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The Philippine Research Reactor-1 (PRR-1) will be revived as a subcritical assembly for training, education, and research (SATER) in the field of nuclear science and technology. SATER will utilize the existing slightly irradiated nuclear fuel rods that have been maintained in the PRR-1 facility for more than 30 years. A subcritical arrangement for the fuel rods was chosen considering the inherent safety of this configuration. In this paper, we calculate relevant reactor parameters that characterize different annular and hexagonal subcritical configurations of 44 PRR-1 fuel rods. These parameters include the neutron multiplication factor (), the effective delayed neutron fraction (), and the mean neutron generation time (), which are essential quantities to describe reactor behavior. Calculations were performed using the well-validated Monte Carlo radiation transport code MCNP5 v.1.6 together with the ENDF/BVII. 1 evaluated nuclear data library. The maximum value is at 4.0 cm pitch for the chosen annular arrangement, while the maximum value is at 4.3 cm pitch for the chosen hexagonal arrangement. For these configurations, the reactor kinetic parameters were and for the annular arrangement, while and for the hexagonal arrangement. Results demonstrate that with 44 fuel rods, different fuel arrangements remain subcritical with a subcriticality margin that is at least or maximum of 0.97. The key reactor performance characteristics determined in this study can aid in the analysis of transient behavior and safety assessment of subcritical core configurations with TRIGA fuel rods. Our results provide support in expanding the utilization options for irradiated TRIGA fuel rods, even for other TRIGA facilities. [ABSTRACT FROM AUTHOR]

Determining the neutron source strength is essential to ensure optimal use of the source and to implement adequate radiation protection measures. In this study, the source strength of two (α ,n) neutron sources – a newly acquired 185 GBq 241AmBe and a legacy source 74 GBq 239PuBe – were verified. Source strength verification was performed by measuring the induced activities in irradiated 115In foils and comparing the results to MCNP calculated reaction rates. The results for 241AmBe agree well with a maximum and average deviations of 8.1% and 4.1%, respectively. While for 239PuBe the maximum deviation is 11.7% and the average deviation is 6.3%. Results of this study provide a reliable reference for the application of the sources, in particular, for the 241AmBe which is used as the external neutron source of the PRR1-Subcritical Assembly for Training, Education, and Research (PRR1-SATER), and for the design of a neutron irradiator that will repurpose the 239PuBe. • The source strengths of two (α,n) neutron sources, 241AmBe and 239PuBe were verified through foil activation and Monte Carlo simulation. • Calculated reaction rates and measured saturation activities for the 241AmBe neutron source agree well, with an average deviation of 4.1%. • For the legacy 239PuBe neutron source, calculated and measured results have an average deviation of 5.3%. • The study provides a reliable reference for determining neutron source strength to facilitate efficient utilization of the source. [ABSTRACT FROM AUTHOR]

A Windows-based application software called EpiXS (available at https://www.pnri.dost.gov.ph/index.php/downloads/software/epixs) was constructed for photon attenuation, dosimetry and shielding, and is based on EPICS2017 of ENDF/B-VIII and EPDL97 of ENDF/B-VI.8. The software performs data library interpolation between 1 keV and 100 GeV and calculates partial or total cross sections (σ), mass attenuation coefficients (μ/ρ), linear attenuation coefficients (μ), mean free paths (mfp), half-value layers (hvl), effective atomic numbers (Z eff), electron densities (N eff), and the auxiliary parameters and buildup factors for energy absorption (EABF) and exposure (EBF) via Geometric-Progression, for any user-defined material. It has a user-friendly design and performs quick interactive graphing and data tabulations. Data generated by the software were compared with published results from literature as well as generated results from XCOM and from other programs based on XCOM/NIST libraries. The comparisons were mostly found to agree well, nonetheless the deviations observed may highlight the values from the latest EPICS2017. The EpiXS is introduced as a new tool for instant calculations involving ionizing photon radiation dosimetry and shielding. • Software was made for calculating σ, μ/ρ, ρ, mfp, hvl, Z eff , N eff , EABF and EBF. • All photon energies from 1 keV to 100 GeV can be interpolated. • X-ray absorption edge energies in EPICS2017 and EPDL97 can be viewed. • Data tables can be generated and exported as CSV for MS Excel. • Graphs can be exported as vector image EMF files. [ABSTRACT FROM AUTHOR]

A multi-shell neutron spectrometer with indium foil detector (In-MuNS) was developed to evaluate intense neutron fields that are generated in medical accelerators. The response matrix of this new spectrometer was calculated from 1 meV to 100 MeV using MCNP5 v.1.6 with ENDF/B-VIII.0 nuclear data. An experiment with a252Cf source with known emission rate was performed to validate the computational model of the spectrometer. This included detailed modelling of the irradiation room to evaluate the room-scattered field. The contribution of scattered neutrons to the induced activity in the foil reached 30% for the smallest sphere configuration (diameter 5.0 cm). The quotient between the experimental and simulated foil activity remained satisfactorily constant (1.03 ± 0.04) as the sphere diameter varied, demonstrating the validity of the simulation model. In-MuNS proved to be a portable and compact alternative to conventional Bonner spheres. • A portable and flexible design for passive neutron spectrometry system is presented. • An indium-based multi-sphere neutron spectrometer (In-MuNS) was developed. • The In-MuNS computational model was validated using a252Cf source. • Scattered neutron in the validation experiment was analyzed. [ABSTRACT FROM AUTHOR]

A new type of compact multi-shell neutron spectrometer (MuNS) was developed at PNRI. It consists of a series of concentric polyethylene shells that can be combined to form Bonner spheres with different diameters ranging from 5 cm to 30 cm. With respect to conventional Bonner spheres, this system is more flexible and portable. MuNS can accommodate either passive or active neutron detectors. The response matrix of the MuNS with a cylindrical 3 He proportional counter (He-MuNS) was calculated using Monte Carlo simulations for incident neutrons with energies ranging from 1 meV to 100 MeV. Anisotropy effect was investigated through simulations and is found to be at most 90% for the smallest configuration, indicating that He-MuNS should have a specific orientation during deployment. Validation of the calculated response matrix and calibration of the He-MuNS system were performed using a 252Cf reference source and the shadow cone method. The quotient of the measured and calculated neutron counts in the 3 He detector was computed for every sphere and the value was found to be constant within 5.1%. Results confirm the validity of the He-MuNS model and the simulated response matrix. [ABSTRACT FROM AUTHOR]

A portable neutron spectrometry system was designed based on thermal neutron detectors embedded in concentric polyethylene spherical shells. The system is flexible and can accommodate the use of either active or passive neutron detectors in different configurations. In this work, the response matrix of the system with In-115 foil detectors was calculated with MCNP5 v.1.6. Activation foils were chosen as an ideal detector for the planned use of the system in medical accelerator environments. Calculations were performed using ENDF/B VII.0 and ENDF/B VIII.0 data libraries. The response functions calculated with the two libraries differ by as much as 11.6% in the thermal energy region for the largest moderator. A sensitivity analysis was also performed to evaluate the effect of main design parameters on the response matrix. • A portable neutron spectrometry system (PNSS) that can accommodate neutron detectors in various configurations is designed. • The current work focused on indium activation foils due to its advantages as a passive thermal neutron detector. • Response functions calculated with ENDF/B VIII.0 and ENDF/B VII.0 deviate by as much as 11.6% for the largest moderator. • The effect of relevant design parameters on the PNSS response matrix is negligible within a certain range. • PNSS provides an alternative flexible configuration for Bonner sphere spectrometers. [ABSTRACT FROM AUTHOR]

The mass attenuation coefficient (MAC), effective atomic number (Zeff), and effective electron density (Neff) values of soils in mangrove forests across the Philippine islands were investigated. In addition, ENDF/B-VI.8 cross-section data library was used for interpolating Zeff and Neff values. Photon energies considered in this study ranged from 59.5-1332 keV. MAC values were obtained using MCNP5 and PHITS Monte Carlo (MC) simulations and were found to be in good agreement with XCOM and values in literature. Results for Zeff were in good agreement with values obtained using Phy-X/ZeXTRa program. Each shielding parameter was investigated as a function of SiO2 concentrations. [ABSTRACT FROM AUTHOR]

A method for estimating source strength measurements of californium neutron sources is presented, based on the model of 252Cf, 250Cf, and 248Cm decay. This is combined with the Monte Carlo method (MCM) of propagating uncertainties. Californium sources were categorized into two types: Sort–A are those with most input quantities known while Sort–B are sources with only the mass at a certain reference date known. For Sort–A, the spread of all input quantities was estimated with Gaussian distribution and the deterministic 1st order GUM uncertainty propagation is applied to validate the MCM results. While, for Sort–B with inputs that have non-Gaussian distributions, only MCM is applied to evaluate uncertainties. Results show that for californium sources that are 25 y or older, a simple 252Cf decay correction is imprecise due to the contribution of 250Cf and 248Cm. The MCM was also shown to be a robust technique for uncertainty analysis that provides results for both Gaussian and non-Gaussian distributions. Moreover, the time-dependence of the contributors in the source strength and the corresponding uncertainties are presented. When exceedingly low uncertainties are not required, the calculation techniques presented in this work may serve as an alternative to actual measurements, which tend to be expensive. • A method of estimating neutron source strength was investigated. • Monte Carlo method in GUM supplement 1 was used for uncertainty propagation. • The consequences of the nuclear data uncertainties in literature were evaluated. • The consequence of an unknown isotopic impurity content was evaluated. • The consequences of unknown dates of isotopic analysis and Cm separation were evaluated. [ABSTRACT FROM AUTHOR]

A fast neutron mass removal cross sections (Σ R / ρ) database covering the elemental range of Z = 1 to 92 is generated for source neutron energy distributions based on 241Am–Be, 252Cf, and 235U. Monte Carlo simulations using ENDF/B-VIII.0 nuclear data were performed with the fast neutron removal theory's ideal spherical representation. The ENDF/B-VIII.0-based Σ R / ρ ′s derived were compared with values from the Oak Ridge National Laboratory's Lid Tank Shielding Facility (LTSF) experiments, values from software Phy-X/PSD and MRCsC, values obtained by an empirical model in Zoller, and values derived using the older ENDF/B-VII.0 nuclear data library. Subsequently, the generated Σ R / ρ ′s were evaluated for convergence under varying moderator thickness. Results indicate that the Σ R / ρ quantity for fission and (alpha, n) spectra is universal, with slight deviations for low Z elements. The Σ R / ρ values generated from ENDF/B-VIII.0 agreed with LTSF experiments with an average relative deviation of 8.58%. Moreover, using ENDF/B-VIII.0-based Σ R / ρ ′s, the derived fast neutron effective removal cross sections (Σ R 's) for several light compounds and dry concrete have better agreement on average with LTSF experiments compared with values derived from the available Σ R / ρ ′s in the literature. Significant differences were found for Σ R / ρ ′s that were generated with ENDF/B-VIII.0 as compared with the superseded ENDF/B-VII.0. Furthermore, ENDF/B-generated Σ R / ρ ′s through spherical model showed the expected convergence towards singular values at moderator thicknesses practical for fast neutron removal theory applications. Owing to the relatively high unavoidable uncertainties of the LTSF experimental Σ R / ρ ′s, the ENDF/B-VIII.0-based Σ R / ρ ′s present a new and potentially favorable database in Σ R general calculations for homogeneous multi-element materials for point-kernel shielding applications. • A Σ R / ρ database was produced using the latest ENDF/B via multi-layer sphere model. • Produced Σ R / ρ ′s led to calculated Σ R values close to LTSF experimental measurement. • All Σ R / ρ ′s for fission and (α,n) neutron source energy spectra were similar. • Differences were found in Σ R / ρ ′s produced using 7th and 8th versions of ENDF/B. • Convergence of Σ R / ρ ′s at increasing moderator thickness was demonstrated. [ABSTRACT FROM AUTHOR]