应变幅对国产锻造奥氏体不锈钢环境疲劳寿命影响的试验研究

为验证模拟压水堆核电站冷却剂服役环境对国产锻造主管道用奥氏体不锈钢疲劳寿命的影响,采用高温高压循环水疲劳测试系统对从产品锻件取样加工后的标准试样进行了低周疲劳试验,分析了试验数据与美国机械工程师学会(American Society of Mechanical Engineers,ASME)规范平均/设计疲劳曲线的关系,获得了应变幅对奥氏体不锈钢环境疲劳寿命的影响规律,并初步评价了ASME规范设计疲劳曲线和环境疲劳修正系数的适合性。     

Hydrogen water chemistry for BWRs: a status report on the EPRI development program

Many boiling water reactors (BWRs) have experienced extensive intergranular stress corrosion cracking (IGSCC) in their austenitic stainless steel reactor coolant system piping, resulting in serious adverse impacts on plant capacity factors, O&M costs, and personnel radiation exposures. A major research program to provide remedies for BWR pipe cracking was co-funded by EPRI, GE, and the BWR Owners Group for IGSCC Research between 1979 and 1988. Results from this program show that the likelihood of IGSCC depends on reactor water chemistry (particularly on the concentrations of ionic impurities and oxidizing radiolysis products) as well as on material condition and the level of tensile stress. Tests have demonstrated that the concentration of oxidizing radiolysis products in the recirculating reactor water of a BWR can be reduced substantially by injecting hydrogen into the feedwater. Recent plant data show that the use of hydrogen injection can reduce the rate of IGSCC to insignificant levels if the concentration of ionic impurities in the reactor water is kept sufficiently low. This approach to the control of BWR pipe cracking is called hydrogen water chemistry (HWC). This paper presents a review of the results of EPRI's HWC development program from 1980 to the present. In addition, plans for additional work to investigate the feasibility of adapting HWC to protect the BWR vessel and major internal components from potential stress corrosion cracking problems are summarized.     

Electrochemical noise measurements of stainless steel in high temperature water

Corrosion in a high purity aqueous environment simulating a boiling water reactor (BWR) is addressed in this work. This condition necessitates autoclave experiments under high pressure and temperature.Long-term electrochemical noise measurements were explored as a mean to detect and monitor stress corrosion cracking phenomenon. An experimental set up, designed to insulate the working electrode from external interference, made possible to detect and monitor stress corrosion cracking in slow strain rate tests for sensitized and solution annealed 304 stainless steel at 288 °C. Time-series analysis showed variations in the signature of the current density series due to transgranular stress corrosion cracking (TGSCC) and intergranular stress corrosion cracking (IGSCC).     

碳钢对核主泵用奥氏体不锈钢的污染研究

奥氏体不锈钢在加工、运输和装配过程中如果与碳钢直接接触,就会被碳钢污染,而导致奥氏体不锈钢耐蚀性能的改变。众所周知,核主泵用奥氏体不锈钢对耐蚀性有着非常严格的要求,本文以Z2CN18-10核主泵用奥氏体不锈钢为例,通过FeCl3腐蚀试验和电化学方法测试了被碳钢污染后其耐腐蚀性能的变化。试验结果表明:附着在不锈钢表面的碳钢对其长期总体腐蚀速率影响不大;嵌入式的碳钢颗粒会显著降低奥氏体不锈钢的点蚀电位,增大发生点蚀的倾向;硝酸钝化可部分抵消被污染不锈钢点蚀电位的降低,但该值仍远低于同样经过硝酸钝化,而未被污染的不锈钢的点蚀电位。此外,还针对碳钢污染对核电站辐射场的影响和对燃料包壳热传导效率的影响进行了讨论。     

射流装置启泵过程的瞬态特性

射流装置由射流泵和主泵组成,引入MRX(Marine Reactor X)压水堆一回路系统中,有助于提升反应堆的固有安全性。反应堆启泵过程中,流量急剧上升导致堆芯温度变化,影响堆芯运行安全。通过计算流体力学(Computational Fluid Dynamics,CFD)方法对引入射流装置MRX一回路10%满功率(Full Power,FP)、17.5%FP和25%FP堆芯功率下启泵进行三维瞬态模拟,分析MRX一回路中射流装置流场瞬态特性。结果表明,射流装置的加入可以改善一回路自然循环能力,提高启泵工况下冷却剂初始变化流量,减缓变化趋势,改善过渡安全性;启泵过程中一回路温度存在波动现象,且堆芯功率越大,波动幅度越大,时间越长;启泵完成后射流泵喷嘴处流速较大。验证了压水堆中引入射流装置提升反应堆固有安全性的可行性,同时为进一步优化设计方案提供方向参考。     

The feasibility of small size specimens for testing of environmentally assisted cracking of irradiated materials and of materials under irradiation in reactor core

Environmentally assisted cracking (EAC) or, in other words, stress corrosion cracking (SCC) of in-core materials has become an increasingly important reason for the downtime and maintenance costs of nuclear power plants (NPPs). Use of small size specimens for stress corrosion testing of irradiated materials is necessary because handling of high dose rate materials is difficult and the availability of these materials is limited. A drawback of using small size specimens is that they do not in some cases fulfil the requirements of the relevant testing standards and sometimes their limited load-bearing capacity prevents corrosion fatigue tests and tests with static loading at reasonable K_I values. The test results show that the ductile fracture resistance curves of a Cu–Zr–Cr alloy are, to some extent, independent of the specimen geometry and size. However, the curves of small specimens deviate from the curves of larger specimens at high J values (large plastic zone relative to the remaining ligament) or when the crack growth exceeds about 30% of the remaining ligament. The size dependency of the tested Cu–Zr–Cr alloy seems to be a consequence of decreasing stress triaxiality as the size of the specimen is decreased. The results of the SCC tests of sensitized SIS 2333 stainless steel (equal to AISI 304) specimens in simulated boiling water reactor (BWR) water show that the plastic deformation of the remaining ligament of the specimen has no significant effect on the environmentally assisted crack growth rate. This indicates that stress corrosion testing is not limited by the specimen size. The size dependency in SCC tests should be further studied by conducting tests using various specimen sizes.     

A semiempirical prediction of the decay ratio for the boiling water reactors start-up process

During the start-up of a commercial boiling water reactor (BWR), the power and the coolant flow are continuously monitored. In order to prevent power instability events, the decay ratio (DR) could also be monitored. The process can be made safer if the operator could anticipate the DR too. DR depends on the power, the flow and many other quantities such as axial and radial neutron flux distribution, feed water temperature, void fraction, etc. A simple relationship for DR is derived. Three independent variables seem to be enough: the power, the flow and a single parameter standing for all other quantities which affect the DR. The relationship is validated with data from commercial BWR start-ups. A practical procedure for the start-up of a BWR is designed; it could help preventing instability events.     

Water Chemistry Experience of Nuclear Power Plants in Japan

Japanese LWRs have experienced several troubles caused by corrosions of structural materials in the past ca. 20 years of their operational history, among which are increase in the occupational radiation exposures, intergranular stress corrosion cracking (IGSCC) of stainless steel piping in BWR, and steam generator corrosion problems in PWR. These problems arised partly from the improper operation of water chemistry control of reactor coolant systems. Consequently, it has been realized that water chemistry control is one of the most important factors to attain high availability and reliability of LWR, and extensive researches and developments have been conducted in Japan to achieve the optimum water chemistry control, which include the basic laboratory experiments, analyses of plant operational data, loop tests in operating plants and computer code developments. As a result of the continuing efforts, the Japanese LWR plants have currently attained a very high performance in their operation with high availability and low occupational radiation exposures. A brief review is given here on the R & D of water chemistry in Japan     

Demonstration of piping integrity with SMA technology

The safe function of a new pipe whip restraint device has been demonstrated in a full scale test. The restraint is based on using a shape memory alloy to protect a pipe and its environment in the event of a double-ended-guillotine-break. The evaluation test has been performed at boiling water reactor (BWR) operating pressure and temperature using a pipe representing BWR primary piping.     

Flow instabilities in boiling channels of pressure-tube-type reactor

The core of an advanced thermal reactor (ATR), which is a light-water-cooled, heavy-water-moderated, pressuretube-type reactor, consists of many parallel channels. For this reason the ATR was believed to exit from flow instability easily. Hence the flow instability conditions were investigated with the heat transfer loop (HTL) and safety experiment loop (SEL), which simulate the flow system of an ATR at full scale. The present data were compared with those for a boiling water reactor (BWR) system. The effect of the outlet pipe, which is not provided in a BWR, on the flow oscillation under low power conditions was also studied.     

Piping Cracks in JPDR, (IV)

Circumferential cracks detected in the JPDR (BWR) near welded joints connecting the nozzle safe-end to pipe (austenitic stainless steel) were studied in reference to the stresses applied in service, the conditions of welding, environment (O₂ and Cl^? concentration, water flow, temperature etc.), metallurgical structure and operating records, to determine the cause of cracking. Fatigue tests were also undertaken with simulated welded pipe of size to examine possible contribution of fatigue to the cracking.

The analysis indicated stress corrosion to be the principal cause of cracking, and that it had initiated in the heat-affected zone which had been sensitized by the excess heat input of welding. The factors contributing to the stress corrosion were the presence of oxygen in the fluid, which had at times attained a level exceeding 0.2 ppm, and which combined with stress, at times exceeding the yield strength. This had caused the cracks once initiated, to propagate along the grain boundary.     

The development of a closed-form analytical model for the stability analysis of nuclear-coupled density-wave oscillations in boiling water nuclear reactors

A state-of-the-art one-dimensional thermal-hydraulic model has been developed to be used for the linear analysis of nuclear-coupled density-wave oscillations in a boiling water nuclear reactor (BWR). This model accounts for phasic slip, distributed spacers, subcooled boiling, space/time-dependent power distributions and distributed heated wall dynamics. In addition to a parallel channel stability analysis, a detailed model was derived for the BWR loop analysis of both the natural and forced circulation modes of operation.The model for coolant thermal-hydraulics has been coupled with the point kinetics model of reactor neutronics. Kinetics parameters for use in the neutronics model have been obtained by utilizing self-consistent nodal data and power distributions.The computer implementation of this model, NUFREQ-N, was used for the parametric study of a typical BWR/4, as well as for comparisons with existing in-core and out-of-core data. Also, NUFREQ-N was applied to analyze the expected stability characteristics of a typical BWR/4.     

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.     

Simulation model of a nuclear reactor turbine

In applying optimal control theory to a boiling water nuclear reactor (BWR) system which includes the primary recirculation loop, the turbine and their associated auxiliaries, it is necessary to have a linearized mathematical model. Nonlinear and linearized models of a turbine coupled to a BWR, and open-loop responses for specific disturbances are presented.     

Irradiation devices and irradiation programmes at the high flux reactor (HFR) Petten for the investigation of the irradiation behaviour of stainless steels

Irradiation testing of stainless steels represents an important share in the utilization spectrum of the high flux reactor (HFR) Petten. These tests address in particular austenitic and martensitic steels for thermo-nuclear fusion reactors, austenitic steels for fast breeder reactors, and stainless steels and Cr---Ni-alloys for light water reactors.A number of well-proven irradiation devices covering an extensive range of applications and operated by a vastly experienced group are available at the HFR. In addition, at the Petten hot cells, many diversified tools exist for irradiation testing of stainless steel. The irradiation devices are tailored to provide typical, reactor specific environment conditions to the samples. Such conditions are maintained within narrow specifications. In this respect, the long lasting operational experience originating from the many materials irradiation programmes performed at the HFR, the high and consistent availability of the HFR and its reliability with regard to provision of predictable and reproducible irradiation conditions, are the keys to these many demonstrable experiments. The adjustment of typical neutron spectrum at the irradiation samples is performed through an appropriate choice of structural materials and their arrangement within the irradiation device. Many irradiation devices are reloadable, being specially suited for intermediate unloading and hot cell measurement of the irradiated samples. Other devices are equipped with integral loading and measuring systems for in-pile measurement of characteristic irradiation parameters (e.g. the creep behaviour). The neutron fluence is calculated and then measured after irradiation from neutron fluence monitor sets which are integrated into the sample carriers.The post irradiation examination and testing of the irradiated samples is performed either at the customer's hot cells or at the Petten hot cells, in particular, when intermediate measurements are required. Dedicated and specialized testing equipment for tensile, creep, Charpy, fatigue and crack propagation testing, various microscopy systems (SEM, TEM) and other test equipment (gamma-scanning, chemical analysis, image analysis, etc.) are all available at the Petten hot cells.     

Z3CN20.09M奥氏体不锈钢热老化冲击性能试验研究

采用GB/T19748-2005钢材夏比V型缺口摆锤冲击试验仪器化试验方法,对压水堆核电厂用离心铸造Z3CN20.09M奥氏体不锈钢主管道样品进行了实验室热老化的冲击性能研究。冲击试验数据的统计分析表明,热老化对Fiu/Fm比值不产生影响,而对冲击载荷有显著影响,对冲击能量的影响则更为显著。透射电子显微分析表明,热老化导致铁素体中出现沉淀物,并引发了奥氏体中位错组态的改变。与热老化时间lg t之间也满足线性关系。     

Development and application of filters for air cleaning in nuclear power plants

The development of filter systems for air cleaning in nuclear power plants will be briefly described. The result of research work on iodine filters was the basis for the use of gasketless deep-bed filters and of the multiway sorption filter in German reactor stations. The composition of the iodine release to the environment was validated with the “discriminating iodine species sampler”. The main sources for the release of elemental iodine from BWR and PWR were discovered and finally suppressed. The mechanical strength of HEPA filters is being tested at high temperatures and humidities. Prototype HEPA filters have been developed with much higher resistance against humidity, differential pressure, and corrosion. A filter for the removal of particles during extreme operating conditions such as containment venting was developed using stainless steel fibers for the filter medium. The first filter of this type has already been built and installed in a modern German PWR as a part of the containment-venting system for serious accidents.     

Fractographic and microstructural characterization of irradiated 304 stainless steel intergranularly fractured in inert gas

Fractographic and microstructural examinations were performed by scanning and transmission electron microscopy, respectively, and correlated, for the thermally sensitized 304 stainless steel (SS) irradiated to 1.2×10²¹ n/cm² (E>1 MeV) in BWR condition and fractured intergranularly in 290 °C inert gas. Intergranular (IG) cracks were present in the specimen surface region and the fracture surface periphery. The fractography showed IG facets decorated with various patterns of linear features/steps. The microstructures of the surface region revealed linear features/deformation twinning near grain boundaries and microtwins at grain boundaries. The linear features identified on the [1 1 1] habit plane varied depending on deformation levels. The high number density of microtwins evidences a high local stress and strain concentration, which may nucleate and initiate at the impingement of deformation twins and grain boundaries. Therefore we conclude that a mechanism causing the IG cracking mechanically in non-aqueous environment is present in the highly irradiated austenitic SS.     

Assessment of thermal aging embrittlement in a cast stainless steel valve and its effect on the structural integrity

This paper analyzes the thermal aging embrittlement occurred in a cast stainless steel valve, which is part of the reactor water clean-up (RWCU) system of a Spanish boiling water reactor (BWR) nuclear power plant. The aim is to estimate the current and future state of the material and the corresponding structural integrity of the valve. Given that there is no data available for the experimental characterization of the material, the evolution of the mechanical properties (fracture toughness, yield stress, flow stress and Ramberg-Osgood parameters) has been estimated using the ANL procedure.With the obtained estimations, the critical crack size has been calculated using the European procedure FITNET FFS and the ASME Code.This analysis considers not only the evolution of the mechanical properties up to now but also its future evolution in case there is a life extension of the plant until year 2029.