Fast Neutron Spectrum Research Articles

An overview of activation cross-section measurements of some neutron and charged-particle induced reactions in Bangladesh

Abstract Over the last several decades, our research group has been working on nuclear data measurements using a J-25 Neutron Generator, a TRIGA Mark II Research Reactor and a Tandem Accelerator at Savar, as well as an Am/Be neutron source at Rajshahi. The spectrum of fast neutrons having energies from 0.5 to 20 MeV in the core of the TRIGA reactor as well as from 1.5 to 11 MeV of the Am/Be source was well characterized for validation of evaluated data of neutron threshold reactions. Spectrum-averaged cross sections were determined by integral measurements. The integral values measured at both the reactor and the Am/Be source were compared with the integrated data deduced from measured and evaluated excitation functions of some nuclear reactions given in some data libraries. As example, the details on the integral and integrated cross sections of the 64Zn(n,p)64Cu and 60Ni(n,p)60Co reactions are discussed. In both cases, integral test shows that further improvement of the evaluated excitation function reported in data libraries is needed. The use of a newly developed facility at the 3 MV Tandem Accelerator in measurement of proton induced reaction cross sections in the energy region below 5 MeV is outlined, and new results on the 64Ni(p,n)64Cu and natNi(p,x)61Cu processes below 3 MeV are discussed.

Conceptual core design of the molten metal fueled microreactor with self-regulating capability

The microreactor concept has received significant attention in the United States for its lower capital investment, siting flexibility and high mobility. A microreactor aims to provide reliable electricity in remote communities, mining areas, or isolated islands, which will save the high fuel transportation cost. Thus, minimizing the staff level for operating and maintaining a microreactor is desirable. The objective of the study is to design a self-regulated Molten Metal fueled microReactor core (MMR) and to demonstrate the self-regulating capability of the MMR core. The control of the reactor is solely relying on reactivity feedbacks from fuel temperature perturbation. Molten UMn (liquid form at the temperature over 720 °C) was utilized as the fuel material in MMR due to the larger thermal expansion coefficient compared to the conventional solid fuel. The MMR core has a fast neutron spectrum, which corresponds to a small burnup reactivity swing. The core design parameters were selected based on DAKOTA, REBUS, and ANLHTP calculations. The MMR is designed to run for 10 years at a power level of 15 MWth. No refueling is needed during the reactor lifetime. The system dynamic analyses were performed, and it was found that the reactor can be self-regulating within a temperature range of 800 ± 80 °C for a loss of heat sink transient.

MSR Simulation with cGEMS: Fission Product Release and Aerosol Formation

The release of fission products and fuel materials from a molten-salt fast-reactor fuel in hypothetical accident conditions was investigated. The molten-salt fast reactor in this investigation features a fast neutron spectrum, operating in the thorium cycle, and it uses LiF-ThF4-UF4 as a fuel salt. A coupling between the severe accident code MELCOR and thermodynamical equilibrium solver GEMS, the so-called cGEMS, with the updated HERACLES database was used in the modeling work. The work was carried out in the frame of the EU SAMOSAFER project. At the beginning of the simulation, the fuel salt is assumed to be drained from the reactor to the bottom of a confinement building. The containment atmosphere is nitrogen. The fission products and salt materials are heated by the decay heat, and due to heating, they are evaporated from the surface of a molten salt pool. The chemical system in this investigation included the following elements: Li, F, Th, U, Zr, Np, Pu, Sr, Ba, La, Ce, and Nd. In addition to the release of radioactive materials from the fuel salt, the formation of aerosols and the vapor-phase species in the modeled confinement were determined.

Self-sustainability and energy-balance in a fast Fusion–Fission Hybrid Reactor (FFHR) based on Dense Plasma Focus (DPF) and multiplicative cascade

A Fusion–Fission Hybrid Reactor (FFHR) is made of a subcritical reactor driven by the neutrons generated in a nuclear fusion device. The fast neutron spectrum produced by this system could be able to burn actinides from the spent nuclear fuel. Unfortunately, current fusion devices are unable to provide an adequate neutron yield with reasonable spending of energy. However, it is possible to design the subcritical set to multiply considerably the neutron flux by using the so-called multiplying cascades. In addition, given its simplicity, the Dense Plasma Focus (DPF) is an interesting candidate as a neutron source. In this work, we analyze a concept of a metallic-fuel hybrid reactor based on a subcritical set and driven by a DPF as a neutron source. Their spectra are evaluated to determine if they are compatible with the operation of a fast reactor. The energy balance is also studied to find its ability to be self-sufficient when using an existing DPF. Finally, its capability to generate tritium from its lithium blanket is also analyzed. The system is self-sustainable.

The Versatile Test Reactor Project: Mission, Requirements, and Description

The mission of the U.S. Department of Energy (DOE) Office of Nuclear Energy is to advance nuclear power to meet the nation’s energy, environmental, and national security needs. Advanced nuclear technology development, as well as support for the current nuclear power industry, requires a robust infrastructure for experimentation, testing, design evolution, and component qualification. The current lack of fast neutron spectrum testing capabilities has been identified as a significant gap in the U.S. infrastructure that impedes the development of next-generation nuclear reactors—many of which require a fast neutron spectrum for operation—and equally impacts the United States’ ability to regain global technology leadership in this arena. To close the gap and support advanced technology development, the DOE has established the Versatile Test Reactor (VTR) project to provide high-performance testing capability, specifically, a fast neutron source to develop, test, and qualify advanced fuels and materials for the next generation of advanced reactors and existing commercial reactors. This paper describes the establishment of the project, the identification of its mission and requirements, and the design approach and status.

Preliminary study on TRU transmutation in VVER-1000 fuel assembly using MCNP6

Abstract Transmutation technology has been developed to address the issue of long-term safety of nuclear waste disposal in geological repository. Transuranic (TRU) transmutation in thermal neutron spectrum can be considered more beneficial at the present time due to the larger fission cross section of TRU elements than in fast neutron spectrum. This paper discusses preliminary study on TRU transmutation in Vodo-Vodyanoi Energetichesky Reactor (VVER)-1000 fuel assembly using MCNP6 code. The fuel assembly is configured by 312 fuel cells consist of 300 UO2 fuel rods with 3.7 wt% 235U and 12 UGD fuel rods containing a mixture of UO2 with 3.6 wt% 235U and 4.0 wt% Gd2O3. The calculation results show that the k inf value is higher with the increase in the TRU thickness at the beginning of cycle and at the end of cycle, which means that the addition of TRU will increase the fuel cycle length. The total temperature coefficient of reactivity is negative for all fuel assemblies both without TRU and with TRU. In general, the computed β eff values of the assembly with and without TRU addition are not significantly different. It shows that the coating of the TRU layer in the UGD fuel cell will not complicate the reactor control. The transmutation of TRU recycled from spent nuclear fuel in the VVER-1000 fuel assembly can be considered feasible from the viewpoint of excess reactivity and control safety characteristics. The total transmutation rate of ∼55.73% can be achieved with 239Pu and 241Am isotopes are taking the largest portion of transmuted nuclides.

On the equivalence of irradiation conditions on present and future facilities for fusion materials research and qualification: A computational study

We performed a computational study to assess the suitability and equivalence of the irradiation conditions on several test irradiation facilities (either currently operating or planned to deploy in the future) aimed at the qualification of materials for nuclear fusion reactors such as ITER and DEMO. The degradation of the material's properties is driven by the changes in its microstructure and chemistry (transmutation). The primary objective of this study is thus to perform a comparison of the microstructural pattern as predicted by means of simulations. The focus of the study is put on two materials: Eurofer97 steels and tungsten. We considered operation conditions in fusion reactors (i.e. ITER and DEMO) and in test irradiation facilities such as material test reactors (fast and mixed neutron spectrum), IFIMIF-DONES, ESS and proton accelerators. Typical irradiation conditions are addressed according to the currently available design (for DEMO) and expected operation modes (for ITER). The study is realized by means of object kinetic Monte Carlo which is parameterized and configured using state-of-the-art knowledge on irradiation spectra, neutron cross-sections, primary damage states, lattice defects mobility/cohesion and interaction of the material's microstructure with lattice defects. The irradiation defects are singled out using a dedicated post-processing tools to enable a comparison with expected findings in transmission electron microscopy (TEM) and atom probe tomography (APT). The results are discussed in terms of the equivalence of the emerging irradiation microstructure predicted to occur in test irradiation facilities if compared with the one simulated in the nuclear fusion reactors. The summary and discussion provide information on the equivalence and deviations of the microstructural patterns suggesting the suitability of the test irradiation facilities for certain irradiation regimes, as well as pointing at some limitations, e.g., originating from the difference in the neutron spectra or flux.

Microdosimetry of an accelerator based thermal neutron field for Boron Neutron Capture Therapy

The MUNES project (MUltidisciplinary NEutron Source) aims at the realization of an intense accelerator-based source of thermal neutrons, suitable for Boron Neutron Capture Therapy (BNCT). This exploits the interaction of 5 MeV protons onto a beryllium target, producing a fast neutron spectrum, which is moderated to the thermal range by a large assembly made of a Polytetrafluoroethylene (PTFE) tank filled with heavy water, surrounded by graphite blocks. The thermal neutron field is extracted through a bismuth beam port. The microdosimetric characterization of this field was performed using a cylindrical avalanche-confinement Tissue Equivalent Proportional Counter (TEPC) equipped with interchangeable cathode walls, positioned in front of the beam port. Measurements were taken both with a boron-doped wall and with an undoped one. The comparison of the two microdosimetric distributions allows to distinguish the relative dose contribution due to alpha particles and lithium ions from the BNC reaction from that of photons and other particles from neutron interactions on the cathode walls. The Relative Biological Effectiveness (RBE) was also calculated from the convolution of the measured spectra with a biological weighting function. This paper describes the experimental microdosimetric approach and the results of measurements with a boron-loaded cathode performed for the first time at an accelerator-based BNCT source.

Flow and heat transfer characteristics of liquid metal and supercritical CO2 in a twisted tube heat exchanger

The Lead-cooled Fast Reactors (LFRs) are fast neutron spectrum reactors cooled by molten lead or lead alloys, which are recommended as one of candidates for the next generation of nuclear reactors due to high performance in sustainability, thermal-hydraulics, and security features. As the critical equipment linking the primary circuit and secondary circuit in LFR, the performance of the main heat exchanger plays a significant role in the operating efficiency of the reactor. In order to improve the heat transfer efficiency and optimize the temperature distribution between the coolant in the primary circuit and secondary circuit of LFR, the flow and heat transfer characteristics of fluids in twisted elliptical tube are investigated in the present study, since the specific spiral cross flow is expected occurring in both inside of the tubes and the interfering area outside the tubes. Theoretical calculation is conducted firstly based on the periodical unit model of the twisted tube heat exchanger. Then the numerical simulation is carried out to study the characteristics of fluid and heat transfer in both shell and tube sides. The results show that the cross flow in twisted tubes improves the overall heat transfer performance of the heat exchanger by strengthening the mixing and turbulence of working fluid. Compared with straight tubes, the entire performance evaluating factor of twisted tubes is around 1.596 times. The theoretical and numerical studies are helpful to the design of the main heat exchanger in LFR.

Genetic algorithm-based optimization of a target for the production of atmospheric-like neutrons via 100 MeV proton beam

This paper presents the optimization of a white neutron target for 100 MeV proton beam to produce a fast neutron spectrum shape similar to that of the atmospheric neutron at ground level. The neutron target consists of 0.5 mm-thick discs consecutively placed in front of a cooling water channel. A sensitivity analysis is performed with different materials to assimilate resulting neutron spectrum with the ground-level cosmic ray-induced neutron spectrum by applying the genetic algorithm (GA). Optimization of the spectrum is conducted in consideration of neutron impact to semiconductor devices. To reconstruct a target-generated neutron spectrum, parametrization on neutron generation and stopping power by target material and position is configured. The loss function in the GA is defined to evaluate similarity between the target neutron spectrum and the atmospheric neutron spectrum. The GA is modeled with a package of consecutive discs as a gene for optimization, and the length and material of target for each disc are decided throughout generations. As a result, the target is configured with six different materials, with approximately half of the weighted deviation of atmospheric neutrons being compared with any other single-solid neutron target. The GA-based optimization process shown here suggests further use for designing targets with other incident proton energies or neutron spectra for various applications.

Assessment of radiation damage of the first wall of a fusion neutron source DEMO-FNS with a blanket for the transmutation of minor actinides

The design of the demonstration fusion neutron source (DEMO-FNS) demonstration fusion–fission hybrid facility is carried out at the National Research Center ‘Kurchatov Institute’ in Russia. For the implementation of the DEMO-FNS project, it is important to assess the effect of the fast neutron spectrum on radiation-induced damage to the device materials, primarily the materials of the first wall, which is the most problematic unit of the device. Due to the lack of real hybrid fusion devices, the study can only be carried out using computer simulations of the experiment. A computer simulation using the Monte Carlo method was carried out to calculate the developed three-dimensional full-scale model of the DEMO-FNS reactor with a blanket for the transmutation of minor actinides. The MCNP-4 code was used with cross-sections from the FENDL-2.1 and ENDF/B-6 files, as well as with cross sections for calculating radiation displacements. The neutron spectrum in the first wall of DEMO-FNS was determined at the power of a volumetric deuterium–tritium neutron of 1.42 × 1019 n s−1. It was found that beryllium as a plasma-facing material has too short lifetime under the neutron load on the first wall and divertor plates of DEMO-FNS. The replacement beryllium to tungsten is problematic and requires special study. Copper alloys have sufficient resistance to the effects of neutron radiation of DEMO-FNS for one full power year, but will require several replacements when operating the device for more than 10 years. The steels used in fast reactors can meet the operational requirements of the first wall of DEMO-FNS. To solve the problem of choosing materials for the first wall of DEMO-FNS, new experimental researches on changes in the physical properties of these materials in the fast neutron spectrum are needed.

Core Configuration Analysis for Modular Gas-cooled Fast Reactor (GFR) using OpenMC

The core configuration analysis of modular Gas-cooled Fast Reactor (GFR) has been done to understand GFR performance. The modular GFR used a fast neutron spectrum and high-temperature helium gas, providing higher thermal efficiency than the other generation IV reactor candidates. In this paper, the variation of core configuration and dimension for core design has been applied in radial, axial, and radial-axial directions. The Monte Carlo method, named OpenMC code, has been used for the criticality and isotope evaluation of design core GFR. The OpenMC code provides the probabilistic solution to solve the neutron transport equation in a 3D model and non-homogenous physical volumes using Evaluated Nuclear Data File (ENDF/B-VII.b5) and continuous energy spectrum. The neutronics parameters characterized are the value of keff, fission rate and neutron flux distribution, and fissile material evolution to know of GFR core design’s performance. The analysis showed that the core configuration in radial direction gives a good understanding of the feasibility of GFR core design.

Neutron spectrum determination of accelerator-driven d(10)+Be neutron source using the multi-foil activation technique

New neutron field of the compact accelerator-driven fast neutron source based on the cyclotron and beryllium target station was studied and developed at the NPI CAS. Neutrons were generated using the deuteron bombardment on 8 mm thick beryllium layer by the 10.3 MeV deuteron beam. Broad neutron energy spectra of the d(10)+Be source reaction at two source-to-sample distances (14 mm and 154 mm) were determined for the first time at the NPI. Sets of eight activation materials comprising Au, Co, Ti, In, Al, Fe, Ni, and Nb were irradiated in the d(10)+Be neutron field, and activated dosimetry foils were analysed using the nuclear gamma-spectroscopy technique (HPGe detector) and reaction rates were measured. Neutron energy spectra of the d + Be source reaction for 10.3 MeV deuterons were reconstructed employing the modified version of the SAND-II unfolding code. For 9.7 μA deuteron beam, the fast neutron flux of developed d(10)+Be neutron field reached the value of 1.4 × 1010 cm−2s−1 at a distance of 14 mm from the source target and 4.1 × 108 cm−2s−1 at a distance of 154 mm. The novel d(10)+Be neutron field with energy range up to 15 MeV is convenient for neutron cross-section data validation, experimental simulation of the fast neutron spectrum of experimental nuclear reactors, applications of neutron activation analysis, and radiation hardness tests of materials important for nuclear energetics.

The dependency of fast neutron flux and spectra on the collimation-field size and geometry for fast neutron therapy purposes

The aim of this work is to optimize and investigate the relationship between the collimation-filed size and their geometries and the fast-neutron intensity and their energy-spectrum. Advantageously for these purposes, a prototype of collimator was designed and constructed to simulate different geometrical and materials compositions. Beams of fast neutrons generated by the reaction 9Be(d, n) with 13.6 MeV deuteron energy were investigated. The average energy of the emitted neutrons is about 6–7 MeV. The maximum intensity was at neutron energy of 4–5 MeV. In the process of the study, experimental works were carried out with Al, Fe, Cu and Cd foils which positioned on the collimator aperture as neutron detectors. The neutron activation method and Gamma-spectrometer were deployed to determine the neutron intensities within the energy-range from thermal energies to 14 MeV. Experiments were conducted for ten different combinations of materials and geometries which can be adjusted by the removable-polyethylene collimator parts and the metal parts. It was concluded that the thermal and fast neutron flux were depended on the collimation-field size and materials, where the fast neutron flux output was about 100% more for bigger collimator 9 × 9 cm2 comparing with the smaller ones 5 × 5 cm2 . And more than 60% in case of adding Lead layers into the collimator. These results could have influential implications on the fast neutron intensities and improvement of collimation systems for fast neutron therapy. Also, it could enhance the irradiation fluxes in the irradiation channels in nuclear reactors for medical isotopes productions and material testing and other applications.

Uncertainty propagation from n+56Fe nuclear reaction model parameters to neutron multiplication factor

Nuclear data are the main uncertainty sources for modern neutron transport calculations. However, complete correlation matrices among different nuclear data are not presented in the current evaluations. In order to completely consider that correlation between nuclear data, the present work propagates the covariance matrix of continuum n+56Fe reaction model parameters to the effective neutron multiplication factor (keff) of a thermal neutron spectrum benchmark and a fast neutron spectrum benchmark. Results show that the uncertainty propagated from angular distribution is not negligible. The general correlations between angular distributions and cross sections of 56Fe for neutron energies above 850 keV are respectively −0.39 ± 0.05 and −1.02 ± 0.02 for keff calculation in the thermal and fast benchmarks, respectively. These correlations reveal the importance of a complete correlation matrix of nuclear data for propagating uncertainties to neutronic calculations.

Integral experiments of technetium-99 using fast-neutron source reactor ‘YAYOI’

ABSTRACT The present study performed integral experiments of 99Tc by an activation method using a fast-neutron source reactor ‘YAYOI’ of the University of Tokyo to validate evaluated nuclear data libraries: JENDL-4.0, ENDF/B-VII and VIII. Technetium-99 samples were irradiated with reactor neutrons using a pneumatic system, and repeated 74 times to accumulate statistics. Reaction rates of 99Tc were obtained by measuring decay gamma rays emitted from 100Tc. The fast-neutron flux spectrum at the YAYOI was checked with flux monitors and MCNP calculations. The experimental reaction rate of 99Tc was compared with the reaction rates given using both the fast-neutron flux spectrum and evaluated nuclear data libraries. The present integral experiments supported the evaluated nuclear data library JENDL-4.0.

Analysis of radiation emission from MYRRHA spent fuel and implications for non-destructive safeguards verification

The radionuclide composition of, and emitted radiation in, spent nuclear fuel from the future MYRRHA facility have been studied using depletion simulations to understand potential consequences for safeguards verification using non-destructive assay. The simulations show that both the gamma-ray and neutron emission rates in spent MYRRHA assemblies are lower than in spent PWR UO2 and MOX assemblies. In addition, gamma-ray emission rates from 134Cs and 154Eu are considerably lower, and the total neutron emission rate in MYRRHA fuel is much less sensitive to fuel burnup and cooling time. The main reason is that the fast neutron spectrum in MYRRHA affects the radionuclide production in the fuel. One result is that 244Cm, the main contributor to the neutron emission in spent light water reactor fuel, has a limited production in MYRRHA. Consequently, neutron-detection techniques could be used to more directly assay the plutonium content of spent MYRRHA fuel.

ПРОБЛЕМЫ ФИЗИКИ И ТЕПЛОГИДРАВЛИКИ В РЕАКТОРАХ, ОХЛАЖДАЕМЫХ ВОДОЙ СВЕРХКРИТИЧЕСКОГО ДАВЛЕНИЯ

The supercritical Water-cooled Reactor (SCWR) was adopted as one of the promising IV-generation reactors within the framework of the international forum “Generation-IV” (MFP). More than 45 organizations in 16 countries with advanced nuclear power are developing SCWR concept proposals for this program. The SCWR concept is based on the implementation of a direct-flow single-circuit scheme of a nuclear power plant, cooled by SCP water. The introduction of this type of nuclear power plant will increase the efficiency up to 45 %, increase the fuel reproduction coefficient, reduce metal consumption and construction volumes, and improve economic and environmental performance. Countries participating in the SCWR MFP consider the development of a reactor with a thermal neutron spectrum and uranium fuel as a priority task, but in the subsequent stages, with increasing problems with the storage of spent nuclear fuel (SNF) and small actinides (SA), it is possible to switch to a reactor with a fast neutron spectrum, MOX fuel and a closed fuel cycle (CFC). Within the framework of the MFP, various versions of SCWR are being developed differing in the parameters of the coolant and its circulation schemes in the core. Groups have been created to study the issues of physics, thermohydraulics, heat transfer, materials, personnel training. Water-cooled reactors research carried out during ~15 years in A.I. Leypunsky Institute for Physics and Power Engineering (IPPE), OKB “Gidropress”, NRC “Kurchatov Institute” with supercritical thermal and fast neutron spectra, it seems more promising to develop a reactor with fast spectrum of neutrons. For ~10 years, IPPE and OKB “Gidropress” have been working together on the VVER-SKD concept project - a single-circuit RC with a coolant SCP with a fast-resonance neutron spectrum with a capacity of Ne = 1700 MW. This rector is recognized as a prospect for the development of VVER technology with the possibility of using uranium fuel and switching in the future to MOX-based fuel (U-Pu-Th) and to SNF. When developing VVER-SKD, it is necessary to solve a complex of scientific and technical problems: development and verification of calculation codes of improved estimation for neutron physics, hydrodynamics and water heat transfer of SCP in fuel assemblies (FA) of the core and throughout the reactor; development of fuel elements and FA structures, justification of their operability; analysis of reactor stability under transient and emergency conditions; selection of heat-resistant structural materials for fuel rods and reactor vessel with high corrosion resistance; justification and development of optimal water-chemical regime, etc. Some of these problems are investigated in bench and loop tests, but to solve most of them and justify the technology for subsequent licensing, it is necessary to create an experimental test reactor. In relation to the VVER-SKD reactor Ne = 1700 MW, the paper presents the results of calculations of fuel cycles with MOX and nitride fuels, justifies the use of NPS in the ZTC, discusses the problems of heat exchange and thermal hydraulics, and suggests a test reactor option.

Calculation and experimental analysis of benchmark experiments with a fast neutron spectrum and models of sodium and lead cooled fast reactors using different evaluated nuclear data libraries

The paper presents the results of a comparative analysis of criticality calculations using a Monte-Carlo code with the BNAB-93 and BNAB-RF neutron group constants, as well as with evaluated neutron data files from the Russian ROSFOND evaluated nuclear data library and other evaluated nuclear data libraries (ENDF, JEFF, JENDL) from different years. A set of integral experiments on BFS critical assemblies carried out in different years at the Institute of Physics and Power Engineering (60 different critical configurations) was analyzed. The considered integral experiments are included in the database of evaluated experimental neutronic data used to justify the neutronic performance of sodium and lead cooled fast reactors, to verify codes and nuclear data as well as to estimate uncertainties in neutronic parameters due to the nuclear data uncertainties. It has been shown that the ROSFOND evaluated nuclear data library is a library that minimizes the calculation and experimental discrepancies for the considered set of integral experiments. The paper also presents the results of criticality calculations for models of sodium and lead cooled fast reactors based on different evaluated neutron data libraries and provides estimates for the uncertainty in criticality associated with nuclear data.

Development and experimental validation of a fast neutron spectrometry system based on Superheated Drop Detectors (SDDs) operating under different external pressures

In this paper, the spectrum of 241Am-Be neutron source was measured by developing and using a neutron spectrometer system based on Superheated Drop Detectors (SDDs) operating under various external pressures. The system consisted of a set of SDDs fabricated by in situ polymerization method. The external pressures were controlled by the amount of Freon-12 added onto the surface of detectors. Also, this system includes a Geant4 simulation application capable of calculating the response of SDDs made from various materials at various temperatures and pressures. Ultimately, the system contained a customized program to design and train the Adaptive Network-based Fuzzy Inference System (ANFIS) for fast neutron spectrum unfolding. This program has the capability of designing the ANFIS with different rule structures and various parameters of the fuzzy inference system.In the present study, the simulation application was used to calculate the response matrix of five SDDs under external pressures of 0.86, 1.93, 2.72, 3.48, and 3.85 atm in 20 energy bins from 0.5 MeV to 514 MeV at 28 °C. Furthermore, the customized program was utilized to design a FIS structure using the FCM method with six membership functions for each input.Neutron spectra published in the IAEA technical reports series no. 403 and 318 were utilized as the Target data. The response matrix was used to obtain the input data of ANFIS by multiplying the target data by the response matrix. 99% of data (373 out of 377) was considered as the train data and the rest as the test data.After fabrication of SDDs, they were exposed to an 241Am-Be neutron source. The average numbers of the formed bubbles in SDDs with different external pressures were introduced to the trained ANFIS, and finally, the unfolded spectrum was the output of the ANFIS. The RMSE for the measured neutron spectrum was the acceptable value of 0.016. The reasonable agreement between the unfolded and reference 241Am-Be neutron spectra shows that the developed neutron spectrometry system can be a suitable candidate for fast neutron spectrometry.