Reactor Neutron Spectrum Research Articles

Toward the Development of an InSb-Based Neutron-Resistant Hall Sensor

We describe a study concerning two irradiation series of InSb thin films in the form of Hall structures, using the natural neutron spectrum of a thermal-type fission reactor. We used mono- and polycrystalline InSb thin films with various levels of donor doping. The range of electron concentration values of the samples (from $10^{{16}}$ to $10^{{18}}$ cm $^{{-3}}$ ) allowed us to study the impact of predominant thermal neutrons on various InSb structures. Preirradiation long-term annealing of the samples at high temperatures ensured the thermal stabilization of the electrical parameters and made it possible to assess only defects introduced by neutrons in postirradiation heat treatments, without the influence of postevaporation defects. Neutron fluences of $8.7\times 10^{ {17}}$ cm $^{ {-2}}$ and $1.2\times 10^{ {18}}$ cm $^{ {-2}}$ were applied sequentially, and the total fluence was above $2\times 10^{ {18}}$ cm $^{ {-2}}$ . Annealing at high temperatures was performed after each irradiation process to estimate the quantity of donors created and to remove defects arising in the InSb structure. The results demonstrate the possibility of application of the InSb thin film in neutron-resistant Hall sensor (NRHS) for modern technological settings: fusion and fission reactors, particle accelerators, cosmic, and other scientific research.

Investigation of an Online Reactor Neutron Spectrum Measurement Method with Ionization Chambers

In this paper, an online reactor neutron spectrum measurement method is presented. The basic theory of this method is based on the unfolding of few-channel data, in which three miniature ionization chambers are applied. The neutron spectrum can be unfolded with the count rates and response functions of the three detectors through an unfolding program. In order to investigate the feasibility of this method, simulation tests have been performed with the reference neutron spectra and neutron spectra from the China LEAd-based Reactor (CLEAR). The research results show that this method can provide an alternative means for an online neutron spectrum measurement in the reactors. This method is suitable to be applied in fast neutron reactors due to the miniature size of ionization chambers and fission threshold of 238U.

How to Produce a Reactor Neutron Spectrum Using a Proton Accelerator

A method for reproducing the neutron energy spectrum present in the core of an operating nuclear reactor using an engineered target in an accelerator proton beam is proposed. The protons interact with a target to create neutrons through various (p,n) type reactions. Spectral tailoring of the emitted neutrons can be used to modify the energy of the generated neutron spectrum to represent various reactor spectra. Through the use of moderators and reflectors, the neutron spectrum can be modified to reproduce many different spectra of interest including spectra in small thermal test reactors, large pressurized water reactors, and fast reactors. The particular application of this methodology is the design of an experimental approach for using an accelerator to measure the betas produced during fission to be used to reduce uncertainties in the interpretation of reactor antineutrino measurements. This approach involves using a proton accelerator to produce a neutron field representative of a power reactor, and using this neutron field to irradiate fission foils of the primary isotopes contributing to fission in the reactor, creating unstable, neutron rich fission products that subsequently beta decay and emit electron antineutrinos. A major advantage of an accelerator neutron source over a neutron beam from a thermal reactor is that the fast neutrons can be slowed down or tailored to approximate various power reactor spectra. An accelerator based neutron source that can be tailored to match various reactor neutron spectra provides an advantage for control in studying how changes in the neutron spectra affect parameters such as the resulting fission product beta spectrum.

Geochemical Investigations for Uranium in Some Areas of Jharkhand State Using Fission Track Technique

Human population is always exposed to ionizing radiations from natural sources present in the earth crust. Hence the analysis of uranium in soil sample has great significance due to health hazards to human beings. For the purpose, some different soil samples collected from some area of Jharkhand have been analysed for trace uranium concentration using the fission track technique. Lexan polycarbonate was used as detector for recording fission tracks. As reactor neutron spectra is associated with both thermal and fast neutron fluxes; correction to the present uranium data due to fast neutron fission of 232Th was also applied. The uranium contents were estimated by comparing the track densities detectors immersed in the sample and the standard uranium solutions, irradiated along with the samples under the same irradiation conditions. The uranium in the soil samples were found to vary from 209 ng/g to 991 ng/g.

Effective Atomic Displacements in α-Zirconium under Neutron Irradiation

A computer simulation of atomic displacements cascades in α-zirconium is performed by molecular dynamics method. The dependence of the fraction of surviving defects on the PKA energy (in the range from 0.1 to 20 keV) is revealed for the initial temperature of the material of 0 and 600 K. Group neutron cross sections of the effective displacement under neutron irradiation are prepared. The effective dpa rates for zirconium in neutron spectra of different reactors are evaluated. These values are about four times lower than the corresponding values of conventional dpa rates for neutron energies from 0.5 MeV and higher.

Radiochemical measurement of neutron-spectrum averaged cross sections for the formation of 64Cu and 67Cu via the (n,p) reaction at a TRIGA Mark-II reactor: Feasibility of simultaneous production of the theragnostic pair 64Cu/67Cu

Abstract Integral cross sections of the 64Zn(n, p)64Cu and 67Zn(n, p)67Cu reactions were measured for the fast neutron spectrum of TRIGA Mark-II reactor at Savar, Dhaka, Bangladesh. A clean radiochemical separation was performed to isolate the copper radionuclides from the target element zinc. The radioactivities produced in the irradiation were measured by HPGe γ-ray spectroscopy. The neutron flux over the energy range 0.5–20 MeV was determined using the 58Ni(n, p)58Co monitor reaction. The measured results amount to 28.9 ± 2.0 mb and 0.84 ± 0.07 mb for the formation of 64Cu and 67Cu, respectively. These values are slightly lower than the respective values for a pure fission spectrum. The present results were compared with data calculated using the neutron spectral distribution and the recently critically analysed excitation function of each reaction given in the literature. The good agreement validates the reliability of those excitation functions. The feasibility of simultaneous production of 64Cu and 67Cu with fast neutrons is discussed.

Compilation of Existing Neutron Screen Technology

The presence of fast neutron spectra in new reactors is expected to induce a strong impact on the contained materials, including structural materials, nuclear fuels, neutron reflecting materials, and tritium breeding materials. Therefore, introduction of these reactors into operation will require extensive testing of their components, which must be performed under neutronic conditions representative of those expected to prevail inside the reactor cores when in operation. Due to limited availability of fast reactors, testing of future reactor materials will mostly take place in water cooled material test reactors (MTRs) by tailoring the neutron spectrum via neutron screens. The latter rely on the utilization of materials capable of absorbing neutrons at specific energy. A large but fragmented experience is available on that topic. In this work a comprehensive compilation of the existing neutron screen technology is attempted, focusing on neutron screens developed in order to locally enhance the fast over thermal neutron flux ratio in a reactor core.

55Fe effect on enhancing ferritic steel He/dpa ratio in fission reactor irradiations to simulate fusion conditions

This study evaluated methods for increasing the helium production rate in ferritic steel irradiation in a fission reactor neutron spectrum in order to increase the helium to atomic displacement ratio to values typical of fusion reactor first wall conditions. An early experiment showed that the accelerated He(appm)/dpa ratio of about 2.3 was achieved for 96% enriched 54Fe in iron in the High Flux Isotope Reactor (HFIR), ORNL. In the current work, the ferritic steel He(appm)/dpa ratio was studied in the neutron spectrum of HFIR with the 55Fe thermal neutron helium production taken into account. A benchmark calculation for the same sample, as used in the aforementioned experiment, was then used to adjust and evaluate the 55Fe (n, a) cross section values in TALYS-based Evaluated Nuclear Data Library (TENDL). The analysis showed that a decrease of a factor of 6700 for the TENDL 55Fe (n, a) cross section in the intermediate and low energy regions was required in order to fit the experimental results. The best fit to the cross section value at thermal neutron energy was about 27mb. With the adjusted 55Fe (n, a) cross sections, calculation showed that the 54Fe and 55Fe isotopes could be enriched by the isotopic tailoring technique in a ferritic steel sample irradiated in HFIR to significantly enhance the helium production rate. This new calculation can be used to guide future isotopic tailoring experiments designed to increase the He(appm)/dpa ratio in fission reactors. A benchmark experiment is suggested to be performed to evaluate the 55Fe (n, a) cross section at thermal energy.

Nuclear physical properties of zirconium alloys E110 and E635 under long-term neutron irradiation in VVER-1000 reactor

For zirconium alloys of the type E110 (Zr-Nb) and E635 (Zr-Nb-Sn-Fe), the dynamics of accumulation (variation) of concentrations of alloying elements and impurities, including hydrogen and helium, the behavior of primary radiation damage and radioactivity under irradiation up to 50 years in the neutron spectrum of a VVER-1000 fission reactor, and the decrease (cooling) for a period of up to 1000 years after ending of the irradiation are computed using the ACDAM computational system, which includes libraries of recommended microscopic cross sections of neutron reactions, activation cross sections, and decay data. Alloys E110 and E635 when irradiated for more than four years do not meet “low activation material” criteria; they cannot be put in a recycling process for re-fabrication but must be packaged and buried for a time exceeding 1000 years.

Enhancement of thermal neutron self-shielding in materials surrounded by reflectors

Materials containing from 41 to 1124 mg chlorine and surrounded by polyethylene containers of various thicknesses, from 0.01 to 5.6 mm, were irradiated in a research reactor neutron spectrum and the 38Cl activity produced was measured as a function of polyethylene reflector thickness. For the material containing the higher amount of chlorine, the 38Cl specific activity decreased with increasing reflector thickness, indicating increased neutron self-shielding. It was found that the amount of neutron self-shielding increased by as much as 52% with increasing reflector thickness. This is explained by neutrons which have exited the material subsequently reflecting back into it and thus increasing the total mean path length in the material. All physical and empirical models currently used to predict neutron self-shielding have ignored this effect and need to be modified. A method is given for measuring the adjustable parameter of a self-shielding model for a particular sample size and combination of neutron reflectors.

MAX phase carbides and nitrides: Properties for future nuclear power plant in-core applications and neutron transmutation analysis

Abstract A family of ternary carbides and nitrides, known as MAX phases, combine attractive properties of both ceramics and metals, and has been suggested for potential nuclear reactor applications. The unirradiated materials properties of importance for in-core structural materials and as fuel pellet coatings for several leading MAX phase materials have been summarized from literature. The materials show high mechanical damage tolerance in terms of creep, thermal/mechanical fatigue and fracture resistance, and very good chemical compatibility with select coolants such as molten lead and sodium. Neutron activation has been calculated for commercial purity materials exposed to both idealized fast and thermal reactor neutron spectra for 10, 30, and 60 years of exposure. The specific activities of Ti 3 SiC 2 , Ti 3 AlC 2 , and Ti 2 AlC were compared to those of SiC and Alloy 617, two leading candidate materials for next generation reactor components. The specific activities of MAX phases were similar to SiC and three orders of magnitude less than Alloy 617 after 10–60 years decay for all three activation times in both the fast and thermal spectra. As with SiC, the main radioisotopes after a decay period of 10 years for all three activation times in the MAX phases are tritium and C 14 . Neutron irradiation results of Ti 3 SiC 2 , Ti 3 AlC 2 , and Ti 2 AlC experimentally confirmed the neutron transmutation analysis.

Influence of the Neutron Flux Characteristic Parameters in the Irradiation Channels of Reactor on NAA Results Using k<sub>0</sub>-Standardization Method

An approximation method using to estimate the influence of the uncertainties of the neutron flux characteristic parameters in the irradiation positions on the NAA results using k0-standardization technique was presented. Those are the epithermal reactor neutron spectrum shape-factor α, the effective resonace energy Ε for a given nuclide and the thermal to epithermal neutron flux ratio f. The method is applied to estimate the effect of the uncertainties in the determination of α Ε, and f on final NAA results for some irradiation channels of the Dalat reactor. It also shows that presented method is suitable in practical use for the estimation of the errors due to the uncertainty of the neutron flux characteristic parameters at the irradiation position.

Development and implementation of Høgdahl convention and Westcott formalism for k0-INAA application at Malaysian Nuclear Agency reactor

The present work shows the development of k0-instrumental neutron activation analysis (k0-INAA) method at the Malaysian Nuclear Agency research reactor. To use the k0-INAA method, two formalisms were regulated according to 1/ν and non-1/ν (n, γ) reaction nuclides. The reactor neutron spectrum parameters, the thermal to epithermal neutron flux ratio (f), the epithermal neutron flux shape factor (α) were measured using the bare bi-isotopic monitor and bare triple monitor methods, respectively, based on the Hogdahl convention. In addition, the modified spectral index $$ r(\alpha )\sqrt {T_{n} /T_{0} } $$ , the Westcott $$ g_{Lu} (T_{n} ) $$ factor and the absolute neutron temperature T n parameters were determined using the Westcott formalism. 176Lu was used as non- 1/ν monitor while 197Au, 96Zr and 94Zr were used as 1/ν monitors. The average values of $$ r(\alpha )\sqrt {T_{n} /T_{0} } ,\,g_{Lu} (T_{n} ) $$ and T n were determined to be 0.1795 ± 0.0044, 1.9729 ± 0.0234 and 50.12 ± 3.21°C, respectively. The accuracy of the method was evaluated by analysing IAEA-Soil 7, IAEA-SL 1, NBS SRM 1633A-1 and IAEA-Soil 375 as reference materials. The results show an acceptable level of consistency.

Neutron self-shielding in irradiation channels of small reactors is isotropic

In order to predict the self-shielding effect in neutron activation of non-spherical samples in reactor neutron spectra, it is important to know whether the neutron field is sufficiently anisotropic to cause significant variations between horizontal and vertical sample orientation or with orientation relative to the direction towards the reactor core. Metal wires with significant neutron self-shielding were irradiated in several channels of the SLOWPOKE reactor at Ecole Polytechnique Montreal and the TRIGA reactor at the Jožef Stefan Institute. In all cases, the amount of thermal or epithermal self-shielding was found to be identical, within the experimental uncertainty, regardless of the orientation of the wire, indicating that the neutron field is essentially isotropic. Models used to predict neutron self-shielding need to be adjusted accordingly. In our Monte Carlo model, the tube-shaped neutron source was moved back into the moderator and reflecting materials near the sample location were included, which produced an isotropic neutron field at the sample location.

Determination of trace amount of nickel in cosmic spherules by photon activation analysis

Trace amount of nickel in high-purity iron and iron oxide samples was determined by photon activation analysis (PAA) using the Electron Linear Accelerator at the Research Center for Electron Photon Science, Tohoku University. It was confirmed that about 0.015 μg of nickel in the high-purity iron oxide sample could be determined with an experimental uncertainty of 20% by PAA, that is, it was possible to determine 1% of nickel at least in a few μg of cosmic spherules. PAA was applied to the nickel-determination in cosmic spherules. Nickel in cosmic spherules was determined by PAA more accurately than by instrumental neutron activation analysis done in a reactor neutron spectrum with the fluence rate ratio of thermal neutrons to fast of 4–5 which has so far been applied to the analysis of spherule samples. Additionally, trace amount of nickel in the reference material Giant Clam JCt-1 supplied by the Geological Survey of Japan was determined by PAA with radiochemical treatment.

Irradiation tests of ITER candidate Hall sensors using two types of neutron spectra

We report on irradiation tests of InSb based Hall sensors at two irradiation facilities with two distinct types of neutron spectra. One was a fission reactor neutron spectrum with a significant presence of thermal neutrons, while another one was purely fast neutron field. Total neutron fluence of the order of 10(16) cm(-2) was accumulated in both cases, leading to significant drop of Hall sensor sensitivity in case of fission reactor spectrum, while stable performance was observed at purely fast neutron spectrum. This finding suggests that performance of this particular type of Hall sensors is governed dominantly by transmutation. Additionally, it further stresses the need to test ITER candidate Hall sensors under neutron flux with ITER relevant spectrum.

Investigation of the impurity concentrations and product activities of some targets used in radioisotope production: a neutron activation analysis study

Neutron activation analysis using the k0 standardization method (k0-NAA) and fast neutron activation analysis using reactor fission neutrons, were used to determine the impurity concentrations in WO3, MoO3, SnO2 and TeO2 targets. The radioimpurities 124Sb, 134Cs, 60Co, 87Rb, 182Ta, 233Pa, 65Zn, 59Fe, 110mAg, 51Cr, 95Zr, 75Se and 114mIn were found in the irradiated targets and their origin either neutron capture reactions, or threshold reactions or both were identified. The specific activities of 187W, 188Re, 99mTc, 113mIn, 117mSn and 131I radioisotopes were determined. It is shown that the epithermal neutron flux significantly contributes to the isotopes characterized with high Q0 values. It is shown that the 117mSn isotope can be produced with a high specific activity using the fast neutron component of the reactor neutron spectrum.

Measurement of the fission neutron spectrum averaged cross sections for the 95Mo(n, p)95Nb, 92Mo(n, α)89Zr, 90Zr(n, 2n)89Zr and 60Ni(n, p)60Co reactions

The average cross-sections in a fission-type reactor neutron spectrum are determined experimentally for the reactions: (92)Mo (n, a) (89)Zr, (90)Zr (n, 2n) (89)Zr and (60)Ni (n, p) (60)Co. The determined cross sections are in the ranges of the reported values in literature. A method for determining the cross sections in relative way is introduced. It replaces the cross sections and the nuclear data by a composite nuclear data constant which can be calculated and determined experimentally. It is shown that the uncertainties of determining these constants experimentally are mainly lower than the calculated ones. The consequences of using this constant in neutron activation analysis are briefly discussed.

Modified method of α determination in the 1/E1+α epithermal neutron spectrum of reactor.

A modified method has been developed for the α determination in the 1/E 1+α epithermal neutron spectrum of reactor. It is based on the Cd-covered and without Cd-covered irradiations of two monitors. This method was applied to determine the α value in the channels of Dalat reactor and the results were compared with those obtained by the other methods. It appeared that the results of the modified method were in quite good agreement with those of other methods. It also showed that the modified method was simple in practical uses and a good application in the experiment of α determination in the reactor irradiation channels.

Measurements of the thermal neutron cross-sections and resonance integrals for 186W (n,γ) 187W and 98Mo (n,γ) 99Mo reactions

The thermal neutron cross-sections and resonance integrals of the 186W (n,γ) 187W and 98Mo (n,γ) 99Mo reactions in the thermal and 1/E regions, respectively, of a thermal reactor neutron spectrum have been experimentally determined by the activation method using 197Au (n,γ) 198Au reaction as a single comparator. The high purity natural W, Mo, and Zr foils; and Au wire diluted in aluminum, were irradiated without Cd shield in two neutron irradiation sites, characterized with different values for the thermal-to-epithermal flux ratios, f at the Second Egyptian Research Reactor (ETRR-2). The induced activities in the samples were measured by high-resolution γ-ray spectrometry with a calibrated germanium detector. Thermal neutron cross-sections for 2200 m/s neutrons and resonance integrals for the 186W (n,γ) 187W and 98Mo (n,γ) 99Mo reactions have been obtained relative to the reference values, σ0 = 98.65 ± 0.09 b and I 0 = 1500 ± 28 b for the 197Au (n,γ) 198Au reaction. The necessary correction factors for thermal neutron and resonance neutron self-shielding effects, and the epithermal flux index (α) were taken into account in the determinations. The results obtained were: σ0 = 38.43 ± 0.4 b and I 0 = 502 ± 65 b for 186W (n,γ) 187W, and σ0 = 0.137 ± 0.014 band I 0 = 6.47 ± 0.8 for 98Mo (n,γ) 99Mo. These results are discussed and compared with previous measurements and evaluated data in literature. The traditional method of determining thermal cross-sections and resonance integrals via neutron irradiation with and without Cd shield in one irradiation position was avoided in this work by neutron irradiation without Cd shield in at least two different neutron irradiation positions. This method provides alternative way for determining thermal cross-sections and resonance integrals simultaneously.