Investigation of neutron converters for production of optically stimulated luminescence (OSL) neutron dosimeters using Al2O3:C

This paper presents the optically stimulated luminescence (OSL) properties of neutron dosimeters in powder and in the form of pellets prepared with a mixture of Al₂O₃:C and neutron converters. The neutron converters investigated were high density polyethylene (HDPE), lithium fluoride (LiF), lithium fluoride 95% enriched with ⁶Li (⁶LiF), lithium carbonate 95% enriched with ⁶Li (⁶Li₂CO₃), boric acid enriched with 99% of ¹⁰B and gadolinium oxide (Gd₂O₃). The proportion of Al₂O₃:C and neutron converter in the mixture was varied to optimize the total OSL signal and neutron sensitivity. The neutron sensitivity and dose-response were determined for the OSL dosimeters using a bare ²⁵²Cf source and compared to the response of Harshaw TLD-600 and TLD-700 dosimeters (⁶LiF:Mg,Ti and ⁷LiF:Mg,Ti). The results demonstrate the possibility of developing an OSL dosimeter made of Al₂O₃:C powder and neutron converter with a neutron sensitivity (defined as the ratio between the ⁶⁰Co equivalent gamma dose and the reference neutron absorbed dose) and neutron–gamma discrimination comparable to the TLD-600/TLD-700 combination. It was shown that the shape of the OSL decay curves varied with the type of the neutron converter, demonstrating the influence of the energy deposition mechanism and ionization density on the OSL process in Al₂O₃:C.     

Post irradiation examination of uranium inclusions in inert matrices.

The inert matrix materials CeO₂, MgO, Y₂O₃, MgAl₂O₄ and Y₃Al₅O₁₂ were selected as candidates for inert matrices for the EFTTRA²-T3 neutron irradiation experiment. Most targets contain 20% enriched ²³⁵U fissile inclusions with an average size of roughly 150 μm. The volume fraction of the fissile phase is either 2.5 vol% UO₂ or 19.6 vol% of Y_5.78UO_x in the inert matrices. The samples were irradiated for 198.9 full power days in the High Flux Reactor in Petten. The calculated burn-up is between 17.3 and 19.5% FIMA. The temperature of the cladding was kept at 600 ± 25 K. A dimensional change of at least +5 % is measured after neutron irradiation for Y₃Al₅O₁₂ and MgAl₂O₄ with macro dispersions of UO₂; the other targets with a macro dispersion of UO₂ show a volume change of less than + 1 vol%. The fractional release of the fission gas Xe is more than 40% for the MgAl₂O₄ and Y₃Al₅O₁₂ matrices with a macro dispersion of UO₂, the other targets show a fractional release of Xe of less than 15%. Cracks are observed in MgO and MgAl₂O₄ targets which is possibly related to the stress caused by swelling of the UO₂ inclusions.     

Measurement of the cross sections for the reactions Cr(n,2n)Cr, Zn(n,2n)Zn, Y(n,2n)Y and Zr(n,2n)Zr from 13.5 to 14.8 MeV

Cross section measurements for the reactions ⁵²Cr(n,2n)⁵¹Cr, ⁶⁶Zn(n,2n)⁶⁵Zn, ⁸⁹Y(n,2n)⁸⁸Y and ⁹⁶Zr(n,2n)⁹⁵Zr were carried out in the neutron energy range 13.47–14.79 MeV applying the activation technique. Neutrons were produced via the T(d,n)⁴He reaction, making use of the variation of neutron energy with the emission angle. The neutron fluences incident on the samples were determined relative to the well-evaluated cross section for the reaction ⁹³Nb(n,2n)^92mNb.

The induced γ-ray activities of the irradiated Zn, Zr and Y₂O₃ samples and their monitor foils were measured by means of a calibrated Ge(Li) γ-ray detector at the KFI, Debrecen. At the IRK, Vienna, relative γ-ray measurements using a high-purity Ge detector were combined with integral γ-ray counting by means of a NaI(TI) well-type detector on the Cr, Zn and Zr foils of highest activity and on some Nb monitor foils; integral γ-ray counting only was applied in the case of the Y₂O₃ samples. All necessary corrections were taken into account.

The results are compared to the corresponding results of cross section measurements published in the literature. The uncertainties obtained in this work are considerably smaller in most cases than the uncertainties given by other authors.     

Neutron beam design for both medical neutron capture therapy and industrial neutron radiography for TRIGA reactor

Neutron beam designs were studied for TRIGA reactor with a view to generating thermal, epithermal and fast neutron beams for both medical neutron capture therapy (NCT) and industrial neutron radiography (NR). The beams are delivered from thermal and thermalizing columns, and also horizontal beam hole. Several prospective neutron filters (high-density graphite (G), bismuth (Bi), single-crystal silicon (Si), aluminum (Al), aluminum oxide (Al₂O₃), aluminum fluoride (AlF₃) and lead fluoride (PbF₂)) were examined for obtaining sufficiently intense neutron beam for various applications. Monte Carlo calculations indicated that with a suitable neutron filter arrangement, thermal and epithermal neutron beams attaining 2×10⁹ and 7×10⁸ n cm⁻²S⁻¹, respectively, could be obtainable from thermal and thermalizing columns with the reactor operating at 100 kW. These neutron beams could be adopted for boron neutron capture therapy. Compared with these columns, horizontal beam port would deliver neutron fluxes of 10⁻² 10⁻³ lower intensity, but produced thermal and neutron beams would be adequate for different application of nondestructive inspection by neutron radiography.     

316NG不锈钢在硼-锂溶液中的电化学腐蚀行为研究

利用电化学方法获得316NG不锈钢（316NG）在300℃、pH=5~8硼-锂溶液中的极化曲线和交流阻抗（EIS），并绘制相应水化学条件下的电位-pH图。结果表明:碱性条件下钝化区尤其是二次钝化区极化电流急剧减小，pH=7~8时阳极极化表现出3次钝化现象，偏碱性条件极化阻抗显著高于偏酸性和近中性条件，说明碱性条件下316NG表面钝化膜保护效果更佳。pH=5时电化学极化后样品表面主要生成Cr₂O₃和Fe₂O₃；碱性条件下（pH=6~8）样品表面氧化膜为分层结构:最外层为Fe₃O₄，随深度增加开始出现NiFe₂O₄，内层成分主要为FeCr₂O₄。随着电导率升高，溶液电阻、电荷传递电阻和钝化膜电阻均显著降低。依据极化曲线绘制的电位-pH图与文献结果相吻合。      

掺杂Cr2O3对UO2芯块烧结行为的影响

提高燃料燃耗的一个有效手段是通过增大UO₂晶粒尺寸来减少元件内部气体压力，在大晶粒UO₂芯块中，裂变气体到达晶界表面的距离增加，因而裂变气体的释放速率降低，元件内部气体压力的增高缓慢。本文研究了添加Cr₂O₃对UO₂晶粒尺寸的影响。对纯UO₂、添加0.5% Cr₂O₃及5% Cr₂O₃ 3种配方的芯块进行了试验，在5%H₂Ar保护下，以10 ℃/min和5 ℃/min的升温速率升温至1 700 ℃，然后烧结2 h或4 h，对比纯UO₂芯块与添加Cr₂O₃的芯块发现，添加Cr₂O₃可有效增大晶粒尺寸；较长的烧结时间可促进晶粒长大；较低的升温速率也使晶粒长大。烧结过程产生液相烧结，液相浸润晶粒边界，促进晶粒长大。     

Post-irradiation examination of uranium-doped rock-like oxide fuels

Post-irradiation examinations of rock-like oxide fuels were performed in JAERI to evaluate irradiation behavior and geochemical stability. Five kinds of fuels were prepared using 20% enriched U instead of Pu; a single-phase fuel of an yttria-stabilized zirconia containing UO₂ (U-YSZ), two particle-dispersed type fuels of U-YSZ particles + MgAl₂O₄/Al₂O₃ powder, two homogeneously blended type fuels of U-YSZ powder + MgAl₂O₄/Al₂O₃ powder. The fuels were irradiated in JRR-3 for about 100 days and estimated irradiation conditions were as follows; linear power was 15 kW m⁻¹, thermal neutron fluence was 7 x 10²⁴ m⁻² and fuel temperatures at the surface were 740–1130 K. From the results of non-destructive examinations, the stainless steel cladding surfaces were partially discolored by oxidation and no remarkable deformation of the pins was observed. Significant pellet fragmentation was not observed in spite of the crack formation as observed in irradiated LWR UO₂ fuels. Nonvolatile FPs were observed only in pellets but volatile Cs moved partly to the plenum. From these examinations, no significant difference in macroscopic irradiation behavior was distinguished among 5 fuels.     

GdI3:Ce闪烁体中子探测性能测试与改进研究

为改善GdI₃:Ce闪烁体在探测中子过程中的γ抑制能力，使用Geant4和XCOM计算了其γ线性吸收系数，并通过模拟计算与实验测量研究了铅屏蔽法抑制γ的有效性。结果表明：GdI₃:Ce闪烁体在探测中子过程中易受低能γ射线的干扰；随着铅层厚度的增加，100 keV~1 MeV的γ射线对中子探测的干扰减小，而3~10 MeV的γ射线的干扰呈先增加后减小的趋势。对²⁵²Cf中子源的实验测试发现，在碘化钆闪烁体外围添加铅层后，中子峰得以显现；随着铅层厚度的增加，中子峰内净计数减小，而净计数与本底计数的比值上升。模拟和实验结果均表明，在使用GdI3:Ce闪烁体探测中子时，应根据中子探测效率和信噪比的优化确定γ屏蔽铅层的厚度。     

添加纳米Y2O3的ODS-HT9钢的显微结构和性能

为了研究纳米Y₂O₃对HT9钢的显微结构和力学性能的影响，采用粉末冶金工艺，制备了纳米Y₂O₃含量为0.1%~0.9%的ODS-HT9钢样品，测定了样品的抗拉强度、伸长率、维氏硬度等力学性能，利用透射电子显微镜（TEM）观察和分析了样品中纳米Y₂O₃颗粒的分布状况、形状和相结构，利用扫描电子显镜（SEM）观察了样品拉伸断口的形貌。研究表明，球磨和热压烧结后，纳米Y₂O₃颗粒能够均匀地分布于基体中，相结构和形状未发生明显变化。弥散分布的纳米Y₂O₃硬质颗粒，具有明显的弥散强化作用，导致ODS-HT9钢的抗拉强度和维氏硬度随Y₂O₃含量的增加而显著增加，伸长率显著降低。Y₂O₃含量低于0.7%时，样品以韧性断裂为主，进一步增加含量，断裂方式将由韧性断裂转变成脆性断裂。纳米Y₂O₃含量为0.3%~0.5%的ODS-HT9钢，抗拉强度达到了913~936 MPa，伸长率为10.7%~11.2%，具有良好的综合力学性能。本文研究结果有助于ODS-HT9钢高温性能的研究及其在反应堆中的实际应用。      

Neutron-irradiation effect in ceramics evaluated from macroscopic property changes in as-irradiated and annealed specimens

Macroscopic length (linear swelling) and thermal diffusivity changes were measured for heavily neutron-irradiated -Al₂O₃, AlN, β-Si₃N₄ and β-SiC that were irradiated under the same capsule to compare the difference between these materials. And in addition, several capsules were irradiated under different temperatures (646–1039 K) and to different neutron doses (0.4–8.0 × 10²⁶ n/m²) in the Japanese experimental fast reactor JOYO. The swelling and the thermal diffusivity of as-irradiated specimens showed some dependence on the neutron-irradiation dose and the irradiation temperature, and that indicates stability under neutron-irradiation environments. Alpha-Al₂O₃ and AlN showed relatively large swelling and degradation of thermal diffusivity than β-Si₃N₄ and β-SiC. This difference is related to the crystal structure of each material. The dependence of swelling on irradiation dose, that is, -Al₂O₃ showed linear inclination but β-Si₃N₄ and β-SiC showed saturation, supports the model of defect structures. In addition, annealing behaviors of swelling and thermal diffusivity were compared to analyze the behavior of defects at higher temperature.     

Low-temperature diffusion in inert-gas bombarded Al2O3

Previous work on diffusion in inert-gas bombarded Al₂O₃ has revealed the presence of four diffusion processes, of which two take place well below the temperatures for self-diffussion, one agrees with self-diffusion, and one occurs at temperatures well above those for self-diffusion. The present work serves to explore in greater detail the two low-temperature processes. It is shown that the first, which is found in -Al₂O₃, Al(OH)₃, and γ-Al₂O₃beginning at about 100° C, is consistent with a range of ΔH's of 28 to 50 kcal/mole. The mechanism of the process is hinted at by the fact that it overlaps the temperatures both for Al(OH)₃decomposition and for point-defect motion in -Al₂O₃; the correlation with point defects is believed, however, to be the more significant. The second process, which is found only in -Al₂O₃ beginning at 500–650° C, implies an essentially single ΔH lying between 69 and 79 kcal/mole. It was suggested previously by Matzke and Whitton on the basis of electron diffraction that the process could be attributed to the amorphous-crystalline transition of -Al₂O₃. Further aspects of low-temperature diffusion in Al₂O₃ were revealed by comparing autoradiographs of specimens of -A₂O₃which were bombarded to various doses and then either heated to 850° C or immersed in unheated 12N NaOH. Thus regions exposed to a high dose and which would be expected to be amorphous, had an increased sticking factor, a greater tendency to lose gas during heating, and an enhanced chemical reactivity.     

含钆中子屏蔽材料的设计及其力学性能分析

设计了一种具有良好中子屏蔽能力、高强度及高韧性的新型中子屏蔽材料，用于吸收核电站乏燃料储存格架和乏燃料运输过程中的热中子辐射。材料通过蒙特卡罗粒子传输(输运）软件MCNP进行设计，并通过放电等离子烧结设备及热轧的方式制成了板材。MCNP模拟结果及材料热中子屏蔽测试结果表明：铝基Gd₂O₃复合材料的热中子屏蔽性能与铝基碳化硼相当。Gd₂O₃颗粒球磨后呈现μm、亚μm级甚至有些颗粒达到了nm级。随球磨时间的增加，材料的力学性能逐渐增强。X射线衍射检测发现了钆-铝合金相的生成。经TEM分析表明：材料的强化机制主要是位错强化和nm级Gd₂O₃颗粒的弥散强化，拉伸强度和伸长率分别达到了240 MPa和16%，其断口主要为韧性断裂。     

Prediction of (n,3He) cross-sections around 14 MeV

A comprehensive review of the neutron-induced cross-sections for (n,³He) reactions has been made for the interval of 14⩽Z⩽84 around 14 MeV neutron energy. For practical purposes, an empirical expression has been found by using the experimental (n,³ He) cross section values as a function of (N-Z) and (E_n-E_th) where (N-Z) is the neutron excess of the target nucleus, E_n and E_th are the incident neutron energy and the (n,³He) threshold energy, respectively. The derived empirical relation gives a good fit with the experimental values     

Mo在Al2O3上的吸附、解吸性能及其与U、Sr、Cs、I等杂质元素的分离

目前国际⁹⁹Mo面临供应危机，急需新技术和新反应堆。医用同位素生产堆是以²³⁵UO₂(NO₃)₂溶液为燃料的专用反应堆，生产成本低、三废少、经济效益高。本工作利用Al₂O₃为分离材料，从模拟的医用同位素生产堆（MIPR）燃料溶液中分离和纯化Mo。结果表明，经两次分离，Mo的总回收率大于 60%， U、Sr、Cs、I等杂质可以被除去，采用Al₂O₃从MIPR燃料溶液中提取⁹⁹Mo工艺可行。     

铁磷摩尔比对含锆铁磷酸盐玻璃陶瓷结构和化学稳定性的影响北大核心CSCD

以P_(2)O_(5)-Fe_(2)O_(3)-Al_(2)O_(3)-Na_(2)O四元玻璃作为基础,掺入x=6%(摩尔分数)的ZrO_(2),改变基质玻璃的Fe/P摩尔比(r(Fe/P)),采用熔淬法制备一系列玻璃陶瓷样品,研究含ZrO_(2)的玻璃陶瓷结构和化学稳定性随Fe/P摩尔比的变化。X射线衍射用于物相分析,拉曼光谱、X射线光电子能谱和穆斯堡尔谱用于结构分析。Fe/P摩尔比小于0.18时,样品中Fe^(3+)的相对含量(n(Fe^(3+))/(n(Fe^(2+))+n(Fe^(3+)))摩尔比)随着r(Fe/P)增加而增多;r(Fe/P)介于0.18~0.32时,n(Fe^(3+))/(n(Fe^(2+))+n(Fe^(3+)))在60%左右波动。r(Fe/P)介于0.2~0.3时,样品的化学稳定性最好。     

Design of a gadolinia bearing mixed-oxide fuel assembly for pressurized water reactors

A study on neutronics design of a gadolinia (Gd₂O₃) bearing mixed-oxide (MOX) fuel assembly (MOX-UO₂ (Gd₂O₃) assembly) was performed for the purpose of suppressing the use of fresh lumped burnable poison rods (BPRs). The MOX-UO₂ (Gd₂O₃) assembly investigated consists of MOX and UO₂ (Gd₂O₃) fuel rods, which have already been verified through both fabrication and irradiation experiences. In all, 16 UO₂ (10 wt% Gd₂O₃) fuel rods are located at every corner and the peripheral region of the MOX-UO₂ (Gd₂O₃) assembly in order to reduce the power peaking of MOX fuel rods due to the thermal neutron inflow, and to reduce the reactivity penalty at the end of cycle (EOC). Since fresh BPRs are not expected to be inserted and UO₂ (Gd₂O₃) fuel rods are located at every corner of the assembly, the number of splits in plutonium (Pu) content can be only two, which is less than three splits required for a standard MOX assembly. Core characteristics of an equilibrium core loaded with MOX-UO₂ (Gd₂O₃) assemblies are evaluated and it is verified that adoption of the MOX-UO₂ (Gd₂O₃) assembly is effective to avoid the use of fresh BPRs with securing both the core safety and cycle length. The simplication of the splits in Pu content is also supposed to be beneficial, since it has the possibility of reduce MOX fuel fabrication costs.     

Na_2O/Al_2O_3摩尔比对模拟高放废液硼硅酸盐玻璃固化体结构和性能的影响

研究了Na2O/Al2O3摩尔比(n)对模拟高放废液硼硅酸盐玻璃固化体结构和性能的影响。利用红外光谱分析了不同Na2O/Al2O3摩尔比时硼硅酸盐玻璃固化体的结构变化,并用溶解速率法(DR)和全谱直读等离子发射光谱(ICP-OES)表征了所制备出固化体的化学稳定性。结果表明:在研究组分范围内,当n1.0时,硼硅酸盐玻璃固化体结构中Al以[AlO4]四面体的形式存在,但[BO3]三角体的量较大;随着Na2O/Al2O3摩尔比的增加(n=1.0),固化体结构中[BO3]三角体向[BO4]四面体转变,Al仍以[AlO4]四面体的形式存在,固化体结构稳定性增加;Na2O/Al2O3摩尔比继续增加(n=1.5或2.0),固化体成分中由于Al含量已很少而使[AlO4]含量过少,对固化体结构网络致密性的影响起主要作用,且此时成分中存在过多的碱金属离子在结构中起断网作用,玻璃固化体网络结构变疏松。在Na2O/Al2O3摩尔比为1.0时,玻璃固化体有相对较佳的结构稳定性和化学稳定性,浸泡56d后的失重速率为10-9 g/(cm2·min)数量级,且浸出液中各浸出离子的平均浓度最低。     

Effects of xenon implantation in spinel-zirconia/ceria composites

Two prethinned spinel specimens containing either Y_0.15Zr_0.85O₂ or Ce_0.5Zr_0.5O₂ particles were implanted with 200–400 keV Xe ions at 873 K using the IVEM-Tandem Facility at Argonne National Laboratory. In situ transmission electron microscopy (TEM) was conducted during the implantation in order to follow the evolution of the microstructure. At an ion fluence between 2.4x10²⁰ to 3x10²⁰ m⁻² (up to 50 dpa and 4.7 at %), large Xe bubbles of 50–100 nm developed at the boundaries of the small oxide particles, while a high density of dislocation loops (up to 8 nm in diameter) and much smaller bubbles (up to 4 nm in diameter) formed in the spinel matrix. No large bubbles were observed at the boundaries between the spinel grains. These results suggest that the boundaries between spinel and oxide particles are preferred sites for fission gas accumulation.     

DBD plasma assisted atomic layer deposition alumina barrier layer on self-degradation polylactic acid film surface

In this work, the plastic of polylatic acid(PLA) film is coated by alumina(Al_2O_3)through dielectric barrier discharge plasma assisted atomic layer deposition(DBD PA-ALD) for the proposal of the barrier property enhancement. The influence of ALD Al_2O_3 thickness on properties of barrier, mechanical, optical and degradation is investigated in detail. It is obtained that the growth rate of Al_2O_3 in DBD PA-ALD is as quick as 0.12 nm/cycle. After coated～40 nm Al_2O_3, the water vapor transmission rate of PLA is reduced by two orders of magnitude.Additionally, it is noticed that the tension strength of the coated film is improved slightly,whereas the light transmission rate is decreased with the increase of Al_2O_3 thickness. The degradation test shows that Al_2O_3 coating almost does not affect the self-degradation rate of PLA film.