CR-39 detector based thermal neutron flux measurements, in the photo neutron project

PhoNeS (photo neutron source) is a project aimed at the production and moderation of neutrons by exploiting high energy linear accelerators, currently used in radiotherapy. A feasibility study has been carried out with the scope in mind to use the high energy photon beams from these accelerators for the production of neutrons suitable for boron neutron capture therapy (BNCT). Within these investigations, it was necessary to carry out preliminary measurements of the thermal neutron component of neutron spectra, produced by the photo-conversion of X-ray radiotherapy beams supplied by three LinAcs: 15 MV, 18 MV and 23 MV. To this end, a simple passive thermal neutron detector has been used which consists of a CR-39 track detector facing a new type of boron-loaded radiator. Once calibrated, this passive detector has been used for the measurement of both the thermal neutron component and the cadmium ratio of different neutron spectra. In addition, bubble detectors with a response highly sensitive to thermal neutrons have also been used. Both thermal neutron detectors are simple to use, very compact and totally insensitive to low-ionizing radiation such as electrons and X-rays. The resultant thermal neutron flux was above 10⁶ n/cm²s and the cadmium ratio was no greater than 15 for the first attempt of photo-conversion of X-ray radiotherapy beams.     

Neutron flux determination in irradiation sites of an Am–Be neutron source at NNRI

Neutron flux measurements and flux distribution parameters for two irradiation sites of an Am–Be neutron source irradiator were measured by using gold (Au), zirconium (Zr) and aluminum (Al) foils. thermal neutron flux Φ_th = 1.46 × 10⁴ n cm⁻² s⁻¹ ± 0.01 × 10², epithermal neutron flux Φ_epi = 7.23 × 10² n cm⁻² s⁻¹ ± 0.001, fast neutron flux Φ_f = 1.26 × 10² n cm⁻² s⁻¹ ± 0.020, thermal-to-epithermal flux ratio f = 20.5 ± 0.36 and epithermal neutron shaping factor α = −0.239 ± 0.003 were found for irradiation Site-1; while the thermal neutron flux Φ_th = 4.45 × 10³ n cm⁻² s⁻¹ ± 0.06, the epithermal neutron Φ_epi = 1.50 × 10² n cm⁻² s¹ ± 0.003, the fast neutron flux Φ_f = 1.17 × 10 n cm⁻² s⁻¹ ± 0.011, thermal-to-epithermal flux ratio f = 29.6 ± 0.94, and epithermal neutron shaping factor α = 0.134 ± 0.001 were found for irradiation Site-2. It was concluded that the Am–Be neutron source can be used for neutron activation analysis (NAA). The Am–Be source can be used for neutron activation analysis thereby reducing the burden on GHARR-1 and increasing the research output of the nation.     

Measurements of the Y(n,γ)Y cross-section in the neutron energy range of 13.5-14.6 MeV

The ⁸⁹Y(n,γ)^90mY cross-section has been measured at three neutron energy points between 13.5 and 14.6 MeV using the activation technique and a coaxial HPGe γ-ray detector. The data for the ⁸⁹Y(n,γ)^90mY cross-sections are reported to be 0.39 ± 0.02, 0.43 ± 0.02, and 0.38 ± 0.02 mb at 13.5 ± 0.2, 14.1 ± 0.1, and 14.6 ± 0.2 MeV incident neutron energies, respectively. The first data for the ⁸⁹Y(n,γ)^90mY reaction at neutron energy points of 13.5 and 14.1 MeV are presented. The natural high-purity Y₂O₃ powder was used as target material. The fast neutrons were produced by the T(d,n)⁴He reaction. Neutron energies were determined by the method of making cross-section ratios of ⁹⁰Zr(n,2n)^89m+gZr and ⁹³Nb(n,2n)^92mNb reactions, and the neutron fluencies were determined using the monitor reaction ⁹³Nb(n,2n)^92mNb. The results obtained are compared with existing data.     

Flux measurements in a nuclear research reactor by using an aluminum nitride detector

A small polycrystalline aluminium nitride detector with a thickness of 381 μm was used to measure a 200,000 Ci Co⁶⁰ source and to measure the flux in a research reactor where the neutron flux is about 10¹⁴/cm² s, which is nearly the same order as in the commercial power plant. If the applied voltage is greater than or equal to 2000 V and if the measurements are done in a short period of time so that the heat energy does not build up in the aluminium nitride, then the measured electric current is linearly proportional to the input flux. It is assumed of course that the energy spectrum of the input flux remains constant. This linearity relation is illustrated by the results of a measurement in which the reactor power has been controlled so that the flux becomes a step function.     

Measurements of activation cross-sections for the F(n, α)N reaction for neutrons with energies between 13 and 15 MeV

In this study, activation cross-sections were measured for the ¹⁹F(n, α)¹⁶N reaction at six different neutron energies from 13.5 and 14.9 MeV. The fast neutrons were produced via the ³H(d,n)⁴He reaction on SAMES T-400 neutron generator. The cyclic activation technique was used. Induced gamma activities were measured by a high-resolution gamma-ray spectrometer with high-purity germanium (HpGe) detector. Measurements were corrected for gamma-ray attenuations, random coincidence (pile-up), dead time and fluctuation of neutron flux. Results were compared with the previous works.     

Studies on the overall safety aspects during irradiation of TeO2 in the central thimble of the TRIGA research reactor

The purpose of this study is to provide a detailed safety analysis of overall system and components in terms of their ability to provide optimum output from the irradiation of TeO₂ in the central thimble of the 3 MW TRIGA MARK II research reactor at Atomic Energy Research Establishment (AERE), Savar, Dhaka, Bangladesh. It identifies safety issues relevant to ¹³¹I radioisotope production and ensures that safety analysis and design are consistent. It evaluates threats developed within the facility during the irradiation process and ultimately ensures establishment of in-core safety limits and conditions at all stages of ¹³¹I production. In-core irradiation safety not only ensures the safe operation of the reactor but also strengthens the production of radioisotopes (RI). This study attempts to review and modify all safety related events and aspects relating to RI production. The three-dimensional continuous energy Monte Carlo code MCNP is used to develop a versatile and accurate full-core model of the TRIGA core. The cross-section library and fission product inventory are generated by using NJOY and ORIGEN computer codes. The methodology to evaluate heat generation and other relevant parameters necessary to provide enough information for thermal hydraulic analysis are discussed. The neutron flux distribution inside the dry and water filled central thimble is determined in order to locate the highest neutron flux trapping position. The thermal hydraulic and safety analysis are performed by elaborate numerical analysis as well as by using GENGTC computer code. A mock-up facility has also been developed to supplement and verify the theoretically predicted results. The total energy generated during irradiation of 50 gm TeO₂ sample in dry condition is found to be 113.84 w of which 75% energy is due to neutron heating and rest of the amount is from gamma heating. Around 11.28 w of heat energy is also generated in the quartz vial. When the total generated-heat transfer is considered through conduction and radiation mechanisms, the calculated temperature of 50 g of TeO₂ reaches at 970 °C. Considering simultaneous heat transfer mechanisms, (conduction, radiation and convection) the calculated maximum temperature of the 50 g of TeO₂ powder comes down at 680 °C. It may be pointed out that very high amount of heat is generated during the irradiation of TeO₂ at 3 MW reactor power in dry condition which is nearly the melting point of TeO₂ and may be termed as unsafe mode of irradiation.     

Neutron energy spectrum flux profile of Ghana’s miniature neutron source reactor core

The total neutron flux spectrum of the compact core of Ghana’s miniature neutron source reactor was understudied using the Monte Carlo method. To create small energy groups, 20,484 energy grids were used for the three neutron energy regions: thermal, slowing down and fast. The moderator, the inner irradiation channels, the annulus beryllium reflector and the outer irradiation channels were the region monitored. The thermal neutrons recorded their highest flux in the inner irradiation channel with a peak flux of (1.2068 ± 0.0008) × 10¹² n/cm² s, followed by the outer irradiation channel with a peak flux of (7.9166 ± 0.0055) × 10¹¹ n/cm² s. The beryllium reflector recorded the lowest flux in the thermal region with a peak flux of (2.3288 ± 0.0004) × 10¹¹ n/cm² s. The peak values of the thermal energy range occurred in the energy range (1.8939–3.7880) × 10⁻⁰⁸ MeV. The inner channel again recorded the highest flux of (1.8745 ± 0.0306) × 10⁰⁹ n/cm² s at the lower energy end of the slowing down region between 8.2491 × 10⁻⁰¹ MeV and 8.2680 × 10⁻⁰¹ MeV, but was over taken by the moderator as the neutron energies increased to 2.0465 MeV. The outer irradiation channel recorded the lowest flux in this region. In the fast region, the core, where the moderator is found, the highest flux was recorded as expected, at a peak flux of (2.9110 ± 0.0198) × 10⁰⁸ n/cm² s at 6.961 MeV. The inner channel recorded the second highest while the outer channel and annulus beryllium recorded very low flux in this region. The flux values in this region reduce asymptotically to 20 MeV.     

Measurements of neutron total cross-sections and resonance parameters of erbium at the Pohang Neutron Facility

We measured neutron total cross-sections of natural erbium in the neutron energy region from 0.2 to 120 eV by using the neutron time-of-flight method at the Pohang Neutron Facility, which consists of an electron linear accelerator, a water-cooled tantalum target with a water moderator, and a 12-m-long time-of-flight path. A ⁶Li-ZnS(Ag) scintillator with a diameter of 12.5 cm and a thickness of 1.6 cm was used as a neutron detector, and a group of high-purity natural erbium metallic plates with various thickness was used for the neutron transmission measurements. The present measurement was compared with the existing experimental and the evaluated data. The resonance parameters of ¹⁶⁶Er, ¹⁶⁷Er, ¹⁶⁸Er, and ¹⁷⁰Er in the neutron energy region below 120 eV were extracted from the transmission by using the multilevel R-matrix SAMMY code and were compared with the evaluated data from ENDF/B VII.0 and other previous reported results.     

On the impact of neutron beam divergence and scattering on the quality of transmission acquired tomographic images

The impact of the divergence of a thermal neutron beam and the scattered neutrons on the quality of tomographic images acquired by transmission have been evaluated by using a third generation tomographic system incorporating neutron collimators under several different arrangements. The system equipped with a gaseous position sensitive detector has been placed at the main channel outlet of the Argonauta Research Reactor in Instituto de Engenharia Nuclear (CNEN-Brazil) which furnishes a thermal neutron flux of 2.3 × 10⁵ n cm⁻² s⁻¹. Experiments have then been conducted using test-objects with well-known inner structure and composition to assess the influence of the collimators arrangement on the quality of the acquired images. Both, beam divergence and scattering - expected to spoil the image quality - have been reduced by using properly positioned collimators between the neutron source and the object, and in the gap between the object and the detector, respectively. The shadow cast by this last collimator on the projections used to reconstruct the tomographic images has been eliminated by a proper software specifically written for this purpose. Improvement of the tomographic images has been observed, demonstrating the effectiveness of the proposed approach to improve their quality by using properly positioned collimators.     

Measurement of thermal neutron cross section and resonance integral for (n,γ) reaction in Sm

This study implies that ⁵⁵Mn(n,γ)⁵⁵Mn monitor reaction may be a convenient alternative comparator for the activation method and thus, it was used for the determination of thermal neutron cross section (TNX) and the resonance integral (RI) of the reaction ¹⁵²Sm(n,γ)¹⁵³Sm. The samples of MnO₂ and Sm₂O₃ diluted with Al₂O₃ powder were irradiated within and without a cylindrical 1 mm-Cd shield case in an isotropic neutron field obtained from the ²⁴¹Am–Be neutron sources. The γ-ray spectra from the irradiated samples were measured by high resolution γ-ray spectrometry with a calibrated n-type Ge detector. The correction factors for γ-ray attenuation, thermal neutron and resonance neutron self-shielding effects and epithermal neutron spectrum shape factor (α) were taken into account in the determinations. The thermal neutron cross section for ¹⁵²Sm(n,γ)¹⁵³Sm reaction has been determined to be 204.8 ± 7.9 b at 0.025 eV. This result has been obtained relative to the reference thermal neutron cross section value of 13.3 ± 0.1 b for the ⁵⁵Mn(n,γ)⁵⁶Mn reaction. For the TNX, most of the experimental data and evaluated one in JEFF-3.1, ENDF/B-VI, JENDL 3.3 and BROND 2.0, in general, agree well with the present result. The RI value for ¹⁵²Sm(n,γ)¹⁵³Sm reaction has also been determined to be 3038 ± 214 b, relative to the reference value of 14.0 ± 0.3 b for the ⁵⁵Mn(n,γ)⁵⁶Mn monitor reaction, using a 1/E^1+α epithermal neutron spectrum and assuming Cd cut-off energy of 0.55 eV. In surveying literature, the existing experimental and evaluated data for the RI values are distributed from 1715 to 3462 b. However, when the Cd cut-off energy is defined as 0.55 eV, the present RI value agrees with some previously reported RI values, 3020 ± 163 b by Simonits et al., 3141 ± 157 by Van Der Linden et al., and 2962 ± 54 b by Kafala et al., within the limits of error.     

Monte Carlo calculation of the core of the heavy water zero power reactor (HWZPR) using MCNP4c

The MCNP4c code, based on the probabilistic approach, was used to simulate 3D configuration of the core of the heavy water zero power reactor (HWZPR). In present work, first, all of the constituents of the core such as fuel pellets, fuel element, moderator (D₂O) and annular graphite reflector were modeled using MCNP4c code. Then calculations of axial and radial neutron fluxes were performed in three energy groups such as thermal (0-0.625 eV), epithermal (0.625-550 eV), and fast (0.550-20 MeV). The cadmium ratio was calculated as well and the neutron flux parameters such as extrapolated height (H_e), extrapolated radius (R_e) and physical center of the core (z₀) were computed using cadmium ratio. Comparison of the neutron flux parameters with the experimental data showed that the MCNP4c model of the HWZPR was validated.     

Measurements of Neutron Capture Cross Section of 237Np for Fast Neutrons

The neutron capture cross section of ²³⁷Np has been measured for fast neutrons supplied at the center of the core in the Yayoi reactor. The activation method was used for the measurement, in which the amount of the product ²³⁸Np was determined by γ-ray spectroscopy using a Ge detector. The neutron flux at the center of the core calculated by the Monte Carlo simulation code MCNP was renormalized by using the activity of a gold activation foil irradiated simultaneously. The new convention is proposed in this paper to make possible a definite comparison of the integral measurement by the activation method using fast reactor neutrons with differential measurements using accelerator-based neutrons. “Representative neutron energy” is defined in the convention at which the cross section deduced by the activation measurement has a high sensitivity. The capture cross section of ²³⁷Np corresponding to the representative neutron energy was deduced as 0:80 ± 0:04b at 214 ± 9 keV from the measured reaction rate and the energy dependence of the cross section in the nuclear data library ENDF/B-VII.0. The deduced cross section of ²³⁷Np at the representative neutron energy agrees with the evaluated data of ENDF/B-VII.0, but is 15% higher than that of JENDL-3.3 and 13% higher than that of JENDL/AC-2008.     

Simulation of e–γ–n targets by FLUKA and measurement of neutron flux at various angles for accelerator based neutron source

A 6 MeV Race track Microtron (an electron accelerator) based pulsed neutron source has been designed specifically for the elemental analysis of short lived activation products where the low neutron flux requirement is desirable. The bremsstrahlung radiation emitted by impinging 6 MeV electron on the e–γ primary target, was made to fall on the γ–n secondary target to produce neutrons. The optimisation of bremsstrahlung and neutron producing target along with their spectra were estimated using FLUKA code. The measurement of neutron flux was carried out by activation of vanadium and the measured fluxes were 1.1878 × 10⁵, 0.9403 × 10⁵, 0.7428 × 10⁵, 0.6274 × 10⁵, 0.5659 × 10⁵, 0.5210 × 10⁵ n/cm²/s at 0°, 30°, 60°, 90°, 115°, 140° respectively. The results indicate that the neutron flux was found to be decreased as increase in the angle and in good agreement with the FLUKA simulation.     

Measurement of thermal neutron cross-section and resonance integral for the Ho(n,γ)Ho reaction using electron linac-based neutron source

The thermal neutron cross-section and the resonance integral of the ¹⁶⁵Ho(n,γ)^166gHo reaction have been measured by the activation method using a ¹⁹⁷Au(n,γ)¹⁹⁸Au monitor reaction as a single comparator. The high-purity natural Ho and Au foils with and without a cadmium shield case of 0.5 mm thickness were irradiated in a neutron field of the Pohang neutron facility. The induced activities in the activated foils were measured with a calibrated p-type high-purity Ge detector. The correction factors for the γ-ray attenuation (F_g), the thermal neutron self-shielding (G_th), the resonance neutron self-shielding (G_epi) effects, and the epithermal neutron spectrum shape factor (α) were taken into account. The thermal neutron cross-section for the ¹⁶⁵Ho(n,γ)^166gHo reaction has been determined to be 59.7 ± 2.5 barn, relative to the reference value of 98.65 ± 0.09 barn for the ¹⁹⁷Au(n,γ)¹⁹⁸Au reaction. By assuming the cadmium cut-off energy of 0.55 eV, the resonance integral for the ¹⁶⁵Ho(n,γ)^166gHo reaction is 671 ± 47 barn, which is determined relative to the reference value of 1550 ± 28 barn for the ¹⁹⁷Au(n,γ)¹⁹⁸Au reaction. The present results are, in general, good agreement with most of the previously reported data within uncertainty limits.     

Thermal neutron cross-section and resonance integral of the W(n, γ)W reaction

We measured the thermal neutron cross-section and the resonance integral of the reaction ¹⁸⁶W(n, γ)¹⁸⁷W by the activation method using a ¹⁹⁷Au(n, γ)¹⁹⁸Au monitor reaction as single comparator. The high-purity natural W and Au metallic foils with and without a cadmium shield case of 0.5 mm thickness were irradiated in a neutron field of the Pohang neutron facility. The induced activities in the samples were measured by high-resolution γ-ray spectrometry with a calibrated p-type high-purity Ge detector. The necessary correction factors for γ-ray attenuation (F_g), thermal neutron self-shielding (G_th), and resonance neutron self-shielding (G_epi) effects, and the epithermal neutron spectrum shape factor (α) were taken into account. The thermal neutron cross-section for the ¹⁸⁶W(n, γ)¹⁸⁷W reaction has been determined to be 37.2 ± 2.1 barn, relative to the reference value of 98.65 ± 0.09 barn for the ¹⁹⁷Au(n, γ)¹⁹⁸Au reaction. The present result is, in general, in good agreement with most of the experimental data and the recently evaluated value of ENDF/B-VII.0 by 5.7%. By assuming the cadmium cut-off energy of 0.55 eV, the resonance integral obtained is 461 ± 39 barn, which is determined relative to the reference values of 1550 ± 28 barn for the ¹⁹⁷Au(n, γ)¹⁹⁸Au reaction. The present resonance integral value is in general good agreement with the recently measured values by 9%. The present result is lower than the evaluated ones by 10-13%.     

Sensitivity of fuel failure detection by the delayed neutron measurement in the primary cooling circuit of HANARO

The sensitivity of the fuel failure detection system based on the delayed neutron measurement in the primary cooling circuit of a research reactor, HANARO is investigated. The neutrons around the primary cooling pipe during normal operation of HANARO are measured with BF₃ detector, and their count rate is 900 cps. They are regarded as photoneutrons due to the high energy gamma-rays from N-16 and delayed neutrons from the fission of the uranium contaminated on the fuel surface. The contribution of each neutron source is analyzed by measuring the changes of the neutron counts before and after the abrupt shutdown of reactor. In order to estimate the sensitivity of the fuel failure detection, the neutron count rate of BF₃ detector is predicted by Monte Carlo calculation. The generation, transportation and detection of the photoneutrons and the delayed neutrons are simulated for the geometry similar to the experiments. From the calculations and experiments, it is ascertained that the photoneutron contribution to the total count rate is about 20–30%, and that the delayed neutron count rate is expected to about 720 cps. The fission rate in the flow tube of the reactor core by the surface contamination is obtained from the deduced delayed neutron count rate, and it is estimated to 1.66 × 10⁵ fissions/cm³ s. From the MCNP calculation, it is confirmed that this fission rate can originate from the contaminated uranium of 120 μg, which is about 13% of the maximum allowable surface contamination on the fuel surface. The sensitivity of U-235 mass detection by the delayed neutron measurement can be concluded to about 0.2 μg-U²³⁵/cps. Thus, it is confirmed that the delayed neutron detection is sensitive enough to monitor the fuel failure, and that the neutron count rate is high enough for stable signal with short counting time.     

The limitations of associated alpha particle technique for contraband container inspections

Inspection of a shipping container for the presence of the threat materials has been investigated in the laboratory by using a 14 MeV neutron beam, a BaF₂ gamma detector and the associated alpha particle technique. The associated alpha particle technique is proposed as a part of a two sensor system for contraband container inspections. This method is effective in the reduction of background radiation with the possibility of collimating electronically the neutron beam.The intrinsic time resolution has been experimentally estimated to be 1.3 ns (FWHM), which allows inspection of a minimum voxel having 7 cm depth along the neutron flight path. The neutron beam intensity plays a crucial role as a limiting factor for the acquisition time reduction. Single counting rates of the gamma and alpha detector were investigated as a function of the neutron intensity, distance between the gamma detector and the neutron source and the type of shielding. The time and the energy spectra for different neutron intensities were evaluated.     

Measurements of activation cross sections for Lu(n, α)Tm,Lu(n, α)Tm and Lu(n, p)Yb reactions induced by neutrons around 14 MeV

The cross sections for the ¹⁷⁵Lu(n, α)¹⁷²Tm, ¹⁷⁶Lu(n, α)¹⁷³Tm and ¹⁷⁵Lu(n, p)^175m+gYb reactions have been measured in the neutron energy range of 13.5–14.8 MeV using the activation technique. The first data for ¹⁷⁵Lu(n, α)¹⁷²Tm reaction cross sections are presented. In our experiment, the fast neutrons were produced via the ³H(d, n)⁴He reaction on K-400 Neutron Generator at Chinese Academy of Engineering Physics (CAEP). Induced gamma activities were measured by a high-resolution (1.69 keV at 1332 keV for ⁶⁰Co) gamma-ray spectrometer with high-purity germanium (HPGe) detector. Measurements were corrected for gamma-ray attenuations, random coincidence (pile-up), dead time and fluctuation of neutron flux. The neutron fluences were determined by the cross section of ⁹³Nb(n, 2n)^92mNb or ²⁷Al(n, α)²⁴Na reactions. The neutron energy in the measurement was by the cross section ratios of ⁹⁰Zr(n, 2n)^89m+gZr and ⁹³Nb(n, 2n)^92mNb reactions. The results were discussed and compared with experimental data found in the literature and with results of published empirical formulae.     

Determination of thermal neutron cross-section and resonance integral for the As (n, γ)As reaction by activation method using Mn (n, γ)Mn as a monitor

Nuclear constants for use in reactor activation analysis especially (n, γ) cross-sections and absolute gamma intensities, are known to show a rather large scatter in literature. Thermal and resonance cross-sections for the ⁷⁵As (n, γ)⁷⁶As reaction is determined by the method of foil activation using ⁵⁵Mn (n, γ)⁵⁶Mn as a reference reaction. The experimental sample with and without a cadmium cover of 1-mm wall thickness was irradiated in the isotropic neutron field of the outer irradiation sites 7 of Ghana Research Reactor-1 facility which is a miniature neutron source reactor designed by the Chinese. The irradiation channel used has a neutron spectral parameter (α) found to be (0.037 ± 0.001). The induced activity in the sample was measured by gamma ray spectrometry with a high purity germanium detector. A standard solution of Arsenic was used for the analysis. The necessary correction for gamma attenuation, thermal neutrons and resonance neutron self-shielding effects were not taken into account during the experimental analysis because they were negligible. By defining cadmium cut-off energy of 0.55 eV, the result for ⁷⁵As (n, γ)⁷⁶As reaction was found to be: thermal neutron cross-section σ₀ = (4.28 ± 0.19) b and resonance integral I₀ = (61.88 ± 1.07) b.     

Thermal neutron cross-section and resonance integral of the Mo(n,γ)Mo reaction

We measured the thermal neutron cross-section and the resonance integral of the ⁹⁸Mo(n,γ)⁹⁹ Mo reaction by the activation method using a ¹⁹⁷Au(n,γ)¹⁹⁸ Au monitor reaction as a single comparator. The high-purity natural Mo and Au metallic foils with and without a cadmium shield case of 0.5 mm thickness were irradiated in a neutron field of the Pohang neutron facility. The induced activities in the activated foils were measured with a calibrated p-type high-purity Ge detector. The necessary correction factors for the γ-ray attenuation (F_g), the thermal neutron self-shielding (G_th) and the resonance neutron self-shielding (G_epi) effects, and the epithermal neutron spectrum shape factor (α) were taken into account. In addition, for the ⁹⁹Mo activity measurements, the correction for true coincidence summing effects was also taken into account. The thermal neutron cross-section for the ⁹⁸Mo(n,γ)⁹⁹Mo reaction has been determined to be 0.136 ± 0.007 barn, relative to the reference value of 98.65 ± 0.09 barn for the ¹⁹⁷Au(n,γ)¹⁹⁸ Au reaction. The present result is, in general, in good agreement with most of the experimental data and the recently evaluated values of ENDF/B-VII.0, JENDL-3.3, and JEF-2.2 by 5.1% (1σ). By assuming the cadmium cut-off energy of 0.55 eV, the resonance integral for the ⁹⁸Mo(n,γ)⁹⁹Mo reaction is 7.02 ± 0.62 barn, which is determined relative to the reference values of 1550 ± 28 barn for the ¹⁹⁷Au(n,γ)¹⁹⁸Au reaction. The present resonance integral value is in general good agreement with the previously reported data by 8.8% (1σ).