The effect of tube rupture location on the consequences of multiple steam generator tube rupture event

A multiple steam generator tube rupture (MSGTR) event has never occurred in the commercial operation of nuclear reactors while single steam generator tube rupture (SGTR) events are reported to occur every 2 years. As there has been no occurrence of a MSGTR event, the understanding of transients and consequences of this event is very limited. In this study, a postulated MSGTR event in an advanced power reactor 1400 (APR1400) is analyzed using the thermal-hydraulic system code, MARS1.4. The APR1400 is a two-loop, 3893 MWt, PWR proposed to be built in 2010. The present study aims to understand the effects of rupture location in heat transfer tubes following a MSGTR event. The effects of five tube rupture locations are compared with each other. The comparison shows that the response of APR1400 allows the shortest time for operator action following a tube rupture in the vicinity of the hot-leg side tube sheet and allows the longest time following a tube rupture at the tube top. The MSSV lift time for rupture at the tube-top is evaluated as 24.5% larger than that for rupture at the hot-leg side tube sheet.     

Improvement of steam generator tube failure propagation analysis code LEAP for evaluation of overheating rupture

As a part of safety assessment or design of steam generators of sodium-cooled fast reactors, it is necessary to evaluate the water leak rate under sodium–water reaction accident. The computer code called LEAP-II calculating a design basis water leak rate during long-time event progress including self-wastage, target-wastage, wastage-type failure propagation, water leak detection, and water/steam blowdown was developed for the prototype fast reactor in the past studies. In this study, a numerical analysis method to predict occurrence of overheating tube rupture was constructed and integrated into this code to expand its application range. The newly constructed method consists of the elemental analysis models for temperature distribution formed by a reacting jet, water-side thermal hydraulics, heat transfer at the tube wall, temperature and stress of the tube, and failure of the tube. Applicability of the method was investigated through the numerical analysis of the experiment on water vapor discharging into liquid sodium pool under the actual condition of the steam generator. The numerical analysis demonstrated that the method could provide the appropriately conservative result on the overheating-rupture-type failure propagation.     

Structural integrity evaluation method for overheating rupture of FBR steam generator tube

This paper describes a structural integrity evaluation method for a SG tube of FBR in case of sodium–water reaction and creep rupture tests to obtain the strength of the tube material. In the SG of FBR, if intermediate size of water/steam leak (1–2 kg s⁻¹) would occur from a tube, it could cause overheating rupture of the multiple tubes surrounding the initially failed tube due to generated sodium–water reaction heat. In the ultra-high temperature condition, the creep strength of the material is one of the dominant factors for failure behavior. Accordingly, we tried to apply the creep failure criterion for the overheating rupture of the SG tube. The creep rupture tests have been performed at ultra-high temperature conditions ranging from 1223.2 to 1323.2 K. The test material is ‘Mod .9Cr–1Mo steel’ which is one of the candidate materials for the tubes of the future SG of FBR. The test results have shown that tube rupture depends on the creep strength of the material; hence, instantaneous rupture does not occur even if the stress exceeds the design value of ultimate tensile strength. The test data have been suitably expressed using the Larson–Miller Parameter, and a structural integrity evaluation method based on the sum of the use-fraction associated with the creep damage has been proposed. Based on this method, the structural integrity of the tube in the sodium–water reaction flame has been evaluated. The results show that it is important to detect the initial leak of the tube within a short period and to reduce the steam pressure more rapidly by SG blowdown.     

蒸汽发生器工作过程建模及仿真分析

基于分布参数热工对象的集总参数化动力学模型,对自然循环蒸汽发生器进行了控制体划分并建立了数学模型,并用MATLAB语言和SIMULINK仿真软件对其进行了仿真,文章采用了Runge-Kutta (4,5)求解器,得到不同功率装置运行时,一次侧,二次侧,以及管束的温度分布,并得到一回路给水扰动时,传热量以及冷却剂出口焓的响应曲线.     

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.     

Level 2 analysis to estimate LERF risk from a thermally induced rupture of a steam generator tube defect

This paper outlines the Level 2 portion of a methodology for determining the incremental induced steam generator tube rupture large early release fraction caused by an actual through-wall defect. This defect was responsible for the minor steam generator tube leak that occurred in September 2002 at the Comanche Peak Steam Electric Station Unit 1. In order to quantify the performance of the defect over the operating cycle, a range of defect lengths were input to the PROBFAIL computer code [Kenton, M., 2001. PROBFAIL: A Computer Code for Evaluating the Likelihood of Steam Generator Tube Rupture in Severe Nuclear Power Plant Accidents, CREARE TM-2138], using appropriate boundary conditions derived from MAAP4 [Henry, R., et al., May 1994. MAAP4—Modular Accident Analysis Program for LWR Power Plants, Computer Code Manual, EPRI Research Project 3131-02] runs. From the analysis of the calculated times of burst for each assumed defect length, the minimum through-wall defect length necessary for tube burst to occur prior to hot leg or surge line creep rupture was calculated. The probability that the defect would actually have this length was then estimated by determining the fraction of the cycle for which the defect would be at least that long. The methodology development and implementation relied on MAAP4 runs, which are discussed extensively in connection with their role in: (1) guiding the construction of the accident progression event tree, (2) generating relevant information for probability assignments in the various underlying fault trees and (3) obtaining boundary conditions of pressure and temperature for use in PROBFAIL. The overall increment in LERF due to the existence of the defect was calculated to be approximately 4E−08.     

Potential steam generator tube rupture in the presence of severe accident thermal challenge and tube flaws due to foreign object wear

This study develops a methodology to assess the probability for the degraded PWR steam generator to rupture first in the reactor coolant pressure boundary, under severe accident conditions with counter-current natural circulating high temperature gas in the hot leg and SG tubes. The considered SG tube flaws are caused by foreign object wear, which in recent years has emerged as a major inservice degradation mechanism for the new generation tubing materials. The first step develops the statistical distributions for the flaw frequency, size, and the flaw location with respect to the tube length and the tube's tubesheet position, based on data of hundreds of flaws reported in numerous SG inservice inspection reports. The next step performs thermal-hydraulic analysis using the MELCOR code and recent CFD findings to predict the thermal challenge to the degraded tubes and the tube-to-tube difference in thermal response at the SG entrance. The final step applies the creep rupture models in the Monte Carlo random walk to test the potential for the degraded SG to rupture before the surge line. The mean and range of the SG tube rupture probability can be applied to estimate large early release frequency in probabilistic safety assessment.     

Three-dimensional behaviors of the hydrogen and steam in the APR1400 containment during a hypothetical loss of feed water accident

During a hypothetical severe accident in a nuclear power plant (NPP), hydrogen is generated by an active reaction of the fuel-cladding and the steam in the reactor pressure vessel and released with the steam into the containment. In order to mitigate hydrogen hazards which could possibly occur in the NPP containment, a hydrogen mitigation system (HMS) is usually adopted. The design of the next generation NPP (APR1400) developed in Korea specifies that 26 passive autocatalytic recombiners and 10 igniters should be installed in the containment for a hydrogen mitigation. In this study, an analysis of the hydrogen and steam behavior during a total loss of feed water (LOFW) accident in the APR1400 containment has been conducted by using the computational fluid dynamics (CFD) code GASFLOW. During the accident, a huge amount of hot water, steam, and hydrogen is released into the in-containment refueling water storage tank (IRWST). The current design of the APR1400 includes flap-type openings at the IRWST vents which operate depending on the pressure difference between the inside and outside of the IRWST. It was found from this study that the flaps strongly affect the flow structure of the steam and hydrogen in the containment. The possibilities of a flame acceleration and a transition from deflagration to detonation (DDT) were evaluated by using the Sigma–Lambda criteria. Numerical results indicate that the DDT possibility was heavily reduced in the IRWST compartment by the effects of the flaps during the LOFW accident.     

Evaluation of a steam generator tube rupture accident in an accelerator driven system with lead cooling

A postulated steam generator tube rupture (SGTR) accident in a lead cooled accelerator driven transmuter (ADT) is investigated. The design of the ADT without intermediate loops bears the risk of water/steam blasting into the primary coolant. As a consequence a nuclear power excursion could be triggered by steam ingress into the ADT core which has a significant positive void worth. A thermal coolant–coolant interaction (CCI) might initiate a local core voiding too and additionally could lead to sloshing of the lead pool with mechanical impact of the heavy liquid on structures. The steam formation will also lead to a pressurization of the cover gas. The problems related to an SGTR are identified and investigated with the SIMMER-III accident code.     

CAP1400蒸汽发生器干燥器疏水槽实验研究

采用常温常压的空气和水,通过实验模拟研究了核电站CAP1400机组蒸汽发生器中疏水槽的疏水特性。实验中设计了不同倾角的疏水槽底板(3.5°-5°),并设置了不同的疏水入水孔数(36-136个),通过控制进入疏水槽的疏水流量(13.8-138.1 m3·h-1),研究不同底板倾角以及不同疏水入水孔数对疏水槽的疏水能力的影响。采用摄像仪对不同工况下疏水槽的水位进行记录,通过MATLAB图像处理的方法对疏水槽水位进行识别,提取出水位高度数据用于实验分析。实验结果表明,疏水槽中最大液位高度随流量呈线性升高;疏水槽入水孔数量对于疏水槽内最大液位高度的影响不明显;疏水槽底面倾角对于疏水能力的影响较小。     

压水堆核电厂蒸汽发生器传热管道破裂事故源项的计算分析

当压水堆核电厂发生事故后,带有放射性的核素会通过破损处释放到环境中,从而危害核电厂周边环境及相关人员的安全,因此对事故后释放到环境中的放射性源项分析,对于核电厂的辐射防护具有重要意义。本文根据事故发生的频率以及后果严重程度,选取蒸汽发生器传热管破裂事故(Steam Generator Tube Rupture,SGTR)进行分析。事故分为事故前碘尖峰释放和事故并发碘尖峰释放两种事故工况,建立事故后放射性核素迁移和扩散计算模型,同时使用先进压水堆AP1000参数进行计算验证,并重点关注惰性气体和挥发性核素碘在环境中的放射性活度。计算结果显示:使用文中计算模型计算的放射性源项与设计源项比较一致,在两种工况下,惰性气体的释放活度与设计源项吻合较好,但碘的释放活度有明显差别。     

Stress analysis and fatigue life prediction for a U-bend steam generator tube

An analysis is carried out to determine the stresses in a steam generator tube that failed by fatigue. Using data available for the failed tube and for failures in two similar steam generators, the magnitudes of the alternating and mean stresses produced during operation are estimated. The cause for the early failure is shown to be the high mean stress caused by denting of the tube in the location where it passed through the tube sheet.     

De-agglomeration mechanisms of TiO2 aerosol agglomerates in PWR steam generator tube rupture conditions

A steam generator tube rupture (SGTR) in a pressurized water reactor (PWR) might be a major source of accidental release of radioactive aerosols into the environment during severe accident due to its potential to by-pass the reactor containment. In the ARTIST program, tests were carried out at flow conditions typical to SGTR events to determine the retention of dry aerosol particles inside a steam generator tube. The experiments with TiO₂ agglomerates showed that for high velocities in the range of 100-350 m/s, the average particle size at the outlet of the tube was significantly smaller than at the inlet due to particle de-agglomeration. Earlier, particle de-agglomeration has not been considered significant in nuclear reactor severe accidents. However, the tests in ARTIST program have shown that there is a possibility that TiO₂ aerosol particles de-agglomerate inside a tube and in the expansion zone after the tube exit under SGTR conditions.In this investigation, we measured TiO₂ aerosol de-agglomeration in the tube with very high flow velocities with two different TiO₂ aerosols. The de-agglomeration was determined by measuring the size of the agglomerates at the inlet and outlet of the test section. The test section was composed of tubes with three different lengths, 0.20, 2.0 and 4.0 m, followed by an expansion zone.The main results were: (i) the de-agglomerate process was relatively insensitive to the initial particle size distribution, (ii) the agglomerates were observed to de-agglomerate in all the tubes, and the resulting particle size distributions were similar for both TiO₂ aerosols, (iii) at high flow rates, increasing the gas mass flow rate did not produce further de-agglomeration, and (iv) the agglomerates did not de-agglomerate to primary particles. Instead, after de-agglomeration the particles had a median outer diameter D_c = 1.1 μm. Based on analysis using computational fluid dynamics (CFDs), the de-agglomeration was caused by the turbulent shear stresses due to the fluid velocity difference across the agglomerates in the viscous subrange of turbulence.It has to be noted that the particles used in this investigation were TiO₂ agglomerates, and not prototypical nuclear aerosols with significantly different characteristics. Therefore, the results of this investigation cannot be directly used to determine whether the nuclear aerosol particles may de-agglomerate in SGTR sequences. However, this investigation highlights the possibility of particle de-agglomeration under SGTR conditions, and identifies the mechanism of the de-agglomeration inside the broken tube and when the aerosol is discharged to an open space.     

MELCOR 1.8.4 sensitivity analysis of the severe accident evolution during the APR 1400 LOCA

Unlike most other systems in which the emergency core cooling (ECC) water is injected into the cold-legs, the Advanced Power Reactor (APR) 1400 employs a concept of a direct vessel injection (DVI) to reduce the bypass effects of the ECC water via a break during a design basis LOCA. For this, the DVI piping is designed so that the ECC water taken from an in-containment refueling storage tank (IRWST) directly flows into the reactor pressure vessel (RPV) down-comer. The main objective of this paper is to provide the MELCOR 1.8.4 sensitivity analysis results on the evolution of the severe accidents that can be expected during the APR 1400 LOCA and the insights gained from the analysis. For this purpose, the present sensitivity analysis mainly focuses on: (1) the impact of the foregoing engineering features (i.e., DVI and IRWST) in mitigating a severe core degradation and (2) the APR 1400-specific impacts of different break locations and sizes, and an operation of the containment spray systems on the timings of the key thermal-hydraulic responses, the severe degradation of the core, and the evolution of the core materials. No significant accident management strategy that plays a great role in mitigating a further progression of severe accidents has been taken into account in the present analysis. As a result, the present analysis results can be taken as the technical basis for assessing the effectiveness of a potential severe accident management.     

Seismic analysis of the APR 1400 reactor vessel internals using the model reduction method

To secure reliability of the seismic design of the reactor vessel internals (RVIs) through the finite element analysis, it is important to develop the accurate analysis model which can represent the geometric complexity of the RVIs. However, the seismic analysis requires too large computation cost to solve the complex equations; thus, it needs to reduce the overall size of the analysis model. Here, we apply a model reduction method based on the fixed-interface component mode synthesis (CMS) method to practical RVIs to solve complex numerical problems efficiently. To verify the model reduction method, several cases of the RVIs with different conditions are analyzed for the static and dynamic problems. Finally, the seismic analysis was performed with the suggested reduced model with the CMS method. The time history analysis is performed to extract important seismic responses at the specified locations, and the stress analysis is also performed to identify that the RVIs satisfy the seismic design. In the last part of the paper, an example of the design modification is suggested to reduce the stress intensity at the support locations.     

Overview of steam generator tube degradation and integrity issues

The degradation of steam generator tubes in pressurized water nuclear reactors, and, in particular, the stress corrosion cracking (SCC) of Alloy 600 tubes, continues to be a serious problem. Primary water SCC is commonly observed at the roll transition zone (RTZ), at U-bends and tube denting locations, and occasionally in plugs and sleeves. Outer-diameter SCC (ODSCC) and intergranular attack (IGA) commonly occur near tube support plate (TSP) crevices, near the tube sheet in crevices, or under sludge piles, and occasionally in the free span. A particularly troubling recent trend has been the increasing occurrence of axial and circumferential cracking at the RTZ on both the primary and secondary sides. Outer-diameter stress corrosion cracking in TSP crevices, commonly consisting of segmented axial cracks with interspersed uncracked ligaments, is also becoming more common. Despite recent advances in inservice inspection (ISI) technology, a clear need still exists for quantifying and improving the reliability of ISI methods with respect to the probability of detection of the various types of flaws and their accurate sizing. These improvements are necessary to permit an accurate assessment of the consequences of leaving degraded tubes in service over the next reactor operating cycle. Degradation modes such as circumferential cracking, intergranular attack, and ODSCC at the TSP have affected a large number of tubes. New regulatory guidance is being developed that requires the development and implementation of a steam generator management program that monitors tube condition against accepted performance criteria to ensure that the tubes will perform the required safety function over the next operating cycle. In formulating new guidance for the implementation of alternate repair criteria, the U.S. Nuclear Regulatory Commission is also evaluating the contribution to overall plant risk from severe accidents. Preliminary analyses are being performed for a postulated severe-accident scenario that involves station blackout and loss of primary feedwater, in order to determine the probability of failure for degraded tubes.     

Estimation of temperature-induced reactor coolant system and steam generator tube creep rupture probability under high-pressure severe accident conditions

A severe accident has inherently significant uncertainties due to the complex phenomena and wide range of conditions. Because of its high temperature and pressure, performing experimental validation and practical application are extremely difficult. With these difficulties, there has been few experimental researches performed and there is no plant-specific experimental data. Instead, computer codes have been developed to simulate the accident and have been used conservative assumptions and margins. This study is an effort to reduce the uncertainty in the probabilistic safety assessment and produce a realistic and physical-based failure probability. The methodology was developed and applied to the OPR1000. The creep rupture failure probabilities of reactor coolant system (RCS) components were evaluated under a station blackout severe accident with all powers lost and no recovery of steam generator auxiliary feed-water. The MELCOR 1.8.6 code was used to obtain the plant-specific pressure and temperature history of each part of the RCS and the creep rupture failure times were calculated by the rate-dependent creep rupture model with the plant-specific data.     

Detection of steam generator tube leaks in pressurized water reactors

This report addresses the early detection of small steam generator tube leaks in pressurized water reactors. It discusses the third, and final, year's work on an NRC-funded project examining diagnostic instrumentation in water reactors. The first two years were broad in coverage, concentrating on anticipatory measurements for detection of potential problems in both pressurized- and boiling-water reactors, with recommendations for areas of further study. One of these areas, the early detection of small steam tube leaks in pressurized water reactors (PWRs), formed the basis of study for the last year of the project.Four tasks are addressed in this study of the detection of steam tube leaks.

1. (1) Determination of which physical parameters indicate the onset of steam generator tube leaks.

2. (2) Establishing performance goals for diagnostic instruments which could be used for early detection of steam generator tube leaks.

3. (3) Defining the diagnostic instrumentation and their location which satisfy Items 1 and 2 above.

4. (4) Assessing the need for diagnostic data processing and display.

Parameters are identified, performance goals established, and sensor types and locations are specified in the report, with emphasis on the use of existing instrumentation with a minimum of retrofitting. A simple algorithm is developed which yields the leak rate as a function of known or measurable quantities. The conclusion is that leak rates of less than one-tenth gram per second should be detectable with existing instrumentation.     

Analysis of hydrogen flame acceleration in APR1400 containment by coupling hydrogen distribution and combustion analysis codes

This study was conducted as part of the construction of an integrated system to mechanistically evaluate flame acceleration characteristics in a containment of a nuclear power plant during a severe accident. In the integrated analysis system, multi-dimensional hydrogen distribution and combustion analysis codes are used to consider three-dimensional effects of the hydrogen behaviors. GASFLOW is used for the analysis of a hydrogen distribution in the containment. For the analysis of a hydrogen combustion in the containment, an open-source CFD (computational fluid dynamics) code OpenFOAM is chosen. Data of the hydrogen and steam distributions obtained from a GASFLOW analysis are transferred to the OpenFOAM combustion solver by a conversion and interpolation process between the solvers. The combustion solver imports the transferred data and initializes the containment atmosphere as an initial condition of a hydrogen combustion analysis. The turbulent combustion model used in this study was validated by evaluating the F22 test of the FLAME experiment. The coupled analysis method was applied for the analysis of a hydrogen combustion during a station blackout accident in an APR1400. In addition, the characteristics of the flame acceleration depending on a hydrogen release location are comparatively evaluated.     

An integral loop test and MARS code analysis for a DVI line break LOCA in the APR1400

The transient thermal-hydraulic phenomena of a DVI (Direct Vessel Injection) line break LOCA (Loss-of-Coolant Accident) in pressurized water reactor, APR1400, were investigated. In order to understand the phenomena during the LOCA transient, a reduced-height and reduced-pressure integral loop test facility, the SNUF (Seoul National University Facility), was constructed with scaling down the prototype. For the appropriate test conditions in the experiment with the SNUF, the energy scaling method was suggested with scaling the coolant mass inventory and the thermal power for the reduced-pressure condition. According to the conditions determined by the method, the experimental study was performed with the SNUF. The experimental results showed that the phenomenon of the downcomer seal clearing played a dominant role in the reduction of the system pressure and the recovery of the coolant level in the core. That phenomenon occurred when the steam incoming from cold legs penetrates the coolant in the upper downcomer toward the broken DVI line. The experimental results were compared with the prototype analysis to estimate the energy scaling method, so that the experiment reasonably simulated the phenomena in the prototype. For the analytical investigation, the experiment was simulated with MARS code to validate the calculation capability of the code, especially for the downcomer seal clearing, which showed good agreement with the results of experiment.