Dimensional changes and thermal conductivity by annealing and its relation to the defect concentration and stored energy release of neutron-irradiated graphite

The dimensional changes and thermal conductivity with the annealing of fine-grained isotropic graphite IG-110U and ETP-10 irradiated to 0.02 and 0.25 dpa (1.38 x 10²³ and 1.92 x 10²⁴ n/m², E > 1MeV) at a design temperature of <200°C were studied. The irradiated graphite exhibited a small volume expansion and large degradation in thermal conductivity. Post-irradiation annealing experiments were carried out on dimensional changes and thermal conductivity up to 1700°C, and the results were analyzed in terms of changes in the defect concentration of graphite crystals. The rapid recovery of thermal conductivity observed below 200°C in the graphite irradiated to 0.02 dpa was attributed to the annihilation of Frenkel defects, whereas the recovery observed in both dimension and thermal conductivity above 200°C in the graphite irradiated to 0.02 dpa and 0.25 dpa was caused by the annihilation of small interstitial clusters of 4 ± 2 atoms. The role of large clusters of interstitials and vacancies in the changes to smaller dimension than pre-irradiation at high annealing temperatures are discussed. The temperature dependence of stored energy release was estimated from the changes in defect concentration calculated from the recovery of thermal conductivity.     

Thermal conductivity degradation of ceramic materials due to low temperature, low dose neutron irradiation

The thermal conductivity degradation due to low-temperature neutron irradiation is studied and quantified in terms of thermal resistance terms. Neutron irradiation is assumed to have no effect on umklapp scattering. A theoretical model is presented to quantify the relative phonon-scattering effectiveness of the three dominant defect types produced by neutron irradiation: point defects, dislocation loops and voids. Several commercial ceramics have been irradiated with fission reactor fast neutrons at low temperatures to produce defects. Materials include silicon carbide, sapphire, polycrystalline alumina, aluminum nitride, silicon nitride, beryllium oxide, and a carbon fiber composite. The neutron dose corresponded to 0.001 and 0.01 displacements per atom (dpa) for a 60 °C irradiation and 0.01 and 0.1 dpa for a 300 °C irradiation. Substantial thermal conductivity degradation occurred in all of the materials except BeO following irradiation at 60 °C to a dose of only 0.001 dpa. The data are discussed in terms of the effective increase in thermal resistance caused by the different irradiation conditions. Evidence for significant point defect mobility during irradiation at 60 and 300 °C was obtained for all of the ceramics. The thermal stability of the radiation defects was investigated by isochronal annealing up to 1050 °C.     

Interpolation and Extrapolation Method to Analyze Irradiation-Induced Dimensional Change Data of Graphite for Design of Core Components in Very High Temperature Reactor (VHTR)

Graphite materials are used as core components in the High-Temperature Gas-Cooled Reactor (HTGR) and Very High Temperature Reactor (VHTR). The authors prepared technical documents for design, material, products, in-service inspection and maintenance of the graphite components for the HTGR/VHTR, which were summarized as a draft of standard for the graphite components through discussion made in a “Special committee on research on preparation for codes for graphite components in HTGR” set up within AESJ. The draft of standard contains graphical expressions for the irradiated material properties of IG-110 graphite. It is possible to use the graphical expressions for the components design of VHTR. The graphs were obtained based on the interpolation and extrapolation of the irradiation data. The irradiation-induced dimensional change of IG-110 graphite was obtained through the interpolation and extrapolation of the irradiation data with a quadratic equation of fast neutron fluence. The irradiation data for H-451 and ATR-2E graphites were used for the evaluation of the interpolation and extrapolation of irradiation data for IG-110. It was shown in this study that the proposed interpolation and extrapolation method is reasonable for IG-110 with regard to the database available at present.     

Thermal conductivity of highly neutron-irradiated beryllium in nuclear fusion reactors

Beryllium will be used as a neutron multiplier in Helium Cooled Pebble Bed (HCPB) DEMO blankets. The beryllium thermal conductivity is determining the maximum pebble bed temperature and, therefore, is very important for blanket design. Different grades of beryllium discs were neutron-irradiated at temperatures between 343 and 673 K and at fluences up to 1.6 × 10²³ cm⁻². At lower irradiation temperatures a significant drop of the beryllium thermal conductivity occurs even after small neutron fluences. With increasing neutron fluence, further moderate decreases of the conductivity are observed. With increasing irradiation temperature, the thermal conductivity further decreases. If the thermal conductivity of the irradiated beryllium is known, the conductivity of irradiated beryllium pebble beds can be assessed using the model suggested in this study.     

Pressure Recovery in Diffuser for Discharge Flows from Gas Centrifuges Rotating at High Mach Numbers

This paper deals with the irradiation effects on the thermal shock resistance Δ = σ_tk/Eα (σ_t : tensile strength, k: thermal conductivity, E: Young's modulus, α: thermal expansivity) and the thermal shock fracture toughness ?=K_Ick/Eα (K_Ic : fracture toughness value of mode I) in addition to other mechanical properties such as the diametral compressive strength and fracture toughness of two kinds of near isotropic graphite neutron irradiated at 750～1,000°C to a dose of (1.1 ～ 1.5)10²¹ n/cm² (>29 fJ). One of the graphite specimen is an isostatically molded graphite IG-11 with fine grain petroleum coke and the other is a binderless molded graphite HCB-18 which was prepared using mesophase pitch carbon with very fine grain size. These measurements are carried out by means of disk testing method developed by us. Results show that both the thermal shock resistance and the thermal shock fracture toughness of the two kinds of graphite after irradiation decrease considerably in contrast with increasing trends of the usual mechanical strengths.     

Effect of Fast Neutron Irradiation on the Properties of Boron Carbide Pellet

Boron carbide pellets were irradiated in the experimental fast reactor “JOYO” to ¹⁰B burnup of up to 170x10²⁶cap/m³, fluences of 2x10²⁶/m²(E>0.1MeV), and maximum temperatures of about 1,200°C. Post irradiation examinations were made of microstructural changes, helium release, swelling, and thermal conductivity.

Boron carbide pellets irradiated to high burnups developed extensive cracking. Helium release from the pellets was initially low, but enhanced helium release was observed at high burnups and high temperatures. The swelling linearly increased with burnup, and when boron carbide was irradiated at high temperatures, the swelling rate began to decrease corresponding to the beginning of enhanced helium release. The correlation between swelling and the helium release was studied and the swelling was interpreted in terms of accumulation of helium in the boron carbide pellet. The thermal conductivity of the boron carbide pellets decreased rapidly by neutron irradiation accompanied with loss of temperature dependence.     

Fracture mechanical properties and neutron irradiation effects of fuel compacts for the HTTR

To simulate the nuclear fuel for High Temperature Engineering Testing Reactor (HTTR), fuel compact models using SiC-kernel coated particles instead of UO₂-kernel coated particles were prepared under the same conditions as those for the real fuel compact. The mechanical and fracture mechanics properties were studied at room temperature. The thermal shock resistance and fracture toughness for thermal stresses of the fuel compact were experimentally assessed by means of arc discharge heating applied at a central area of the disk specimens. These model specimens were then neutron irradiated in the Japan Material Testing Reactor (JMTR) for fluences up to 1.7 × 10²¹n/cm² (E ·> 29 fJ) at 900°C ± 50°C. The effects of irradiation on a series of fracture mechanical properties were evaluated and compared with the cases of graphite IG-110 used as the core materials in the HTTR.     

氦离子辐照烧结碳化硅损伤效应研究

利用中国科学院近代物理研究所320 kV高压平台提供的氦离子辐照烧结碳化硅,辐照温度从室温到1 000 ℃,辐照注量为1015~1017 cm-2。辐照完成后,进行退火处理,然后开展透射电子显微镜、拉曼光谱、纳米硬度和热导率测试。研究发现,烧结碳化硅中氦泡形核阈值注量低于单晶碳化硅。同时,氦泡形貌和尺寸与辐照温度、退火温度有关。另外,对辐照产生的晶格缺陷、元素偏析进行了研究。结果表明,辐照产生了大量的缺陷团簇,同时氦泡生长也会发射间隙子,在氦泡周围形成间隙型位错环。在晶界处,容易发生碳原子聚集。辐照导致材料先发生硬化而后发生软化,且热导率降低。     

Neutron induced thermal properties changes in carbon fiber composites irradiated from 600 to 1000°C

An experimental study on the neutron irradiation induced thermal properties changes of various carbon fiber composites has been carried out using the High Flux Reactor of Petten. The investigated materials include: A 05 (2D felt-type CFC), CX 2002U (2D felt-type CFC), DMS 678 (2D woven CFC) and N 112 (3D CFC). The irradiation temperatures range from 610 to 1030°C and the neutron doses are between 0.8 to 1.8 dpa. This paper presents the experimental results on the neutron induced changes of the specific heat capacity, thermal diffusivity, and thermal conductivity.     

Fluence and temperature dependence of swelling in irradiated molybdenum

Volume changes have been analyzed in molybdenum which has been neutron irradiated to various fluences over the temperature range 50 to 1300°C. This data together with all previously reported data has been compiled into a three-dimensional plot of swelling versus temperature versus fluence. Significant low temperature swelling, < 400°C, is observed and is consistent with the presence of isolated immobile vacancies. Void swelling only becomes significant at temperatures > 400°C. The nature of the dislocation and void microstructure at high irradiation temperatures are analyzed quantitatively as a function of irradiation temperature and the results are reasonably consistent with a recent model of Brailsford and Bullough. The same model is also consistent with an observed trend towards saturation in the void swelling at high fluences at the low temperature end of the void region.     

Neutron Irradiation Effect on Isotopically Tailored 11B4C

The present investigation was carried out to elucidate the effects of neutron irradiation on the dimensional change and thermal conductivity of isotopically tailored ¹¹B₄C. The specimens used in the present investigation are 99% ¹¹B-enriched ¹¹B₄C, 91% ¹⁰B-enriched ¹⁰B₄C, and β-SiC. ¹⁰B₄C was sampled from an irradiated material used as a neutron absorber in a “JOYO” MK-II control rod. The ¹¹B₄C and SiC specimens were irradiated in the experimental fast reactor “JOYO” to fluences of 1:94 × 10²⁶ n/m² (E > 0:1 MeV) at 530°C and 3:12 × 10²⁶ n/m² (E > 0:1 MeV) for ¹⁰B₄C, with a ¹⁰B burnup of 47:3 × 10²⁶ cap/m³. Measurements on changes in dimensions and thermal conductivity, as well as postirradiation annealing up to 1400°C, were carried out. The results of such measurements indicated that the changes in the dimensions and thermal conductivity of neutron-irradiated ¹¹B₄C were substantially smaller than those of ¹⁰B₄C and SiC. Postirradiation annealing measurements for ¹¹B₄C showed that the thermal conductivity was almost completely recovered at 1400°C. The changes in thermal conductivity by annealing were analyzed in terms ofphonon scattering theory. The onset of recovery of the thermal conductivity of ¹¹B₄C agreed well withirradiation temperature; however, the recovery of length did not coincide with irradiation temperature.     

Development of Pt-Mo Alloy Thermocouples for Incore Temperature Measurement in HTGRs

The Pt-Mo alloy thermocouples with element wires of Pt-5%Mo and Pt-0.1%Mo were developed for incore temperature measurements at a maximum temperature of 1,200°C for a long term and of 1,350°C for a short term in VHTRs. The thermocouples were tested at high temperature conditions of 1,000-1,400°C in an outpile electric furnace and in an incore capsule in the JMTR. The total thermal neutron fluence irradiated was 8.1×10²⁰ n/cm². The changes in their electromotive forces due to those exposures to high temperatures and radiation in the reactor core were evaluated to be less than ±2%. This meets the specifications that is required for the incore temperature measurements in VHTRs. Hence, the developed thermocouples can be used also in the incore temperature measurement in advanced type HTGRs and in incore irradiation experiments in the HTTR, in future.     

Investigation of kinetic recovery process in low dose neutron-irradiated nuclear graphite by thermal annealing

To investigate the kinetic recovery process of low dose neutron-irradiated graphite, nuclear-grade isotropic graphite IG-110U and ETP-10 were neutron irradiated using the JMTR up to 1.38 × 10²³ n/m² (E_n > 1 MeV) at ~473 K. In-situ measurement of macroscopic length was conducted during the isothermal and isochronal annealing process from room temperature up to 1673 K. From room temperature to 773 K for IG-110U, and to 1023 K for ETP-10, macroscopic lengths, lattice parameters, and unit cell volumes of both specimens recovered to their pre-irradiation values, and this recovery process subdivided into two stages. The activation energies of macroscopic volume recovery at 523–673 K and 673–773 K were determined to be ~0.22 eV and ~0.57 eV for IG-110U, respectively; ~0.13 eV and ~2.59 eV at 523–923 K and 923–1023 K for ETP-10, respectively. The migration of not only single interstitials but also interstitials dissociated from submicroscopic interstitial groups along basal planes followed by vacancy-interstitial recombination play a dominant role in the first stage. The second stage is suggested to proceed via the motion of carbon groups along basal planes for IG-110U, and migration of single interstitials along the c-axis for ETP-10. During 773 K or 1023 K up to 1673 K, macroscopic length continuously shrank with decreasing shrinking rate, even with a turnaround to swell at 1173 K for IG-110U.     

Observation and analysis of damage structure in Al and Al/Mg (N4) alloy after irradiation with 100 and 400 keV aluminium ions

A comparative TEM study has been made of ion irradiation damage structure in pure aluminium, commercial aluminium (grade 1100) and in a modified N4 (Al/2.95% Mg) alloy of the type used in the construction of the calandria of the Winfrith prototype SGHW Reactor. Atom displacements equivalent to many years neutron irradiation were simulated by bombardment with 100 and 400 keV Al⁺ ions to doses of up to 200 dpa at temperatures between 30 and 250°C. Dynamic observations of damage formation were made during irradiation with 100 keV ions in a linked heavy-ion accelerator/200 keV electron microscope, and further results were obtained by 400 keV Al⁺ ion bombardment in a Cockcroft-Walton accelerator. Dislocation structure and voids were seen in aluminium irradiated with 100 and 400 keV A1⁺ at temperatures between 30 and 250°C. Void swelling of 8.7% at 104 dpa was a maximum at 1&#x0303;50°C in type 1100 aluminium. No voids were found at temperatures μ 250°C. No voids were seen in the Al/Mg N4 alloy after bombardments up to 200 dpa with 100 keV A1⁺, and 150 dpa with 400 keV Al⁺ at temperatures between 50 and 170°C. The void-resistant property is consistent with observations in the USA of neutron-irradiated 5052 Al alloy which has a similar magnesium content to the modified N4 alloy. The 1100 alloy and N4 results have been analysed using the rate theory of swelling. The absence of voids in the N4 alloy indicates an effective vacancy annihilation mechanism, which possibly occurs at small precipitates formed during irradiation.     

Approach to Improvement of Thermal Efficiency in BWR Plant Based on Exergy Analysis

This paper presents finite element analytical results of the thermal and irradiation induced stresses occurred in the target plate of the divertor structures which are exposed to neutron irradiation at 0-1 dpa with high heat flux up to 15MW/m². Three types of target plate models are studied in the present work, e.g., flat plate, saddle and mono-block type models which have bonded structures of one-dimensional high thermal conductivity Carbon-Carbon (C/C) composite material and Oxygen-Free-High Conductivity Copper (OFHC). With these results, it is concluded that flat model exhibits the longer life-time up to 1 dpa under 15 MW/m², whereas serious troubles, such as fracture in bonded interface or drop-off of armor tile from bonded substrate, are expected in saddle and mono-block type models before 1 dpa. Design change with reduction of the volume of armor region in the model is also effective to prolong the life-time by the reduction of sublimation of armor material.     

ASME规范概率评价方法对细颗粒石墨的适用性研究

石墨由于其高中子散射截面和低中子吸收截面特性,被广泛应用于第四代高温气冷堆中作为慢化剂、反射层和堆芯结构,故保证其结构完整性对反应堆的安全运行非常重要。由于石墨材料强度分散,概率论方法评价其失效较常用的确定论评价方法更为合适。目前,美国ASME规范采用的概率方法主要针对NBG-18这种大颗粒石墨,对我国高温气冷堆核电站工程项目采用的细颗粒石墨IG-110的适用性未知。同时,我国成都碳素生产的高温堆备选石墨NG-CT-01颗粒大小与IG-110相似,也为细颗粒石墨。因此,文章研究ASME规范概率方法对细颗粒石墨的适用性,并通过实验数据加以验证。结果表明,对于细颗粒石墨,ASME规范过于保守,低估了材料的强度性能。     

Anisotropy of coefficient of thermal expansion of nuclear graphite under compressive stresses

The coefficient of thermal expansion (CTE) of nuclear graphite IG-110 and NBG-18 under compressive stresses of 20 MPa, 30 MPa and 40 MPa has been measured by strain gauge method and corresponding anisotropies of CTE under stresses were investigated. With the increasing compressive stresses, the CTE of IG-110 and NBG-18 parallel and perpendicular to the loading directions increased significantly and decreased gradually respectively. The corresponding CTE anisotropies of IG-110 and NBG-18 almost maintain below 1.05 and keep their original near-isotropic properties under compressive stresses maybe due to the homogeneous sensitivity of CTE to the stresses, perfect crystallites in the grains and homogeneous alignment of grains in graphite. The constant isotropic properties of graphite IG-110 and NBG-18 under stresses are beneficial for the integrity and safety of the graphite used in the reactor.     

The effect of neutron irradiation on mechanical properties of YAG laser weldments using previously irradiated material

The objective of this study is to make clear the effect of neutron irradiation on mechanical properties of laser weldments using irradiated material. This estimation is necessary for the application to joining coolant piping of the ITER blanket. Irradiation testing was performed at Japan Material Testing Reactor (JMTR). On the irradiation condition for weldments using irradiated material, fast neutron fluence was 1.4 × 10²⁴ n/m², which corresponds to a displacement damage rate of 0.26 displacement per atom (dpa) and irradiation temperature 200 °C. The results of this study show that tensile properties of all weldments changed into that of base material by the effect of neutron irradiation. The results of hardness tests show that irradiation hardening at an irradiation damage dose of 0.3 dpa is almost same as that at irradiation damage 0.6 dpa. It is concluded that irradiated weldments using irradiated material were moved toward irradiated base material on tensile and hardness properties up to 0.6 dpa. On the other hand, tensile properties of base material were changed by the effect of neutron irradiation up to about 0.3 dpa, and with much less change from 0.3 dpa to 0.6 dpa. It is inferred that the effect of neutron irradiation of SS316LN-IG almost saturated up to 0.3 dpa.     

基于SAXS方法的核石墨微孔结构的温度影响

X射线小角散射(Small Angle X-ray Scattering,SAXS)是研究纳米尺度微观结构的重要手段。本文利用同步辐射SAXS技术测量了25oC、100oC、200oC、300oC和400oC时,IG-110和NBG-18核石墨在纳米尺度范围内孔隙的数量分布及其分形特征的变化。实验结果表明,IG-110和NBG-18核石墨的微观结构中存在微小尺寸上的不均匀区域,且核石墨孔隙的固气结构具有明锐的界面。但随着温度的升高,固气界面的变化并没有呈现出明显的规律性。此外,在纳米尺度上,IG-110和NBG-18核石墨的孔隙数量随温度呈现增加的趋势,且IG-110核石墨孔隙数量的增加幅度大于NBG-18核石墨,其平均孔隙尺寸的减小幅度大于NBG-18核石墨。在核石墨的微孔结构内,其固气界面的分形维数随温度升高逐渐减小,且NBG-18核石墨分形维数的变化幅度小于IG-110核石墨。这表明核石墨的分形结构随温度的升高逐渐光滑。     

Contribution of thermal neutrons to radiation hardening of pure copper

The paper presents the results of an experiment the aim of which was to estimate directly the effect of the thermal neutron fluence on pure copper hardening. Identical specimens were irradiated in two reactors (SM-2 and RBT-6) in the dose range 10⁻³-10⁻¹ dpa at T_irr=80 °C under substantially different, by a factor of 5, thermal neutron fluences, with other irradiation parameters being close. The results show that the elevated thermal fluence in the SM-2 reactor increases the radiation hardening of pure copper by 50% at a dose of about 10⁻³ dpa as compared with specimens irradiated in the RBT-6 reactor. The contribution of thermal neutrons proved to be much more considerable than the theoretical estimates.