Neutron Reaction Cross Sections Research Articles

Research on a Neutron Detector with a Boron-Lined Multilayer Converter

3He is a splendid neutron detection material due to its high neutron reaction cross section, gaseous state, and nonelectronegative and nonpoisonous nature. With the worldwide problem of the “3He supply crisis” arising, boron-lined gaseous neutron detectors are being widely used in neutron detection to replace 3He neutron detectors. In this work, to reduce the scattering neutron background coming from the substrate of a boron-lined neutron detector in the application of neutron scattering, a new design of the boron-lined gaseous neutron detector composed of a boron-lined multichip converter and a multiwire proportional chamber was proposed. The electron drift efficiency matrix simulated by Garfield++ (Version 2023.4) and the values and positions of electron energy deposition simulated by Geant4 were obtained. The α, 7Li, and total charged particle energy deposition spectra were acquired via coupling calculations of the electron drift efficiency matrix and the values and positions of electron energy deposition, and the width of the slit was selected as 3 mm. The boron-lined multilayer converter neutron detector (BMCND) was tested using a 241Am–239Pu mixture α source, and the total count rate of α charged particles was measured as 599.5 s−1, which is 89% of the theoretical α particle emission rate of 672.9 s−1. The drift voltage experiments showed that 1200 V is enough to acquire a relatively ideal count, and a 2500 V drift voltage was confirmed, considering the higher count and instrument safety. We also performed the neutron detection experiments using a photo-neutron source, and a characteristic spectrum shape of “two stairs” was measured. When borated polyethylene was used to shield the BMCND, the detected total count decreased while keeping the characteristic spectrum shape, demonstrating that the BMCND was equipped with the ability to detect neurons, indicating that BMCNDs have the potential to be an outstanding 3He alternative neutron detector.

Performance assessment and further optimization of the activation based criticality dosimetry system at the Belgian nuclear research centre SCK CEN

In places where fissile materials are being handled, the risk of a criticality accident can never be absolutely eliminated. Criticality accidents require specialized dosimetry techniques, which differ markedly from those used in routine radiological protection. This is caused by the fact that one should be able to measure very high doses up to 10 Gy in mixed gamma and neutron fields with very high instantaneous dose rates up to 105 Gy/s and complex energy spectra. The criticality dosimetry system at the Belgian Nuclear Research Centre SCK CEN is accredited according to ISO 17025 by the Belgian accreditation body BELAC and is used to monitor about 4000 nuclear workers. The system is based on a dedicated criticality dosemeter with four activation detectors (gold foils with and without cadmium shielding, an indium foil and sulphur pellets) and an iterative unfolding algorithm starting from a guess neutron fluence energy spectrum. This work describes the design of the SCK CEN criticality dosemeters and the iterative unfolding algorithm. Furthermore, it also discusses two new developments. Firstly, effective neutron reaction cross sections were calculated by MCNP6.2 Monte Carlo radiation transport simulations to be used in the unfolding algorithm. These effective cross sections take into account the effects of the dosemeter, the potential presence of the body and the neutron angle of incidence. These effects were demonstrated to be important, especially for the gold foils and if the dosemeter is worn against the body. Secondly, a Bayesian unfolding algorithm based on a parametric description of the neutron fluence energy spectrum was developed and tested. The results of this new algorithm were in good agreement with the iterative algorithm. The uncertainty assessment was more realistic with the Bayesian algorithm because it includes also the uncertainty associated with the unfolding process. However, for spectra that cannot be well described by the parametric function, the Bayesian algorithm can lead to significant under- or overestimations. In case sufficient information is available, it is better to make use of the iterative algorithm with a realistic guess spectrum to start from. Finally, this work also demonstrated the compliance of the SCK CEN criticality dosimetry system with the IAEA criteria for criticality dosimetry systems.

Pure CaF2 crystal for fast neutron detection

Pure CaF2 crystals have been widely used for optical functional crystals due to their good optical properties, mechanical properties, and chemical stability. This study discovered the crystal's pulse shape discrimination (PSD), proving it a promising scintillator for fast neutron detection via the 40Ca neutron capture reaction. Luminescence emission spectra were studied by X-ray luminescence measurements. The maximum emission peak was observed between 250 and 300 nm, corresponding to self-trap exciton emission. The crystal response to gamma, alpha, and neutron were studied under 137Cs, 241Am, and 252Cf radiation sources, respectively. The difference in gamma and alpha decay times enabled the PSD capability of pure CaF2. Large neutron reaction cross-sections, PSD capability, and good chemical stability make pure CaF2 crystal candidates for fast neutron beam monitoring detectors. Moreover, the crystal can be used for 40Ca(n,p)40K and 40Ca(n,α)37Ag neutron capture cross-section study.

Spectral analysis techniques in measuring neutron-induced gamma production cross-section

In neutron reaction cross-section measurements, the prompt gamma ray method is a method of obtaining cross-section data by measuring the characteristic gamma rays emitted by a nuclear reaction, thereby avoiding the interference generated by competing reaction channels. However, the prompt gamma ray method is an on-line experiment with abundant background sources, high background counts of the obtained experimental spectra, and numerous interferences such as weak peaks, overlapping peaks, Compton scattering peaks, and neutron effect peaks of Ge in HPGe, which cause the difficulty in analysing the on-line experimental spectra and the high uncertainty in the results. In this work, we study and summarise the spectrum analysis techniques of the prompt gamma ray method that can be used for measuring the neutron cross-section, and comprehensively consider the physical processes of the formation of different characteristic peaks of the prompt gamma ray method, so as to reduce the uncertainty of calculating the net area of the effect peaks in the process of on-line experimental spectrum processing. The Compton edge, weak peaks, overlapping peaks, and the neutron response peaks of the HPGe detector on-line experiment are discussed and analysed, and the net area of the effect peaks is accurately extracted by combining several reasonable functions to fit the total energy peak, the background, and the interferences. For the net area of weak peaks, this method can reduce the peak area selection caused fluctuation from 30% to less than 1%, and the difference between the fitted value of the net area and the theoretical value is comparable to the statistical uncertainty; for the overlapping peaks’ decomposition, the difference between the results obtained by this method and the theoretical value is significantly lower than 1%. The reliability of the spectral analysis method is simultaneously verified by efficiency curve analysis and goodness-of-fit calculation.

Fabrication of a Gd2O3-Based Film to Shield from Space Radiation inside Aircraft and Its Effectiveness

Aircraft are exposed to cosmic radiation depending on their flight altitude and latitude. Therefore, flight attendants are exposed to radiation for long periods. In this study, a 0.3 mm thick fabric was designed with which to manufacture crew clothes to shield them against external exposure to space radiation, and the shielding performance was analyzed based on empirical experiments in a real environment. Gadolinium oxide, which has a high neutron reaction cross-section, and tungsten, which is useful for gamma-ray shielding, were proposed as the main raw materials for the shielding fabric, and the shielding performance was evaluated using detectors on Arctic flight routes. Composite (KG-01) and single (KG-02) shielding materials were used. In the case of KG-01, the transmission dose rate was 90.7 ± 5.6% compared with the unshielded case, showing an average space-radiation dose reduction of 9.3%. With KG-02, the transmission dose rate was 103.1 ± 2.0% compared with the unshielded case, and the average dose rate increased by 3.1%; therefore, there was no shielding effect against space radiation. Considering the statistical error of the environmental radiation at aircraft flight altitudes, KG-01 had a shielding effect of at least 5%; however, KG-02 yielded no significant shielding effects.

Proof-of-concept studies of novel protocols for producing highly pure 48V from a 48Cr/48V generator

The quest to improve the quality of nuclear data, such as half-lives, transition yields, and reaction cross-sections, is a shared endeavor among various areas of nuclear science. 48V is a vanadium isotope for which experimental data on neutron reaction cross-sections is needed. However, traditional isotope production techniques cannot produce 48V with high enough isotopic purity for some of these measurements. “Isotope harvesting” at the Facility for Rare Isotope Beams (FRIB) is a new isotope production technique that could potentially yield 48V with the necessary purity for such studies. In this case, 48Cr would be collected and allowed to generate 48V that can be separated from undecayed 48Cr to yield highly pure 48V. Thus, any protocol for producing pure 48V via isotope harvesting would involve utilizing a separation technique that can effectively separate 48Cr and 48V. In this study, the radiotracers 51Cr and 48V were used to develop possible radiochemical separation methodologies, which can be translated to obtain high purity 48V via this novel isotope production method. The developed protocols utilize either ion exchange or extraction chromatographic resins. Separations of 51Cr and 48V with AG 1-X8 anion exchange resin respectively resulted in recoveries of 95.6(26)% and 96.2(12)% with radionuclidic purities of 92(2)% and 99(1)%. An even more effective Cr and V separation was obtained with an extraction chromatographic resin (TRU resin) and 10 M HNO3 loading solution. Here, 51Cr and 48V respectively had recoveries of 94.1(28)% and 96.2(13)% with high radionuclidic purities (100(2)% and 100(1)%) in small volumes (8.81(8) mL and 5.39(16) mL). This study suggests that, to maximize the yield and isotopic purity of 48V, the best production protocol would involve utilizing two separations with TRU resin and 10 M HNO3 to isolate 48Cr and purify the generated 48V.

Regression analysis of experimental reaction cross-section data of 241Am(n, 2n)240Am

Pre-processing of neutron reaction cross-section is essential in the nuclear data evaluation. This work aims to pre-process experimental cross-section data of 241 Am (n, 2n) 240 Am neutron reaction. Pre-processing of the experimental data includes re-normalization, removal of the outliers, integrating multiple cross-section values at single energy to single cross-section value, and regression on the cleaned experimental data. To remove outliers from the data, standardized residual and studentized residual have been used. For integration of multiple cross-section values to single cross-section value, the weighted average method has been used. Regression on the cleaned experimental data has been accomplished using the Gaussian Process Regression (GPR) and Polynomial Regression (PR), and the performance of both regression methods has been studied using statistical indices such as the determination of coefficient (R2) and the sum of the square of residual (SSres).

Synthesis and characterization of Polyethylene/B4C composite, and its neutron shielding performance in cementitious materials: Experimental and simulation studies

Boron compounds are effective neutron shielding materials because of their high neutron reaction cross-section. However, their use is limited in cementitious materials because of the boric acid forming, which reduces the durability. In this study, a polyethylene/B4C composite (Poly-B), for use as a neutron shielding material suitable for cementitious systems, was synthesized using a simple mechanical activation process. The physicochemical properties and neutron shielding performance of the Poly-B-incorporated cement paste were evaluated via experimental and computational methods. The Poly-B exists in a form in which polyethylene coats the outer surface of B4C. The Poly-B-incorporated cement paste, boric acid forming is prevented, is more advantageous in long-term strength than that of B4C. Furthermore, Poly-B, which is a removal-diffusion-based material, improved the neutron shielding performance of cement paste than B4C.

Measurements of the 107 Ag neutron capture cross sections with pulse height weighting technique at the CSNS Back-n facility

Silver indium cadmium (Ag–In–Cd) control rod is widely used in pressurized water reactor nuclear power plants, and it is continuously consumed in a high neutron flux environment. The mass ratio of 107Ag in the Ag–In–Cd control rod is 41.44%. To accurately calculate the consumption value of the control rod, a reliable neutron reaction cross section of the 107Ag is required. Meanwhile, 107Ag is also an important weak r nucleus. Thus, the cross sections for neutron induced interactions with 107Ag are very important both in nuclear energy and nuclear astrophysics. The (n, γ) cross section of 107Ag has been measured in the energy range of 1–60 eV using a back streaming white neutron beam line at China spallation neutron source. The resonance parameters are extracted by an R-matrix code. All the cross section of 107Ag and resonance parameters are given in this paper as datasets. The datasets are openly available at http://www.doi.org/10.11922/sciencedb.j00113.00010.

LOW TEMPERATURE EFFECTS ON PWR FUEL ASSEMBLY CRITICALITY CALCULATIONS

One of the parameters affecting the neutron multiplication factor keff of a system containing fissile material is the system temperature. Therefore, the effect of temperature on criticality safety analyses is an area of international interest. In this context, the Working Party on Nuclear Criticality Safety (WPNCS) of the OECD Nuclear Energy Agency (NEA) formed a subgroup to define and execute a code-to-code comparison benchmark to investigate the effect of temperature on keff for PWR fuel assemblies. Two configurations of a generic water-moderated PWR fuel assembly were analysed at different temperatures between 233 K and 588 K, and with different assembly burnups. Based on this benchmark, GRS performed an additional study to investigate the impact of the moderator densities, the neutron reaction cross sections and the thermal scattering data on keff separately. The benchmark results show the expected decrease of keff with temperature and a considerable jump in keff at the phase transition of the moderator. The additional investigation demonstrates that the jump in keff is mainly caused by the change of the moderator density due to the phase transition. The change of the thermal scattering data due to the phase transitions leads to a similar but smaller jump in keff. Furthermore, the actual impact of the different parameters on keff depend strongly on the considered fuel assembly configuration.

Analysis of the first core of the Indonesian multipurpose research reactor RSG-GAS using the Serpent Monte Carlo code and the ENDF/B-VIII.0 nuclear data library

This paper presents the neutronics benchmark analysis of the first core of the Indonesian multipurpose research reactor RSG-GAS (Reaktor Serba Guna G.A. Siwabessy) calculated by the Serpent Monte Carlo code and the newly released ENDF/B-VIII.0 nuclear data library. RSG-GAS is a 30 MWth pool-type material testing research reactor loaded with plate-type low-enriched uranium fuel using light water as a coolant and moderator and beryllium as a reflector. Two groups of critical benchmark problems are derived on the basis of the criticality and control rod calibration experiments of the first core of RSG-GAS. The calculated results, such as the neutron effective multiplication factor (k) value and the control rod worth are compared with the experimental data. Moreover, additional calculated results, including the neutron spectra in the core, fission rate distribution, burnup calculation, sensitivity coefficients, and kinetics parameters of the first core will be compared with the previous nuclear data libraries (inter-library comparison) such as ENDF/B-VII.1 and JENDL-4.0. The C/E values of ENDF/B-VIII.0 tend to be slightly higher compared with other nuclear data libraries. Furthermore, the neutron reaction cross-sections of 16O, 9Be, 235U, 238U, and S(α,β) of 1H in H2O from ENDF/B-VIII.0 have substantial updates; hence, the k sensitivities against these cross-section changes are relatively higher than other isotopes in RSG-GAS. Other important neutronics parameters such as kinetics parameters, control rod worth, and fission rate distribution are similar and consistent among the nuclear data libraries.

ОЦЕНКА СКОРОСТИ НАРАБОТКИ ИЗОТОПОВ ЭЙНШТЕЙНИЯ И БЕРКЛИЯ В РЕАКТОРЕ ПИК

The production of einsteinium and berklium isotopes at reactor PIK is considered. Presently the design of the reactor core for reactor operation after full power startup is under consideration. The additional places for isotopes production in the reactor core can be organized besides the central experimental channel. The irradiation conditions in the central water trap and in the core are comparing. For this purpose we are comparing calculated here the neutron spectrums, neutron fluxes and cross sections of the main reactions which are influence on the isotope production. We maid the approximation of the averaged over reactor spectrums neutron reaction cross sections using thermal cross sections, resonance integral and Westcott factors. For the realistic isotope targets we have calculated the amound of produced einsteinium and berclium. Also we compare the accumulation speed of these two elements in the central water trap and in the reactor core. We have made the verification of the MCNP code calculations of the isotope production using analitic calculations and code SCALE. We have demonstrated that for targets with initial isotopes placed in the core the amount of produced einsteinium and berclium can be 4-5 times greater then for targets placed in the central experimental channel. However the time for isotope production in the core can be about two times longer. We are discussing also the main ambiguities in the reaction cross section which influences on the amount of the produced einsteinium in the target. The main ambiguity in the calculations of the produced amount of rather long-lived isotope Es-254 is introduced the neutron reaction cross sections of Es-253 nuclei espessially the isomeric ratio of the reaction of the neutron radiation capture.

Quantification of trace amounts of transmuted nuclides in transmutation detectors using SIMS

Neutron detectors based on neutron induced transmutation of nuclides of high-purity materials into other stable nuclides has been recently proposed to improve accuracy of long-term measurement of neutron fluence in fission nuclear reactors. The inductively coupled plasma mass spectrometry (ICP-MS) and the prompt gamma neutron activation analysis (PGNAA) have been used for the measurement of concentration of transmuted nuclides, which was then converted into neutron fluence using known neutron-reaction cross sections.In this work, the concentrations of transmuted elements were measured with a magnetic-sector secondary ion mass spectrometer (SIMS) instead of ICP-MS or PGNAA. Single crystals of Si, Ge and a polycrystalline metal foil of Ni were used as detector materials. The concentrations were measured down to sub ppb range in Si and Ge crystals. 9 prospective nuclides for transmutation detectors — 25Mg, 26Mg, 27Al, 31P, 67Zn, 69Ga, 71Ga, 59Ni, 63Ni were measured and compared to a reference neutron activation detector (NAD) method with neutron reaction cross section values taken from two different databases. With exception of 26Mg, the values determined using SIMS were within 20 % compared to those obtained using NADs with a cross section value from one or both databases. In case of 26Mg the difference between the results of SIMS and NAD methods was higher than 45 % for data from both databases. This discrepancy for 26Mg measurement was assigned tentatively to uncertainties of the neutron reaction cross section values. Nevertheless, the experiments proved distinct advantages of SIMS over ICP-MS in sample preparation and throughput, and in consuming micro-volumes of material and identified at least one well performing transmutation detector for thermal neutrons (Ge - single crystal) with agreement within 3 % between SIMS-TMD and NADs.

The systematics of neutron reaction cross sections

The parameterized theoretical formulae of excitation functions for (n, 2n) and (n, γ) reactions have been established, and for (n,tot), (n,non), (n,3n), (n,p), (n,d), (n,t), (n,3 He) and (n,α) have been recommended. According to these formulae, the SEF code have been developed for systematics calculation of these reactions. The calculated results with the systematics of the corresponding reactions of the discretional nucleus can be provided by the SEF code in the applied range quickly. At the same time, the comparison of calculated results with experimental and evaluated data can be given graphically.

MicroMegas-based detectors for time-of-flight measurements of neutron-induced reactions

MicroMegas detectors are versatile gaseous detectors which are used for ionizing particle detection. A MicroMegas detector consists of two adjacent gas-filled volumes. One volume acts as a drift region with an electric field operating in the ionization chamber regime, the second volume is the amplification region acting as a parallel-plate avalanche counter. The use of the microbulk technique allows the production of thin, radiation resistant, and low-mass detector with a highly variable gain. Such MicroMegas detectors have been developed and used in combination with neutron time-of-flight measurements for in-beam neutron-flux monitoring, fission and light-charged particle reaction cross section measurements, and for neutron-beam imaging. An overview of MicroMegas detectors for neutron detection and neutron reaction cross section measurements and related results and developments will be presented.

Feasibility study on medical isotope production using a compact neutron generator

Compact neutron generators can provide high flux of neutrons with energies ranging from thermal (0.025 eV) to 14 MeV. Recent measurements demonstrated high neutron yields from the D–7Li fusion reaction at an interaction energy of 500 keV. Using the D−7Li reaction and applying new advancements in high flux neutron generator technology along with the commercial availability of high voltage DC power supplies enables the production of useful quantities of radioisotopes for medical applications. Using the known neutron reaction cross-sections, it has been estimated that hundreds-to-thousands MBq (or tens-to-hundreds mCi) of 99Mo, 225Ac, 64Cu and 67Cu can be obtained from a compact high flux neutron generator.

Modeling of burnup express-estimation for UO2-fuel

The paper presents the developed engineering model of cesium isotopes production as function of UO2-fuel burnup and an assessment of their activity ratios. The model considers the evolution of linear power of gadolinium-doped fuel rods and fuel rods surrounding them in fuel assemblies with high enrichment fuel, harder neutron spectrum, and the changes in cross-sections of neutron reactions in thermal and epithermal energy areas. Parametrical dependences in the model are based on the fuel operation data for nuclear power plants and on the detailed neutronic-physical calculations of the core. Presented are the results of the model calculations for the 134Cs/137Cs activity ratios in fuel taking into account the parameter of hardness of the neutron spectrum during the first irradiation cycle for fuel with enrichment ranging from 3.6wt% in 235U.

Fission Fragment Angular Distribution measurements of235U and238U at CERN n_TOF facility

Neutron-induced fission cross sections of 238U and 235U are used as standards in the fast neutron region up to 200 MeV. A high accuracy of the standards is relevant to experimentally determine other neutron reaction cross sections. Therefore, the detection effciency should be corrected by using the angular distribution of the fission fragments (FFAD), which are barely known above 20 MeV. In addition, the angular distribution of the fragments produced in the fission of highly excited and deformed nuclei is an important observable to investigate the nuclear fission process.In order to measure the FFAD of neutron-induced reactions, a fission detection setup based on parallel-plate avalanche counters (PPACs) has been developed and successfully used at the CERN-n_TOF facility. In this work, we present the preliminary results on the analysis of new 235 U(n,f) and 238 U(n,f) data in the extended energy range up to 200 MeV compared to the existing experimental data.

Intergranular fracture in irradiated Inconel X-750 containing very high concentrations of helium and hydrogen

In recent years, it has been observed that Inconel X-750 spacers in CANDU reactors exhibits lower ductility with reduced load carrying capacity following irradiation in a reactor environment. The fracture behaviour of ex-service material was also found to be entirely intergranular at high doses. The thermalized flux spectrum in a CANDU reactor leads to transmutation of 58Ni to 59Ni. The 59Ni itself has unusually high thermal neutron reaction cross-sections of the type: (n, γ), (n, p), and (n, α). The latter two reactions, in particular, contribute to a significant enhancement of the atomic displacements in addition to creating high concentrations of hydrogen and helium within the material. Microstructural examinations by transmission electron microscopy (TEM) have confirmed the presence of helium bubbles in the matrix and aligned along grain boundaries and matrix–precipitate interfaces. Helium bubble size and density are found to be highly dependent on the irradiation temperature and material microstructure; the bubbles are larger within grain boundary precipitates. TEM specimens extracted from fracture surfaces and crack tips provide information that is consistent with crack propagation along grain boundaries due to the presence of He bubbles.

Cross section measurements of a few threshold reactions induced by fast neutrons from an Am/Be source: integral tests of differential neutron reaction cross section data

Abstract Integral cross sections of the reactions 24Mg(n,p)24Na, 27Al(n,p)27Mg, 27Al(n,α)24Na and 95Mo(n,p)95Nb were measured for the first time with fast neutrons (E n > 1.5 MeV) from an Am/Be source. Furthermore, upper cross section limits for the reactions 98Mo(n,α)95Zr and 103Rh(n,p)103Ru were deduced. The results were compared with data calculated using the neutron spectral distribution and the excitation function of each reaction given in the data files and libraries ENDF/B-VII, JEFF-3.1.2, JENDL-4.0, IRDF-2004, TENDL-2010, ROSFOND-2010 and CENDL-3.1. In general, the integral measurement and the integrated value agreed within ± 10%. The fast neutron field (with E n > 1.5 MeV) of the Am/Be source is thus well suited for integral tests of excitation functions of neutron threshold reactions.