Neutron Flux Ratio Research Articles

Measurement of potassium in rats using an In vivo neutron activation analysis system

Abnormal levels of potassium are linked to several health conditions, including high blood pressure, cardiac dysfunction, kidney damage, and osteoporosis. Given the limited availability of in vivo measurement techniques, there is a need for novel methods to measure potassium to enhance the diagnosis and management of potassium metabolism related diseases. This study aimed to evaluate the feasibility of compact neutron generator based in vivo measurement system for quantification of potassium using rat carcasses. A cohort of thirty-nine rats (n = 20 males and 19 females, average weight 255 ± 15 and 163 ± 7 g) were sacrificed, and their carcasses were placed in polyethylene bottles. The rats were then positioned and irradiated in a carefully designed irradiation cave built alongside the neutron generator with an optimized thermal neutron flux and radiation dose ratio. The irradiation time was 10 min, followed by a 5-min decay and 2-h measurement using a high efficiency high purity germanium detector(HPGe). ResultsThe average potassium concentration in male and female rats was found to be comparable (male 2874 ± 161 and female 2866 ± 144 μg/g). A marginally positive correlation between potassium concentration and weight was found in female rats only (male(20) = 0.07, P = 0.76 and female r(19) = 0.34, P = 0.15). We assessed the influence of manganese toxicity on potassium levels and observed no significant impact. These results were consistent with our previous study in mice. ConclusionThis study suggests that in vivo neutron activation analysis could serve as a promising method to quantify potassium and to investigate the storage and metabolism of potassium in human and in animals.

Comparative analysis on small modular reactor (SMR) with uranium and thorium fuel cycle

The utilization of Small Modular Reactors (SMRs) and micro-reactors holds significant potential advantages. The primary benefits include their smaller size, accessibility to remote regions, and reduced land requirements. One noteworthy SMR is the NuScale Power Module (NPM), which has a thermal capacity of 160 MWt and operates on a 2-year cycle. However, there is a growing interest in exploring alternative fuels to enhance SMR performance and safety, and thorium has emerged as an appealing option due to its superior neutron characteristics and potential for higher fuel efficiency. This research aims to do preliminary research into adapting fuel for a thorium fuel cycle in the integral PWR (iPWR). The thorium cycle excels with an initial neutron multiplication factor of 1.134, almost the same as uranium (1.129). Additionally, the thorium cycle boasts superior neutron flux and Conversion Ratio (CR) in the thermal neutron energy range, enhancing sustainability with an average CR of 0.765, compared to 0.574 and 0.692 for uranium and uranium–thorium cycle. The addition of thorium improves efficiency in converting fertile material into fissile material with an enough criticality, where fuels incorporating thorium have a higher CR (Conversion Ratio). Furthermore, thorium fuel cycle has significantly lower production of plutonium isotopes compared to other fuel cycles. However, it faces a challenge with a high power peaking factor (PPF) of 1.923 at the Beginning of Cycle (BOC), requiring further research for mitigation.

Improvement of the quality of neutron radiography beamlines in Isfahan MNSR

Recently, two external neutron beam lines have been installed in the Isfahan MNSR. This reactor is a tank in pool type with a maximum power of 30 kW, which has been in operation since 1994, and in its original design it did not have a beam tube. These two neutron beam lines are designed and installed out of the reactor tank for irradiation experiments of large objects outside the reactor core, such as neutron radiography. The results of the evaluations in the previous researches showed that these two beam lines have acceptable quality according to the ASTM-E545 standard. However, their gamma contamination were relatively high. In this study calculations and experiments have been done to reduce the gamma content of the neutron beams and to increase the ratio of thermal neutron flux to gamma dose rate. This was done through the design and installation of gamma filters in the beam lines. The results of the evaluations showed that the quality of the neutron beams has been greatly improved in terms of reducing gamma content, so that the quality rating of the beams has increased by at least two ranks.

Application of small perturbation theory for assessing variations of prompt neutron lifetime in a lead-cooled fast reactor

The paper considers the applicability of small perturbation theory to assessing the variations of the prompt neutron lifetime caused by variations in the isotope composition of a lead-cooled fast reactor. The generalized small perturbation theory formulas have been developed to calculate derivatives of the prompt neutron lifetime regarded as a bilinear neutron flux and neutron worth ratio. A numerical algorithm has been proposed for the step-by-step application of the small perturbation theory formulas to assess the prompt neutron lifetime variations caused by a major perturbation in the reactor isotope composition, e.g. by the complete change of the material used earlier as the neutron reflector. The advantage of the proposed approach has been shown which consists in that it is basically possible to determine the role of different neutron reactions, isotopes and energy groups in and their contributions to the total prompt neutron lifetime variation caused by major changes in the reactor isotope composition.

Machine learning approaches to equilibrium burnup analysis for Molten Salt Reactor

During the operation of a liquid-fueled molten salt reactor (MSR), the deep depletion to an equilibrium state can be attained with online refueling and reprocessing techniques. However, the equilibrium burnup simulation of an MSR using current depletion codes requires considerable computational cost and time, and even fails when the equilibrium conditions are unsatisfied. To efficiently evaluate the specific characteristics of the equilibrium burnup in MSRs, a fast-filtering model for equilibrium state and a fast prediction model for equilibrium neutronic properties are developed based on the machine learning (ML) technique. A database of 4860 cases including the neutronic information of equilibrium and non-equilibrium of a fuel assembly with variable geometric parameters in an MSR is first generated, and then the neutronic properties of 232Th-233U, 232Th-EU (enriched uranium) and 232Th-Pu fuel cycles are evaluated. Based on the MSR database, fifteen different ML classification algorithms are implemented to establish the fast-filtering model for equilibrium burnup state. Moreover, a model for the fast prediction of equilibrium neutronic parameters of neutron flux (Φ) and breeding ratio (BR) is developed with fifteen ML regression algorithms. Considering the various performance metrics for measuring the predicted performances of ML models in the classification and regression, the LightGBM (LGBM) model is the most favorable for filtering the burnup state and predicting the neutronic parameters in MSRs. With the test data, the LGBM model can obtain the accuracy of about 0.98 for classification and the relative error of about 1.10% for regression.

Design of an in-tank thermal neutron beam for PGNAA application at Isfahan MNSR

Calculations have been done in order to achieve a thermal neutron beam for Prompt Gamma Neutron Activation Analysis (PGNAA) applications using Miniature Neutron Source Reactor (MNSR) as neutron source. The MNSR is a 30 kW tank-in-pool reactor and there are ten small occupied in-tank irradiation positions. Also, there are four positions in order to install or remove aluminum Regulating Rod Tubes (RRTs) that are bigger than that ten irradiation positions. In this study, the eastern RRT has been selected for more simulations of thermal neutron beam. Thermal neutron flux (φth), the ratio of thermal neutron flux to gamma dose rate (φthDγ), and the ratio of thermal neutron flux to total neutron flux (φthφtot) have been considered in calculations as important neutron beam parameters. Some designs have been done and simulated using MCNP code. The effect of the height of beam inlet, the effect of the different materials and thickness of moderators, and the effect of different thickness and height of gamma filters on the beam parameters at the beam outlet have been investigated. The results of different designs showed that using appropriate heights and thicknesses, by using bismuth as gamma filter and heavy water as moderator, φth, φthφtot and φthDγ, are 1.64E + 06n.cm−2.s−1, 0.789, and 4.36E + 05n.cm−2.μSv−1 respectively. Also, by using bismuth as gamma filter and graphite as moderator, these parameters are 1.57E + 06n.cm−2.s−1, 0.761, and 4.69E + 05E + 05n.cm−2.μSv−1, respectively.

Neutron collimator optimization for 14.1 MeV DT neutrons using Monte Carlo and Genetic algorithms

The fast neutrons generated by Deuterium-Tritium (DT) fusion reaction have been widely applied in prompt gamma ray neutron activation analysis measurements. In this study, a multi-layer neutron collimator for DT neutron generator was developed. Genetic algorithm combined with Monte Carlo simulation was used to design a collimator made of iron, lead, graphite, and borated polyethene. Copper foil activations were conducted to determine the fast neutron flux ratios between the beam port and its nearby area and agreed well with those predicted by the simulations. The results demonstrated that a narrower beam was obtained. The fast neutron beam flux was 568 ± 14 s−1 cm−2. The neutron flux ratio of the collimator was improved by a factor of 2.36, which could provide a better neutron beam.

Measurements of epithermal neutron properties of ITU Triga Mark II research reactor

The k0 standardization method is one of the most commonly used neutron activation analyses to determine the concentrations of elements in the sample. To perform this method, it is required to know the neutron spectrum parameters such as epithermal neutron flux shape factor (α) and thermal to epithermal neutron flux ratio (f). In this study, the α and f parameters were determined using the Cd-ratio dual monitor method and bare triple monitor method. In both methods, irradiation studies were performed in the central thimble at the ITU Triga Mark II research reactor. A gamma-ray counting system (HpGe detector) was used for measuring foil activation values. The selected monitors were 197Au and 94Zr for the Cd-ratio dual monitor method and 197Au/94Zr/96Zr for the bare triple monitor method. The α and f parameters were −0.221 ± 0.018 and 41.826 ± 4.701 for 197Au/94Zr (724 keV) monitors, −0.231 ± 0.019 and 42.626 ± 4.791 for 197Au/94Zr (756 keV) monitors, 0.228 ± 0.025 and 43.917 ± 3.596 for 197Au/94Zr/96Zr monitors, respectively.

Measurement of thermal neutron flux and cadmium ratios using neutron activation analysis

Irradiation of known foil samples in a reactor and subsequent measurement of induced activity allows determination of neutron flux. This is an important information for many purposes. It can be used reciprocally to determine contents of unknown samples irradiated at the same position in the reactor. On the other hand, neutron flux information is used in the assessment of a safe operation of nuclear power reactors. In the frame of this work, bare and cadmium covered foils of gold and dysprosium have been irradiated at different positions of the reactor. Induced activity measurements were made using a standard procedure. Thermal neutron flux and cadmium ratio values were determined for the irradiation positions.

Characterization of neutron spectra for NAA irradiation holes in H-LPRR through Monte Carlo simulation

The Korea Atomic Energy Research Institute (KAERI) has designed a Hybrid-Low Power Research Reactor (H-LPRR) which can be used for critical assembly and conventional research reactor as well. It is an open tank-in-pool type research reactor (Thermal Power: 50 kWth) of which the most important applications are Neutron Activation Analysis (NAA), Radioisotope (RI) production, education and training. There are eight irradiation holes on the edge of the reactor core: IR (6 holes for RI production) and NA (2 holes for NAA) holes. In order to quantify the elemental concentration in target samples through the Instrumental Neutron Activation Analysis (INAA), it is necessary to measure neutron spectrum parameters such as thermal neutron flux, the deviation from the ideal 1/E epithermal neutron flux distribution (α), and the thermal-to-epithermal neutron flux ratio (f) for the irradiation holes. In this study, the MCNP6.1 code and FORTRAN 90 language are applied to determine the parameters for the two irradiation holes (NA-SW and NA-NW) in H-LPRR, and in particular its α and f parameters are compared to values of other research reactors. The results confirmed that the neutron irradiation holes in H-LPRR are designed to be sufficiently applied to neutron activation analysis, and its performance is comparable to that of foreign research reactors including the TRIGA MARK II.

Metallic materials analysis by DT neutron generator-based inelastic neutron scattering system: Measurement and Monte Carlo simulation

In this paper, an inelastic neutron scattering system was used for the analysis of metallic materials. Iron and chromium measurements were conducted with the system, and the copper foil activation was used for determining fast neutron flux. A Monte Carlo model was proposed for comparing with experimental measurements. The performance parameters including the signal-to-background ratio, sensitivity and mass detection limit, as well as ratio of fast neutron flux were compared. Comparisons show the simulation results agree well with experimental measurements. The results demonstrate that the proposed Monte Carlo model can well represent the system and be used for improving system performance in the further application.

A neutronics investigation simulating fast reactor environments in the thermal-spectrum advanced test reactor

Developing advanced fuel and materials for fast-spectrum nuclear reactors is an immense challenge with tremendous potential in maximizing the value of advanced nuclear reactors. Historical developments and modern materials show promise for overcoming these challenges, but the lack of fast neutron test capabilities in the United States is a significant obstacle for meaningful progress. Modern modeling techniques were utilized to investigate a proposed method for fast neutron irradiations in an existing thermal-spectrum reactor, the Advanced Test Reactor (ATR). This method builds upon past ideas, where fast flux is increased by surrounding the specimens with fissionable “booster fuel” but diverges from historical approaches by using an already developed fuel element design used in the Belgian Reactor 2 (BR2) as the booster fuel while leveraging modern 3-D modeling and simulation techniques. Design evaluations and neutronics predictions were performed to evaluate the performance of a BR2 fuel element irradiated in an ATR flux trap with test pins in the central channel of the BR2 fuel element. These efforts have yielded promising results.Adding a BR2 fuel element in the northeast (NE) flux trap of ATR was predicted to result in a 150% increase in the incident fast neutron flux with a fast (>0.1 MeV) to thermal (<0.625 eV) neutron flux ratio ranging from approximately 50 to 150, dependent on the material used for thermal neutron filtering and volume of moderator within the central channel of the BR2 element.The predicted annual fast neutron fluence (>0.1 MeV) ranges from 7.9 × 1021 to 9.1 × 1021 n/cm2. Given the relatively large fast to thermal neutron flux ratio, the calculated radial power profiles within 4.3 mm outer diameter fueled specimens irradiated within the BR2 booster fuel element are fairly representative of fast neutron reactors. These predicted radial power profiles are not flat, but they are more prototypic than those seen in advanced fuels tests (Capriotti et al., 2017) which began in ATR in 2003. Another distinct advantage of this experiment design is that full-scale test pins can be irradiated to supplement the ongoing series of reduced scale advanced fuels tests (Beausoleil et al., 2020). The proposed experiment design irradiated within a BR2 fuel element in a flux trap of ATR offers an improved alternative to the current testing of advanced reactor fuels in ATR. Selected results from this design evaluation are documented herein and substantiate this conclusion.

Standardization and validation of k0-based Neutron Activation Analysis using Apsara-U reactor and its application to pure iron metal and coal sample for trace element determination

The k0-NAA method has been standardized by characterizing the G4 irradiation position of the upgraded Apsara (Apsara-U) reactor. The sub-cadmium to epithermal neutron flux ratio (f) and epithermal neutron flux shape factor (α) were determined taking dual (197Au and 94Zr), triple (197Au, 94Zr and 96Zr) and multi (197Au, 59Co, 64Zn, 68Zn, 96Zr and 94Zr) flux monitors using Cadmium ratio method. The value of f was found to be 19.94 ± 1.02 with average value of α as −0.027 ±0.004. The method was validated by analysing SMELS-II (synthetic multielemental standards) and geological certified reference material (Andesite, AGV-1), wherein results are in good agreement with certified or recommended values. The k0-based NAA method was applied for chemical characterization of pure Iron metal and coal samples. The results obtained were further compared with ICP-OES analysis and the values are found to be in good agreement (t<tcritical) for all the elements determined. F-test was also performed in order to show the reliability of the results in terms of standard deviations.

Determination of thermal neutron flux distribution at the rotary rack served for elemental concentration analysis using the k0-INAA method

The accuracy of elements concentration determination using the k0-standardization method directly depends on irradiation and measurement parameters including Non-1/E epithermal neutron flux distribution shape α (ϕ epi ≈1/E1+α ) , thermal-to-epithermal neutron flux ratio f, efficiency ε, peak area… In the case of the irradiation position at the rotary rack of the Dalat Nuclear Research Reactor (DNRR), the difference of thermal neutron flux between the bottom (3.54x1012 n.cm-2.s-1) and the top (1.93x1012 n.cm-2.s-1) of the 15 cm aluminum container is up to 45%. Therefore, it is necessary to accurately determine above-mentioned parameters in the sample irradiation position. The present paper deals with the determination of the distribution of thermal neutron flux along the sample irradiation container by using 0.1% Au–Al wire activation technique. The thermal neutron flux was then used to calculate the concentration of elements in the Standard Reference Material 2711a and SMELS type III using k0-INAA method at different positions in the container. The obtained results with the neutron flux correction were found to be in good agreement with the certified values. In conclusion, the proposed technique can be applied for activation analyses without sandwiching flux monitors between samples during irradiations.

Characterization of (252Cf-ZrH2) Monte Carlo model for detection of nitrogen and chlorine by thermal neutron-capture PGNAA

Monte Carlo calculations were conducted to explore the potentials of zirconium hydride (ZrH2) as possible alternative moderator for high density polyethylene (HDPE) and water (H2O) in a simple (100MBq) 252Cf-based Prompt Gamma Neutron Activation Analysis (PGNAA) assembly. The thickness of moderators was optimized against thermal neutron flux (ϕth) as well as the ratio of (thermal/fast) neutron flux (ϕth/ϕfast), calculated at the sample cavity. The capabilities of moderating 252Cf source neutrons and minimizing the γ-background due to 2525Cf fission γ-rays, 1H(nth,γ)2H γ-rays originates in the moderator as well as other (nth,γ) capture reactions from surroundings were investigated. The optimal selected moderator with sufficient (ϕth) at the sample cavity and minimum γ-contamination was selected as ZrH2 for use in a252Cf-PGNAA for detection of nitrogen and chlorine in bulk sample simulants of melamine (C3H6N6) and sodium chloride (NaCl). The prompt γ-rays spectra were analyzed to detect the 10.83 MeV and 6.11 MeV prompt γ-peaks released by 14N(nth,γ)15N and 35Cl(nth,γ)36Cl, respectively. The results showed that the (252Cf-ZrH2)-PGNAA model has positively identified nitrogen with net counts (1.25 × 105 n.cm−2.s−1) and Chlorine with net counts of (4.19 × 106 n.cm−2.s−1), achieving minimum detectable concentrations of about (350 ppm at 10.83 MeV) and (640 ppm at 6.11 MeV), respectively.

Neutron guide optimization for the Moroccan PGAA system

The Moroccan TRIGA Mark II research reactor is to be equipped with a PGAA (Prompt Gamma Activation Analysis) facility to assist the country's progression in socioeconomic areas such as the environment and geochemistry, agriculture, health, industry, cultural heritage, and human sciences. The requirements of the PGAA facility include a very high thermal to fast neutron flux ratio and a focused, thin beam. Supermirror neutron guides are generally used to meet such requirements due to their ability to reflect neutrons with a specific energy and scattering angle. This study was undertaken to determine a suitable neutron guide for achieving optimal performance for the PGAA facility, where such a determination needs to be made according to both the available space and the minimum performance needed for PGAA analysis. To better parameterize the neutron guide, a set of simulations were performed with the aim of establishing the best guide specifications. Three types of neutron guides were studied, namely straight, curved, and bender. McStas code was used in this study, and the simulations showed that the guides must be 6 m long and that the coating value m must be equal to 5 for the bender and straight guides and 6 for the curved guide. Among the three guide types, the thermal/epithermal (Фth/Фep) and thermal/fast (Фth/Фfast) ratios were found to be much better when using the curved guide.

Measurement of ambient neutrons in an underground laboratory at Kamioka Observatory and future plan

Ambient neutrons are one of the most serious backgrounds for underground experiments in search of rare events. The ambient neutron flux in an underground laboratory of Kamioka Observatory was measured using a 3He proportional counter with various moderator setups. Since the detector response largely depends on the spectral shape, the energy spectra of the neutrons transported from the rock to the laboratory were estimated by Monte-Carlo simulations. The ratio of the thermal neutron flux to the total neutron flux was found to depend on the thermalizing efficiency of the rock. Thus, the ratio of the count rate without a moderator to that with a moderator was used to determine this parameter. Consequently, the most-likely neutron spectrum predicted by the simulations for the parameters determined by the experimental results was obtained. The result suggests an interesting spectral shape, which has not been indicated in previous studies. The total ambient neutron flux is [1]. In this paper, we explain our method of the result and discuss our future plan.

Determination of TRIGA 2000 reactor parameters for NAA absolute methods

Neutron Activation Analysis (NAA) is a nuclear analytical technique for elemental analysis that has been widely applied in various fields such as environment, industry, health and food. It has high specificity and sensitivity for the analysis of elements with very low concentrations up to the picogram order. In Indonesia, NAA technique was developed in BATAN that has a research reactor as a neutron source. Relative method of NAA that uses the multi-elements standard as a comparison is widely used in Indonesia. At present, BATAN is developing an absolute method k0-NAA that has several advantages such as faster and more economics due to the reduction of multi-elements standard dependence. BATAN k0 Utility (BAkUL) beta version software has also been developed for data processing of the absolute methods. k0-NAA requires reactor parameters (α and f) as well as detector parameters for data analysis. The parameter f is the ratio of the thermal neutron flux to the epithermal whereas α is the epithermal spectrum deviation parameter from the 1/E distribution. The aim of this study is to measure parameters a and f using the triple bare monitor method by irradiating simultaneously Au and Zr without Cd cover. Measurement of reactor TRIGA 2000 parameters (α and f) were carried out at the irradiation positions CT1-1 and G17-2, and the results for α and f parameters at the CT1-1 irradiation position were -0,100 and 8.22 respectively, while at the G17-2 position was - 0.087 and 13.8. Each irradiation position has different α and f values so it is necessary to determine the reactor parameters for other irradiation positions. The measurement results α and f can be used for the next samples irradiation and as a database to support the application of the k0-NAA method in Center for Applied Nuclear Science and Technology (PSTNT) BATAN.

Gamma-ray and neutron shielding features for some fast neutron moderators of interest in 252Cf-based boron neutron capture therapy

Monte Carlo simulations (MCNP) were carried out to model suitable 252Cf-based gamma transmission and neutron attenuation geometries to evaluate the gamma-ray shielding features and neutron attenuation capabilities for Al2O3, MgF2, Fluental and AlF3, as fast neutron moderators of potential interest in 252Cf-based Born Neutron Capture Therapy (BNCT) facilities. The mass attenuation coefficient (μ/ρ), Half-Value Layer (HVL), Mean Free Path (MFP) and gamma transmission factor (I/Io) for Al2O3, MgF2, Fluental and AlF3 were simulated at gamma energies, 0.3 MeV, 0.364 MeV, 0.662 MeV, 1.25 MeV and 7.12 MeV. The MCNP simulated μ/ρ values were compared to theoretical XCOM values and good agreement was observed. The ratio of total thermal neutron flux to the total fast neutron flux (φ total thermal/φ total fast) and the total flux of gamma rays (φ total (n,γ) gamma) resulting from thermal neutron capture (n,γ) reactions, were simulated and compared as a function of thickness for the four sample moderators. The values of the effective fast neutron removal cross section ΣR (cm−1) for the four sample moderators were calculated and compared.

Investigations on a typical small modular PWR using coupled neutronic-thermal-mechanical evaluations to achieve long-life cycle-length

The main purpose of this study was to design small modular PWR with power of 150 MWth for long life operation, based on the coupled neutronic and thermal-mechanical evaluations. In this regard, pitch and fuel rod diameter were changed from 0.9 cm to 2.5 cm and 0.6 cm–0.96 cm, respectively; and desirable pairs (pitch - fuel rod diameter) were obtained. Neutronic parameters such as excess reactivity, maximum to average ratio of the radial neutron flux and fuel burnup were calculated using MCNP code. Also, thermal-mechanical parameters of the fuel rod including fuel centerline temperature, cladding hoop stress and fuel rod internal gas pressure were computed using FRAPCON code. Independency of results from number of axial nodes in FRAPCON code was analyzed. The number of axial nodes was considered 40 to ensure the accuracy of the FRAPCON results. Neutronic and thermal-mechanical cost values for each pair were calculated and Pareto front of the desirable pairs were obtained. Based on the results, eight desirable pairs were selected. By assuming equal importance for neutronic and thermal-mechanical parameters, total cost value was defined and pair of (1.1, 0.64) was selected as the most desirable one. Finally, fuel cycle lifetime was assessed for desirable pair considering neutronic (criticality and burnup) and thermal-mechanical (cladding strain, oxidation and hydriding as well as fuel rod internal gas pressure) limits. It was observed that fuel cycle lifetime was enhanced to 1750 days (40 GWD/ton).