Neutron Dosimetry with Boron-Doped CR-39 Plastic

As a highly sensitive, simple and non-radioactive neutron dosimeter, CR-39 plastic plates doped with a boron compound: ortho-carborane were prepared. After thermal neutron irradiation, the plates were etched in an aqueous solution of 30%KOH, at 60°C for 2 to 16 h. The etch-pits generated by ¹⁰B(n, α)⁷Li reactions were then counted using an optical microscope or an automatic track counting system. The density of the etch-pits on an irradiated plate increases with the etching time. When the etching time is kept constant, the etch-pit density is proportional to the irradiated thermal neutron fluence. The proportional constant is termed “sensitivity”, which is 4.2 x 10^?4 for a plate containing ortho-carborane at a concentration of 0.5% by weight and for etching time of 16 h. By considering background counts, a thermal neutron dose of 0.025 mSv can be measured with this plate. The plates are insensitive to visible-, UV-, X-, β- and γ-rays and are easy to handle because the detector and converter are incorporated. There is no possibility of underestimating the dose equivalent due to fading. Furthermore, the isotopes of boron are not radioactive and thus are radio- logically safe.     

Confirmation of Track Formation in Cellulose Nitrate Detector by Lithium Particles from 10B(n, α)7Li Reaction

Two experiments were performed, which confirmed that the lithium particles produced by the ¹⁰B(n, α)⁷Li reactions form etchable latent tracks in cellulose nitrate resin. The first experiment was made on a pair of resin detector plates sandwiching between them a boron-bearing layer. Upon irradiation and etching, the tracks appearing on the two plates were found to constitute matching pairs, some of which appeared to be shorter than others, suggesting their correspondence respectively to the ⁷Li and α particles generated by the ¹⁰B(n, α)⁷Li reaction. This surmise was further confirmed by a second experiment, in which the pairing of etch-pits was examined between the two faces of a film detector of thickness below the range of a but above that of ⁷Li particles. The ratio between paired and single etch-pits counted in a unit area of the two faces of film after irradiation and etching proved to be quite close to the value derived by calculation based on the assumption that ⁷Li as well as α particles leave etchable tracks on the resin with 100% efficiency, whereas the observed ratio distinctly differed from the corresponding calculation undertaken assuming that α particles alone contributed to the track formation.

This finding is consistent with the estimations of etchable track formability obtained by applying criteria given in three independent theoretical studies found in published literature.     

Evaluation of Neutron Cross Sections of 241Pu in Unresolved Resonance Region

Abstract

Damage structure produced in stainless steel by “in-situ” irradiation with 1 MV electron in a high voltage electron microscope has been observed as a function of irradiation temperature. Small hardly resolvable dislocation loops were produced by the irradiation at 300°C and below. Hexagonal-, diamond- and irregular-shaped loops were observed at 400°C irradiation. During prolonged irradiation the diamond-shaped loops directly developed into dislocation network, while the other faulted hexagonal and irregular shaped loops grew to be dislocation networks after unfaulting. At 500°C irradiation, dislocation loops grew rapidly to be dislocation lines and void nuclei appeared with strain field images. On further irradiation, the void grew larger with little increase in the density. Grain boundary migration toward or away from the highly damaged region was observed at 500°C irradiation.     

Void Swelling of Solution-Annealed Type 316 Stainless Steel under Long Time Proton Irradiation

Solution-annealed type 316 stainless steel was irradiated by 150 keV proton to a dose of about 6 dpa at the irradiation temperature ranging 450–700°C. To examine the effect of aging during irradiation, the present proton irradiation was carried out for about 25 h at a low dose rate of 7×10–^?5dpa/s. The specimens without He preinjection showed much smaller void swelling than those preinjected with He to the content of 10 at.ppm. Similarly to the case of neutron irradiations, the void swelling in the He preinjected specimens showed the temperature dependence with double peaks, and the peak swelling temperatures were about 550 and 650°C. In these specimens with He preinjection. void number density decreased and average void diameter increased with the increase of irradiation temperature in the range of 450–600°C, but these trends were reversed between 600 and 650°C. The volume of the grain boudary M₂₃C6 precipitates increased with the increase of irradiation temperature from 600 to 700°C, and it was concluded that the decrease of soluble carbon due to the precipitation of M₂₃C6 caused the second swelling peak at 650°C.     

Formation of platinum nanoparticle colloidal solution by gamma-ray irradiation

Platinum nanoparticle colloidal solution with a small amount of impurities was formed from a suspension of hexahydroxy platinic acid (SHHPA) by gamma-ray irradiation to suppress changes in water chemistry such as electrical conductivity and concentration of impurities in the reactor water during noble metal chemical addition in plant operation. The SHHPA was prepared from sodium hexahydroxyplatinate solution by using an H-type cation exchange resin. Optimum conditions for formation of the platinum nanoparticle colloidal solution were the following: absorbed dose of gamma-ray irradiation, >6 kGy; pH of solution, >8.2; air saturation; no methanol addition. Characteristics of the formed platinum nanoparticles were as follows: mean particle size, 2.3 ± 0.5 nm; particle charge, negative; isoelectric point at a pH of 3.5 ± 0.1; the chemical compound consisted mainly of platinum dioxide without platinum metal. No precipitation of platinum nanoparticles was observed after storage time of 1 year without any stirring in a room where the temperature varied from about 10 to about 35 °C.     

Three Phase Region in Ternary System UO2-PUO2-PU2O3

The thermal decomposition of PuF₆ in the temperature range from 100° to 160°C was studied. At 100° and 120°C, no decomposition was detected. At 140° and 160°C, the rate of decomposition during reaction lasting up to 20 hr, and the dependence of the decomposition on the initial PuF₆ pressure were determined. The results are explained by assuming that the rate of decomposition is proportional to the 0.4-th power of PuF₆ pressure. The activation energy of the reaction was found from calculation to be 14.8 kcal/mole.     

国产CR-39记录高注量的He++和H+的蚀刻特性

研究了国产ＳＹ－１型ＣＲ－３９记录加速器加速的高注量的Ｈｅ＾＋＋和Ｈ＾＋的蚀刻特性,对辐照过程中靶室内的环境本底做了观察实验。结果表明,在靶室内直接照射束斑位置以及远至４０ｍｍ处仍能记录到与束斑处相同的离子,而且在高注量离子的束斑区蚀刻后几乎辨不出径迹。进行了刻度实验,给出了ＣＲ－３９记录不同能量的Ｈｅ＾＋＋和Ｈ＾＋的径迹直径响应蚀刻曲线,对高注量束斑区蚀刻后注入新离子的实验表明记录特性不改变。     

Dose and Temperature Dependence of Void Swelling in Electron Irradiated Stainless Steel

In a high voltage electron microscope, solution treated Type 316 stainless steel was electron-irradiated at temperatures in the range of 370–630°C to a dose of about 30 dpa. The swelling (ΔV/V) induced by the irradiation beyond about 5 dpa is well described by an empirical equation, ΔV/V=A(dpa) ⁿ , under constant void and dislocation densities. With increasing irradiation temperature, the fluence exponent n increases and the pre-exponent term A decreases. At 550°C irradiation, the fluence exponent takes the value of 1.5 due to the diffusion-limited void growth. The value of n larger than 1.5 at higher temperature (>550°C) is attributable to the surface reaction-limited void growth. The smaller value of n for the low temperature (?500°C) irradiation appears to arise from the dislocation assisted vacancy diffusion. The peak swelling temperature of the specimen irradiated to 30 dpa is about 570°C, which shifts to a higher temperature with increase in electron dose.     

The fission gas bubble distribution in a mixed oxide fast reactor fuel pin

The fission gas bubble distribution has been studied in a mixed oxide fast reactor fuel pin irradiated in DIDO MTR to 2.8% burn-up at centre and surface temperatures of 2000 and 1000°C. The intragranular fission gas bubbles are very small (<6 nm diameter) and this is a consequence of the high re-solution rate at fast reactor ratings. The bubbles nucleate heterogeneously and linear arrays of bubbles, due to nucleation on fission tracks, are observed up to irradiation temperatures of 1900°C. At 1980°C ~4% of the fission gas produced is present in intragranular bubbles. There is no definite evidence for gas bubble mobility or coalescence. Apart from any effects of columnar grain growth fission gas release in fast reactor fuel pins seems to occur predominantly by the diffusion of single gas atoms, at least up to irradiation temperatures of 2000°C.     

Attachment and spreadout study of 3T3 cells onto PP track etched films

Polymer surface modifications are obtained by the application of radiation treatments and other physico-chemical methods: fission fragment (ff) irradiation and etching. The biocompatibility of the surface is then observed by cell seeding and cell adhesion experiments. Approaches to improvement of the cell adhesion are obtained by different methods: for example, in PS, cell adhesion is improved after ion implantation; in PMMA, after bombarding the polymer, the surface is reconditioned with surfactants and proteins and in PVDF, cell adhesion is assayed on nuclear tracks membranes. In this work, we obtained important cell adhesion improvements in PP films by irradiation with swift heavy ions and subsequent etching of the nuclear tracks. We use BOPP (isotactic −25 μm thickness). Irrradiations were performed with a Cf-252 californium ff source. The source has a heavy ff and a light one, with 160–200 MeV energy divided among them corresponding to ff energies between 1 and 2 MeV/amu. A chemical etching procedure consisting of a solution of sulphuric acid and chromium three oxide at 85 °C was used. The 3T3 NIH fibroblast cell line was used for the cell adhesion experiment. Here we report for the first time, the results of a series of experiments by varying the ff fluence and the etching time showing that attachment and spreadout of cells are very much improved in this cell line according to the number of pores and the pore size.     

Synthesis of Amino Acids by 60Co Gamma Irradiation

Measurements of creep by the indentation with Knoop indenter were performed for sintered UO₂ pellets before and after irradiation and a non-irradiated single crystal with an applied load of 50 g. Room temperature creep of non-irradiated sinters included grain boundary sliding. The creep at the lowest dose (7.8×10¹⁴ fiss./cm³) also involved the grain boundary sliding, while no creep was observed after irradiation of 2.9×10¹⁶ fiss./cm³. Irradiated sinters at a dose between 1.4×10¹⁶ and 1×10¹⁸ fiss./cm³ yielded not only primary creep without the grain boundary sliding but also secondary creep, in which cracks were formed when a dose was less than 4.8×10¹⁷ fiss./cm³. At a dose of 1×10¹⁸ fiss./cm³ the enhanced secondary creep independent of the temperature between 100 and 150°C was observed.     

Transient Behavior of Low Enriched Uranium Silicide Plate-Type Fuel for Research Reactors during Reactivity Initiated Accident Conditions

Abstract

This paper describes the results of transient experiments using a low enriched uranium silicide miniplate fuel for research reactor. The pulse irradiation was performed in the Nuclear Safety Research Reactor (NSRR) at the Japan Atomic Energy Research Institute (JAERI).

The results obtained in this study are summarized as follows :

(1) The tested fuel plates were damaged with energy depositions above 94 cal/g·fuel, but remained intact below 82cal/g·fuel. A failure threshold should therefore exist between these two values.

(2) Four of the fuel plates that showed peak cladding surface temperatures <330°C were damaged by the thermal stress during quenching. These damaged fuel plates revealed small intergranular cracks that propagated perpendicularly to the axial direction of the plate, from the Al cladding surface to the fuel core, without significant dimensional changes. On the other hand, when peak cladding surface temperatures were >400°C, the test fuel plates were damaged mainly by melting of the Al cladding, accompanying significant dimensional changes.

(3) The thermal stress of the damaged fuel plates calculated on the basis of the maximum transient temperature drop during quenching was greater than the tensile stress that occurred during fabrication.     

Influence des conditions de trempe sur le gonflement par irradiation aux neutrons d'un acier austenitique au niobium

Steels of type AISI 316 and 318 which had been subjected to different heat-treatments were neutron-irradiated at 450 °C to a fluence of 5 × 10²¹ n/cm². Density measurements after irradiation reveal a swelling of 0.15% for AISI 316 and 318 steels quenched from 1050 °C and tempered at 700 °C. A homogeneous dispersion of niobium carbide does not appear to influence the amount of swelling. On the other hand, the AISI 318 steel hyperquenched from 1270 °C and tempered at 700 °C swells to an extent of only 0.025%. Examination by electron microscopy showed that this reduction in swelling is associated with the dispersion of the fine niobium carbide precipitates on extrinsic stacking faults, as a consequence of the heat-treatment. It was noted that voids were absent in the neighbourhood of these stacking faults with their associated precipitates.     

Dissolution of ThO2Based Oxides in Nitric Acid Solutions at Elevated Temperatures

Highly-dense spherical particles of thorium-based oxides, ThO₂ and (Th, U)O₂, prepared by the sol-gel method were subjected to dissolution with nitric acid containing 0–0.05 mol/l NaF at high temperatures above 120°C. The dissolution rate depended upon temperature, fluoride concentration and UO₂ content. High-temperature in the range of 120–200°C enhanced the dissolution of the ThO₂-based fuels. At low temperatures and/or low U0₂ concentrations, insoluble tetrafluoride precipitates were formed on the particle surfaces and they resulted in the decrease of the dissolution rates. In the present study, the apparent activation energies for the high-temperature dissolution were obtained.     

Study on the structural evolution of modified phenol–formaldehyde resin adhesive for the high-temperature bonding of graphite

A novel adhesive for carbon materials composed of phenol–formaldehyde resin, boron carbide and fumed silica, was prepared. The adhesive property of graphite joints bonded by the above adhesive treated at high-temperatures was tested. Results showed that the adhesive was found to have outstanding high-temperature bonding properties for graphite. The adhesive structure was dense and uniform even after the graphite joints were heat-treated at 1500 °C. Bonding strength was 17.1 MPa. The evolution of adhesive structure was investigated. The results indicated that the addition of the secondary additive, fumed silica, improved the bonding performance greatly. Borosilicate phase with better stability was formed during the heat-treatment process, and the volume shrinkage was restrained effectively, which was responsible for the satisfactory high-temperature bonding performance of graphite.     

Thermal Recovery of Radiation Defects and Microstructural Change in Irradiated UO2 Fuels

Abstract

Thermal recovery of radiation defects and microstructural change in UO₂ fuels irradiated under LWR conditions (burnup: 25 and 44 GWd/t) have been studied after annealing at temperature range of 450-1,800°C by X-ray diffractometry and transmission electron microscopy (TEM). The lattice parameter of as-irradiated fuels increase with higher burnup, which was mainly due to the accumulation of fission induced point defects. The lattice parameter for both fuels began to recover around 450-650°C with one stage and was almost completely recovered by annealing at 850°C for 5 h. Based on the recovery of broadening of X-ray reflections and TEM observations, defect clusters of dislocations and small intragranular bubbles began to recover around 1.150–1,450°C. Complete recovery of the defect clusters, however, was not found even after annealing at 1,800°C for 5h. The effect of irradiation temperature on microstructural change of sub-grain structure in high burnup fuels was assessed from the experimental results.     

Determination of alpha-particle track depths in CR-39 detector from their cross-sections and replica heights

A challenging task in the application of solid-state nuclear track detectors (SSNTDs) is the measurement of depths of the tracks. One approach involves breaking and polishing the side of SSNTDs to reveal the cross-sections of the tracks for direct measurements. Recently, surface profilometry was used to measure the heights of the replicas of alpha-particle tracks to give the track depths. In the present work, systematic comparisons among the track depths for alpha-particles with normal incidence and different incident energies were made for these two methods. After irradiation, the detectors were etched in a 6.25 N aqueous solution of NaOH at 70 °C. Both long etching time of 15 h (to produce spherical-phase tracks) and short etching time from 1 to 8 h (to produce sharp-phase tracks) were used. Good agreement was achieved between the two methods for spherical-phase tracks but not for sharp-phase tracks. It has been found that the surface profilometry method only works for replicas for spherical-phase tracks. Replicas for sharp-phase tracks are easier to collapse or deform, so the surface profilometry method may not give correct results.     

Hardening and Recovery of Neutron-Irradiated Fe-Cu-C Alloy

Internal friction measurements were performed on neutron-irradiated and annealed Fe-0.2w/0Cu-0.0066w/0C alloy, in an attempt to gain information on the role of interstitial C atoms in radiation effects. It was revealed that C atoms are trapped by defects produced by neutron-irradiation, and released upon annealing at about 400°C. The radiation hardening induced in the neutron-irradiated Fe-Cu-C alloy recovers by annealing in at least three steps of 150°–250°C, 350°–450°C and above 550°C. The second step corresponds to the release of C atoms from traps.     

Effect of Condition of Recrystallization Heat Treatment on High-Temperature Mechanical Properties of an Irradiated AISI 316 Austenitic Steel

A study was made of the effects brought upon the radiation-induced high-temperature embrittlement of AISI 316 austenitic steel by different conditions to the specimens to produce recrystallization.

Cold worked specimens were recrystallized at temperatures 950°–1,100°C held for periods ranging 2~60 min. The specimens were then irradiated to 1.7×10²¹nvt(< MeV) at 55°C after which they were subjected to tensile testing at 650°C.

In the specimens recrystallized at 950°C, carbide precipitation was observed to have occurred, and these specimens were found less liable to show radiation-induced embrittlement. On the other hand, specimens recrystallized at 1,025° and 1,100°C became completely solution treated, and exhibited severe radiation-induced embrittlement. It is surmisted that the carbides precipitated on the grain-boundaries tend to inhibit propagation of grain-boundary cracks, and hence contribute to lessing radiation-induced embrittlement.     

Hardening and Recovery of Neutron-Irradiated Pressure Vessel Steel (ASTM A533B)

Specimens of ASTM A533B steel were studied to gain information on the annealing process following irradiation, through measurements of internal friction and of hardness.

The specimens were quenched from 900°C and tempered at 650°C, then irradiated in the JMTR reactor at 65°–75°C to a neutron dose of 1.4–1.7×10²⁰ n/cm² (E _n >1MeV).

Peaks were observed on the internal friction curves from unirradiated specimens. These peaks disappeared upon irradiation, but reappeared with annealing treatment at 150°C.

Radiation-anneal hardening was observed at 250°C. The recovery of radiation hardening begins at a temperature between 250° and 350°C, but is not completed even at 550°C.