Neutron Spectrum Research Articles

Performance of NB-CTMFD Detector Versus Ludlum 42-49B, and Fuji NSN3 Detectors for Hard (Am-Be) and Soft (Cf-252 Fission) Energy Spectra Neutron Sources Within Lead/Concrete Shielded Configurations

Abstract This paper presents the results of neutron detection efficiency and dosimetry between a nonborated centrifugally tensioned metastable fluid detector (NB-CTMFD) configured to detect fast (>0.1 MeV) versus two widely used combined thermal and fast neutron detection devices of similar form factors: moderated He-3 based Ludlum-42-49BTM neutron detector, and, the Fuji NSN3TM pressurized nitrogen-methane filled neutron detector. The one-on-one performance comparisons were conducted using hard (Am-Be) and soft (Cf-252 fission) neutron spectra isotope neutron sources positioned behind various levels of lead and concrete shielding ranging in thickness from 0 cm (unshielded) to up to 30 cm. The comparisons were conducted together with Monte Carlo N-Particle (MCNP) code simulations to account for 3-D effects and to relate the detection sensitivity with the dose rate. The MCNP model results for the neutron energy spectrum were validated versus experimental measurements using the H*TMFD. While the Ludlum and NSN3 detectors operate at a fixed sensitivity setting, the NB-CTMFD detector could be operated at various sensitivity levels by varying the tensioned metastable negative pressure (Pneg) from 0.3 MPa to 0.7 MPa. The NB-CTMFD (configured for fast: >0.1 MeV neutron detection at Pneg = 0.7 MPa) offered relative sensitivity enhancements of up to 15x greater than the Ludlum and ∼5x greater than the NSN3 detector for low shielding thicknesses. For larger thicknesses, the advantage factor for NB-CTMFD unit reduces with increased fast neutron removal via down-scattering, which benefits the Ludlum and NSN3 detectors. Our companion paper compares performance with a Cf-252 spectrum using a borated CTMFD (covering dual energy bins: thermal-epithermal and fast energy ranges) wherein, it is demonstrated that the advantage factor for the CTMFD (if borated) remains elevated at the 5–10x higher level for low and large shielding thicknesses and soft and hard neutron spectra.

Performance of Borated Centrifugally Tensioned Metastable Fluid Detector Detector Versus Ludlum 42-49B, Fuji NSN3 Detectors for Fission Energy Spectrum Neutron Detection With the Source Within Lead/Concrete Shielded Configurations

Abstract This paper presents the results of neutron detection efficiency and dosimetry between a borated centrifugally tensioned metastable fluid detector (B-CTMFD) configured to detect thermal to fast neutrons versus two widely used neutron detection devices of similar form factors: moderated He-3 based Ludlum-42-49B neutron detector, and, the Fuji Electric's NSN3TM (NNSN3) pressurized nitrogen-methane filled neutron detector. The one-on-one performance comparisons were conducted using soft (Cf-252 fission) neutron spectrum isotope neutron sources positioned behind various levels of lead and concrete shielding ranging in thickness from 0 cm (unshielded) to up to 30 cm. The comparisons were conducted with Monte Carlo N-Particle transport code (MCNP) code simulations to account for three-dimensional effects and to relate the absolute detection rate with the dose rate—to derive the sensitivity factor (cpm/μSv/h). While the Ludlum 42-49B and NSN3 detectors operate at a fixed sensitivity setting, the centrifugally tensioned metastable fluid detector (CTMFD) can be (and was) operated to detect and separate the contributions from epithermal (0.02 eV–0.2 MeV) and from fast (>0.2 MeV) neutrons at various sensitivity levels by varying the tensioned metastable negative pressure (Pneg) from 0.3 MPa to 0.7 MPa. The B-CTMFD (configured for detecting the full 0.02 eV to 12 MeV neutrons >0.1 MeV neutron detection at Pneg = 0.7 MPa) offered relative sensitivity enhancements of up to ∼22× greater than Ludlum and over 5× greater than NSN3, over the 0-15 cm range of Pb shielding, and the 0–30 cm concrete shield thickness. The contribution from detecting down scattered neutrons increases with increased thickness, especially for concrete shielding. The B-CTMFD design overcomes the detection penalty (up to 60% depending on shielding type and thickness) inherent in the nonborated centrifugally tensioned metastable fluid detector (NB-CTMFD), designed only to detect fast-energy neutrons—as described in the companion (Part-1) paper. However, unlike the NB-CTMFD, which used 100% decafluorapentane (DFP) (C5H2F10), the B-CTMFD requires the use of an azeotropic mixture of DFP, methanol, and tri-methyl borate (TMB—C3H9BO3, using natural boron) in 80:4:16 proportion. The B-CTMFD was about 6 times more sensitive than NB-CTMFD under the most heavily shielded condition and, taken together, also offered 2-energy bin neutron spectroscopic enablement, together with 22-5× higher absolute efficiency and relative sensitivity compared with the nonspectroscopic Ludlum (He-3) and NSN3 (methane-nitrogen) based detectors. From an intrinsic efficiency standpoint, the B-CTMFD operating at Pneg = 0.7 MPa state demonstrated ∼103× higher intrinsic efficiency over Ludlum 42-49B and NSN3 detectors, respectively.

Analysis of the relationship between ambient dose, ambient dose equivalent and effective dose in operational neutron spectra.

This study examines the relationship between ambient dose H*, ambient dose equivalent H*(10), and effective dose E for external neutron irradiation over 163 operational spectra from different workplaces. The results show that H* provides a reasonable estimate with a controlled margin, even if overestimated, to assess E compared with H*(10), which can lead to a significant overestimation or underestimation of E depending on the neutron spectra. The results highlight the limitations of H*(10) and the superiority of H* in estimating E according to the requirements of the operational quantities, especially in environments with high-energy neutrons.

Spectra measurement of 238Pu-13C source

The precise description of the neutron spectrum is crucial for the measurement of cross sections and well defined integral experiments. Radioisotope sources such as PuC are one of the most widely used neutron sources for different purposes. The neutron spectrum of this source was measured with narrow energy bins for energies greater than 1.3MeV in this experiment with the method of stilbene scintillation spectrometry and the NGA-01 system. The measured neutron spectrum was also compared to simplified simulations in MCNP6 with satisfactory agreement. Effect of the internal structure of the neutron source was tested thoroughly with simulations and is also presented here.

MTIR-SCP nuclear safety concept

The MTIR-SCP reactor is a research reactor with a light-water coolant having a fast neutron spectrum in nominal operation. None of the existing or projected reactors has a set of these features, therefore, special attention should be paid to the choice of the design of control rods and their location. The article considers various designs of control rods characteristic of the VVER-SCP power reactor currently under development, as well as existing research reactors with a liquid metal coolant having a fast neutron spectrum, taking into account the features of the MTIR-SCP: a seven-element control rod assembly design, as well as a ring-type absorbing element. Efficiency calculations are carried out for each of the structures under consideration and it was determined that the design of a control rod with an annular type of absorbing element has the highest specific efficiency. This design is chosen as the main one for the MTIR- SCP reactor. Criteria are proposed and optimization of the placement of control rods in the reactor core is carried out in order to reduce the uncertainty of obtaining neutron-physical characteristics in the central autonomous loop channel and increase the flow density in it. When optimizing the placement of control rods in the research reactor, their influence on the main neutron-physical characteristics, and especially on their distribution during operation under nominal conditions, is taken into account. Emergency protection and compensating rods that compensate for temperature and density effects when the reactor reaches the nominal power level have been removed, therefore their effect on the neutron-physical characteristics of the reactor and the loop channel during operation at power is not significant, and the compensating rods are located in the core and are removed as the fuel burns out, therefore their effect on the spatial The distribution of neutron-physical characteristics will be significant. To reduce this effect, burnout compensators and automatic regulators are proposed to be moved outside the core into the reflector. For various reactor conditions, the efficiency of the control and protection system was evaluated. It is shown that in various states, the control rods transfer the reactor to the required level of subcriticality, in accordance with NP-009-17.

Evaluation of source terms using the BESNA code for reference spent fuels for future underground disposal system of South Korea

In this work, the source-terms such as nuclide inventories, radioactivities, neutron and gamma spectra, and decay heats were evaluated using new code BESNA in which the burnup dependent one-group cross sections were evaluated based on MCNP6. This work started with the selection of the reference spent fuels through a statistical analysis of all the accumulated PWR spent fuels in South Korea for safety analysis of a geological repository, which led to the two reference spent fuels for each of 16x16 and 17x17 fuel assembly types considering representative low and high burnups. Finally, the source terms were evaluated using our inhouse code BESNA for the reference spent fuels, and they were compared with those from ORIGEN in SCALE6.2 for validation purposes. The results of the comparison showed that BESNA gives good agreements with ORIGEN for all the considered source terms.

Physical model of neutron scattering by clathrate hydrate and C 60 hosting paramagnetic oxygen molecules

This paper describes the physical modelling of neutron scattering in two polycrystalline inclusion compounds, fully deuterated clathrate hydrate and C60 , each with paramagnetic oxygen as guest molecules. For studying the suitability of these materials for neutron moderation to very low energies, the model includes, in addition to the magnetic neutron scattering by the oxygen, the nuclear scattering by all constituents. The theoretical total cross sections are calculated based on the phonon density of states obtained by density functional theory and molecular dynamics simulations. At low temperatures, the developed scattering kernels are in good agreement with experimental neutron scattering data reported in the literature. At 20 K and above, a Lorentzian distribution for the zero-field splitting of the magnetic substates of the spin triplet of the oxygen molecules helps to reproduce magnetic peaks observed in inelastic neutron scattering experiments better than the original theory based on a single-valued splitting constant. Neutron spectra obtained by Monte Carlo simulations in infinite media are presented, highlighting the potential use of O2 -containing fully deuterated clathrate hydrate as a neutron moderator for the production of very cold neutrons.

Independent isomeric yield ratios of fission products in the epi-cadmium neutron induced fission of 239Pu

Measurement of independent isomeric yield ratios (IR) of 128,130,132Sb, 131,133Te, 132.134,136I, 135Xe and 138Cs have been carried out for the first time in the epi-cadmium neutron induced fission of 239Pu by using an off-line gamma-ray spectrometric technique. The average neutron energy (<En>) of the epi-cadmium reactor neutron spectrum is 1.9 MeV. From the IR values, root mean square fragment angular momenta (JRMS) were deduced by using spin dependent statistical model analysis. Effect of nuclear structure on JRMS values was examined. The present data in the epi-cadmium neutron induced fission of 239Pu were compared with the similar data in the thermal neutron induced fission of 239Pu to examine the role of excitation energy on JRMS values.

Analyses of hydrogen local environments in metals and intermetallic compounds using inelastic neutron scattering calculations based on first-principles hydrogen adiabatic potentials

This study re-evaluates the theoretical approach to analyzing inelastic neutron spectra of hydrogen-containing metals and intermetallic compounds. Previously, these analyses utilized hydrogen quantum nuclear states, modeled as solutions to the Schrödinger equation. The potential surfaces in these models were approximated from the total energies derived from first-principles electronic structure calculations. The current study improves upon this method by employing more efficient and accurate treatments for sampling the potential surface. It utilizes symmetrically irreducible sampling points arranged on densely populated mesh grids for the first-principles calculations. A comparative analysis of the theoretical predictions with experimental spectra for hydrides of Ti2Sb and Ti3Sb, as well as a LaNi5 hydrogen primary solid solution, demonstrates that this approach is promising for elucidating the unknown local environments of hydrogen atoms in systems where the approximate potential well describes the hydrogen quantum states.

The Use of High-Density UN Fuel in Heat-Pipe Microreactors

Heat-pipe microreactors (HPMRs) are very small-scale nuclear reactors that employ heat pipes (HPs) for heat removal. HPMRs can be easily integrated with other forms of renewable energies, can be used for emergency responses to disaster relief zones, can be deployed in remote locations not connected to the grid, and can be removed from sites and replaced by new ones. HPMRs can also be used for space missions as HPs do not rely on gravity for heat transfer. Conventional fuel materials, such as uranium oxide (UO2) and uranium oxycarbide (UCO), are currently considered in most existing HPMR designs, but ceramic uranium nitride (UN) fuel that has high uranium density, high thermal conductivity, and high melting point may become a better fuel candidate. Through neutronics calculations, this paper assesses the impact of using UN fuel in HPMRs with two different neutron spectra (fast and thermal) and two different fuel forms [traditional solid fuel pellets and TRi-structural-ISOtropic (TRISO) fuel compacts]. It was concluded that retrofitting HPMRs with UN fuel has the potential to reduce the initial 235U enrichment requirement by ~3 wt% (to keep the same cycle length) or increase the cycle length (by keeping the same initial 235U enrichment), which enables more compact and transportable HPMR core designs. However, using UN fuel decreases the control element worth [by up to 20% for the Special Purpose Reactor (SPR) and 5% for HP-MR] and is up to 80% more costly. Increasing 15N enrichment can further decrease the initial 235U enrichment requirement and increase the control element worth but is more costly. Compared to fast-spectrum HPMRs fueled with solid pellet fuels, retrofitting UN fuel is more suitable for thermal-spectrum HPMRs fueled with TRISO fuel compacts, where the neutron spectrum hardening caused by using UN is less significant.

Long term reactivity operation for a high temperature testing reactor (HTTR) loaded with uranium nitride fuel

A High Temperature Testing Reactor (HTTR) is one of the fourth generation nuclear reactors, the reactor uses UO2 as typical fuel with different axial and radial enrichments. In the present research, uranium nitride (UN) is proposed as a fuel for the reactor for the same core enrichment distribution. The core multiplication factor, cycle length, discharge burnup are calculated and the results of UO2 fuel are compared with UN fuel results. Neutron spectrum, fissile isotopes behavior and Breeding capability are compared for the two fuel types. Kinetic parameters, Control rods and Burnable poisons worth are also evaluated. The results indicated that uranium nitride (UN) has a longer fuel cycle, higher breeding capability and harder thermal flux compared to UO2 fuel.

Analysis of the neutron spectrum in the infinite homogeneous circulating fuel reactor

An analysis of the behavior of the neutron spectrum in transient condition in circulating fuel reactors, based on an extension of the inhour equation to the energy dimension, is presented. This approach enables the description of the time evolution of the spectrum as a combination of eigenstates associated to the roots of the inhour equation. The differences between the inhour equation for the conventional and circulating fuel reactors are discussed. An example compares the neutron spectra inside a conventional and a circulating fuel reactor. It is found that spectrum differences are of the order of magnitude of 0.5%. Stability features are also discussed.

A passive single-cylindrical neutron spectrometer for photoneutron spectrometry in 10 MV Elekta Precise Linac head

Measuring the neutron spectrum on the Linac head using a passive type neutron detector is relatively easier than an active type detector because it does not require the involvement of operators who work in a high radiation area. Neutron spectrum measurements with passive foil activation neutron detectors in a single-moderator are more practical and easier than multi-foil techniques. This technique is quite simple, easy to make, low cost, and the measurement time is relatively short. This spectrometer is designed by placing seven passive Au detectors at varying depths in a single-cylindrical moderator. The response function for each detector is calculated by MCNPX 2.7 program and used ENDF7 as a nuclear data library. Detector foil activity was measured using a gamma spectrometer, and the neutron spectrum was obtained using the unfolding technique using the UMG 3.3 program. The results of measuring the neutron spectrum measurements have been carried out at a position 10 cm below the head of the Elekta Precise 10 MV Linac was 2.73 × 105 n/cm2·s, where the fast neutron from these results are 4% higher when compared to the simulation results using PHITS.

Real-space local self-motion of protonated and deuterated water.

We report on the self-part of the Van Hove correlation function, the correlation function describing the dynamics of a single molecule, of water and deuterated water. The correlation function is determined by transforming inelastic scattering spectra of neutrons or x rays over a wide range of momentum transfer Q and energy transfer E to space R and time t. The short-range diffusivity is estimated from the Van Hove correlation function in the framework of the Gaussian approximation. The diffusivity has been found to be different from the long-range macroscopic diffusivity, providing information about local atomic dynamics.

Fast flux spectrum unfolding of PFTS of KAMINI: an investigation into the viability of radioisotope production

The fast component (0.7–18 MeV) of the neutron spectrum at the Pneumatic Fast Transfer System (PFTS) in the KAMINI reactor was determined by irradiating threshold activation foils. The acquired reaction rates are subsequently unfolded using SAND-II code. The KAMINI reactor was modelled in the MCNP-4B code for an estimate of neutron spectrum at the PFTS location, which was subsequently used as the initial apriori solution for the SAND-II code. The unfolded fast component spectrum at the PFTS was found to be similar to that of the fast benchmark GODIVA and JEZEBEL-23. Moreover, the investigation is substantiated by incorporating modified spectra to estimate the theoretical yield of radioisotopes 32P at PFTS using a Monte Carlo simulation against its experimental yield. The results agree with an accuracy of 6%, affirming the credibility of the derived spectrum. Theoretical yields of 89Sr, 64Cu, and 47Sc were also estimated at the same location.

Experiments on Criticality and Reactivity Worths in the FCA-XXII-1 Assembly Simulating Highly Enriched MOX Fueled Tight Lattice LWR Cores

A series of integral experiments were conducted at the fast critical assembly (FCA) of the Japan Atomic Energy Agency, simulating light water reactor cores with a tight lattice cell of highly enriched mixed-oxide (MOX) fuel containing >15% fissile plutonium (Pu). The three experimental configurations of the FCA-XXII-1 assembly were constructed using foamed polystyrene with different void fractions (45%, 65%, and 95%) to clarify the prediction accuracy of neutronics calculation codes and nuclear data libraries among various neutron spectra. The hydrogen-to–nuclear fuel atomic ratio varied from 0.1 to 0.8. The nuclear characteristics measured in the experiments were criticality (keff), moderator void reactivity worths, and sample reactivity worths using boron carbide (20%, 60%, and 90% 10B enrichment) and Pu (92%, 81%, and 75% fissile Pu ratio). Preliminary analyses on experiments were conducted using a deterministic calculation code system conventionally used for fast reactors and the Japanese evaluated nuclear data library of JENDL-4.0. The calculated keff values overestimated the experiments beyond the experimental uncertainties. However, most reactivity worth calculations agreed well with the experimental values. Even beyond the experimental uncertainties, discrepancies between the calculation and the experiment were <13%. Specifically in the reactivity worth analyses of the softer neutron spectra configurations, the treatment of ultrafine energy groups obviously improved the prediction accuracy of the deterministic calculations. Furthermore, reference calculations for criticality and large reactivity worths were performed with the Monte Carlo calculation code MVP3 by modeling the experimental configurations in detail, confirming that the deterministic calculations closely agreed with the reference values.

Calibration and validation measurements of the Extended-range Bonner sphere spectrometer

An Extended-range Bonner sphere spectrometer (EBSS) has been developed to investigate the neutron spectra of the China initiative Accelerator Driven System. This paper presents the design, calibration, and validation measurements of the EBSS system using a standard 241Am-Be neutron source and the cosmic ray neutrons. The EBSS system was simulated using the PHITS code, and the geometric structures were designed using the response functions obtained by simulations. The EBSS system comprises of seven polyethylene-only spheres and seven extended-range spheres encased in lead, copper, or tungsten shells along with a bare 3He proportional counter. To verify the accuracy of the simulated response functions, the EBSS system was calibrated with monoenergetic neutron beams of 565 keV at the China Institute of Atomic Energy in Beijing. The EBSS system was used to measure neutron count rates of the 241Am-Be neutron source and cosmic ray neutrons. Subsequently, the measured neutron count rates were converted into neutron spectra by the unfolding software. Experimental findings demonstrate the precise capability of the EBSS system to measure neutron spectra across a wide range of neutron radiation fields.

Performance of deconvolution codes (MAXED, GRAVEL, MLEM) for neutron spectrometry of radioactive source using plastic scintillator simulated data

Neutron spectrometry plays a crucial role in decommissioning of nuclear sites and in ensuring the safety of workers from radiation exposure. In this study, we associate a proton recoil spectrometer made by a homemade plastic scintillator with triple discrimination capabilities (fast neutrons, thermal neutrons, and gamma rays) with deconvolution methods to obtain accurate measurements of neutron spectra. The purpose of this work is to test and evaluate the performance of three distinct deconvolution codes, namely MAXED (Maximum Entropy Deconvolution), GRAVEL, and MLEM (Maximum Likelihood Expectation Maximization), using two types of input spectra (flat and Watt spectrum). These deconvolution codes are applied to simulated data using the reference MCNP6.2 Monte Carlo code. Comparing the calculated mean squared error (MSE) performed by the three unfolding methods, we find that MLEM seems to perform better than MAXED and GRAVEL. Furthermore, given the calculated MSE values, the unfolded spectrum of Cf-252 is in better agreement with the standard reference spectrum by using Watt spectrum as input spectrum (MSE<1.2 × 10−6) than using a flat spectrum (MSE<1.3 × 10−6).

Neutronic Analysis of Accident-Tolerant Cladding Materials in 3D Full Core BEAVRS PWR Benchmark Using OpenMC Code

After the Fukushima Daiichi nuclear accident in 2011, the performance of nuclear fuel during accidents became a matter of great concern. To address this, a new type of fuel technology called accident-tolerant fuel (ATF) has been developed with the goal of enhancing the ability of light water reactors (LWRs) to withstand severe accident conditions. Iron-based alloys have been suggested as potential candidates for fuel cladding due to their favourable thermomechanical properties, lower reactivity with steam, and lower hydrogen generation. This study evaluates the neutronic performance of C26M (a 2nd generation nuclear grade FeCrAl alloy), APMT™, 310SS, and 304SS cladding materials by comparing them with Zircaloy-4 cladding in a 3D PWR core at the beginning of the cycle (BOC) using OpenMC code. The results revealed that the neutronic penalty varied for different alternative cladding materials where C26M exhibited the lowest neutronic penalty value of -12551 pcm, while 310SS demonstrated the highest with a value of -17855 pcm. Additionally, important parameters in the reactor core such as neutron spectrum, reactivity coefficients, boron worth, control rod bank worth, power distribution, and radial thermal neutron flux distribution are evaluated and discussed. The analysis results showed that C26M provided a significantly higher level of neutronic performance compared to APMT™, 304SS, and 310SS. Although this study primarily focused on the neutronic performance of PWRs at BOC, future research should encompass fuel depletion analysis to delve deeper into the potential of alternative cladding materials.

Unfolding of mono-energy neutron spectra using artificial neural network based on LMBP training algorithm

In this work, neutron spectra are unfolded using artificial neural networks (ANNs). The neutron response of the NE213 scintillator detector, characterized by the pulse height distribution, is calculated to obtain the necessary data for unfolding the energy spectrum. This is achieved using both analytical response functions and response functions generated by the MCNPX-PHOTRACK code. In this query, the Levenberg-Marquardt method (LMM), which has a high computational speed in the learning method, is used to train the network. The performance of the ANN for unfolding the neutron energy spectrum of the NE213 scintillation detector was evaluated by comparing its results to the established Gravel method. The ANN method consistently produced spectra with a single peak closely matching the incident energy, while the Gravel method showed additional peaks and distortions. Quantitative analysis revealed a lower relative energy peak difference (indicating higher accuracy) for the ANN method compared to Gravel, particularly when noise was introduced into the data. These results suggest that ANNs offer a more robust and accurate approach for neutron spectrum unfolding.