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.     

Thermal and structural properties of low-fluence irradiated graphite

The release of Wigner energy from graphite irradiated by fast neutrons at a TRIGA Mark II research reactor has been studied by differential scanning calorimetry and simultaneous differential scanning calorimetry / synchrotron powder X-ray diffraction between 25 and 725 °C at a heating rate of 10 °C min⁻¹. The graphite, having been subject to a fast-neutron fluence from 5.67 × 10²⁰ to 1.13 × 10²² n m⁻² at a fast-neutron flux (E > 0.1 MeV) of 7.88 × 10¹⁶ n m⁻² s⁻¹ and at temperatures not exceeding 100 °C, exhibits Wigner energies ranging from 1.2 to 21.8 J g⁻¹ and a Wigner energy accumulation rate of 1.9 × 10⁻²¹ J g⁻¹ n⁻¹ m². The differential-scanning-calorimeter curves exhibit, in addition to the well known peak at ∼200 °C, a pronounced fine structure consisting of additional peaks at ∼150, ∼230, and ∼280 °C. These peaks correspond to activation energies of 1.31, 1.47, 1.57, and 1.72 eV, respectively. Crystal structure of the samples is intact. The dependence of the c lattice parameter on temperature between 25 and 725 °C as determined by Rietveld refinement leads to the expected microscopic thermal expansion coefficient along the c axis of ∼26 × 10⁻⁶ °C⁻¹. At 200 °C, coinciding with the maximum in the differential-scanning-calorimeter curves, no measurable changes in the rate of thermal expansion have been detected - unlike its decrease previously seen in more highly irradiated graphite.     

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.     

Thermal,structural, and radiological properties of irradiated graphite from the ASTRA research reactor – Implications for disposal

The release of Wigner energy from the graphite of the inner thermal column of the ASTRA research reactor has been studied by differential scanning calorimetry and simultaneous differential scanning calorimetry/synchrotron powder X-ray diffraction between 25 °C and 725 °C at a heating rate of 10 °C min⁻¹. The graphite, having been subject to a fast-neutron fluence from ∼10¹⁷ to ∼10²⁰ n cm⁻² over the life time of the reactor at temperatures not exceeding 100 °C, exhibits Wigner energies ranging from 25 to 572 J g⁻¹ and a Wigner energy accumulation rate of ∼7 × 10⁻¹⁷ J g⁻¹/n cm⁻². The shape of the rate-of-heat-release curves, e.g., maximum at ca. 200 °C and a fine structure at higher temperatures, varies with sample position within the inner thermal column, i.e., the distance from the reactor core. Crystal structure of samples closest to the reactor core (fast-neutron fluence >1.5−5.0 × 10¹⁹ n cm⁻²) is destroyed while that of samples farther from the reactor core (fast-neutron fluence <1.5−5.0 × 10¹⁹ n cm⁻²) is intact, with marked swelling along the c-axis. The dependence of the c lattice parameter on temperature between 25 °C and 200 °C as determined by Rietveld refinement for the non-amorphous samples leads to the expected microscopic thermal expansion coefficient along the c-axis of ∼ 26 × 10⁻⁶ °C⁻¹. However, at 200 °C, coinciding with the maximum in the rate-of-heat-release curves, the rate of thermal expansion abruptly decreases indicating a crystal lattice relaxation. The ¹⁴C activity in the inner thermal column graphite ranges from 6 to 467 kBq g⁻¹. The graphite of the inner thermal column of the ASTRA research reactor has been treated by heating to 400 °C for 24 h in a hot-cell facility prior to interim storage.     

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     

Spatial distribution of γ-rays and moderated neutrons in the graphite column of the RFT reactor

Measurements have been made of the spatial distributions of neutrons of various energies and the -rays in the graphite thermal column of the RFT reactor. The fluxes of thermal, resonance, and fast neutrons were measured using the activity induced in various indicators. The decay in the intensity of the /gg-radiation was determined by small condenser type ionization chambers. At distances from 80 to 160 cm the fluxes of resonance and fast neutrons fall off approximately exponentially with a decay length of /R~ 13 cm and 15.7 cm, respectively. At large distances the fluxes of fast and resonance neutrons are in equilibrium. The flux of slowed-down neutrons in this region falls off exponentially with a decay length of 18.2 cm as determined by the penetrating fast-neutron component with an energy of approximately 6.6 Mev. The intensity of the -radiation in the graphite column falls off in an almost exponential manner with an attenuation coefficient = 3.78 · 10^–2 cm^–1. The theoretical predictions are found to be in satisfactory agreement with the experimental data.     

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.     

Measurements of Pseudo-Decay-Constants of Trapped Neutrons in Small Graphite Systems at 77 and 194°K

Abstract

Neutron pulse die-away experiments for small graphite assemblies by using a pulsed “cold” neutron source technique were carried out for the purpose to measure the pseudo-decay-constants of trapped neutrons at the lowest Bragg-peak energy E _B of coherent scattering in graphite, and also to estimate the temperature dependence of inelastic scattering cross section of neutrons at E _B . Experiments were carried out for 77 and 194°K graphite systems with dimensions from 40x40x40 cm³ to 30x30x20 cm³.

The experimentally determined pseudo-decay-constants showed distinct temperature dependence and good agreement with the theoretical inelastic scattering cross sections by Young-Koppel model of phonon frequency distribution of graphite at these low temperatures.     

Measurement of Neutron Slowing Down Time in Graphite

A series of pulsed neutron experiments was performed to investigate the chemical binding effect on the neutron slowing down time. Bursts of D-T neutrons of 1 μsec width were generated in a hexagonal prism with 240 cm high with 120 cm flanks, made of reactor grade graphite with a density of 1.54. The slowing down neutrons were detected by bare as well as energy-selective filter-covered BF₃ counters, and analyzed with a 256 channel time analyzer. Slowing down times in graphite were determined by interpreting the increment of the difference of events between the two counters to be due to the contribution made by the fraction of the slowing down neutrons at the time of measurement that was below the cut-off energy of the filter. The results of measurements showed good agreement with calculation based on the crystal model.     

重水研究堆热柱石墨潜能释放分析研究

重水研究堆堆内石墨构件在长期中子辐照下将会累积潜能,为确保重水研究堆堆内石墨构件安全退役及处理处置,本文采用差示扫描量热仪对重水研究堆3个不同位置所取热柱石墨样品进行了潜能测量,扫描温度范围为10～550 ℃、升温速率为10 ℃/min。结果表明：3个位置的样品在80～500 ℃温度积分区间内潜能释放量分别为70.690、42.167、18.158 J/g;潜能释放率曲线峰值温度均大于300 ℃,未辐照石墨样品的比热容较热柱石墨样品释放率dS/dT（S为潜能释放量（J/g）,T为温度（℃））高,表明本实验所取石墨样品不会发生潜能释放导致石墨自身温度上升的情况;3个位置样品的快中子注量分别为6.75×10¹⁶、6.10×10¹⁴、1.89×10⁷ cm^-2;获得了潜能释放分数曲线与潜能释放速率曲线,1#和2#位置样品的潜能释放速率曲线具有至少2个释放峰,表明潜能释放过程中具有至少2个动力学过程。     

Spectra of fast reactor neutrons after passing through graphite,lead, and iron

By means of a fast neutron scintillation spectrometer with one hydrogen-containing detector, the spectra of fast reactor neutrons after passing through various thicknesses of lead, graphite, and iron were measured in the range 0.7–11 MeV. The measurements were carried out in a water-moderated water-cooled experimental reactor in barrier geometry. The results of the experiments enabled us to determine the deformation of the neutron spectrum in relation to the penetration through the layers of material, and to calculate the relaxation lengths and the removal Cross sections. These quantities were punished earlier for fission spectrum neutrons in the energy range 0.7^–3 MeV.Translated from Atomnaya Énergiya, Vol. 16, No. 1, pp. 32–40, January, 1964     

Radiation-Induced Change in the Properties of Isotropic Structural Graphite

Three types of samples of isotropic graphite with different grain density and size were irradiated in a BOR-60 reactor up to neutron fluence (1.7–2.8)·10²⁶ m^–2 (E > 0.18 MeV) at 360–400°C. After irradiation, the change in the dimensions, resistivity, linear thermal expansion coefficient and dynamic elastic modulus were investigated. It was determined that the density in the range 1.67–1.76 g/cm³ results in an increase of the maximum weight and depth of volume shrinkage of isotropic fine-grain graphite. An equation was proposed for fitting the temperature dependence of the critical neutron fluence in the range 380–780°C for the experimental graphite samples.     

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.     

Computational-experimental studies of the neutronics of the SM reactor with different variants of the neutron-trap arrangement

The results of experimental studies of the neutronics of the high-flux SM reactor with different arrangements of the neutron trap are presented. The MCU series of high-precision computer programs implementing the Monte Carlo method is used for computations. Experimental data on reactivity effects, the effectiveness of safety and control rods, and the coefficients of nonuniformity of energy release in the core have been obtained in experiments on a critical assembly – a physical model of the SM reactor – and directly in experiments in the reactor. The error is 4.2–10% in determining the reactivity parameters and 5–10% for the relative energy release in the fuel elements. Information on the neutron field formed in the volume of the neutron trap has been obtained for two arrangements of the beryllium and water moderators. The differential and integral energy spectra of the neutrons in the energy interval from 0.5 eV to 20 MeV are obtained for three points inside the trap (external, central series, center). The flux density of thermal, superthemal, and fast neutrons are determined.     

Properties of the final product of thermochemical reprocessing of irradiated reactor graphite wastes in the Al-TiO<Subscript>2</Subscript>-C system

Laboratory investigations of the strength and chemical resistance of the final product of thermochemical reprocessing of reactor graphite wastes in the Al-TiO₂-C system are presented. The ¹³⁷Cs and ⁹⁰Sr leaching rate, which is determined for samples synthesized from a charge with real irradiated graphite from an AM research reactor, does not exceed 10⁻⁶ g/(cm²·day) at the 28th day. __________ Translated from Atomnaya énergiya, Vol. 104, No. 4, pp. 224–227, April, 2008.     

The moderation length of neutrons

The neutron moderation length is an important constant without which is is impossible to deal with the design of nuclear reactors in all their aspects. A knowledge of the moderation lengths is especially necessary for the determination of the space distribution of the neutrons in a reactor, and for the calculation of the energy spectrum of the moderated neutrons.In the present work is given an approximate solution of the integral equations satisfied by the space moments of the neutron distribution function inan infinite medium with an infinite, plane, isotropic source. The energy-angle moments of the neutron scattering function are expressed in terms of experimentally determined angular distributions of neutrons of various energies in the case of anisotropic elastic scattering from nuclei. By using experimental results for the total cross section and for the angular distribution in the case of elastic scattering from the nuclei H¹, D², Be⁹, C¹², O¹⁶ the neutron moderation lengths were calculated for the moderators: water, heavy water, graphite, beryllium, and beryllium oxide.The results of the calculation were found to be in satisfactory agreement with experimental values.In conclusion the authors wish to thank Doctor of Physical and Mathematical Science G. I. Marchuk for useful discussions, and also B. S. Gudkov, Z. P. Drobyshev, and Z. I. Shemetenko for carrying out the calculations.     

Absolute Fission-Rate Distributions in Lithium and Hybrid Fusion Blanket Assemblies, (II)

The absolute fission rates of ²³⁵U, ²³⁷Np, ²³⁸U and ²³²Th were measured in four types of spherical blanket assemblies containing lithium and/or natural uranium and/or graphite. The results of measurement are compared with those of one-dimensional transport calculations employing 100-group neutron cross-sections obtained from the ENDF/B-IV data file. It is shown that the ratios between calculated and experimental values of ²³²Th, ²³⁸U and ²³⁷Np fission rates decrease with distance from the assembly center, where D-T neutrons are generated. An overestimation of about 50% is observed in the calculated ²³⁵U fission rate for the graphite reflector region.

One of the main sources of the disagreement proves to lie in the inability of the codes adopted for generating the multi-group cross-section to take account of the angular distributions of the secondary neutrons emanating from nonelastic reactions. The results of the analysis indicate that the method of calculation currently employed in fusion reactor neutronics overestimates the reflection of neutrons and underestimates the penetration of fast neutrons when a graphite reflector is used.     

<Superscript>59</Superscript>Ni Accumulation in Corrosion-Resistant Steel under the Action of Neutron Flux

A γ-ray line with energy E_γ = 11.3 MeV was detected during an experiment, performed on a nuclear reactor, investigating the characteristics of the energy spectrum of γ-rays. The most likely source of this line is radiative capture of thermal neutrons by ⁵⁹Ni nuclei, which accumulated in the corrosion-resistance steel as a result of the more than 20 years of irradiation in the reactor, via the reaction ⁵⁸Ni(n, γ)⁵⁹Ni. It was found that for thermal-neutron fluence 10²¹ cm⁻² the ⁵⁹Ni concentration is 0.47% of the ⁵⁸Ni concentration. __________ Translated from Atomnaya Energiya, Vol. 99, No. 4, pp. 268–272, October, 2005.     

Influence of the nearest environment of the core on the power pulse dynamics in the IBR-2 reactor

An analysis of the IBR-2 reactor power pulse shape measured over the entire dynamic range of neutron flux variation (10⁴), i.e. from the maximum pulse power to the background power between pulses, has been carried out. Three variants of the model describing the reactor dynamics during the power pulse have been investigated. The best approximation to the experimental data has been obtained by adding to six equations describing the effect of delayed neutrons on the power pulse of two analogous ones describing the effect of the neutrons reflected from the structural elements of the reactor. It is shown that the most probable source of additional groups of neutrons may be the neutron moderators enveloping the core as well as the elements of the biological concrete shielding that are the closest to the core. These additional groups of neutrons influence essentially the formation of the power pulse.     

Fission Density in Plutonium-Uranium Mixed Oxide Fuel Irradiated by Pulsed Power in NSRR Experiments

Fuel behavior during a reactivity initiated accident condition is recognized to be predo minantly related to energy deposition in the fuel. The first stage of NSRR in-pile experiments addressing the behavior of PuO₂-UO₂ mixed oxide fuels evaluated the energy deposition per unit integrated reactor power by γ-ray spectrometry. Solid samples were used to measure the γ-rays because the facility is permitted to handle solid plutonium only. Determination of the penetration ratios of γ-rays from the actinides contained in the fuels allowed correction for the self-attenuation of γ-rays in the solid samples. Evaluation of the effect of epithermal neutron fissions was also necessary since the fissile nuclides of ²⁴¹Pu and ²³⁹Pu have high resonance cross sections in the epithermal energy region. For this evaluation, the fission density was first calculated for the fission products as a function of the contribution ratio of the fissions of epithermal neutrons. Accurate fission density was then determined using the contribution ratio which minimized the deviation of the calculated values for the fission density. The fission densities determined by this simplified method agreed well with the values calculated using the computer codes CITATION and GGC-4.