Radiation Resistance of Reactor Graphite with Enhanced Operating Properties

Neutron fluence dependences of the dimensional changes, the thermal expansion coefficient, the dynamic elastic modulus, and the maximum strength of samples of high-strength structural graphite GR-1 and GSP-50 after irradiation in BOR-60 at 360–400°C up to fluence 6.5·10²⁶ m^–2 (E > 0.18 MeV) were determined. The shape change of graphite reached the secondary-swelling stage, and the state of the material is characterized by disseminated fracturing.The influence of the initial density in the range 1.58–1.9 g/cm³ is determined. It is shown that GSP-50 graphite, based on pyrolytic carbon matrix, possesses a higher radiation resistance than GR-1 graphite based on a composite filler and granular binder.     

Neutron Irradiation Induced Changes in the Structure and Properties of Glassy Carbon

The results of an investigation of SU-1300 and SU-2000 glassy carbon samples after irradiation in a BOR-60 reactor at 360–400°C are presented. It is established that substantial radiation compression of glassy carbon under irradiation remains up to neutron fluence 3.6·10²⁶ m^–2 (E > 0.18 MeV). It is shown that radiation shrinkage is caused by compaction of packets of graphite-like layers and decrease of molecular porosity. Increasing the production temperature of glassy carbon from 1300 to 2000°C decreases the graphitizability of the material under irradiation.     

Influence of the neutron flux density on the radiation embrittlement of VVér-440/213 vessel materials

In work on minisamples of the fifth complex of the No. 3 unit of the Kola nuclear power plant it is shown that for neutron fluence 410²³ m^–2 (operation for approximately 10 yr), neutron flux density 310¹⁵ sec^–1m^–2 and copper content 0.03% and 0.09% in the metal the shifts of the cold-brittleness temperature are 50 and 120°C, respectively. Under the same irradiation conditions but with neutron flux density 310¹⁶ sec^–1m^–2, this shift for standard samples is 50°C. These results attest to the state of the vessel material at a given moment in time.Translated from Atomnaya Énergiya, Vol. 97, No. 3, pp. 177–182, September, 2004.     

Investigation of radiation creep of reactor graphite

The results of investigations of the radiation creep of GR-280 graphite under a high compression load (about 15 MPa) after irradiation in a BOR-60 reactor at 520°C to fast-neutron fluence 1.2·10²² cm⁻² are presented. It is shown that the fluence dependence of the creep deformation, calculated using the standard relation as the difference of the change in the dimensions of loaded and control samples, is anomalous. The linear thermal expansion coefficients of loaded and control samples are found as functions of the neutron fluence under the same conditions. It is noted that the linear thermal expansion coefficient of the samples irradiated under a load is much higher than that of the control samples. Simmons' theorem is used to take account of the effect of a load on the linear thermal expansion coefficient, and the dimensional changes of graphite exposed to radiation and the dependence of the true creep deformation on the neutron fluence are calculated. It is shown that these dependences are close to linear in the experimental fluence range (0.4–1.2)·10²² cm⁻². Translated from Atomnaya énergiya, Vol. 105, No. 2, pp. 83–87, August, 2008.     

Computational-Experimental Investigation of the Wigner Energy and the Physical Properties of Plunger Graphite in the Safety and Control System of an RBMK Reactor

The total amount of stored energy (Wigner energy) and the physical-mechanical properties of the graphite plunger, which exhausted its total service life (6 yr), in the No. 2 unit of the Kursk nuclear power plant were estimated experimentally. The results showed that the total accumulated energy was 180–220 cal/g ((8–10)·10⁵ J/ kg). The real tempearture of the graphite plunger was found to be much lower – 70–80°C compared with the computed value 180–200°C. The energy was nonuniformly distributed over the cross section and azimuth of the plunger.Measurements showed that the thermal conductivity of the graphite in the plunger is low (no greater than 14–15 W/(m·K) at the measurement temperature 70°C) and that the temperature dependence is clearly nonmonotonic and contains stages with accelerated variation followed by moderation. These stages of nonmonotonic behavior correlate with stages where energy is released in experiments with linear heating of the irradiated graphite samples.     

Behavior of mechanical properties of nickel-alloyed reactor pressure vessel steel under neutron irradiation and post-irradiation annealing

The effect of neutron irradiation and post-irradiation thermal annealing on tensile and impact properties of Cr–Ni–Mo steel used for WWER-1000 reactor pressure vessel (RPV) manufacturing was studied. A gap in yield stress and ultimate tensile stress fluence dependence at the fluence range of 0–3×10²³ neutrons m⁻² was observed while ductile-to-brittle transition temperature (DBTT) was continuously increasing with damage dose. The post-irradiation annealing recovery of tensile properties was found to be higher than the one of impact properties. Over-recovery of tensile properties due to 460 and 490°C post-irradiation annealings were observed. The annealing effectiveness of WWER-440 and WWER-1000 grades was compared. Nickel was supposed to affect both the radiation sensitivity and the post-irradiation residual DBTT shift of WWER-1000 type steel.     

Neutron irradiation studies on low density pan fiber based carbon/carbon composites

Carbon has been extensively used in nuclear reactors and there has been growing interest to develop carbon-based materials for high-temperature nuclear and fusion reactors. Carbon-carbon composite materials as against conventional graphite material are now being looked into as the promising materials for the high temperature reactor due their ability to have high thermal conductivity and high thermal resistance. Research on the development of such materials and their irradiation stability studies are scant. In the present investigations carbon-carbon composite has been developed using polyacrylonitrile (PAN) fiber. Two samples denoted as Sample-1 and Sample-2 have been prepared by impregnation using phenolic resin at pressure of 30 bar for time duration 10 h and 20 h respectively, and they have been irradiated by neutrons. The samples were irradiated in a flux of 10¹² n/cm²/s at temperature of 40 °C. The fluence was 2.52 × 10¹⁶ n/cm². These samples have been characterized by XRD and Raman spectroscopy before and after neutron irradiation. DSC studies have also been carried out to quantify the stored energy release behavior due to irradiation. The XRD analysis of the irradiated and unirradiated samples indicates that the irradiated samples show the tendency to get ordered structure, which was inferred from the Raman spectroscopy. The stored energy with respect to the fluence level was obtained from the DSC. The stored energy from these carbon composites is very less compared to irradiated graphite under ambient conditions.     

Investigation of the Effect of Radiation and High-Temperature Treatment on the x-Ray Luminescence of Quartz Glasses

The x-ray luminescence of KI, KV, and KU-1 quartz glasses, irradiated with and n– radiation in the dose range 10²–10⁷ Gy and neutron fluence range 10¹⁵–10¹⁷ cm^–2 and subjected to high-temperature annealing in air at 450 and 900°C is investigated. It is shown that the spectra of the nonirradiated and the and n– irradiated glasses of the first two types are a superposition of bands with _max = 410 and 460 nm, which are due to an impurity center initially present in the glasses (_max = 410 nm) and the initial and radiation-generated with dose 10⁶ Gy and fluence 10¹⁶ cm^–2 E' centers (_max = 460 nm). X-Ray luminescence is not observed in nonirradiated KU-1 glasses; a band with _max = 460–470 nm, due to radiation-generated E' centers, appears in the spectra of and n– irradiated glasses. As the radiation dose and the neutron fluence increase, the number of impurity centers decreases and the number of E' centers increases. It is established that the 410 nm band is due to the component of the n– radiation. High-temperature annealing in air at 900°C induces in the spectra new bands with _max = 470 and 520–540 nm, which are believed to be due to interstitial defects of the type O^– and O₂ ^–, formed when oxygen from air diffuses into the glass and localizes in interstices. 6 figures, 7 references.     

Irradiation Behaviors of Nuclear Grade Graphite in Commercial Reactor, (III) Mechanical Properties

Bending and compressive strengths, and Young's modulus were measured for Pechiney nuclear grade graphite irradiated in the temperature range 220～400°C in the environment of CO₂ in a commercial reactor, up to the neutron fluence 6.2 × 10¹⁹ and 2.2 × 10²⁰n/cm² (E>0.85 MeV), respectively.

All of them increased owing to neutron irradiation, and the changes in both strengths were almost similar in the whole range of irradiation temperature, however the changes in Young's modulus depended on irradiation temperature.

It was clarified in the present experiment that both strengths were related with Young's modulus and the relation could be expressed by the formula σ=kE ⁿ, where σ and E are strength and Young's modulus, respectively, and n is constant which has different value for bending or compressive strength and also for their measured direction.     

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.     

Flux of secondary neutrons produced by 660 MeV protons in shielding

With the help of a set of threshold and resonance detectors, measurements were made of the spatial and energy distribution of secondary neutrons in graphite and nickel blocks. Absolute values of the neutron flux as a function of depth in an infinite slab were obtained for a plane, monodireetionat proton source. The energy distribution of the secondary neutrons in the energy range 2.5·10^–8 to 6.6·10² Mev was represented by seven groups. The magnitude of the dose behind plane nickel and graphite shielding as a function of thickness was also determined. The results are discussed.Translated from Atomnaya Énergiya, Vol. 18, No. 6, pp. 573–578, June, 1965     

Stored energy in the graphite stack of the IR reactor

Storage of Wigner energy and the nature of stored energy release were investigated in the graphite stacks of the IR reactor. The investigations were carried out on graphite bricks removed from the reactor stack as the latter was being dismantled. Other samples were cut out from the reactor stack with the aid of a special drill cutter during the two years after the stack was dismantled, when an additional integral flux of thermal neutrons 2.2 ·10²¹ neutrons/cm² developed. The samples were so chosen that the pattern of stored-energy distribution throughout the stack might be obtained. Stored Wigner energy was determined by the technique of two successive heatings of the samples to temperatures of 600–650 ° C in vacuum calorimeters, making possible an energy release during the anneal. The total quantity of stored energy was determined by combustion of the graphite in a standard solid fuel testing calorimeter.     

Method based on isotope ratio mass spectrometry for evaluation of carbon activation in the reactor graphite

The general idea of this work is to introduce an evaluation method to restore the irradiation parameters of graphite or other carbonaceous materials using experimental and modelling results of ¹³C generation in the irradiated material. The method is based on coupling of stable isotope ratio mass spectrometry and computer modelling of the reactor core to evaluate the realistic characteristics of the reactor core such as the neutron fluence in any position of the reactor graphite stack or other graphite constructions.The generation of carbon isotopes ¹³C and ¹⁴C in the irradiated graphite of the RBMK-1500 reactor has been estimated by modelling of the reactor core with computer codes MCNPX and CINDER90. Good agreement of simulated and measured Δ¹³C/¹²C values in graphite of the central part of the reactor core indicates that the neutron flux (1.40 × 10¹⁴ n/cm² s) is modelled accurately in the graphite sleeve of the fuel channel. The simulated activity of ¹⁴C is compared with the one measured by the β spectrometry technique. Results indicate that production of ¹⁴C from ¹⁴N in the RBMK-1500 reactor is considerable and has to be taken into account in order to make proper evaluation of ¹⁴C activity. Measured ¹⁴C specific activity values correspond to 15 ± 4 ppm impurity of ¹⁴N in graphite samples from the RBMK-1500 reactor core.     

Effect of neutron irradiation on some properties of heat-resistant concretes

tn contrast to the structural materials of nuclear reactors, the radiation resistances of concretes used in biological shielding have not been sufficiently studied. A tendency has recently arisen for the preferential use of heat-resistant concretes in biological shielding instead of materials such as steel, cast iron, graphite, boron, etc., which are costly and in relatively short supply. In this paper we shall indicate the effect of reactor neutron irradiation on certain properties of Porland-cement and liquid-glass heat-resistant chromite concretes. The integral neutron flux used in this investigation was (2–2.4) x 10²¹ neutrons/cm² and the irradiation temperature up to 550° C.It was found experimentally that these concretes:retain quite high strength and elastic properties. The thermal conductivity and thermal expansion coefficient change very littte. It is concluded that such concretes may be recommended for use in the biological shielding of nuclear reactors.Translated from Atomnaya Énergiya, Vol. 21, No. 2, pp. 108–112, August, 1966.     

Neutron irradiation-induced dimensional changes in MEMS glass substrates

A study is made of radiation-induced expansion/compaction in Pyrex^® (Corning 7740) and Hoya SD-2^® glasses, which are used as substrates for MEMS devices. Glass samples were irradiated with a neutron fluence composed primarily of thermal neutrons, and a flotation technique was employed to measure the resulting density changes in the glass. Transport of Ions in Matter (TRIM) calculations were performed to relate fast (∼1 MeV) neutron atomic displacement damage to that of boron thermal neutron capture events, and measured density changes in the glass samples were thus proportionally attributed to thermal and fast neutron fluences. Pyrex was shown to compact at a rate of (in Δρ/ρ per n/cm²) 8.14 × 10⁻²⁰ (thermal) and 1.79 × 10⁻²⁰ (fast). The corresponding results for Hoya SD-2 were 2.21 × 10⁻²¹ and 1.71 × 10⁻²¹, respectively. On a displacement per atom (dpa) basis, the compaction of the Pyrex was an order of magnitude greater than that of the Hoya SD-2. Our results are the first reported measurement of irridiation-induced densification in Hoya SD-2. The compaction of Pyrex agreed with a previous study. Hoya SD-2 is of considerable importance to MEMS, owing to its close thermal expansivity match to silicon from 25 to 500°C.     

Measurement of neutron emission from Ti and Pd in pressurized D2 gas and D2O electrolysis cells

Experiments using high-efficiency neutron detectors have detected neutron emission from various forms of Pd and Ti metal in pressurized D₂ gas cells and D₂O electrolysis cells. Four independent neutron detectors based on³He gas tubes were used. Both random neutrons (0.05–0.2 n/s) and time-correlated neutron bursts (10–280 n) of 100-s duration were measured using time-correlation counting techniques. The majority of the neutron burst events occurred at –30°C as the samples were warming up from the liquid nitrogen temperature.     

Formation of Gas Pores in Nickel Alloys and Structural Steel under Irradiation by Helium Ions

Transmission electron microscopy is used to study the development of helium porosity in binary alloys of nickel with elements possessing a different dimensional atomic mismatch with nickel – from negative (beryllium and silicon) to positive (molybdenum, tungsten, aluminum, titanium, tantalum, tin, and zirconium), in structural steels ChS-68, ÉP-150, and the nickel alloy KhNM. The gas pores were produced by irradiation with 40 keV He⁺ up to fluence 5·10²⁰ m^–2 at 650 and 20°C followed by annealing at 650°C for 1 h. It is shown that under high-temperature annealing beryllium and silicon, relative to nickel, give rise to the formation of larger bubbles, while elements with a larger positive size mismatch with nickel atoms substantially decrease the size and increase the density of the bubbles. On the whole, as atomic radius and the concentration of the alloying element increases in alloys, the gas swelling of the irradiated layer decreases. Under post-irradiation annealing, bubbles with the largest diameter and the lowest density develop in nickel. Any alloying used decreases the size and increases the density of bubbles. The data obtained are discussed from the standpoint of the formation of various vacancy complexes of helium and their thermal stability.     

Neutron irradiation effects on optical absorption of KU1 and KS-4V quartz glasses and Infrasil 301

In this work we have compared the effects of neutron (10²¹–10²² n/m² fluences) and gamma irradiation (23.8 MGy dose) on the IR–vis–UV optical absorption spectra of high purity silica with different OH content: KU1 (800 ppm), KS-4V (<0.2 ppm), and commercial silica Infrasil 301 (<8 ppm). The results show that the UV–vis optical degradation of the silica, after neutron irradiation at the highest fluence is similar for the three grades studied, while gamma-induced optical absorption depends on the material grade (KS-4V shows the lowest optical absorption). The effects of both types of radiation on the IR band related with the hydroxyl group (3650 cm⁻¹) depend on the silica grade. For KU1, the shape of this band changes with neutron fluence. For Infrasil 301 gamma and neutron irradiated, this band height increases, possibly due to free molecular or hydrogen atoms. The shift to lower energies observed for the 2260 cm⁻¹ band in the three neutron irradiated silica grades, reflects the changes induced by neutrons in the lattice bonding angle distribution.     

Effect of the composition of radiation on the radiation damage to graphite

A statistical analysis is performed of the results on the determination of the critical neutron fluence in MR, SM-2, and BOR-60 with different irradiation temperature. It is shown that the critical neutron fluence depends not only on the irradiation temperature but also, and to an even greater extent, on the radiation composition factor (ratio of the neutron and γ-ray flux densities). Thus the critical neutron fluence for irradiation at 600°C in MR (radiation composition factor 0.13) is 17·10²¹ cm⁻² and in SM-2 (radiation composition factor 0.1) 11·10²¹ cm⁻² at the same temperature. When the same graphite is irradiated in the region of the outer corner of a working block of RBMK, where the radiation composition factor is 0.55, it is expected that the critical neutron fluence will be 31.7·10²¹ cm⁻². In summary, taking account of the effect of γ-radiation introduces substantial corrections: the experimental results obtained in research reactors are found to be at least a factor of 2 too low. This gives hope of substantiating the substantial increase in the service life of the RBMK graphite masonry. 3 figures, 8 references. Scientific-Research and Design Power-Engineering Institute. State Science Center—Scientific-Research Institute of Nuclear Reactors. Translated from Atomnaya énergiya, Vol. 87, No. 1, pp. 24–28, July, 1999.     

Fusion Safety Studies in Russia in 1996

The paper seeks to provide a summary report of observations and results of some Russian fusion safety studies performed in 1996. Release of tritium and helium from neutron irradiated beryllium at relatively high neutron fluences has a burst nature. With the growth of the beryllium temperature-increase rate to 90 K/s, the temperature of tritium burst release decreases from 800 to 450–500°C and for helium decreases from 1200 to 500°C. Characterization of carbon and tungsten dust produced in experiments simulating plasma disruptions revealed that dust particle distribution of sizes for graphites and carbon fiber composites has a bimodal nature with maxima in the range of 0.01–0.03 and 2–4 m for composite UAM and in the range of 0.14–0.18 and 2–4 m for graphite MPG-8. Chemical reactivity of beryllium with air was studied as well. A mathematical model for beryllium weight gain under its chemical interaction with air at temperatures of 700–800°C as a function of beryllium porosity, temperature, and interaction duration was developed.