Seismic analysis and fracture evaluation of stabilized core shroud

This paper deals with the seismic analysis and fracture evaluation of a stabilized core shroud in a boiling water reactor of nuclear power plant. To study the adequacy of original seismic loadings, the dynamic behaviors of core shrouds with cracks, without cracks and with stabilizers are analyzed. Seismic analysis of a lumped-mass model of reactor internals is then performed to obtain the seismic loadings around various weldments of the repaired core shroud. The interaction between the core internals and this repaired core shroud is thus taken into account in this study. Further, fracture analysis of the stabilized core shroud is performed to obtain the stress intensity factors along the crack front of horizontal welds based on these seismic loadings. The computed results show that the influence of existing cracks in the stabilized core shroud is insignificant on the overall structural integrity.     

Modeling of weld residual stresses in core shroud structures

This paper presents a computational model to predict residual stresses in a girth weld (H4) of a BWR core shroud. The H4 weld is a multi-pass submerged-arc weld that joins two type 304 austenitic stainless steel cylinders. An axisymmetric solid element model was used to characterize the detailed evolution of residual stresses in the H4 weld. In the analysis, a series of advanced weld modeling techniques were used to address some specific welding-related issues, such as material melting/re-melting and history annihilation. In addition, a 3-D shell element analysis was performed to quantify specimen removal effects on residual stress measurements based on a sub-structural specimen from a core shroud. The predicted residual stresses in the H4 weld were used as the crack driving force for the subsequent analysis of stress corrosion cracking in the H4 weld. The crack growth behavior was investigated using an advanced finite element alternating method (FEAM). Stress intensity factors were calculated for both axisymmetric circumferential (360°) and circumferential surface cracks. The analysis results obtained from these studies shed light on the residual stress characteristics in core shroud weldments and the effects of residual stresses on stress corrosion cracking behavior.     

Fracture mechanics investigation regarding the effects of cracks on the structural integrity of a BWR-core shroud

As a consequence of core shroud intergranular stress corrosion cracking (IGSCC) detected in the course of inservice inspections, a fracture mechanics analysis was carried out to evaluate the effects of postulated cracks on the structural integrity. In this study, critical crack sizes and crack growth were calculated. Due to the comparatively low stress acting on the core shroud during normal operation, the residual stresses in the welds make up the major proportion of the tensile stresses responsible for IGSCC. In order to consider residual stresses of the lower core support ring welds, a finite element analysis was performed at MPA Stuttgart using the FE-code ANSYS. The crack growth computed on the basis of USNRC crack growth rates da/dt demonstrated that crack growth in depth direction increases quickly at first, then retards and finally comes almost to a standstill. The cause of this ‘quasi-standstill’ is the residual stress pattern across the wall, being characterized by tensile stresses in the outer areas of the wall and compressive stresses in the middle of the wall. Crack growth in circumferential direction remains more or less constant after a slow initial phase. As the calculation of stress intensity factors K_I of surface flaws under normal conditions demonstrated, a ‘lower bound’ fracture toughness value is only exceeded in the case of very long and deep surface flaws. It can be inferred from crack growth calculations that under the assumption of intergranular stress corrosion cracking, the occurrence of such deep and at the same time long flaws is unlikely, regardless of the initial crack length. Irrespective of the above, the calculated critical throughwall crack lengths, which were determined using a ‘lower bound’ fracture toughness value, demonstrated that even long throughwall cracks will not affect the component’s integrity under full load. Moreover, it can be concluded from the studies of crack growth that—assuming intergranular stress corrosion cracking—a sufficiently long period will elapse before a crack which has just been initiated reaches a relevant size. Therefore, it can be stated that these cracks will likely be detected during periodic inservice inspections.     

Effects of thermal load and cooling condition on weld residual stress in a core shroud with numerical simulation

Stress corrosion cracking (SCC) in the heat affected zone is the primary damage form due to weld residual stress, corrosion and neutron irradiation environment in the core shroud of a boiling water reactor. The distribution of weld residual stress around a weld is necessary to be clarified to evaluate the structural integrity of core shroud for SCC. Moreover, studying the effects of welding parameters on residual stress on reducing the residual stress is very important to suppress the initiation and propagation of SCC.In this paper, we used a finite element method (FEM) to clarify the distribution of weld residual stress around the sixth horizontal weld (H6a) between the lower ring and the cylinder in the core shroud. The simulation results of axial stress were consistent with the experimental results at the inside and outside surfaces of the core shroud, respectively. The effects of thermal loads and cooling conditions were also investigated with the same model. We simulated the welding progress with water cooling on the inside and outside surfaces of the core shroud in order to study the influence of cooling conditions on the residual axial stress around the weld. The simulation results indicated that water cooling decreased the residual axial stress at the same side due to changing the temperature-affected fields. Moreover, with fixing the peak temperatures of weld passes, the simulation results of the distribution of residual axial stress by the thermal loads with different heating time were compared. The simulation results suggested that the heating time was expected to be longer and the heat flux to be smaller for reaching the small tension residual axial stress or even compression stress around the H6a weld.     

Probabilistic analysis of inelastic structures

Probabilistic methods for the analysis of linear elastic structures, although sometimes unsatisfactory for the purpose of reliability assessment, are well stated and codified. By contrast, when the beneficial contribution of the redistribution of stresses has to be investigated, numerical procedures of analysis are involved and the randomness of the structural response is generally investigated by simulation procedures as no probabilistic method is applicable.However, an irrational use of simulation techniques does not produce probabilistic results, but only useless numerical results. These procedures are criticized and the adoption of “hybrid approaches” is suggested as a rationalization.Special emphasis is devoted to the dynamic analysis of inelastic structures; the problem of establishing adequate probabilistic failure criteria is also discussed.     

Remaining life prediction of the core shroud due to stress corrosion cracking failure in BWRs using numerical simulations

Stress corrosion cracking (SCC) of the welded joints in a reactor core shroud is the primary result of the residual stresses caused by welding, corrosion and neutron irradiation in a boiling water reactor (BWR). Therefore, the evaluation of SCC propagation is important for the safe maintenance of the core shroud. This paper attempts to predict the remaining life of the core shroud due to SCC failures in BWR conditions via SCC propagation time calculations. First, a two-dimensional finite element method model containing H6a girth weld in the core shroud was constructed, and the weld processing was simulated to determine the weld's residual stress distribution. Second, using a basic weld residual stress field, the SCC propagation was simulated using a node release option and the stress redistribution was calculated. Combined with the J-integral method, the stress intensity factors were calculated at depths of 2, 3, 4, 8, 12, 16, 19, 22, 25 and 30 mm in the crack setting inside the core shroud; then, the SCC propagation rates were determined using the relation between the SCC propagation rate and the stress intensity factor. The calculations show that the core shroud could safely remain in service after 9.29 years even when a 1-mm-deep SCC has been detected.     

Probabilistic methods in plastic structural analysis

The redistribution of stresses in ductile structures, although beneficial from the safety viewpoint, introduces another source of uncertainty, which requires specific methods when the probabilistic approach to reliability evaluation is followed.Practicable procedures have been developed for structures that satisfy the classical assumptions of plastic limit analysis. In particular, two theorems that allow to find rigorous upper and lower bouds on the probability of full plastic collapse under given loads, are presented. Other methods for probabilistic limit analysis are also indicated, including in particular a specifically developed parametric simulation procedure.The last part of the paper is devoted to the reliability analysis of plastic structures subject to loads varying (slowly) in time.It is recalled first that probabilistic limit analysis can be easily extended to the shakedown—incremental collapse problem, provided the loads vary within a finite domain: however, the significance of such an approach for stochastically varying loads is questioned. In fact, as time increases, the probability also increases that the loads cross any given threshold. Therefore, it is more appropriate to speak of “plastic adaption” rather than “shakedown”, and to focus the attention on the probability of reaching, in any given time interval, a certain permanent deformation. Again, only approximate solutions (in the form of upper and lower bounds) can be found to this question, but this appears to be a more rational and promising approach to the problem.     

Core shroud replacement of Fukushima-Daiichi Unit #3

The core shroud replacement of a boiling water reactor (BWR) was successfully completed at Fukushima-Daiichi Unit #3 (1F3) of the Tokyo Electric Power Company (TEPCO) in Japan. The core shroud and other core internal components made of type 304 stainless steel (SS) were replaced with the ones made of low carbon type 316L SS to improve Intergranular Stress Corrosion Cracking (IGSCC) resistance. This project was the first application of the replacement procedure developed for the welded core shroud, and employed various advanced technologies. The outline of the core shroud replacement project and applied technologies are discussed in this paper.     

Probabilistic models for fuel rod puncturing in an LWR core during a LOCA

The number of fuel rods which puncture during an LWR loss-of-coolant accident (LOCA) must be estimated as part of the plant radioactivity release analysis. Due to the great number of fuel rods in the core and the great number of contributing parameters, many of them associated with wide uncertainty and/or truly random variability limits, probabilistic methods are well applicable. A succession of computer models developed for this purpose is described together with applications to a WWER-440 PWR. Deterministic models are shown to be seriously inadequate and even misleading under certain circumstances. A simple analytical probabilistic model appears to be suitable for many applications. Monte Carlo techniques allow the development of such sophisticated models that errors in the input data presently available for practical problems probably become dominant in the residual uncertainty of the core-wide fuel rod puncture analysis.     

Pressure wave propagation in an LWR downcomer and resultant load on the core shroud due to rupture of a primary coolant loop

Pressure differences and the resultant dynamic load act on the core shroud when pressure waves propagate in the downcomer of a light water reactor (LWR) pressure vessel after rupture of the primary pipe has occurred. An equivalent geometry, i.e. a diverging duct is used to solve by Euler and wave equation for acceleration and velocity of the fluid behind the wave front, that the two-dimensional, time-dependent pressure distribution, induced by the wave propagation, can be calculated. The assumptions lead to an approximate but conservative value of the resultant core shroud load.     

Probabilistic methodology for turbine missile risk analysis

A methodology has been developed for estimation of the probabilities of turbine-generated missile damage to nuclear power plant structures and systems. Mathematical models of the missile generation, transport, and impact events have been developed and sequenced to form an integrated turbine missile simulation methodology. Probabilistic Monte Carlo techniques are used to estimate the plant impact and damage probabilities. The methodology has been coded in the TURMIS computer code to facilitate numerical analysis and plant-specific turbine missile probability assessments. Sensitivity analyses have been performed on both the individual models and the integrated methodology, and probabilities have been estimated for a hypothetical nuclear power plant case study.     

Probabilistic seismic hazard analysis for a site

Traditionally, the seismic design basis ground motion has been specified by normalised response spectral shapes and peak ground acceleration (PGA). The mean recurrence interval (MRI) used to be computed for PGA only. The present work develops uniform hazard response spectra, i.e., spectra having the same MRI at all frequencies for Tarapur Atomic Power Station Site. Sensitivity of the results to the changes in various parameters has also been presented. These results determine the seismic hazard at the given site and the associated uncertainties.     

安全壳内置换料水箱子系统设计改进的PSA评价

根据一级概率安全分析(Probabilistic Safety Analysis,PSA)的结果,安全壳内置换料水箱(In-containment Refueling Water Storage Tank,IRWST)子系统的初始设计导致安注管线破裂(Safety Injection Line Break,SI-LB)始发事件对堆芯损坏频率(Core Damage Frequency,CDF)有较大的贡献。本文提出了IRWST子系统的设计改进方案,将IRWST水箱内的滤网由两个(A/B)增加为三个(A/B/C),并通过管线实现滤网之间的相互连接。通过重新构建故障树对改进后的IRWST子系统进行建模分析,并对相应的事件树以及一级PSA模型进行详细的定量化计算。结果表明,IRWST子系统这一改进能够显著降低堆芯损坏风险。IRWST子系统的改进将SI-LB始发事件的CDF降低了53.5%,将整个一级PSA的CDF降低了21.5%。     

Probabilistic fire analysis capabilities, applications and weak points

Fires occur in nuclear power plants with a relatively high frequency, and can cause multiple and simultaneous failures of redundant or diverse plant equipment or systems. The development of a method of probabilistic fire risk analysis, although subject to more uncertainties than internal events, can provide insights complementary to those provided by deterministic analysis. In many cases, it can also provide a more realistic point of view of the risk involved.     

Reactivity space analysis of PWR core depletion

Reactivity space analysis of PWR core depletion is used to investigate solutions to the core burnup and fuel utilization maximization problems. Results of few-region analysis of the SEQUOYAH PWR with the EPRI nodal code simulate-e show that there is an order of magnitude difference in the effect of fuel arrangement and reactivity control in the optimal solutions. It is concluded that emphasis in the core reload design should be given to achieving the optimal fuel arrangement. Furthermore, it is shown that the constant power, Haling depletion strategy is an effective method of isolating the arrangement problem from reactivity control considerations during the core design process.     

Probabilistic seismic risk analysis of CANDU containment structure for near-fault earthquakes

In this study, the seismic risk of a CANDU (CANada Deuterium Uranium) containment structure is estimated by performing the nonlinear seismic analysis for the near-fault earthquakes. Nonlinear seismic analysis is more effective to consider the distinct nonlinear behavior of concrete structures subjected to the near-fault ground motion which has high input energy. In Korea, the seismic fragility analysis has been performed by using the design analysis results which were obtained from a linear elastic analysis.The lumped mass model of the containment structure was used for a nonlinear dynamic time history analysis. The tri-linear skeleton curve was used for the nonlinear behavior of the prestressed concrete containment structure. In order to estimate the inelastic nonlinear response of the containment, the maximum point-oriented model was used for the hysteretic rule of the shear deformation.For the nonlinear seismic analyses, 30 set of real near-fault earthquake records were used as the input motion. The seismic fragility and risk of the containment for the near-fault ground motions are compared with those from the results based on the conventional method.     

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。     

Influence of thermal properties of zirconia shroud on analysis of PHEBUS FPT0 bundle degradation test with ICARE2 code

In the FPT0 test of the PHEBUS/FP program, it was observed that the fraction of liquefied UO₂ reached 50%, which is much larger than the expected maximum value of 20%. Most of the post-test analyses with various computer codes underpredicted the bundle temperature during a late phase and could not reproduce such a large core degradation. In most of the previous analyses, the shroud thermal conductivity evaluated based on the Pears' ZrO₂ specific heat data and the thermal diffusivity measured by JAERI was used. However, recent thermal property data books adopt a lower specific heat than measured by Coughlin and King's at high temperature. The present analyses with ICARE2 showed that the FPT0 bundle behavior could be mostly reproduced by using the shroud thermal conductivity based on Coughlin and King's. If the present calculation is assumed to be correct enough, the shroud thermal conductivity at high temperature could be smaller than the current evaluation based on the Pears' data. Since the shroud thermal conductivity has thus a strong effect on the bundle behavior, further measurement and evaluation of the thermal properties of the shroud are highly recommended.     

BEAVRS堆芯keff的敏感性与不确定性分析

核数据不确定性分析影响着反应堆安全,在反应堆堆芯物理计算过程中具有重要意义。利用SCALE6.1程序包中KENO模块建立反应堆模拟评估和验证基准BEAVRS（Benchmark for Evaluation and Validation of Reactor Simulations）第一循环热态零功率堆芯物理模型,采用TSUNAMI-3D模块开展k_eff的敏感性与不确定性分析,分析了不同燃料富集度、不同温度对k_eff敏感性与不确定性的影响。结果表明:核数据不确定性导致BEAVRS模型的k_eff总的不确定性为0.501 6%;²³⁵U的平均裂变中子数敏感性导致k_eff的敏感性系数最大（0.926 58）;对k_eff不确定性贡献最大的是²³⁸U（n,γ）反应截面,为0.298 14%;在燃料富集度降低、温度上升时,²³⁸U（n,γ）反应截面不确定性会导致k_eff的不确定性增大。因此,在开展反应堆...     

核磁共振岩心实验分析的标准样品研究

按照国家一级标准的规范,制作了核磁共振（NMR）岩心实验分析的系列化标准样品一流体标样、离散体系标样和固体陶瓷标样,通过对这些标样以及天然岩心的实验考察,将其应用于大庆、新疆、大港等油田的NMR岩心实验分析。结果表明,由于流体标样、固体陶瓷标样和天然岩心的弛豫机理不同,其对岩心NMR孔隙度的标定效果不同。流体标样适合于贝瑞砂岩、陆相沉积分选好、泥质含量低的砂岩NMR孔隙度的标定;固体陶瓷标样适合于分选中等的泥质砂岩NMR孔隙度的标定;对粘土含量高的砂岩、含顺磁性物质的砂岩和砾岩等复杂岩性岩样,用流体标样和固体陶瓷标样标定都得不到准确的NMR孔隙度值。为此,建议选用本地区有代表性的天然岩心作为标样标定其孔隙度,或者研究该类岩石内部磁场梯度分布,得到经内部磁场梯度校正后的NMR孔隙度。