species within the crud. The fuel thermal performance is characterized in terms of an effective crud thermal conductivity derived from the use of this model, and the non-linear dependence this effective thermal conductivity has on parameters such as crud thickness and pore density is determined.