Probability of crack-induced failure in BWR recirculation piping

The Lawrence Livermore National Laboratory (LLNL) has estimated the probability of double-ended guillotine break (DEGB) in the reactor coolant piping of Mark I boiling water reactor (BWR) plants. Two causes of pipe break are considered: crack growth at welded joints and the seismically-induced failure of component supports. For the former a probabilistic fracture mechanics model is used, for the latter a probabilistic support reliability model. This paper describes a probabilistic model developed to account for effects of intergranular stress corrosion cracking (IGSCC). The IGSCC model, based on experimental and field data compiled from several sources, correlates times to crack initiation and crack growth rates for Types 304 and 316NG stainless steel against material-specific ‘damage parameters’ which consilidate the separate effects of coolant environment (temperature, dissolved oxygen content, level of impurities), stress (including residual stress), and degree of sensitization. Application of this model to actual BWR recirculation piping shows that IGSCC clearly dominates the probability of failure in 304SS piping, mainly due to cracks that initiate within a few years after plant operation has begun. Replacing Type 304 piping with 316NG reduces failure probabilities by several orders of magnitude.     

Analysis of potential waterhammer at the Ignalina NPP using thermal–hydraulic and structural analysis codes

The RBMK (Russian acronym for ‘channeled large power reactor’)-1500 reactors at the Ignalina nuclear power plant (NPP) have a series of check valves in the main circulation circuit (MCC) that serve the coolant distribution in the fuel channels. In the case of a hypothetical guillotine break of pipelines upstream of the group distribution headers (GDH), the check valves and adjusted piping integrity is a key issue for the reactor safety during the rapid closure of check valve. An analysis of the waterhammer effect (i.e. the pressure pulse generated by the valves slamming closed) is needed. The thermal–hydraulic and structural analysis of waterhammer effects following the guillotine break of pipelines at the Ignalina NPP with RBMK-1500 reactors was conducted by employing the RELAP5 and PipePlus codes. Results of the analysis demonstrated that the maximum values of the pressure pulses generated by the check valve closure following the hypothetical accidents remain far below the value of pressure of the hydraulic tests, which are performed at the NPP and the risk of failure of the check valves or associated pipelines is low. Sensitivity analysis of pressure pulse dependencies on calculation time step and check valve closure time was performed. Results of RELAP5 calculations are benchmarked against waterhammer transient data obtained by employing structural mechanics code BOS fluids.     

Opening and extension of circumferential cracks in a pipe subject to dynamic loads

Nuclear power plants are presently designed to withstand instantaneous pipe severance in combination with the maximum seismic loads. The hypothetical combination of these two unlikely events leads to system designs which are very expensive and require dynamic event devices such as pipe whip restraints which have the potential for deleterious interaction with the piping system during normal operations. These present pipe rupture criteria are based on the a priori hypothesis that the instantaneous guillotine pipe break is possible, rather than from a consideration of the manner in which cracks might open or extend in a real piping system. The objective of this study is to help establish the basis for understanding how cracks which might exist in the primary piping of a pressurized water reactor (PWR) would open and extend so that improved criteria can be developed based on this information.One of the regions where loss of pressure boundary integrity must be postulated is the terminal end of the cold leg at the reactor vessel inlet nozzle. This region (including the effects of the reactor vessel and the primary pump) is modelled for analysis with the MARC general purpose finite element program. A circumferential crack, one-half circumference long, is considered to suddenly occur around the outside of the elbow when the pipe is at normal operating pressure. The most severe part of the safe shutdown earthquake (SSE) loading transient is applied simultaneously with the initiation of the crack.The plastic dynamic analysis of the crack opening effects in the discharge leg pipe is performed using the MARC program until the maximum opening occurs. The J-integral plastic crack extension criterion is computed for all times during the transient. The results indicate that none of the cracks will extend significantly and that the opening areas are small fractions of the flow area of the pipe.     

Erratum

Nuclear power plants are presently designed to withstand instantaneous pipe severance in combination with the maximum seismic loads. The hypothetical combination of these two unlikely events leads to system designs which are very expensive and require dynamic event devices such as pipe whip restraints which have the potential for deleterious interaction with the piping system during normal operations. These present pipe rupture criteria are based on the a priori hypothesis that the instantaneous guillotine pipe break is possible, rather than from a consideration of the manner in which cracks might open or extend in a real piping system. The objective of this study is to help establish the basis for understanding how cracks which might exist in the primary piping of a pressurized water reactor (PWR) would open and extend so that improved criteria can be developed based on this information.One of the regions where loss of pressure boundary integrity must be postulated is the terminal end of the cold leg at the reactor vessel inlet nozzle. This region (including the effects of the reactor vessel and the primary pump) is modelled for analysis with the MARC general purpose finite element program. A circumferential crack, one-half circumference long, is considered to suddenly occur around the outside of the elbow when the pipe is at normal operating pressure. The most severe part of the safe shutdown earthquake (SSE) loading transient is applied simultaneously with the initiation of the crack.The plastic dynamic analysis of the crack opening effects in the discharge leg pipe is performed using the MARC program until the maximum opening occurs. The J-integral plastic crack extension criterion is computed for all times during the transient. The results indicate that none of the cracks will extend significantly and that the opening areas are small fractions of the flow area of the pipe.     

Reliability analysis of pipe whip impacts

A probability-based approach is presented as the integration of probabilistic methods and deterministic modelling based on the finite element method. An existing finite element software package was linked to an existing probabilistic package to analyse the complex mechanics that occur during the transient non-linear analysis of impact problems. This methodology is applied to a pipe whip analysis of a group-distribution-header, which results from a guillotine break, and subsequent impact with the adjacent building wall; this is a postulated accident for the Ignalina Nuclear Power Plant RBMK-1500 reactors. The uncertainties of material properties, component geometry data and loads were taken into consideration. The probabilities of failure of the impacted header and of the header support-wall were estimated given uncertainties in material properties, geometrical parameters and loading. The software ProFES was used for the probabilistic analysis and the finite element software NEPTUNE for deterministic structural integrity evaluation. The Monte Carlo Simulation, First Order Reliability method and Response Surface method were used in the probabilistic analysis.     

压水堆主回路失水事故下水锤与管道结构的相互作用分析

传统的水锤分析和管道动力响应计算是分开的,存在一定的缺陷。本文针对核电站主回路假想双端断裂时系统的受力和力矩分析这一问题,对破裂管道分充体和管道的耦合机制,引入描述流体－管道单元的１４个参数和１４个偏微分方程,利用特征线法对水锤和管道结构的相互耦合作用进行了模拟计算。计算得到了更为准确的水锤波和管道的受力和力矩,其波形和数值均与不考虑耦合作用时有所不同。这些计算结果为压水堆核电站的核安全设计和分析     

Application of probabilistic methods to validate NPP pipewhip impact simulations

Piping in nuclear power plants is vital to the proper operation and safety of these facilities. To assure safety in the unlikely event of a pipe break, it is necessary to evaluate the consequences from the resulting whipping pipe on neighboring components and structures. Numerical simulations allow for rapid evaluation of these consequences. Before simulations can be accepted, however, the methodology and computer codes must be validated against experimental results. This paper uses a probabilistic approach to validate pipe whip simulations against limited experimental results. Probabilistic analysis software was developed and coupled to existing deterministic finite element software. An example of a whipping pipe impacting against a reinforced concrete slab was simulated. The described probabilistic approach was used to validate the numerical simulations. The conclusions obtained were that the numerical simulations of whipping pipe impact were validated, even though the numerical results did not exactly agree with experimental results. The chosen points of comparison – namely, time-to-impact and total reaction force – were within the 95% confidence interval.     

Fracture mechanics and full scale pipe break testing for the Department of Energy's new production reactor-heavy water reactor

Oak Ridge National Laboratory (ORNL) has completed a major task for the US Department of Energy (DOE) in the demonstration that the primary piping of the proposed new production reactor-heavy water reactor (NPR-HWR), with its relatively moderate temperature and pressure, should not suffer an instantaneous double-ended guillotine break (DEGB) under design basis loadings and conditions. The growth of possible small pre-existing defects in the piping wall was estimated over a plant life of 60 years. This worst-case flaw was then evaluated using fracture mechanics methods. It was calculated that this worst-case flaw would increase in size by at least 14 times before pipe instability during a safe shutdown earthquake (SSE) would even begin to be possible. The approach to showing the improbability of an instantaneous DEGB for HWR primary piping required a major facility (pipe impact test facility, PITF) to apply all possible design loads, including an equivalent major earthquake (called the SSE earthquake). The facility was designed and built at ORNL in 6 months. The test article was 6.1 m long, 406 mm diameter, 13 mm thick pipe of stainless steel 316LN material that was fabricated to exacting standards and inspections following the nuclear industry standard practices. A flaw was machined and fatigued into the pipe at a tungsten inert gas (TIG) butt weld (ER316L weld wire) as an initial condition. The flaw-crack was sized to be beyond the worst-case flaw that HWR piping could see in 60 years of service—if all leak detection systems and if all crack inspection systems failed to notice the flaw's existence. Starting October 1991, the first test article was subjected to considerable overloadings. The pipe was impacted 104 times at levels equal to and well beyond the SSE loadings. In addition, over 560 000 fatigue cycles and numerous purposeful static overloads were applied in order to extend the flaw to establish the data necessary to confirm fracture mechanics theories, and more importantly, to demonstrate simply that instantaneous DEGB is highly improbable for the relatively moderate energy system.     

Results of reliability test program on light water reactor piping

The Japan Atomic Energy Research Institute has conducted a piping reliability test program to demonstrate the safety and reliability of light water reactor primary piping. In this program, pipe fatigue test, leak-before-break (LBB) verification test and pipe rupture test were carried out to examine the integrity of piping, to verify the LBB and to demonstrate the effectiveness of protective measures against jet impingement and pipe whip loads under a pipe rupture event.In the pipe fatigue test, a procedure to predict the fatigue crack growth was developed, and the integrity of piping during the plant service life was evaluated. In the LBB verification test, the pipe fracture test and the leak rate test were performed to verify the LBB in the primary piping.In the pipe rupture test, the influence of jet impingement on the target disk and the deformation behavior of whipping pipe and restraint were investigated. Using the test results, the jet impingement behavior and the effectiveness of pipe whip restraint were demonstrated.     

Water-ingress analysis for the 200 MWe pebble-bed modular high temperature gas-cooled reactor

Water ingress into the primary circuit is generally recognized as one of the severe accidents with potential hazard to the modular high temperature gas-cooled reactor adopting steam-turbine cycle, which will cause a positive reactivity introduction, as well as the chemical corrosion of graphite fuel elements and reflector structure material. Besides, increase of the primary pressure may result in the opening of the safety valves, consequently leading the release of radioactive isotopes and flammable water gas. The analysis of such a kind of important and particular accident is significant to verify the inherent safety characteristics of the modular HTR plants.Based on the preliminary design of the 200 MWe high temperature gas-cooled reactor pebble-bed modular (HTR-PM), the design basis accident of a double-ended guillotine break of one heating tube and the beyond design basis accident of a large break of the main steam collection plate have been analyzed by using TINTE code, which is a special transient analysis program for high temperature gas-cooled reactors. Some safety relevant concerns, such as the fuel temperature, the primary loop pressure, the graphite corrosion, the water gas releasing amount, as well as the natural convection influence on the condition of failing to close the blower flaps, have been studied in detail. The calculation results indicate that even under some severe hypothetical postulates, the HTR-PM is able to keep the inherent safeties of the modular high temperature gas-cooled reactor and has a relatively good natural plant response, which will not result in environmental radiation hazard.     

Predicting local distributions of erosion–corrosion wear sites for the piping in the nuclear power plant using CFD models

The erosion–corrosion (E/C) wear is an essential degradation mechanism for the piping in the nuclear power plant, which results in the oxide mass loss from the inside of piping, the wall thinning, and even the pipe break. The pipe break induced by the E/C wear may cause costly plant repairs and personal injures. The measurement of pipe wall thickness is a useful tool for the power plant to prevent this incident. In this paper, CFD models are proposed to predict the local distributions of E/C wear sites, which include both the two-phase hydrodynamic model and the E/C models. The impacts of centrifugal and gravitational forces on the liquid droplet behaviors within the piping can be reasonably captured by the two-phase model. Coupled with these calculated flow characteristics, the E/C models can predicted the wear site distributions that show satisfactory agreement with the plant measurements. Therefore, the models proposed herein can assist in the pipe wall monitoring program for the nuclear power plant by way of concentrating the measuring point on the possible sites of severe E/C wear for the piping and reducing the measurement labor works.     

Stratified flow-induced air-ingress accident assessment of the GAMMA code in HTGRs

An air-ingress accident is a major safety issue pertaining to high-temperature gas-cooled reactors. To see the effect of a stratified flow, which is a multi-dimensional phenomenon that occurs in large broken pipes, we perform 1-D and 2-D air-ingress simulations in the guillotine break of the main coaxial pipe of a 600 MWth GT-MHR with the GAs multicomponent mixture transient analysis (GAMMA) code. We used a 2-D fluid volume to build the coaxial inlet pipe, the lower plenum of the reactor core, and the cavity and simplified the other components as 1-D fluid blocks. After the guillotine break of the main coaxial pipe, the air in the reactor cavity flows into the reactor core in four phases: the blow-down phase, the stratified flow phase, the molecular diffusion phase, and the natural convection phase. In the early stage of a broken pipe, the lower plenum region of the reactor is filled with air within 30 s by a density-driven airflow. In a 1-D simulation, the process of filling the lower plenum with air ingressed from a cavity caused by the diffusion process takes 30 min. However, after 30 s, the flow velocity of air ingressed into the broken pipe decreases and the diffusion phase eventually begins. The natural circulation in this scenario starts after more than 360 h for the 1-D simulation but fails to commence after more than 500 h for the 2-D simulation. The belated natural circulation in the 2-D simulation is mainly attributed to the slower diffusion process in the core region. In turn, the slower diffusion occurs because the temperature of the air in the lower plenum is lower in the 2-D simulation than in the 1-D simulation. The maximum core temperature in the 2-D simulation was by 60 °C lower than that in the 1-D simulation.     

Finite element analysis for creep failure of coolant pipe in light water reactor due to local heating under severe accident condition

During severe accident of a light water reactor (LWR), the piping of the reactor cooling system would be damaged when the piping is subjected to high internal pressure and very high temperature, resulted from high temperature gas generated in a reactor core and decay heat released from the deposit of fission products. It is considered that, under such a condition, short-term creep at high temperatures would cause the piping failure. For the evaluation of piping integrity under a severe accident, a method to predict such high temperature short-term creep deformation should be developed, using a creep constitutive equation considering tertiary creep. In this paper, the creep constitutive equation including tertiary creep was applied to nuclear-grade cold-drawn pipe of 316 stainless steel (SUS316), based on the isotropic damage mechanics proposed by Kachanov and Ravotnov. Tensile creep test data for the material of a SUS316 cold-drawn pipe were used to determine the coefficients of the creep constitutive equation. Using the constitutive equation taking account of creep damage, finite element analyses were performed for the local creep deformation of the coolant piping under two types of conditions; uniform temperature (isothermal condition) and temperature gradient of circumferential direction (non-isothermal condition). The analytical results show that the damage variable integrated into the creep constitutive equation can predict the pipe failure in the test performed by Japan Atomic Energy Research Institute, in which failure occurred from the outside of the pipe wall.     

模块式高温气冷堆进水事故后的氢气行为分析

蒸汽发生器两根传热管双端断裂是模块式高温气冷堆（HTR-PM）典型的超设计基准事故,事故可能会导致氢气在反应堆舱室内的聚集,产生爆燃甚至爆轰的风险。本文使用反应堆流体计算程序GASFLOW,模拟了两根传热管断裂后排放到蒸汽发生器舱室以及反应堆舱室内的气体的输运及分布,并利用氢气燃爆分析程序COM3D进行了舱室内的氢气燃烧分析。计算结果表明,两根传热管断裂事故排放的氢气含量很小,舱室内氢气浓度最大不超过0.1%,如此低浓度的氢气不会发生燃烧爆炸。     

基于RELAP5的中国氦冷固态包层真空室外破口瞬态特性分析

利用RELAP5/MOD3.4对中国氦冷固态包层、氦气冷却剂回路和二次侧水冷系统进行建模和系统热工水力安全评价。依据ITER事故分析制定的事故序列,对设计基准真空室外破口进行了瞬态分析,并对比了不同破口位置、面积和停堆方式对第一壁的影响。结果表明：真空室外破口发生在风机的下游较上游危险,且小破口较大破口更危险;若真空室外破口同时包层第一壁破口,也可通过自然循环和辐射换热带走衰变热冷却包层;真空室外破口事故中采用聚变停堆系统的3s停堆方式,可避免第一壁熔化。     

Experimental studies of 4-inch pipe whip test under BWR loca conditions

Pipe whip tests or jet discharge tests have been performed at the Japan Atomic Energy Research Institute, which simulate the instantaneous circumferential guillotine break of primary coolant piping in nuclear power plants. The present paper describes the results of the pipe whip tests using test pipes of 4 inch diameter, under the BWR LOCA conditions, which were performed from 1979 to 1981. The tests were carried out at an initial pressure of about 6.8 MPa and an initial temperature of about 285°C.The test pipe was 114.3 mm (4 in) in diameter, 8.6 mm in thickness and 4500 mm in length. The four pipe whip restraints used in the tests were the U-bar type of 8 mm in diameter and fabricated from Type 304 stainless steel. The experimental parameters were the clearance (30, 50 and 100 mm) and the overhang length (250, 400, 550 and 1000 mm).The main purpose of these tests is to investigate the effects of the clearance and the overhang length on the pipe whip behavior. It has been clarified from the test results that a smaller clearance and a shorter overhang length causes the deformation of the pipe and restraints to be minimized, and the test pipe collapses near the setting point of the restraints with the overhang length of 1000 mm.     

Air ingress phenomena in a depressurization accident of the very-high-temperature reactor

The inherent properties of the very-high-temperature reactor (VHTR) facilitate the design of the VHTR with high degree of passive safe performances, compared to other type of reactors. However, it is still not clear if the VHTR can maintain a passively safe function during the primary-pipe rupture accident, or what would be a design criterion to guarantee the VHTR with the high degree of passively safe performances during the accident. The primary-pipe rupture accident is one of the most common of accidents related to the basic design regarding the VHTR, which has a potential to cause the destruction of the reactor core by oxidizing in-core graphite structures and to release fission products by oxidizing graphite fuel elements. It is a guillotine type rupture of the double coaxial pipe at the nozzle part connecting to the side or bottom of the reactor pressure vessel, which is a peculiar accident for the VHTR. If a primary pipe ruptures, air will be entered into the reactor if there is air in the reactor containment or confinement vessels. This study is to investigate the air ingress phenomena and to develop the passively safe technology for the prevention of air ingress and of graphite corrosion. The present paper describes the influences of a localized natural circulation in parallel channels onto the air ingress process during the primary-pipe rupture accident of the VHTR.     

Acoustics impact on steam dryer induced by main steam line break in a boiling water reactor

In boiling water reactor (BWR) design, significant acoustic pressure loads impact the steam dryer hood as a result of the main steam line break outside containment (MSLB) event. When a main steam line breaks, it is assumed that the pipe instantaneously breaks completely open to the ambient environment (double-ended guillotine break). Due to the huge pressure difference between the inside of the reactor pressure vessel (RPV) and surrounding ambient environment, a shock wave will form at the break point and burst into the surrounding environment. At the same time, an expansion wave will travel upstream through the main steam line to the RPV, which results in a pressure reduction on the outside of the steam dryer hood. This expansion wave will create a substantial pressure difference between the two sides of the steam dryer hood with a resultant high stress on the hood. This differential pressure load is the acoustic load used in the structure design evaluations for this event. A key design basis requirement for the steam dryer is to maintain structural integrity during transient, and accident conditions. Demonstration that the steam dryers meet this design basis requires a calculation of the magnitude of the acoustic load on the steam dryer during a MSLB. In this study, computational fluid dynamics (CFD) is used as an alternate calculation method to investigate the phenomenon of MSLB. Transient simulations with fine time steps were carried out. The results show that CFD is a useful tool to provide additional information on the acoustic load as compared to the traditional methods. From the CFD results, the minimum pressure value and its distribution area at different flow times was identified. Through the modeling, an understanding of the detailed transient flow field, particularly the acoustic pressure field near the dryer hood during the MSLB was achieved.     

Outline of nuclear piping research conducted in Italy

Following an early phase of limited activity at the University of Pisa on small stainless steel pipes containing axial cracks, in 1981 ENEA, the Italian Committee for Research and Development of Nuclear Energy and Alternative Energies, has started a massive research campaign on fracture of carbon and stainless steel piping containing through and part-through cracks loaded either under pressure or in bending. The purpose of the program was to develop a better understanding on pipes fracture behaviour in order to set new design criteria more realistic, yet conservative, than the guillotine break and prepare acceptance criteria for in-service flaws particularly under the growing pressure of IGSCC that has merciless affected worldwide practically any BWR piping system.The analysis of more than 100 tests carried out at CISE research centre, in Milano, on 4 inch, 6 inch, 8 inch and 10 inch pipes has indicated that unstable fracture requires at least 150° through wall crack under ASME maximum design stresses. The leak area, before instability takes place, is always less than 10% of the net cross section area of the pipe. This has led ENEA to consider a 10% break area as a reasonable value to calculate jet forces. Further, it was found that the net section collapse load criterion by far underestimates the actual collapse load and that 360° part-through cracks tend to switch from ductile to brittle failure mode of a pipe loaded in bending.Further work is planned for the next 3 years including high temperature tests, stainless steel weldments and HAZ tests, high compliance tests and eventually burst tests. Besides the ENEA's research program, Ansaldo AMN, the Italian Nuclear Architect Engineer, is developing theoretical studies and codes to treat the problem of pipe fracture.