Influence of Radiation-Stimulated Grain-Boundary Segregation of Phosphorus on the Operational Properties of Nuclear-Reactor-Vessel Materials

The main mechanisms of radiation embrittlement of reactor vessel materials are considered to be hardening of material as a result of the formation of matrix defects, for example, micropores and second-phase precipitates – copper and others, and a change in the cohesive strength of grain boundaries as a result of the segregation of surface-active impurities, primarily, phosphorus. The question of the degree to which the latter mechanism affects the change in the properties of reactor-vessel materials under irradiation remains open. In the present paper, computational estimates of the kinetics of radiation-stimulated segregation of phosphorus on grain boundaries in reactor-vessel materials and the resulting changes in the mechanical characteristics of steel are compared with corresponding experimental data.     

反应堆压力容器钢中溶质元素空位型扩散机理研究

辐照或热老化导致元素偏析和沉淀析出是反应堆压力容器(reactor pressure vessel,RPV)钢性能退化的主要影响因素,点缺陷与合金/杂质元素结合与扩散是引起元素偏析和沉淀析出的主要原因。本文利用分子动力学方法研究了反应堆压力容器钢中几种主要合金/杂质元素(Cu、Ni、Mn、P)的空位型扩散机理。研究了空位与合金/杂质元素的结合性能;基于多频模型计算了合金/杂质元素的空位风参数和扩散系数。通过计算发现,Cu、P与第1近邻、第2近邻空位均具有较大的结合能,Ni与第2近邻空位具有较大的结合能;溶质元素的空位风均随着温度的升高而增大,表明在高温下合金/杂质元素均倾向通过与空位互换位置而扩散。     

Determination of γ-ray flux in a stopped VVR-SM reactor from the color-center formation in LiF

The absorption and photoluminescence of LiF crystals, irradiated by γ radiation from a stopped reactor and a wet repository, are studied and compared with the corresponding spectra obtained by irradiation with ⁶⁰Co γ rays with known power. The contribution of neutron and γ radiation of the reactor to the formation of point and complex radiation defects is determined. The dose dependences of the optical bands are used to determine the intensity of the γ radiation from the reactor and the repository. __________ Translated from Atomnaya énergiya, Vol. 100, No. 5, pp. 388–393, May, 2006.     

Restoration of Mechanical Properties of Irradiated Steel by Thermal Annealing

During operation a reactor vessel is exposed to neutron irradiation fluxes, which causes degradation of the mechanical properties of the vessel material. This concerns primarily the change in the critical temperature of brittleness and other characteristics of the fracture viscosity of steel. Heating steel above the irradiation temperature, thereby increasing the diffusion mobility of point defects, is a prerequisite for the appearance of thermodynamic instability of various radiation damage of steel and, consequently, creates the conditions for restoration of the mechanical properties. In this paper the factors influencing the effectiveness of the restoration of the critical temperature of brittleness of materials of water-moderated water-cooled power reactor vessels with post-radiation annealing are examined.     

Reconstruction of metal embrittlement in a reactor vessel of the atomic icebreaker Lenin

Data from a study of radiation damage to the vessel of a reactor from the retired atomic icebreaker Lenin are used to determine the radiation embrittlement characteristics of the metal. Irradiation by a low neutron flux of 10¹⁰–10¹¹ cm⁻²sec⁻¹ at the beginning of operation is found to correspond to more intense embrittlement of the metal. Then, apparently, as harmful elements are depleted in the matrix of the metal, embrittlement is reduced until there is a change in sign relative to the standard curve obtained for neutron fluxes above 10¹³ cm⁻²sec⁻¹. It is proposed that, because of irradiation by low fluxes of neutrons in the peripheral zones of reactor vessels during some stages of operation, these zones may be damaged to a greater extent than those lying closer to the core. The irradiating neutron flux is a factor that influences the embrittlement of reactor vessel materials, so there is some interest in studying how material is damaged in the vessels of power reactors with low radiation loads which are under development. This is also needed in order to evaluate the efficacy of measures undertaken to reduce the effect of neutron irradiation on reactor vessels. Translated from Atomnaya énergiya, Vol. 105, No. 4, pp. 201–205, October, 2008.     

高温无机闪烁晶体Ce:YAlO3的缺陷研究

采用同步辐射白光透射形貌术研究了提拉法生长的高温无机闪烁晶体Ce:YAlO3（简称Ce:YAP）中的缺陷。实验发现在Ce:YAP晶体中存在着生长条纹、包裹沉积物、核心、孪晶及位错簇等缺陷，同时对生长缺陷形成的原因进行了讨论。结果表明，离子掺杂浓度、原料的纯度以及生长工艺条件等是影响Ce:YAP晶体缺陷的主要原因。     

Contribution of γ radiation to radiation embrittlement of vessel steel

Data on the contribution of reactor γ rays to the embrittlement of nuclear reactor vessel steel are presented and discussed. It is shown qualitatively that the increase of the shift of the critical brittleness temperature in the region of the outer layers of propulsion-reactor vessels and, conversely, the decrease of this shift at the same locations in power reactors correlates with the γ radiation intensity, which changes as a result of the structural features of the reactors. In addition, it is shown that the previously published paradoxical experimental results on the irradiation of A-537 steel in the EL-3 reactor with and without cadmium foil as well as on the anomalous embrittlement of domestic vessel materials under irradiation in the core of a propulsion reactor are explained by the effect of radiation γ annealing. Translated from Atomnaya énergiya,Vol. 106, No. 1, pp. 22–28, January, 2009.     

A review of irradiation effects on LWR core internal materials - IASCC susceptibility and crack growth rates of austenitic stainless steels

Austenitic stainless steels (SSs) are used extensively as structural alloys in the internal components of light water reactor (LWR) pressure vessels because of their relatively high strength, ductility, and fracture toughness. However, exposure to neutron irradiation for extended periods changes the microstructure (radiation hardening) and microchemistry (radiation-induced segregation) of these steels, and degrades their fracture properties. Irradiation-assisted stress corrosion cracking (IASCC) is another degradation process that affects LWR internal components exposed to neutron radiation. The existing data on irradiated austenitic SSs were reviewed to evaluate the effects of key parameters such as material composition, irradiation dose, and water chemistry on IASCC susceptibility and crack growth rates of these materials in LWR environments. The significance of microstructural and microchemistry changes in the material on IASCC susceptibility is also discussed. The results are used to determine (a) the threshold fluence for IASCC and (b) the disposition curves for cyclic and IASCC growth rates for irradiated SSs in LWR environments.     

Reactor radiation dosimetry using fused quartz

The characteristics of quartz glass for separate dosimetry of γ radiation from a VVR-SM reactor in the presence and absence of neutron fluxes are investigated. Comparing the absorption and photoluminescence spectra of samples irradiated with γ rays from a stopped reactor and mixed with neutron and γ radiation from an operating reactor shows that in both cases oxygen defects are produced and the dose dependences are linear. The dosimetric bands are stable with respect to light and temperatures up to 400°C. The stationary γ-ray flux from the reactor, after the reactor is stopped, is calibrated according to a known source of γ-ray source ⁶⁰Co and is ≈15 Gy/sec. __________ Translated from Atomnaya énergiya, Vol. 100, No. 3, pp. 216–220, March, 2006.     

Nuclear fuel and the characteristics of aerosol formation in the object “Cover”

High-temperature atomization of materials, transfer of elements into the gas phase and the condensation of the elements, occurred during the active stage of the accident at least at a local point of the reactor. As a result of these processes, in some cases the ratio Pu/U in large spherical particles of the dispersed phase differ substantially from the ratio characteristic for the average nuclear fuel in the fourth power generating unit of the Chernobyl nuclear power plant. Therefore, the term “average nuclear fuel” with respect to materials containing nuclear fuel in the object “Cover” is inadequate. Moscow Technological Center “Cover,” Ukrainian Academy of Sciences. Yu. N. Lobach Science Center “Institute of Nuclear Research,” Ukrainian Academy of Sciences. Translated from Atomnaya énergiya, Vol. 82, No 1, pp. 39–44, January, 1997     

Embrittlement of low-alloy structural steel by neutron irradiation

The radiation embrittlement of reactor vessel materials is a complex process, which depends on the conditions of irradiation and the microstructure and chemical composition of the steel. It is universally acknowledged that phosphorus, copper, and nickel intensify the radiation embrittlement of vessel material the most. It is believed that Mn, N, C, Mo, Si, As, Sn, V, and other elements also influence radiation embrittlement, but their effect has not been definitely established and is much less than the effect of phosphorus, copper, and nickel. The presence of a synergetic interaction of elements in the irradiation process and the complex interaction of metallurgical factors and the irradiation conditions make it difficult to determine the degree to which impurities and alloying elements influence radiation embrittlement. The effect of the chemical composition of steel, as one of the most important parameters determining the radiation service life of vessel material, on radiation embrittlement is studied, 5 figures, 1 table, 20 references. Translated from Atomnaya énergiya. Vol. 88, No. 4, pp. 271–276, April, 2000.     

Kinetic Monte Carlo simulations of radiation induced segregation and precipitation

Kinetics of radiation induced segregation and precipitation in binary alloys are studied by Monte Carlo simulations. The simulations are based on a simple atomic model of diffusion under electron irradiation, which takes into account the creation of point defects, the recombination of close vacancy-interstitial pairs and the point defect annihilation at sinks. They can reproduce the coupling between point defect fluxes towards sinks and atomic fluxes, which controls the segregation tendency. In pure metals and ideal solid solutions, the Monte Carlo results are found to be in very good agreement with classical models based on rate equations. In alloys with an unmixing tendency, we show how the interaction between the point defect distribution, the solute segregation and the precipitation driving force can generate complex microstructural evolutions, which depend on the very details of atomic-scale diffusion properties.     

Irradiation-induced precipitation modelling of ferritic steels

In high strength low alloy (HSLA) steels typically used in reactor pressure vessels (RPV), irradiation-induced microstructure changes affect the performance of the components. One such change is precipitation hardening due to the formation of solute clusters and/or precipitates which form as a result of irradiation-enhanced solute diffusion and thermodynamic stability changes. The other is irradiation-enhanced tempering which is a result of carbide coarsening due to irradiation-enhanced carbon diffusion. Both effects have been studied using a recently developed Monte Carlo based precipitation kinetics simulation technique and modelling results are compared with experimental measurements. Good agreements have been achieved.     

Dynamic studies of defect mobility using high voltage electron microscopy

The large variety of observed defect-structure development by electron irradiation in fcc and bcc metals is classified from the view point of point-defect mobilities. Vacancy mobility is obtained from the defect structure change caused by the annihilation of vacancies accumulated by irradiation, and their motion activation energy is obtained from the interstitial cluster growth at high temperatures. Mobility of interstitial atoms and their interaction with impurities are obtained from the variation of interstitial cluster formation. Various observed effects of electron radiation induced motion of point defects are explained by the displacement of atoms by the transfer of energy comparable with the migration activation energy of each point defect. The observed superficial temperature dependence of the induced diffusion of vacancies in fcc metals is attributed to divacancies created by the induced diffusion. An efficient method to obtain the self-diffusion energies is proposed with results for some metals.     

Grain-Boundary Segregation of Phosphorus in Low-Alloy Steel

Grain-boundary segregation of impurity elements, such as phosphorus, arsenic, antimony, and others, decreases the grain-boundary cohesion, which can substantially increase the temperature of the ductile-brittle transition in low-alloy structural steel. The most dangerous surface-active impurity for low-alloy steel employed for nuclear reactor vessels is phosphorus. A change of the cohesive strength of grain boundaries as a result of radiation-stimulated phosphorus segregation is considered to be one of the main mechanisms determining the radiation embrittlement of reactor-vessel materials. Since the mechanisms of embrittlement during development of reversible temper brittleness and radiation-stimulated grain-boundary segregation of phosphorus are the same, the main characteristics of the influence of the latter on the mechanical properties of steel can be determined by investigating steel treated in the range 400–600°C. The present investigation made it possible to develop a relation for determining the change in the temperature of the ductile-brittle transition in low-alloy steel as a result of the development of temper brittleness.     

Atom probe study of radiation induced grain boundary segregation/depletion in a Fe-12%Cr alloy

Ferritic steels are important candidate structural materials for future fission and fusion reactors. However, the effect of radiation induced segregation/depletion (RIS/RID) of Cr in grain boundaries and its effect on microstructures and properties are still not clear. Therefore, a systematic approach is shown in this paper, combining electron backscattered diffraction, focused ion beam specimen preparation and atom probe tomography for analysing a single grain boundary in a Fe-12wt%Cr from the point of grain boundary orientation, microchemistry and impurities. Several samples have been prepared from the same grain boundary and consistent 3D reconstruction with quantitative calculation of segregation demonstrates the high reliability and repeatability of this approach.     

Acoustic method for monitoring the state of VVER vessels

The results of an investigation of the possibility of using acoustic characteristics to evaluate the state of the VVER vessel material are presented. The investigations were performed on template–samples cut from the VVER-440 vessel material and on control samples from VVER-1000 vessels. It is shown that the characteristics of elastic waves depend on the fluence and the radiation embrittlement of VVER vessel materials. Translated from Atomnaya énergiya, Vol. 106, No. 1, pp. 31–35, January, 2009.     

Analyzing the effect of displacement rate on radiation-induced segregation in 304 and 316 stainless steels by examining irradiated EBR-II components and samples irradiated with protons

Recent studies have indicated that, at temperatures relevant to fast reactors and light water reactors, void swelling in austenitic alloys progresses more rapidly when the radiation dose rate is lower. A similar dependency between radiation-induced segregation (RIS) and dose rate is theoretically predicted for pure materials and might also be true in complex engineering alloys. Radiation-induced segregation was measured on 304 and 316 stainless steel, irradiated in the EBR-II reactor at temperatures near 375 °C, to determine if the segregation is a strong function of damage rate. The data taken from samples irradiated in EBR-II is also compared to RIS data generated using proton radiation. Although the operational histories of the reactor irradiated samples are complex, making definitive conclusions difficult, the preponderance of the evidence indicates that radiation-induced segregation in 304 and 316 stainless steels is greater at lower displacement rate.     

Composition of reactor radiation—statistically significant radiation damage factor of diamond

Conclusions  

Application of radiation chemistry research to the operation of nuclear reactors

The theme of this review is the application of radiation chemistry research to improve the operating efficiency of nuclear reactors. The intense radiation fields in reactor cores produce a hostile environment for incore materials; this report describes how recent research helped overcome the chemistry problems caused by the radiation.Examples discussed are the inhibition of graphite moderator corrosion and prevention of carbon deposition in gas-cooled reactors, suppression of radiolysis of the cooling water in concrete pressure vessels, hydrogen formation following a loss of coolant accident in a PWR and improving the stability of decontamination reagents for water reactors.