快中子照射唐菖蒲球茎的损伤作用初探

用快中子脉冲堆(China fast burst reactor Ⅱ,CFBR-Ⅱ)对唐菖蒲"超级"球茎进行不同剂量的照射处理,研究快中子对唐菖蒲球茎表面的刻蚀情况及对蛋白质表达的阻遏作用.扫描电镜观察发现球茎受损伤情况随照射剂量和照射角度的不同而变化,受垂直方向照射的球茎表面刻蚀程度最严重,且刻蚀程度随着剂量增大而增大.通过田间栽培试验和SDS-聚丙烯酰胺凝胶电泳(SDS-polyacrylamide gel electrophoresis,SDS-PAGE)分析发现,随照射剂量的增大,植株生长发育受到明显的抑制,蛋白表达数量显著减少.由此表明,快中子照射对植株能产生明显的影响.     

The microstructure of neutron-irradiated Fe-Cr alloys: A small-angle neutron scattering study

The effect of Cr on the irradiation-induced microstructure of neutron-irradiated Fe-Cr alloys is not yet known in detail. Small-angle neutron scattering was applied in order to provide the characteristics of nm-sized defects averaged over macroscopic volumes. Results are reported for a set of Fe-Cr alloys of Cr levels of 2.5, 5, 9 and 12.5 at.%, irradiated at 300 °C up to neutron exposures of 0.6 and 1.5 dpa. We have found that the incoherent magnetic scattering of the unirradiated alloys exhibits a systematic variation with the Cr content and that there is an irradiation-induced increase of the coherent magnetic scattering for each of the irradiated conditions. The effect of Cr on size and type of irradiation-induced scatterers is discussed. For 12.5 at.%Cr, the scatterers are unambiguously identified as α′ particles. For 2.5 and 5 at.%Cr, the scatterers are tentatively interpreted as clusters enriched with alloying Cr and impurity C. For 9 at.%Cr, a mixture of both kinds of scatterers explains the experimental findings.     

Mechanical properties and microstructure of certain construction materials after neutron irradiation

The article presents data on the influence of irradiation during normal operation of the reactor at the First Atomic Power Station on the mechanical properties and microstructure of a number of austenitic, ferrite, ferritemartensitic steels and molybdenum. The authors demonstrate a sharp falloff in the plasticity of the austenitic and ferrite steels and molybdenum as a result of irradiation with a total flux of (0.9–3.4)·10²⁰ neut/cm² at a temperature of 450–650 ° C. It is found that the causes of steel and molybdenum becoming brittle differ from one another in their nature. The information presented may prove useful in the planning of new atomic power plants.The authors wish in conclusion to express their gratitude to E. V. Chermashentsev and A. Ya. Ladygin for their part in the work undertaken.     

Mechanical properties of ferrous alloys after fast neutron-irradiation at high temperature

The results are given of an international “round-robin” experiment to study the nature of the damage structure in neutron irradiated zirconium and zircaloy-2 using transmission electron microscopy. The damage structure consists entirely of $\text{1}\text{3}\text{α<11}\text{2}\text{?}\text{0>}$ dislocation loops and no evidence has been found for c-component loops. Both vacancy and interstitial loops were found in specimens irradiated at 400 °C, with an excess of vacancy loops. Quantitative measurements of loop size distributions and loop concentrations are reported. All specimens exhibited “corduroy” contrast to varying degrees. The importance of choice of imaging conditions to minimize the contrast from thin foil artefacts such as oxide films and surface hydrides is stressed. The significance of the results is briefly discussed with reference to current theories of irradiation growth.     

锗酸铋闪烁体快中子辐照硬度的γ能谱研究

锗酸铋晶体作为一种优良的无机闪烁体已被广泛用于高能物理和核辐射测量技术中,因此有必要研究其辐照硬度。本文对锗酸铋晶体进行两种不同剂量的14MeV快中子辐照,并将其用作γ能谱仪的探测器,测量其辐照前后的本底能谱和^137Cs的γ能谱,通过能谱的变化和退火恢复情况来研究锗酸铋晶体的辐照硬度。     

Si(Au)探测器的快中子辐射损伤

在核实验中,半导体硅探测器常用来对带电粒子和中子进行探测。其能量响应和时间响应特性均十分良好,但是耐辐射性能却较差,尤其在伴有γ和中子本底的环境中,由于辐照引起探测器晶格损伤,导致半导体探测器性能退化,如探测器漏电流增大、能量分辨本领变差、能量亏损增加、时间响应增长等,严重影响探测器的使用和寿命。     

Irradiation induced dislocation loop and its influence on the hardening behavior of Fe-Cr alloys by an Fe ion irradiation

Nano indentation analysis and transmission electron microscopy observation were performed to investigate a microstructural evolution and its influence on the hardening behavior in Fe-Cr alloys after an irradiation with 8 MeV Fe⁴⁺ ions at room temperature. Nano indentation analysis shows that an irradiation induced hardening is generated more considerably in the Fe-15Cr alloy than in the Fe-5Cr alloy by the ion irradiation. TEM observation reveals a significant population of the a₀<1 0 0> dislocation loops in the Fe-15Cr alloy and an agglomeration of the 1/2a₀<1 1 1> dislocation loops in the Fe-5Cr alloy. The results indicate that the a₀<1 0 0> dislocation loops will act as stronger obstacles to a dislocation motion than 1/2a₀<1 1 1> dislocation loops.     

Ab initio formation energies of Fe-Cr alloys

We have calculated ab initio lattice parameters, formation energies, bulk moduli and magnetic moments of Fe-Cr alloys. The results agree well with available experimental data. In addition to body centered cubic (bcc) alloys, which are representative of ferritic steels used in fast neutron reactors, face centered cubic (fcc) and hexagonal close packed (hcp) phases were considered in order to complete a theoretical database of thermodynamic properties. Calculations were done for the ferromagnetic phase, as well as for a phase with local moment disorder, simulating the magnetic structure at high temperatures. For the latter case, the formation energy of the alloy is strictly positive smooth function of chromium concentration, in agreement with experiments performed at high temperature. In the ferromagnetic case, a negative mixing enthalpy is found for chromium concentrations below 6%. Our observation is consistent with the experimentally observed inversion of the ordering trend, as well as with formation of the chromium rich α^′ phase at Cr-concentrations above 9%, occurring at T<900 K.     

微型堆辐照座内快中子通量谱的测定

用Ａｌ，Ｆｅ，Ｉｎ和Ｎｉ作探测片，用阈探测片活化法测定了中国原子能科学研究院微型堆内，外辐照孔道的快中子能量。用平均截面法求得内外辐照孔道的快中子与热中子能量比分别为０．１９８和０．０７７，用有效阈能法计算了不同能量区间的快中子通量，同时也对四个内辐照等之间及内、外辐照管内径向和轴向快中子通量的不均匀度进行了测定。     

Voids resulting from fast neutron irradiation of a stabilized stainless steel

A 16 Cr-13 Ni-niobium stabilized stainless steel (Type 1.4988) was irradiated as fuel pin cladding in the DFR reactor. The irradiation temperatures ranged from 300 to 650 °C. Neutron fluences from 3.3 to 4.0 n/cm²$\text{(E > 0.1 MeV)}$ were achieved. Swelling behavior was studied in this material by transmission electron microscopy, immersion density determinations and diametral change measurements.Void formation was observed in the temperature range 360–610 °C. Void concentration values at lower temperatures (< 480 °C) are comparable to those cited for the unstabilized stainless steels AISI 304 and 316, however, they decrease more strongly with increasing irradiation temperatures. The mean and maximum diameters show a pronounced maximum at about 530 °C. The decrease of the void diameters at higher temperatures can be explained by the nature of the cavities observed. Annealing experiments with specimens irradiated at 610 °C have revealed a bubble or a mixed void-bubble character for these cavities.Swelling in type 1.4988 was found to be lower than that reported previously for types 304 and 316 at comparable irradiation and material conditions. This is especially pronounced at higher temperature. Diameter changes of the pin at $\text{T >- 610 °C}$ cannot be explained by void-formation.     

In-pile dimensional changes in neutron irradiated zirconium base alloys

A model is developed to predict in-pile growth in zirconium base alloys as a function of neutron flux, neutron fluence, temperature, dislocation density, and texture. The model is based on vacancy and interstitial behavior with respect to straight dislocations, dislocation loops, depleted zones and grain or sub-grain boundaries. Results indicate very little growth dependence on temperature or neutron flux at temperatures below ~320°C, at fluxes above ~10¹³ n/cm² sec, and at fluences below 10²¹ n/cm². As the flux is lowered and the temperature and fluence are raised, the temperature and flux dependencies increase. Comparison between theory and data is given for both growth and dislocation loop size.     

Effect of Cr on the mechanical properties and microstructure of Fe-Cr model alloys after n-irradiation

High-chromium ferritic-martensitic steels are candidate structural materials for high-temperature applications in fusion reactors and accelerator driven systems (ADS). Cr concentration has been shown to be a key parameter which needs to be optimized in order to guarantee the best corrosion and swelling resistance, together with the minimum embrittlement. The behavior of Fe-Cr model alloys with different Cr concentrations (0, 2.5, 5, 9 and 12 wt%Cr) has been studied. Tensile tests have been performed in order to characterize the flow properties in the temperature range from −160 °C to 300 °C. The trend of the yield strength with temperature shows that the strain hardening is the same for all materials at low temperatures, even though they have different microstructures. The same materials have been neutron-irradiated at 300 °C in the BR2 reactor of SCK·CEN, up to three different doses (0.06, 0.6 and 1.5 dpa). The results obtained so far indicate that even at these low doses, the Cr content affects the hardening behavior of Fe-Cr binary alloys. Using the Orowan mechanism, the TEM observed microstructure provides an explanation of the obtained hardening but only at the very low dose, 0.06 dpa. At higher doses, other hardening mechanisms are needed.     

Axial strains in fuel cladding associated with creep and fast neutron irradiation

Commercially produced CANDU-PHWR (CANada Deuterium Uranium-Pressurized Heavy Water Reactor) Zircaloy-4 and Zr—2.5% Nb fuel cladding was biaxially creep-tested in the laboratory and in the WR-1 reactor. The axial strains measured have been interpreted, through a knowledge of the texture, to provide evidence supporting prismatic slip as the major process for axial contraction at high stresses and temperatures and that this is the same process which gives rise to axial lengthening of pressure tubes. At low stresses and temperatures, axial lengthening of fuel cladding appears to be associated with Coble creep in elongated and flattened grains.     

Electrical resistivity recovery in Fe-Cr-Ni alloys after neutron irradiation at low temperature

Isochronal and isothermal annealing processes of austenitic Fe-Cr-Ni alloys and Type 316 steel are studied by electrical resistivity after fast neutron irradiation up to $\text{2.9 × 10}{}^{21}\text{n/m}{}^2$ at 5 K and after quench from 1473 K. Little resistivity change is observed below 80 K in the irradiated specimens and below 280 K in the quenched ones. The resistivity variation of the annealing curves in the irradiated specimens is formed by the resistivity decrease due to defect annihilation and the resistivity increase due to structural change produced by the self-interstitial and vacancy migration. Self-interstitials and vacancies migrate above 100 K and 300 K, respectively. The radiation-induced reesistivity consists of two contributions; the resistivity increase due to defect formation and the resistivity decrease due to disordering of solute atoms by cascade collision during neutron irradiation.