Nuclear Power Plant Piping Research Articles

Fracture Toughness of Z3CN20.09M Cast Stainless Steel with Long-Term Thermal Aging

Accelerated thermal aging tests were performed at 400 °C for nearly 18,000 h on Z3CN20.09M cast stainless steel which was used for primary coolant pipes of nuclear power plants. A series of Charpy impact tests were conducted on Z3CN20.09M after different long-term thermal aging time. The test results indicated that the Charpy impact energy of Z3CN20.09M cast stainless steel decreased rapidly at an early stage and then almost saturated after thermal aging of 10,000 h. Furthermore, J-resistance curves were measured for CT specimens of longitudinal and circumferential pipe orientations. It showed that there was no obvious difference in the fracture characteristics of Z3CN20.09M in different sampling directions. In addition, the observed stretch zone width (SZW) revealed that the value of initiation fracture toughness J SZW was significantly lower than that of fracture toughness J IC, indicating a low actual crack initiation energy due to long-term thermal aging.

Data-Driven Prognostics for Major Piping in Nuclear Power Plants

As operation period of Nuclear Power Plants (NPPs) is getting longer, necessity of reflecting ageing effect is increasing. Especially, when it comes to the piping in NPPs such as reactor coolant system piping or steam generatortubes, it is vulnerable to stress corrosion crack (SCC) or wear due to the fluid with high temperature, high pressure and radiation. Accidents related to such cases have been reported. Since ruptures of the piping can result in severe accidents, it is important to predict and prevent them in advance. Current NPPs ageing management is performed with the physical model based on generic experimental data, which cannot properly consider each NPPs’ different operation environment or history. Prognostics using plant specific data can compensate this limit of ageing management using the physical model. Recently, as usable data of NPPs is increasing with the development of instrumentation technology, applicability of prognostics for NPPs has been increased. Therefore, this paper suggests some prognostics methods such as GPM (General Path Model), MCMC (Markov Chain Monte Carlo) and Particle filter that can consider ageing degradation for the major piping in NPPs. It is expected that prognostics results can be used in Probabilistic Safety Assessment (PSA) considering current or future ageing degradation.

Simulation of water hammer phenomena using the system code ATHLET

Abstract Water Hammer Phenomena can endanger the integrity of structures leading to a possible failure of pipes in nuclear power plants as well as in many industrial applications. These phenomena can arise in nuclear power plants in the course of transients and accidents induced by the start-up of auxiliary feed water systems or emergency core cooling systems in combination with rapid acting valves and pumps. To contribute to further development and validation of the code ATHLET (Analysis of Thermalhydraulics of Leaks and Transients), an experiment performed in the test facility Pilot Plant Pipework (PPP) at Fraunhofer UMSICHT is simulated using the code version ATHLET 3.0A.

Pitting corrosion of thermally aged cast duplex stainless steel for primary coolant pipes of nuclear power plants

ABSTRACTThe pitting corrosion of thermally aged Z3CN20.09M cast duplex stainless steel (CDSS) was investigated using potentiodynamic polarisation and electrochemical impedance spectroscopy. Both pitting potential and charge transfer resistance values of thermally aged specimens became less positive with aging time, while passive current density increased, indicating that thermal aging causes degradation in pitting corrosion resistance. The pitting potential of thermally aged specimens after annealed is higher than that of thermally aged specimens revealing that the pitting corrosion resistance of annealed specimens has been improved. The α’ phase precipitation during thermal aging in ferrite brings the Cr-enriched and Cr-depleted zone microstructures which mainly induced the degradation of pitting corrosion resistance. The microsegregation of Cr element or Cr enriched and Cr-depleted zones will be alleviated or even removed due to the Cr-rich α’ phases dissolution during the process of annealing. So, the pitting corrosion resistance of annealed specimens is improved.

Application of Metallographic Methods for Thermal Ageing Evaluation on Samples from Primary Piping of NPP

During a long-term operation of nuclear power plants (NPP), the changes of structural material properties occur. To ensure the safe and reliable operation, it is necessary to monitor and evaluate these changes mainly on components from primary circuit of NPPs. One of the dominant ageing mechanisms of NPP components besides the radiation embrittlement and the fatigue loads is the thermal ageing. The thermal ageing is the temperature, material and time dependent degradation mechanisms due to long-term exposure at the operating temperature of 570 K.This paper describes the project for thermal ageing monitoring at primary piping in NPP Bohunice Unit 3. There are summarized the results obtained from evaluation of original primary piping material.

Thickness reconstruction of nuclear power plant pipes with flow-accelerated corrosion damage using laser ultrasonic wavenumber imaging

The pipeline is an essential part of nuclear and other power plants. Designed to operate for decades, the pipeline is susceptible to multiple types of damage due to working in a high pressure, high temperature, and highly corrosive environment. Therefore, it requires regular inspection. The issue of many current nondestructive testing systems for pipelines is the use of a one-dimensional scan mechanism along the longitudinal direction of the pipe, which is incompatible with the two-dimensional curvature of elbow pipes, yet it is at these places in which the flow changes directions inside the elbow pipes that a common type of damage occurs: flow-accelerated corrosion damage. In this article, flow-accelerated corrosion damage is visualized using the Ultrasonic Wavenumber Imaging algorithm, which maps the dominant local spectroscopic wavenumber of the wavefield at a particular mode and a particular frequency. In previous studies, the mode and frequency were chosen empirically. In this article, a novel theoretical study is presented to optimize the two input parameters of Ultrasonic Wavenumber Imaging based on a sensitivity analysis of the spectroscopic wavenumber with respect to the change in thickness of the structural base. Additionally, an inverse method is proposed to estimate the thickness of the structure based on the Ultrasonic Wavenumber Imaging map.

Development of a LDIE Prediction Theory in the Condition of Magnetite Formation on Secondary Side Piping in Nuclear Power Plants

It has been thought that wall thinning on the secondary side piping in nu-clear power plants is mostly caused by Flow-Accelerated Corrosion (FAC). Recently, it has been seen that wall thinning on the secondary side piping carrying two-phase flow is caused by not only FAC but also Liquid Droplet Impingement Erosion (LDIE). Moreover, it turns out that LDIE in nuclear power plants does not result from a single degradation mechanism but also from the simultaneous happenings of LDIE and FAC. This paper presents a comparison of the mass loss rate of the tested materials between carbon steel (A106 B) and low alloy steel (A335 P22) resulting from degradation effect. An experimental facility was set up to develop a prediction model for clarifying multiple degradation mechanisms that occur together. The experimental facility allows examining liquid droplet impingement erosion in the same conditions as the secondary side piping in nuclear power plants by generating the magnetite on the surface of the test materials. The magnetite is formed by controlling the water chemistry and the temperature of fluid inside the facility. In the initial stage of the experiments, the mass loss rate of A106 B was greater than that of A335 P22. However, after a certain period of time, the mass loss rate of A335 P22 became greater than that of A106 B. It is presumed that the results are caused by the different yield strengths of the test materials and the different degrees of buffer action of the magnetite deposited on their surfaces. The layer of magnetite on the surface of A106 is thicker than that of A335 P22, due to the different amount of chrome content. In nuclear power plants, carbon steel piping having experienced wall thinning degradation is generally replaced with low-alloy steel piping. However, the materials of pipes carrying two-phase flow should be selected considering their susceptibility to LDIE.

Effect of Grain Size on Corrosion Fatigue Behaviors of Primary Coolant Pipes of Nuclear Power Plants

Corrosion fatigue is one of the most important failure modes of primary coolant pipe used in nuclear power plant, due to the thermal cyclic stress caused by a variety of start-up and shutdown as well as transients when severed in the high temperature and pressure water environment. And 316LN stainless steel is one of the main materials used in pressurized water reactor in nuclear power plant. The mechanisms of fatigue crack initiation and propagation were elucidated by investigating the composition properties of oxide films formed in simulated service environment. The effects of grain size on the fatigue life and crack initiation mechanism of 316LN stainless steel in 320°C water environment were also investigated. The results indicated that the specimens with fine grains (30μm) own the longest fatigue lives than those with intermediate (80μm) and coarse grains (210μm). The fatigue stress amplitudes of the specimens increased with the grain refinement, although the fatigue lives of the specimens with intermediate and coarse grains were close. Fatigue cracks were often initiated at the persistent slip bands on the surfaces of the specimens with intermediate and coarse grains, while it was not easy to be initiated on the fine grain specimens. Finally, combined with the experimental results and the actual production of industry, a suggestion for the production of the primary coolant pipe is put forward.

発電プラントの健全性確保に求められる欠陥検出確率の検討（損傷確率に及ぼす欠陥検出確率の影響評価）

This paper discuss about suitable damage size that should be detected by ultrasonic testing (UT) for nuclear power plant piping when consider about propagation of IGSCC. The power plant piping loop is inspected using UT as an in-service inspection (ISI). The reliability of UT result is affected by examination personnel. In this paper, the reliability of UT is defined simplified POD (Probability of detection) curve. The curve is defined by only three parameters, minimum detectable crack depth (a0), most detectable crack depth (a1) and maximum POD(m). The reliability of plant piping with different diameters are estimated by using this POD curve. The Failure probabilities were calculated with Monte Carlo approach. In case of low carbon stainless steel with IGSCC, a1 and m are less than 6 mm and 0.96 respectively, the piping loop failure probability is under 1x10-7. In another case of sensitized stainless steel with IGSCC, a1 and m are less than 6 mm and 0.96, the piping loop failure probability is under 1x10-3. It was found that the parameters of a1 and m do not affect plant reliability when a1 is between 3 to 6mm, and m is between 1.00 to 0.96. This result can be used to decide the requirement for examination personnel reliability.

Ultrasonic Inspection of Electrofusion Joints of Large Polyethylene Pipes in Nuclear Power Plants

High-density polyethylene (HDPE) pipe has many advantages such as good flexibility, corrosion resistance, and long service life. It has been introduced into nuclear power plants for transportation of cooling water in U.S. and Europe. Recently, four HDPE pipelines (PE4710) were used in essential cooling water system with operating pressure of 0.6 MPa and operating temperature of no more than 60 °C in a newly established AP1000 nuclear power plant in Zhejiang, China. The outside diameter and thickness are 30 in. and 3.3 in., respectively, which are much larger and thicker than traditional HDPE pipe for natural gas. This brought forward a challenge for nondestructive testing (NDT) and safety assessment of such pipes. In this paper, a solution for inspecting electrofusion (EF) joints of thick-walled HDPE pipes is presented, and the results of an on-site inspection of the nuclear power plant are revealed. To expand the thickness up-limit of previously developed ultrasonic-phased array instrument, an optimization method was proposed by calculating weighing effects of different testing parameters and introducing the concept of overall performance according to practical requirement, by comprehensively considering sensitivity, penetration, signal-to-noise ratio (SNR), resolution, and accuracy. Typical defects were found in field inspection. The result shows that the presented technique is capable of inspecting EF joints for connecting large-size HDPE pipes used in nuclear power plants.

Characterization of solid particles sampled from condensates in boiling water reactor

The growth, activation and deposition of corrosion products are the primary sources of radiation buildup on the surface of out-of-core piping in nuclear power plants. The buildup of radiation can have negative effects on the performance of the facility and cause harm to staff during maintenance outages for refueling. This paper reports on the crystalline and amorphous structures of corrosion products sampled in the boiling water reactors in nuclear power plants of Kuo-Sheng and identified using an acid dissolving technique. X-ray diffraction, scanning electron microprobe and inductively coupled plasma-atomic emission spectroscopy were used to analyze the samples. The results indicate that the quantity of amorphous iron oxide at inlet of the condensate demineralizer in Unit 2 is higher than that in Unit 1. The proportion of crystalline to amorphous corrosion products can affect the efficiency of removal. Thus, these results can be used to explain the difference in removal efficiency of condensate demineralizers in different units. Moreover, the iron oxide structures with various properties were observed in different operational periods. It is probable that the higher proportion of amorphous structures with a smaller particle size would reduce efficiency in the removal of condensate demineralization in Unit 2.

Development of Wall-Thinning Evaluation Procedure for Nuclear Power Plant Piping—Part 1: Quantification of Thickness Measurement Deviation

Pipe wall thinning by flow-accelerated corrosion and various types of erosion is a significant and costly damage phenomenon in secondary piping systems of nuclear power plants (NPPs). Most NPPs have management programs to ensure pipe integrity due to wall thinning that includes periodic measurements for pipe wall thicknesses using nondestructive evaluation techniques. Numerous measurements using ultrasonic tests (UTs; one of the nondestructive evaluation technologies) have been performed during scheduled outages in NPPs. Using the thickness measurement data, wall thinning rates of each component are determined conservatively according to several evaluation methods developed by the United States Electric Power Research Institute. However, little is known about the conservativeness or reliability of the evaluation methods because of a lack of understanding of the measurement error. In this study, quantitative models for UT thickness measurement deviations of nuclear pipes and fittings were developed as the first step for establishing an optimized thinning evaluation procedure considering measurement error. In order to understand the characteristics of UT thickness measurement errors of nuclear pipes and fittings, round robin test results, which were obtained by previous researchers under laboratory conditions, were analyzed. Then, based on a large dataset of actual plant data from four NPPs, a quantitative model for UT thickness measurement deviation is proposed for plant conditions.

A Numerical Study on Flow-Accelerated Corrosion in Two Adjacent Elbows

Flow-Accelerated Corrosion (FAC) is a well-known degradation mechanism that attacks the secondary piping in nuclear power plants. Since the Surry Unit 2 event in 1986, most nuclear power plants have implemented management programs to deal with damages in carbon and low-alloy steel piping. Despite the utmost efforts, damage induced by FAC still occurs in power plants around the world. In order to predict FAC wear, some computer programs were developed such as CHECWORKS, CICERO, and COMSY. Various data need to be input to these programs; the chemical composition of secondary piping, flow operating conditions and piping geometries. CHECWORKS, developed by the Electric Power Research Institute (EPRI), uses a geometry code to calculate geometry effects. Such a relatively simple geometry code is limited in acquiring the accuracy of FAC prediction. Recently, EPRI revisited the geometry code with the intention of updating it. In this study, numerical simulations were performed for two adjacent elbows and the results were analysed in terms of the proximity effect between the two adjacent elbows.

Tearing resistance of heterogeneous interface region of a dissimilar metal weld characterised with sub-sized single edge bend specimens

Determination of tearing resistance of heterogeneous dissimilar metal welds (DMW) that are critical structures of pipes in nuclear power plants (NPPs) is important in assessment of the structural integrity. Currently, tearing resistance of heterogeneous materials is determined by standards developed for homogeneous materials. The fracture behaviour of heterogeneous materials differs from fracture behaviour of homogeneous materials. Thus standards developed for homogeneous materials are not necessary directly applicable to heterogeneous materials. To develop a standard for heterogeneous materials the current level of knowledge of fracture behaviour in heterogeneous specimens needs to be increased. To enhance the knowledge of fracture in heterogeneous materials tearing resistance of 10×10×55 mm3 and 10×20×100 mm3 single edge bend (SE(B)) specimens extracted from a NPP ferrite-austenite DMW were measured. The specimens were extracted at three different crack locations in a region close to the fusion line between ferritic steel and weld metal of the DMW. As tearing resistance curves of the two specimen geometries are compared toughness and scatter varies, which can be explained by differing crack location with respect to the fusion line. However, the effect of crack location on tearing resistance was not microstructurally quantified, but the qualitative analysis of crack location done in this work implies that a standard developed for heterogeneous materials shall contain guidelines on characterisation of crack location.

Vibration Control of Nuclear Power Plant Piping System Using Stockbridge Damper under Earthquakes

Generally the piping system of a nuclear power plant (NPP) has to be designed for normal loads such as dead weight, internal pressure, temperature, and accidental loads such as earthquake. In the proposed paper, effect of Stockbridge damper to mitigate the response of piping system of NPP subjected to earthquake is studied. Finite element analysis of piping system with and without Stockbridge damper using commercial software SAP2000 is performed. Vertical and horizontal components of earthquakes such as El Centro, California, and Northridge are used in the piping analysis. A sine sweep wave is also used to investigate the control effects on the piping system under wide frequency range. It is found that the proposed Stockbridge damper can reduce the seismic response of piping system subjected to earthquake loading.

An Itô model of pit corrosion in pipelines for evaluating risk-informed decision making

The purpose of this paper is to present a stochastic model based on the Itô differential equation to the time evolution of pit corrosion depth in nuclear power plant piping systems. The stochastic solution obtained by analytical and numerical methods (for comparison purposes) is compared to the deterministic solution, which is obtained by setting the stochastic term of the Itô equation equal to zero. The study of extended qualified useful life of nuclear power plants has two components, one deterministic and another probabilistic and one complements the other. The evaluation based on deterministic methods defines the difference between the current state and condition of the item in the qualifying phase, but does not define its probability of continuing performing its function adequately for a period longer than the one defined by its qualified life. The corrosion process is modeled by the Itô stochastic equation, which describes the time evolution of pit corrosion depth among different states. The Itô equation is determined by the knowledge of two functions known as the drift and diffusion coefficients. This work shows that these functions can be used to estimate pit corrosion depths from plant inspection data. An Itô modeling of pit corrosion can also be developed for any component of a nuclear power plant for which corrosion depth data is available. The above proposed approach provides a precise and easy way in which pit corrosion depth can be monitored over time, which is critical for developing reliability and risk-informed models for inspection and maintenance planning of corroded pipelines. The stochastic model presented is developed considering the influence of the Gaussian white noise on corrosion. The equations of the stochastic model are solved analytically based on the Itô integral. The maximum operating time of a piping before its wall reaches the minimum allowable thickness is calculated. This study uses pit corrosion depth data from the Angra I nuclear power plant.

EFFECTS OF FORGING AND HEAT TREATMENTS ON STRESS CORROSION BEHAVIOR OF 316LN STAINLESS STEEL IN HIGH TEMPERATURE CAUSTIC SOLUTION

The reactor coolant piping in the third generation nuclear power plants of AP1000 is manufactured by integrally forging. Therefore, it is of vital importance to investigate the effects of forging and heat treatments on the stress corrosion cracking(SCC) resistance of 316 LN stainless steel(316LNSS), which is the candidate material for the reactor coolant piping in AP1000 nuclear power plants. In this work, electron back scattering diffraction(EBSD) and microhardness measurements(HV) were used to characterize the microstructure and residual strain of the as-received 316 LNSS, the forged and solution anneal treated 316 LNSS and the forged and stress relief treated 316 LNSS, respectively. The average grain size of the as- received 316 LNSS was the largest, and the forged 316 LNSS followed by solution anneal treatment and stress relief treatment showed no obvious differences on grain size. The as-received 316 LNSS exhibited the highest residual strain followed by the forged and stress relief treated 316 LNSS and then solution anneal treated 316 LNSS. Besides, the residual strain in the as- received316 LNSS concentrated on grain boundaries, while the residual strain in the forged and stress relief treated316 LNSS was characterized by a band-like distribution. The U-bend specimens were utilized to investigate the SCC behavior of the 3 kinds of 316 LNSS specimens in high temperature caustic solution. After SCC experiments, the crack morphologies of the 3 kinds of 316 LNSS specimens were examined by SEM. Then the macro and micro fracture morphologies were examined by OM and SEM, respectively. Grain morphology, residual strain and grain boundary character distribution near the SCC crack tip of the forged and stress relief treated 316 LNSS were investigated using EBSD. The results showed that the forged and solution anneal treated 316 LNSS exhibited the lowest SCC sensibility, while the as-received the highest, with the most cracks and the highest growth rate. The as-received and the forged and solution anneal treated 316 LNSS showed obvious intergranular cracking, while the forged and stress relief treated 316 LNSS showed a mixed cracking mode. The larger average grain size and higher residual strain, especially concentrating on the grain boundaries, were considered to be responsible for the highest SCC sensibility of the as-received 316 LNSS. Compared with the forged and stress relief treated 316 LNSS, the higher content of coincidence site lattice boundary(CSLB) and lower residual strain contributed to the lower SCC sensibility of forged and solution anneal treated 316 LNSS. The stress relief treatment failed to eliminate the band-like microstructure effectively, which disadvantaged the SCC resistance.

Ethanolamine degradation and energy recovery using a single air-cathode microbial fuel cell with various separators

Ethanolamine (ETA) is commonly used for alkalinization to prevent corrosion of pipes in nuclear power plants. ETA, however, causes water to deteriorate, increasing the concentration of organic matter or nutrient salts in water systems. To generate power and degrade ETA at the same time, ETA was used as carbon source for microorganisms in a single air-cathode microbial fuel cell (MFC). Using a membrane as a separator in an MFC makes them costly to set-up. To reduce the expense and evaluate MFC performance, the experiments were conducted with three different separators: a proton exchange membrane (PEM), a cation exchange membrane (CEM), and polypropylene (PP) felt. The PP felt-MFC resulted in the most efficient COD (94%) and ammonium removal (52%). The CEM-MFC produced the highest power density of 583.7 mW m−2 at a current density of 0.15 mA cm−2. The Coulombic efficiencies (CEs) were 25.1, 23.7, and 10.5% for PEM-, CEM-, and PP felt-MFCs, respectively. Although using PP felt decreased the power generation compared with membrane MFCs in terms of energy recovery, it increased ETA degradation and reduced the cost of initial set-up.

Experimental validation of non-contacting measurement method using LED-optical displacement sensors for vibration stress of small-bore piping

There have been many reports of fatigue failures of small-bore piping systems such as drain piping, vent piping and instrumentation piping in nuclear power plants that arise from vibration sources such as pumps. To prevent the failures, integrity evaluation of piping is conducted by measuring and analyzing vibration stress in the piping. But, a more efficient and economical measurement method is desirable to evaluate the vibration fatigue in small-bore piping. In this study, a non-contacting measurement method was proposed that is based on optical displacement sensors using light emission diodes (LEDs) to measure the vibration stress. The applicability of the method was discussed based on the vibration experiments using pipe elements and a mock-up piping system. From the experimental results, the proposed method was clarified to be sufficiently applicable and practically useful for the vibration measurement and stress evaluation in small-bore piping systems.

원전 배관의 두께 측정 데이터에 대한 신뢰도 분석 방법 및 적용

Pipe wall-thinning by flow-accelerated corrosion and various types of erosion is significant damage in secondary system piping of nuclear power plants(NPPs). All NPPs in Korea have management programs to ensure pipe integrity from degradation mechanisms. Ultrasonic test(UT) is widely used for pipe wall thickness measurement. Numerous UT measurements have been performed during scheduled outages. Wall-thinning rates are determined conservatively according to several evaluation methods developed by Electric Power Research Institute(EPRI). The issue of reliability caused by measurement error should be considered in the process of evaluation. The reliability analysis method was developed for single and multiple measurement data in the previous researches. This paper describes the application results of reliability analysis method to real measurement data during scheduled outage and proved its benefits.