Neutron Fluence Values Research Articles

Bonner sphere measurements of high-energy neutron spectra from a 1 GeV/u 56Fe ion beam on an aluminum target and comparison to spectra obtained by Monte Carlo simulations

The goal of this work is to characterize the secondary neutron spectra produced by 1 GeV/u56Fe beam colliding with a thick cylindric aluminum target and to perform a quantitative comparison with simulated results obtained with Monte Carlo codes. The measurements were performed using extended-range Bonner sphere spectrometers at two positions (15° and 40°) with respect to the beam direction. The secondary radiation field was simulated using four Monte Carlo codes (FLUKA, MCNP6, Geant4 and PHITS) and several physical models of nuclei transport and interaction. Neutron and proton energy distributions were simulated for the experimental measurement positions. The simulated neutron spectra, together with data measured with Bonner sphere spectrometers, after carrying out the correction of the contributions induced by the secondary protons, were used as input for the MAXED spectrum unfolding code to obtain the measured neutron spectra. Unfolded neutron spectra were compared with simulated ones to carry out a quantitative analysis of the performance of the chosen Monte Carlo codes and their corresponding physical models. This comparison showed that, because of experimental uncertainties and physical models, there are no unique solutions for each measurement location, but a range of solutions where the true experimental neutron spectra probably lie. The results showed deviations between 4.23% and 8.42% for some simulated spectra. Regarding the total integral values of neutron fluence and ambient equivalent dose, the unfolded neutron spectra showed deviations lower than 2%.

Prediction of radiation embrittlement of VVER-1000 RPV welded seam material at lifetime extension up to 60 years and more

The correctness of the model assumption about additivity of the contributions of hardening (formation of dislocation loops and radiation-induced precipitates) and non-hardening (formation and accumulation of grain boundary segregations) mechanisms to the final radiation embrittlement of VVER-1000 RPV weld materials has been verified. The paper presents the results of the experiment to obtain predicted values of the critical embrittlement temperature of weld metal with nickel content of 1.59–1.88% using accelerated irradiation in the IR-8 research reactor from states characterized by different starting level of grain boundary segregation. The procedure of accounting for the flux effect for the values of critical embrittlement temperature obtained using accelerated irradiation has been proposed. The conservativity of the weld metal radiation embrittlement dependence has been verified based on the modal assumption of additivity of the contributions of the strengthening and non-strengthening mechanisms up to the values of fast neutron fluence ~75–1022 neutrons/m2.

Estimate of Neutron Fluence Using Isotope Ratio Method for ZR-2.5NB Materials Irradiated in the National Research Universal Reactor

Abstract The isotope ratio method (IRM) is a technique used for estimating the energy production in a fission reactor by measuring isotope ratios in nonfuel reactor components. This method has been successfully demonstrated on the estimation of cumulative energy production as well as plutonium production in graphite-moderated and light-water-moderated reactors for nonproliferation purposes. In this paper, IRM was used to estimate neutron fluence in Zr-2.5Nb materials irradiated in fast neutron (FN) irradiation facilities in the National Research Universal (NRU) reactor. Neutron fluence has been shown to be an important parameter for studying irradiation effects on the performance and properties of critical reactor components. Selected isotope ratios of hafnium and iron were used as indicators of the neutron fluence in Zr-2.5Nb sample materials. Correlations between neutron fluence and the indicator element isotope ratios were generated using the reactor physics simulation codes Winfrith improved multigroup scheme (WIMS)-AECL and SCALE/Oak Ridge Isotope GENeration code (ORIGEN). Inductively coupled plasma-mass spectrometry (ICP-MS) was used to obtain accurate measurements of the isotope ratios. Neutron fluence values estimated using IRM, were in good agreement with the values based on measured irradiation power histories of the NRU reactor. This study proposes a potential application of IRM to the estimation of neutron fluence for critical reactor components in heavy-water-moderated reactors such as pressure and calandria tubes in CANDU® reactors.

Influence of Ni-Mn contents on the embrittlement of PWR RPV model steels irradiated to high fluences relevant for LTO beyond 60 years

Understanding the irradiation induced degradation of the reactor pressure vessel (RPV) steels at high fluences becomes more important in view of safe long term operation (LTO) of existing nuclear power plants. Specifically, the role of Ni, Mn and Si and associated synergetic effects on embrittlement behaviour need to be further investigated to understand accelerated embrittlement at high neutron fluences (ca. 1 × 10 20 n.cm −2 , E>1 MeV) observed in certain types of RPV steels. In addition, it is required to verify the validity of existing embrittlement trend curves (ETCs) at high fluence values. To address these issues, NRG and JRC performed a unique irradiation experiment called LYRA-10 in the High-Flux Reactor (HFR) Petten. In LYRA-10 a variety of specimens (e.g. tensile, Charpy, mini-Charpy and microscopy) of RPV steels (base metal and welds) with systematic variations in Ni, Mn and Si contents, while keeping the base composition representative of VVER-1000 and PWR RPV steels, have been irradiated for 16 HFR cycles to achieve the nominal fast neutron fluence of 1.11 × 10 20 n.cm −2 at a nominal temperature of 286 °C. A large post-irradiation examination campaign on LYRA-10 specimens has been initiated within the STRUMAT-LTO, which is an international collaboration project co-funded by EURATOM. The research reported in this paper includes results from tensile and mini-Charpy impact testing of four PWR RPV model steel specimens irradiated in LYRA-10, which is performed as part of STRUMAT-LTO in-kind pilot project before the start of the EURATOM co-funded project. The combined influence of Ni and Mn on embrittlement behaviour, characterized by irradiation hardening and shifts in transition temperature (T k ) properties, has been evaluated. The measured shifts in T k values for all four model steels have been compared with predicted values from two selected ETCs in order to check their validity for these steels at the above high neutron fluence values.

Pulsed fusion grade deuterium plasma induced structural phase transformation in stainless steel using medium energy plasma focus device

Samples of ferrite -austenite stainless steel (SS) have been exposed to fusion grade plasma and radiations using a medium energy plasma focus device, ‘MEPF-12’. Three different polished samples of SS (10 mm diameter ×2 mm thick) were irradiated using single and multiple ( five and ten) plasma focus discharges. Each plasma focus discharge comprised energetic and high intensity deuterium ions, energetic deuterium plasma streams and 2.45 MeV neutrons produced via D(d, 3He)n fusion reactions. At sample position, typical values of deuterium ion fluence and neutron fluence were measured to be ∼1018ions/m2 and ∼1010 neutrons/m2, respectively, in each plasma focus discharge. Post irradiations, the samples have been characterized using different techniques for study of their surface modification characteristics, specifically, crystallinity, surface morphologies and surface roughness. The XRD analysis of the SS sample exposed to single or more plasma focus discharges revealed structural phase transformation from initial mixed ferrite [α(BCC)] -austenite [γ(FCC)] phase to austenite phase. Effects of deuterium ion concentration and transient rise in surface temperature on structural phase transformation have been discussed in detail. Surface morphologies studied using SEM has revealed erosion, melting and recrystallization of surface layer along with various other defects like pores, blisters, open bubbles and so on. Measurements using surface profilometer confirmed increase in surface roughness of the samples on irradiation. Detail discussion on the variation in the surface damage features with increasing number of plasma focus discharges used for irradiating samples and its comparison with virgin sample have been included in this report.

Complexes defects induced by neutron irradiation of Cz-silicon

The boron-doped Cz-Si was irradiated with 1 MeV neutrons delivered by the Es-Salam reactor at two different neutron fluence values, 1.98 × 1018 and 3.96 × 1018 n/cm2, respectively. Annealing processes were carried out under argon atmosphere, at 550 °C and 950 °C for 60 min, respectively. The effects of isochronal annealing on the neutron-irradiated silicon’s IR absorption spectrum are presented. The obtained results indicate two forms of the Near-Edge Absorption that increase with increasing neutron fluence. One centered at 1000–1500 nm (NEA-1) and second near 1600–1800 nm (NEA-2). After the annealing process at 550 °C, the IR absorption spectrum shows the change of the first Near-Edge Absorption band towards a shorter wavelength at 1000–1200 nm range as well as the disappearance of the (NEA-2) band near 1800 nm. The appearance of the lithium impurity associated with divacancy at 1700 nm that confirms the transmutation of boron into lithium of the samples. At 950 °C, the experimental data show that 1700 nm which is associated with Li divacancy persists for higher neutron fluence, but not for lower fluence.

Evaluation of neutron activation of intermetallic matrices for dispersive nuclear fuel obtained by SH-synthesis

The paper describes a technique for obtaining intermetallic matrix materials based on Zr-Al and Ni-Al systems for dispersive nuclear fuel. This type of fuel is planned to be used in existing and future high-temperature nuclear reactors. Physical and mathematical modelling of the process of activation of matrices by neutron radiation showed that upon reaching the value of thermal neutron fluence of the order of 2.76⋅1021 n⋅cm−2, the activity of Zr-Al and Ni-Al matrices was 1.3 ⋅ 1010 and 0.2⋅1010 Bq/g, respectively. The analysis showed that in terms of exposure of irradiated fuel it is preferable to use a matrix based on the zirconium-aluminium system.

Extension of lifespan of graphite in fuel blocks of high-temperature gas-cooled reactors as the resource for ensuring design values of nuclear fuel burn-up

High-temperature gas-cooled reactor (HTGR) is one of promising candidates for new generation of nuclear power reactors. This type of nuclear reactor is characterized with the following principal features: highly efficient generation of electricity (thermal efficiency of about 50%); the use of high-temperature heat in different production processes; reactor core self-protection properties; practical exclusion of reactor core meltdown in case of accidents; the possibility of implementation of various nuclear fuel cycle options; reduced radiation and thermal effects on the environment, forecasted acceptability of financial performance with respect to cost of electricity as compared with alternative energy sources. The range of output coolant temperatures in high-temperature reactors within the limits of 750–950 °C predetermines the use of graphite as the structural material of the reactor core and helium as the inert coolant. Application of graphite ensures higher heat capacity of the reactor core and its practical non-meltability. Residence time of reactor graphite depends on the critical value of fluence of damaging neutrons (neutrons with energies above 180 keV). In its turn, the value of critical neutron fluence is determined by the irradiation temperature and flux density of accompanying gamma-radiation. The values of critical fluence for graphite decrease within high-temperature region of 800–1000 °C to 1·1022 – 2·1021 cm–2, respectively. The compactness of the core results in the increase of the fracture of damaging neutrons in the total flux. These circumstances predetermine relatively low values of lifespan of graphite structures in high-temperature reactors. Design features and operational parameters of GT-MHR high-temperature gas-cooled reactor are described in the present paper. Results of neutronics calculations allowing determining the values of damaging neutron flux, nuclear fuel burnup and expired lifespan of graphite of fuel blocks were obtained. The mismatch between positions of the maxima in the dependences of fuel burnup and exhausted lifespan of graphite in fuel blocks along the core height is demonstrated. The map and methodology for re-shuffling fuel blocks of the GT-MHR reactor core were developed as the result of analysis of the calculated data for ensuring the matching between the design value of the fuel burnup and expected total graphite lifespan.

Neutron spectrum unfolding based on generalized regression neural networks for neutron fluence and neutron ambient dose equivalent estimations

Neutron fluence and neutron ambient dose equivalent, H*(10), are important physical quantities for neutron radiation protection and monitoring. They can be deduced from neutron spectrum, which is usually measured by multisphere system with proper unfolding methods. Novel unfolding methods on the basis of artificial intelligence, mainly artificial neural networks (ANNs), have been researched and developed. However, without normalization on network inputs, ANNs can not be applied to accommodate demands of various neutron field measurements for neutron spectrum unfolding in practice, because the neutron spectra for training the ANNs are mostly extracted from IAEA (2001), the integrals of which over neutron energy are unit fluences. Moreover, derived from an unfolded normalized spectrum, the true values of neutron fluence and H*(10) are never to know. In this work, three normalization methods—zero-mean normalization method, min-max normalization method, and maximum-divided normalization method were used to process with the inputs of generalized regression neural networks (GRNNs), and a new method was proposed for neutron fluence and H*(10) estimations derived from unfolded neutron spectrum based on GRNNs for the first time. Sixty-three neutron spectra were unfolded based on GRNNs with use of three normalization methods, and the corresponding neutron fluences and H*(10) were obtained and compared. From the testing results, the GRNNs with the maximum-divided method is most effective to unfold neutron spectrum and to evaluate neutron fluence and H*(10). The feasibility of the method was further studied through experiments by using Bonner sphere spectrometer in well characterized 241Am–Be neutron field.

The influence of core power distribution on neutron flux density behind a pressure vessel of a VVER-1000 Mock Up in LR-0 reactor

The neutron flux distribution behind a reactor pressure vessel (RPV) is an important parameter that is monitored to determine neutron fluence in the RPV. Together with mechanical testing of surveillance specimens, these are the most important parts of in-service inspection programs that are essential for a realistic and reliable assessment of the RPV residual lifetime. The fast neutron fluence values are determined by a calculation. These calculation results are accompanied by measurements of induced activities of the activation foils placed in the capsules behind the RPV at selected locations, namely in azimuthal profile. In case of discrepancies between the measured and calculated activities of the activation foils placed behind the pressure vessel, it is difficult to determine the source of the deviation. During such analysis, there arises a question on the influence of power peaking near core boundary on neutron profile behind the RPV.This paper compares the calculated and measured increase of the neutron flux density distribution behind the reactor pressure vessel in the azimuthal profile that has arisen from the replacement of 164 fuel pins located close to reactor internals by pins with the higher enrichment. This work can be understood as the first step in the characterization of the effect of incorrectly calculated pin power or burn-up in the fuel assembly at the core boundary relative to the neutron flux distribution behind reactor pressure vessel. Based on a good agreement between the calculated and experimental values, it can be concluded that the mathematical model used to evaluate the power increase is correct.

Neutron field measurement at the Experimental Advanced Superconducting Tokamak using a Bonner sphere spectrometer

The neutron field measurement was performed in the Experimental Advanced Superconducting Tokamak (EAST) experimental hall using a Bonner sphere spectrometer (BSS) based on a 3He thermal neutron counter. The measured spectra and the corresponding integrated neutron fluence and dose values deduced from the spectra at two exposed positions were compared to the calculated results obtained by a general Monte Carlo code MCNP5, and good agreements were found. The applicability of a homemade dose survey meter installed at EAST was also verified with the comparison of the ambient dose equivalent H*(10) values measured by the meter and BSS.

Structural and optical properties improvements of PVP/gelatin blends induced by neutron irradiation

Blends of polyvinylpyrrolidone and gelatin were prepared in three different concentrations to study the modifications in their structural and optical properties induced by neutron irradiations with different neutron fluence values from ~ 108 up to ~ 1011 neutron/cm2. X-ray spectroscopy revealed that the irradiation has induced a recrystallization phenomenon in the studied blends and the crystallinity index increased by increasing the neutron fluence due to the breaking of the crystallites. Fourier-transform infrared spectroscopy came to confirm the existence of interactions between interchain groups and a higher compatibility for the irradiated blends. The irradiation induced defects inside the material were responsible for the change in their optical and structural properties. The creation of free radicals or ions inside the conduction bands has led to the increase in the number of carriers on localized states; this has caused the increase in optical conductivity of the irradiated blends as a result of decreasing the energy gaps by increasing the neutron fluence. Results may widen the applications of the gelatin based blends to include optoelectronic devices, organic light emitting devices, solar selective and anti-reflectance bio-coatings, optical organic glass and lenses.

Characterization of Radiation Fields for Assessing Concrete Degradation in Biological Shields of NPPs

Life extensions of nuclear power plants (NPPs) to 60 years of operation and the possibility of subsequent license renewal to 80 years have renewed interest in long-term material degradation in NPPs. Large irreplaceable sections of most nuclear generating stations are constructed from concrete, including safety-related structures such as biological shields and containment buildings; therefore, concrete degradation is being considered with particular focus on radiation-induced effects. Based on the projected neutron fluence values (E > 0.1 MeV) in the concrete biological shields of the US pressurized water reactor fleet and the currently available data on radiation effects on concrete, some decrease in mechanical properties of concrete cannot be ruled out during extended operation beyond 60 years. An expansion of the irradiated concrete database is desirable to ensure reliable risk assessment for extended operation of nuclear power plants.

Neutron deep penetration through reactor pressure vessel and biological concrete shield of VVER-1000 Mock-Up in LR-0 reactor

Evaluation of the neutron fluence values in a reactor pressure vessel together with surveillance specimens programs from reactor pressure vessel materials are among the most important parts of in-service inspection programs that are necessary for realistic and reliable assessment of the RPV’s residual lifetime. The neutron fluence values are determined by means of calculation. These calculation results are accompanied by measurements of induced activity of the activation foil placed in capsules behind the reactor pressure vessel at selected locations. These results can be used for neutron flux density adjustment in the corresponding parts of the reactor system. Such information is not complete, because for determination of the radiation damage, the flux on the inner side of reactor pressure vessel has to be known. If the flux attenuation ratio is known, then the neutron flux density on the outer side can be easily recalculated for the inner side. These attenuation factors, applicable to the power reactors, can be determined experimentally in the VVER-1000 Mock-Up placed in the LR-0 reactor. The comparison of experimental results with calculated ones increases the reliability of their estimations. This paper aims to evaluate the measured neutron flux densities at various VVER-1000 Mock-Up positions by means of comparison with the Monte Carlo simulation. The effect of the various nuclear data libraries on the calculation results is also studied.

Characterization of Radiation Fields in Biological Shields of Nuclear Power Plants for Assessing Concrete Degradation*

Life extensions of nuclear power plants to 60 and potentially 80 years of operation have renewed interest in long-term material degradation. One material being considered is concrete, with a particular focus on radiation-induced effects. Based on the projected neutron fluence values (E > 0.1 MeV) in the concrete biological shields of the US pressurized water reactor fleet and the available data on radiation effects on concrete, some decrease in mechanical properties of concrete cannot be ruled out during extended operation beyond 60 years. An expansion of the irradiated concrete database and a reliable determination of relevant neutron fluence energy cutoff value are necessary to ensure reliable risk assessment for extended operation of nuclear power plants.

Experimental characterization of semiconductor-based thermal neutron detectors

In the framework of NESCOFI@BTF and NEURAPID projects, active thermal neutron detectors were manufactured by depositing appropriate thickness of 6LiF on commercially available windowless p–i–n diodes. Detectors with different radiator thickness, ranging from 5 to 62μm, were manufactured by evaporation-based deposition technique and exposed to known values of thermal neutron fluence in two thermal neutron facilities exhibiting different irradiation geometries. The following properties of the detector response were investigated and presented in this work: thickness dependence, impact of parasitic effects (photons and epithermal neutrons), linearity, isotropy, and radiation damage following exposure to large fluence (in the order of 1012cm−2).

A review of the present state of the absolute calibration for zircon fission track geochronometry using the external detector method

AbstractThe authors have developed and demonstrated an independent calibration of zircon fission track dating with the external detector method. During the past two decades, investigations have continued concerning the issues of the absolute age calibration of fission track dating. The derivation of accurate fission track age equations has been made possible by taking into account the relationships between the latent and etched track lengths. Our dating exercise involving 10 zircon reference samples supports our claim that absolute age determinations can be carried out based on the following three parameters: (1) the 238U spontaneous fission decay constant of 8.5 × 10−17 a−1 (λf) which is recommended by the IUPAC; (2) thermal neutron fluence values based on pre‐irradiated IRMM540 dosimeter glass, which is equivalent to the Au and Co metal activation monitors in the 1990 IUGS recommendation; and (3) a [GQR] correction value of 1.35 which has been experimentally determined for the combination of zircon external surfaces irradiated with mica external detectors. Independent measurements of uranium concentration by the fission track method were also tested. A new method of fission track dating that uses mass spectrometry for uranium measurements was simulated with success using our zircon data. We propose that the Fission Track Community reassesses the situation and formulates a new recommendation, in which the established independent method is placed on at least an equal footing with the standard‐based method.

An approach to predicting the limiting state of reactor pressure vessel metal on the basis of the mechanical stability concept

An essentially new approach to predicting a reactor pressure vessel service life under radiation has been elaborated. Its special feature is that the limiting value of neutron fluence is determined directly by the condition of brittle fracture initiation in the irradiated steel in a local region at the tip of a crack-like defect in the vessel wall, rather than by an indirect characteristic (a critical shift of the failure temperature for a Charpy specimen). The authors put forward a criterion for the local fracture initiation due to the loss of stability of plastic state of the irradiated metal in the “process zone” in the vicinity of a macrocrack during the reactor vessel emergency loading (thermal shock). The potentialities of the new approach are illustrated by the example of predicting the critical neutron fluence for WWER-1000 reactor pressure vessel.

Measurement of peak fluence of neutron beams using Bi-fission detectors

Fission fragments and other charged particles leave tracks of permanent damage in most of the insulating solids. Damage track detectors are useful for personal dosimeters and for flux/dose determination of high-energy particles from accelerators or cosmic rays. A detector that has its principal response at nucleon energy above 50 MeV is provided by the fission of Bi-209. Neutrons produce the largest percentage of hadron dose in most high-energy radiation fields. In these fields, the neutron spectrum is typically formed by low-energy neutrons (evaporation spectrum) and high-energy neutrons (knock-on spectrum). We used Bi-fission detectors to measure neutron peak fluence and compared the result with the calculated value of neutron peak fluence. For the exposure to 100 MeV we have used the iThemba Facility in South Africa.

Neutron field measurements of the CRNA OB26 irradiator using a bonner sphere spectrometer for radiation protection purposes

The present work deals with the Bonner sphere spectrometer (BSS) measurements performed, to support the authors' Monte-Carlo calculations, to estimate accurately the main characteristics of the neutron field of the (241)Am-Be-based OB26 irradiator acquired for radiation protection purposes by the Nuclear Research Centre of Algiers. The measurements were performed at a reference irradiation position selected at 150 cm from the geometrical centre of the neutron source. The spectrometric system in use is based on a central spherical (3)He thermal neutron proportional counter. The response matrix of the present spectrometer has been taken to be similar to the original Physikalisch-Technische Bundesanstalt (PTB) (Braunschweig, Germany) BSS's response matrix, with a five bins per decade energy group structure, as there is no significant difference in the BSS's physical characteristics. Thereafter, the authors' BSS measurements were used together with MCNP5 results to unfold the neutron spectrum by means of MAXED and GRAVEL computer codes from the U.M.G. 3.3 package, developed at PTB. Besides, sensitivity analysis has been performed to test the consistency of the unfolding procedure. It reveals that no significant discrepancy was observed in the total neutron fluence and total ambient dose values following the perturbation of some pertinent unfolding parameters except for the case where a 10 bins energy structure was assumed for the guess spectrum. In this latter case, a 5 % difference was observed in the ambient dose equivalent compared with the reference case. Finally, a comparative study performed between different counting systems together with MCNP5 and predictive formulas results shows that they were globally satisfactory, highlighting thereby the relevance of the unfolding procedure and the reliability of the obtained results.