252Cf Source Research Articles

A closer look to dose assessment and radiation shielding characteristics of concrete doped magnetite irradiated with 252Cf mixed radiation radionuclide: A Watt Fission approach and Doppler effect

Nuclear radiation emitted by fusion reactors, nuclear power plants, and medical establishments presents potential risks to living organisms personnel, necessitating the implementation of protective measures. To enhance radiation protection for patients workers, various materials can be utilized. Concrete, augmented with various additives, has historically acted as a shielding material. Hence, recent research has predominantly focused on enhancing concrete's ability to attenuate the harmful energy emitted by nuclear sources through modifications to its composition. Accordingly, in the present work, the dose evaluation and radiation shielding characteristics of a range of concrete magnetite (CM) formulations designated as CM-0 (control sample), CM-25, CM-50, CM-75, and CM-100 have been analyzed using MCNPX Monte Carlo (MC) approach and theoretical computations concerning 252Cf mixed radiation radionuclide. In this work, the Watt Fission distribution was employed to derive the neutron spectrum of CM samples, and findings have been thoroughly elucidated in the presence and absence of the specified samples. Then, utilizing the Doppler Effect, the gamma photon spectrum within shielding materials exposed to a spontaneous fission 252Cf source is extracted and characterized. Estimation of Half Value Thickness (HVT) and Mean Free Path (MFP) are provided across a broad spectrum of energy levels. The analysis confirms the successful development of a new type of concrete magnetite (CM) sample that exhibits lower radiation exposure compared to the control sample. This study offers valuable insights into the use of concrete in shielding against mixed radiation radionuclides and opens the door for future research involving similar materials. Specifically, the CM-100 sample demonstrated the lowest half-value thickness (HVT) and provided the most effective reduction of both neutron and gamma radiation. The findings suggest that increasing the concentration of magnetite in concrete greatly enhances its ability to shield against mixed neutron-gamma radiation. This innovation has promising potential for applications in radiation protection, particularly within nuclear reactors and medical facilities. The CM-100 sample showed a notable improvement, achieving an HVT of 0.012 cm and a dose rate reduction of 2.95 × 10−9 Sv.h−1, in contrast to the control sample (CM-0), which had an HVT of 10.358 cm and an equivalent dose rate of 2.84 × 10−9 Sv.h−1. These results underscore the superior shielding properties of the magnetite-doped concrete formulations.

Solid, structured composite neutron detectors with high dynamic range capability

Neutron detectors are essential across disciplines such as fundamental science, nuclear security, safeguards, and civilian applications. While 3He-filled gas proportional counters have long been revered for their efficacy in detecting thermal neutrons and praised for their efficiency, neutron/gamma discrimination, and stability, the scarcity of 3He has spurred a search for alternatives. Here, we explore a solid structured scintillating particle composite (SPC) consisting of 6Li-containing scintillating glass particles within an acrylic matrix as a neutron detector for high dynamic range applications. We show for the first time that an SPC neutron detector can boast an intrinsic detection efficiency of 0.261% for pure 252Cf fission neutrons and an overall neutron detection efficiency of (0.546 ± 0.003)% at the Neutron Free-in-Air facility while being able to function in an intense gamma-ray environment. We also show that the SPC neutron detector supports fast neutron capture times and enables a dual-readout scheme that extends the detector dynamic range to high incident neutron fluxes. A scalable fabrication process allows for tailoring the SPC detector properties to the requirements of specific applications. Good agreement is found between the experimental results taken with a National Institute of Standards and Technology traceable 252Cf source and the coupled MCNP6 and optical-ray-tracing simulations.

Proton discrimination in CLYC for fast neutron spectroscopy

The Cs2LiYCl6:Ce (CLYC) elpasolite scintillator is known for its response to fast and thermal neutrons along with good γ-ray energy resolution. While the 35Cl(n,p) reaction has been identified as a potential means for CLYC-based fast neutron spectroscopy in the absence of time-of-flight (TOF), previous efforts to functionalize CLYC as a fast neutron spectrometer have been thwarted by the inability to isolate proton interactions from 6Li(n,α) and 35Cl(n,α) signals. This work introduces a new approach to particle discrimination in CLYC for fission spectrum neutrons using a multi-gate charge integration algorithm that provides excellent separation between protons and heavier charged particles. Neutron TOF data were collected using a 252Cf source, an array of EJ-309 organic liquid scintillators, and a 6Li-enriched CLYC scintillator outfitted with fast electronics. Modal waveforms were constructed corresponding to the different reaction channels, revealing significant differences in the pulse characteristics of protons and heavier charged particles at ultrafast, fast, and intermediate time scales. These findings informed the design of a pulse shape discrimination algorithm, which was validated using the TOF data. This study also proposes an iterative subtraction method to mitigate contributions from confounding reaction channels in proton and heavier charged particle pulse height spectra, opening the door for CLYC-based fast neutron and γ-ray spectroscopy while preserving sensitivity to thermal neutron capture signals.

Overview and current challenges at the Calibration Laboratory of the Paul Scherrer Institute

The Calibration Laboratory of the Paul Scherrer Institute (PSI) is a secondary calibration laboratory accredited by the Swiss Accreditation Service (SAS) in accordance with ISO 17025. It also acts as a verification body authorized by the Swiss Federal Institute of Metrology (METAS). The Laboratory is equipped with X-ray, gamma and neutron irradiation facilities, providing characterized reference radiation fields for gamma 137Cs, 60Co, narrow spectrum X-rays from N-15 to N-300 and neutron radionuclide sources 252Cf, 252Cf (D2O-moderated) and 241Am-Be. The laboratory performs routine calibrations, organizes intercomparison measurements for individual monitoring services in Switzerland, and participates in various research projects, e.g., by performing irradiations in reference fields. For this reason, the proper characterization of reference radiation fields is of utmost importance for ensuring high quality irradiations and calibrations of personal dosemeters and radiation protection instruments. In this contribution, we provide an overview of the irradiation facilities at the Calibration Laboratory at PSI and discuss the challenges of characterizing reference radiation fields. In particular, we will discuss the factors affecting the determination of the distance between the source and irradiated dosemeters and detectors, as well as issues related to 252Cf decay and contribution of 250Cf. We will also demonstrate the use of Monte-Carlo simulations in determining the optimal position of the source and irradiation facility within the irradiation room.

Compact and transportable system for detecting lead-shielded highly enriched uranium using 252Cf rotation method with a water Cherenkov neutron detector

The global challenge of on-site detection of highly enriched uranium (HEU), a substance with considerable potential for unauthorized use in nuclear security, is a critical concern. Traditional passive nondestructive assay (NDA) techniques, such as gamma-ray spectroscopy with high-purity germanium detectors, face significant challenges in detecting HEU when it is shielded by heavy metals. Addressing this critical security need, we introduce an on-site detection method for lead-shielded HEU employing a transportable NDA system that utilizes the 252Cf rotation method with a water Cherenkov neutron detector. This cost-effective NDA system is capable of detecting 4.17 g of 235U within a 12 min measurement period using a 252Cf source of 3.7 MBq. Integrating this system into border control measures can enhance the prevention of HEU proliferation significantly and offer robust deterrence against nuclear terrorism.

Assessment of the skin contamination dose coefficients for 252Cf radionuclide: Monte Carlo approach

Handling the 252Cf radionuclide source poses a potential hazard of skin surface contamination in case of an unexpected occurrence. Consequently, there is a growing need to establish precise dose conversion coefficients tailored to each type of emitted primary particle and various radionuclides. Nevertheless, the current body of literature does not provide specific data or methodologies for evaluating skin contamination dose and its associated coefficients, particularly with regard to the 252Cf source. Thus, this study aims to quantify the dose rate received by the skin and its associated coefficients after contamination scenario. Utilizing the established MCNPX environment, the Equivalent dose rate and Absorbed dose, along with Skin contamination dose coefficient (SCDC), have been calculated within the skin tissue. Two methodologies, specifically Watt Fission distribution and the Doppler Effect, are proposed to analyze particle spectra within skin phantom, enabling the calculation of Equivalent dose rate. In accordance with ICRP recommendations regarding the optimal depth for assessing skin doses, the designated scoring volume within the skin is located between depths of 50–100 μm. This volume is tasked with evaluating the dose. The SCDC results were entirely consistent with previously published data from MCNPX, with statistical uncertainties of less than 15%, demonstrating the efficacy of the methodologies employed in this study. This research presents an innovative method for generating data related to skin contamination doses. The novel outcomes in the current research facilitate the assessment of skin dose contamination for the targeted radionuclides and radiotherapy purposes due to staff oversight and radiobiological effects.

Investigation of neutron/gamma-ray distribution in SiPM-based pulse shape discrimination using EJ-276 plastic scintillators

The distribution of neutrons and gamma-rays in mixed fields is a main interest in pulse shape discrimination (PSD) studies, and becomes more significant when it involves techniques such as machine learning which requires true data. One of the surest methods to label neutrons and gamma-rays is neutron Time-of-Flight. This can be applied to the identification of special nuclear materials and by extension to nuclear non-proliferation. Moreover, the integration of silicon photomultiplier (SiPM) technology can make the neutron Time-of-Flight system more compact. Two EJ-276 plastic scintillators, which are capable of PSD, were coupled to SiPMs. The actual portion of neutrons from a 252Cf source intruding the gamma-rays in PSD ratio distribution, labeled by neutron Time-of-Flight, was calculated and then compared with gamma-rejection ratio (GRR) for each threshold in SiPM-based pulse shape discrimination.

Development of indigenous pulse-shape discrimination algorithm for organic scintillation detectors

In this study, we present the development of an indigenous pulse shape discrimination (PSD) algorithm grounded in tail area and total area analysis, effectively obviating the necessity for specialized programmable hardware. Pulse data was collected utilizing a BC501 detector paired with a Pu-Be source and digitized in oscilloscope mode during experiments conducted at IIT-Kanpur. The algorithm developed in this work encompasses essential functions such as pulse normalization, shaping, peak identification and removal, and threshold determination. Notably, the algorithm offers the capability to derive neutron and γ-ray counts, generate scatter plots, and calculate the figure-of-Merit (FoM). It is essential to emphasize that the development of this indigenous Tail-to-Total Area (TtoT) PSD algorithm was a direct response to the unavailability of dedicated hardware for PSD in the experiment conducted at IIT-Kanpur. The introduction of the TtoT algorithm played a pivotal role in enabling offline PSD analysis at IIT-Kanpur, effectively overcoming infrastructure limitations. Importantly, the efficacy of our proposed algorithm was tested by applying it to pulse data from a distinct source-detector arrangement featuring a BC501A detector and an Cf-252 source. This comparative analysis also included the Charge Integration (CI) method and was conducted at IIT-Roorkee. The results obtained from our algorithm and the CI method exhibited a strong concurrence concerning the identification of neutrons and γ-rays, along with FoM metrics.

Experimental study of plastic scintillators array for compact fast neutron-gamma dual-modality imaging system

In a compact multi-radiation imaging system, the substitution of traditional charge-coupled devices with a plastic scintillators array coupled with silicon photomultiplier detectors significantly enhances the radiation tolerance of the detection and imaging module, thereby reducing the system footprint. We present a detailed structure of a compact fast neutron/gamma-ray dual-modality imaging system based on a plastic scintillators array and evaluate its performance in imaging various samples. The detector array consists of nine 14.7 × 14.7 × 100 mm3 EJ-276 plastic scintillators coupled with corresponding silicon photomultiplier units, arranged in a 3 × 3 grid. The usage of EJ-276 plastic scintillators allows for discrimination between neutron and gamma signals through pulse shape discrimination techniques, enabling data separation between neutrons and gamma rays. In this paper, we present the experimental procedure and results of fast neutron-gamma imaging for objects containing aluminum and polyethylene materials. Fast neutrons and gammas are delivered by an Cf-252 source. The experimental results demonstrate that the system achieves dual-modality grayscale imaging and has resolution for sample thickness. Reconstructed image contrast allows qualitative differentiation of sample materials, especially detecting low-Z materials encapsulated by high-Z materials.

A comparison of the neutron detection efficiency and response characteristics of two pixelated PSD-capable organic scintillator detectors with different photo-detection readout methods

We characterize the performance of two pixelated neutron detectors: a PMT-based array that utilizes Anger logic for pixel identification and a SiPM-based array that employs individual pixel readout. The SiPM-based array offers improved performance over the previously developed PMT-based detector both in terms of uniformity and neutron detection efficiency. Each detector array uses PSD-capable plastic scintillator as a detection medium. We describe the calibration and neutron efficiency measurement of both detectors using a 137Cs source for energy calibration and a 252Cf source for calibration of the neutron response. We find that the intrinsic neutron detection efficiency of the SiPM-based array is (30.2 ± 0.9)%, which is almost twice that of the PMT-based array, which we measure to be (16.9 ± 0.1)%.

The spectrum-averaged cross section investigation of 117Sn(n,n')117mSn and 67Zn(n,p)67Cu reactions

The spectrum averaged cross sections (SACS) in standard neutron field, e.g. 252Cf(s.f.), is a preferable tool for cross section evaluation and validation. A set of reaction measurements with high energy thresholds was previously performed. The presented work focuses on lower energy threshold reactions, namely on the inelastic scattering of the tin foil, more specifically the reaction 117Sn(n,n')117mSn, and the zinc foil reaction, namely 67Zn(n,p)67Cu. These reactions are of special interest due to their intermediate energy range, which is essential in classical reactor dosimetry and fast reactor dosimetry. The experiments were carried out in a standard neutron field formed by 252Cf(s.f.) source in Řež. The experimental results were compared with calculations using MCNP6.2, ENDF/B-VII.1 transport library, and ENDF/B-VIII.0 and IRDFF-II cross section data library. Additionally, the calculations using CEA code DARWIN/PEPIN2 using JEFF-3.0/A were executed. The obtained experimental SACS of previously measured reactions were in good agreement with the SACS calculated using the IRDFF-II library. Additionally, the calculational reaction rate of 67Zn(n,p)67Cu was in accordance with the experimental data in case of ENDF/B-VIII.0 nuclear data library. Moreover, the calculational results of 117Sn(n,n')117mSn obtained by DARWIN/PEPIN2 code (using JEFF-3.0/A nuclear data library) are closest to the experimental results.

Flexible, chemically bonded bismuth/tungsten‐based polyvinyl alcohol‐polyvinyl pyrrolidone composite for gamma and neutron shielding application

AbstractIn this paper, the gamma and neutron shielding behavior of chemically bonded high atomic number (Z), tungsten trioxide‐polyvinyl alcohol‐polyvinylpyrrolidone (PW), and bismuth oxide‐polyvinyl alcohol‐polyvinylpyrrolidone (PB) composites were reported at varying thickness. The PW and PB composites were prepared through the dissolution technique rather than the commonly used dispersion technique. The gamma and neutron attenuation characteristics of the composites were studied using 22Na (0.511 and 1.274 MeV energy) and 252Cf (0–10 MeV energy) sources, respectively. The half value layer of PW and PB composite was found to be 61.22 ± 0.914 mm and 48.29 ± 1.52 mm and the linear attenuation coefficient was found to be 0.115 and 0.144 cm−1 at 0.511 MeV, respectively. Further, the experimental gamma‐ray attenuation results were validated theoretically using Phy‐X software and found to be well in agreement with the experimental data. Both the composites also showed similar neutron shielding behavior when compared to the commercially used high‐density polyethylene (HDPE). The developed composites possess good mechanical properties as they did not show any cracks or wrinkles after undergoing 5400 bending cycles during the flex test, showing their high durability of flexibility. Thus, these composites are found to be suitable for shielding both, the gamma‐rays and neutrons.

Synthetic Utilization of 2H-Heptafluoropropane: Ionic 1,4-Addition to Electron-Deficient Carbon-Carbon Unsaturated Bonds.

We have found a novel method for introducing heptafluoro-2-propyl CF(CF3)2 groups into carbon-carbon unsaturated bonds via a nucleophilic reaction using 2H-heptafluoropropane as the source of CF(CF3)2 groups. The reaction involves the nucleophilic addition of a heptafluoro-2-propyl anion, generated by treating 2H-heptafluoropropane with a fluoride ion, to various electron-deficient unsaturated compounds. This allows the easy synthesis of various aliphatic compounds containing heptafluoro-2-propyl groups.

Simultaneous Gamma-Neutron Vision device: a portable and versatile tool for nuclear inspections

This work presents GN-Vision, a novel dual γ-ray and neutron imaging system, which aims at simultaneously obtaining information about the spatial origin of γ-ray and neutron sources. The proposed device is based on two position sensitive detection planes and exploits the Compton imaging technique for the imaging of γ-rays. In addition, spatial distributions of slow- and thermal-neutron sources (<100 eV) are reconstructed by using a passive neutron pin-hole collimator attached to the first detection plane. The proposed gamma-neutron imaging device could be of prime interest for nuclear safety and security applications. The two main advantages of this imaging system are its high efficiency and portability, making it well suited for nuclear applications were compactness and real-time imaging is important. This work presents the working principle and conceptual design of the GN-Vision system and explores, on the basis of Monte Carlo simulations, its simultaneous γ-ray and neutron detection and imaging capabilities for a realistic scenario where a 252Cf source is hidden in a neutron moderating container.

MCNP simulation and experimental study in situ low-grade copper analysis based on PGNAA

A prompt gamma neutron activation analysis (PGNAA) facility utilizing 252Cf source has been developed for in situ analysis of copper samples. Monte Carlo simulation is employed to determine optimal sizes for neutron moderator, gamma-ray shielding material, and thermal neutron absorber. Subsequently, based on the parameters optimized by MCNP, the PGNAA facility was constructed. Five sets of experimental samples containing low-grade copper concentration of 0 %, 0.5 %, 1 %, 1.5 % and 2 % are measured with the PGNAA facility. The results show that the minimum detectable concentration of copper is 0.218 %. The maximum relative deviation of copper is 8.53 %.

Alternatives to NIST Cf-252 Iirradiations for Transfer Calibration of S-32 Neutron Monitors

Gas-flow proportional counting systems are used by the Radiation Metrology Laboratory (RML) at Sandia National Laboratories for reactor fluence monitoring with the 32S(n,p)32P reaction. Calibration of these systems has traditionally been accomplished by fluence-transfer irradiations at the NIST 252Cf facility. Such calibrations have become increasingly difficult as the NIST 252Cf source decayed to unusable levels. To minimize the risk to the testing programs from an inability to properly calibrate these systems, the RML has developed two alternative calibration techniques: 1) development and implementation of certified 32P sources for activity calibrations and subsequent calculation of neutron fluence, and 2) direct counting of non-certified reactor-irradiated sulfur pellets by liquid scintillation counting to determine 32P activity for the subsequent calibration of gas-flow proportional counters. Preliminary comparisons show that the several calibration methods are capable of overall uncertainties within about 5 percent.

International comparison of measurements of neutron source emission rate (2016-2021) - CCRI(III)-K9.Cf.2016

Main textSection III (neutron measurements) of the Comité Consultatif des Rayonnements Ionisants, CCRI, conducted a comparison of primary measurements of the neutron emission rate of a 252Cf radionuclide source. A single 252Cf source was circulated to all participants between 2016 and 2020. Ten laboratories participated -CMI (Czech Republic), KRISS (Republic of Korea), IRD/LNMRI (Brazil), LNE-LNHB (France), NIM (China), NIST (USA), NMIJ (Japan), NPL (UK), NRC (Canada) and VNIIM (Russia) - with NPL making their measurements at the start and repeating them at the end of the exercise to verify the 250Cf content of the source. Each laboratory reported the emission rate into 4π sr together with a detailed uncertainty budget. All participants used the manganese bath technique except NMIJ who used a relative method based on measurements with a 3He detector in a graphite pile. VNIIM also made measurements using an associated particle technique. CMI, KRISS, LNE-LNHB, NIM, NPL and VNIIM also measured the anisotropy of the source although this did not formally form part of the comparison.To reach the main text of this paper, click on Final Report. Note that this text is that which appears in Appendix B of the BIPM key comparison database https://www.bipm.org/kcdb/.The final report has been peer-reviewed and approved for publication by the CCRI, according to the provisions of the CIPM Mutual Recognition Arrangement (CIPM MRA).

The contribution of the air and room scattered neutrons in the PTB calibration facility

Neutron detectors and dosimeters used for radiation protection purposes are commonly calibrated with radioisotope-based neutron sources in dedicated calibration facilities. Due to interactions of the neutrons emitted by the sources with components of the calibration room, the neutron fluence at the measuring point is a sum of several components. As the results of calibration procedures should ideally be independent of the environment in which the measurements are performed, the effects caused by neutrons which are not directly reaching the device under test need to be considered, i.e. the influence of scattered neutrons needs to be determined and corrections have to be made. The contribution of the neutrons scattered from the air and the room to the fluence and dose must be taken into consideration, especially in the case of small and medium sized calibration rooms. The scattered neutron component can be evaluated numerically provided a detailed model of the calibration facility is available. In this work, a realistic MCNP6 model of the calibration room at the Physikalisch-Technische Bundesanstalt (PTB) has been developed in order to evaluate carefully the contribution of scattered neutrons from the air, wall, and equipment in the irradiation room for 252Cf, 241Am–Be and D2O moderated 252Cf (with and without Cd-shielding) neutron sources at different reference points and with and without shadow objects. The calculation also considers new fission spectral data for the neutron reference field and recent cross section data libraries. The results show that the contribution of air out-scattered neutrons depends on the detector position and on the neutron source spectrum. This effect is particularly significant for the moderated 252Cf source. More than a third of the total scored neutrons come from the walls and ceiling and the total contributions of the scattered neutrons to the total fluence at the reference position 170 cm away from the source are 50.6 %, 46.6 %, 55.0 % and 52.5 % for 252Cf, 241Am–Be, D2O-252Cf–Cd and D2O-252Cf sources respectively.

Feasibility of using BeO rods as secondary neutron sources in the long-life fuel cycle high-temperature gas-cooled reactor

External sources of neutron provide stable and sufficient neutron for initial startup of a nuclear reactor. They also provide signals for neutron detectors to monitor the safety of reactor during shutdown. In the high temperature engineering test reactor (HTTR), 252Cf is used as the external neutron source. However, the 252Cf sources must be renewed every approximately 7 years because of its relatively short half-life of 2.6 years. The renewal of 252Cf sources requires a high cost and a very complicated procedure. This study investigated the feasibility of using BeO rods as the secondary neutron sources to avoid renewing the 252Cf neutron sources periodically. The BeO rods could exist in the reactor for a long time so that if the reactor operates long enough, the neutron flux at the wide-range monitoring detectors remains more than 10n.s−1.cm−2 even if the reactor is shutdown for as long as 5 years. The results of this study indicated that using BeO rods as the secondary neutron sources would be an attractive option for the future HTGR design with a long-life fuel cycle.

Neutron dosimetry with a pair of TLDs for the Elekta Precise medical linac and the evaluation of optimum moderator thickness for the conversion of fast to thermal neutrons

IntroductionIn this study, TLD 600 and TLD 700 pairs were used to measure the neutron dose of Elekta Precise medical linac. To this end, the optimum moderate thickness for the conversion of fast to thermal neutrons were evaluated. Materials and methods241Am-Be and 252Cf sources were simulated to calculate the optimum thicknesses of the moderator for the conversion of maximum fast neutrons (FN) into thermal neutrons (TN). Pair TLDs were used to measure F&TN doses for three different field sizes at four depths of the medical linac. ResultsThe maximum thickness of the moderator was optimized at 6 cm. The measurement results demonstrated that the TN dose increased with the expansion of field size and depth. The FN dose, which was converted TN, exhibits behaviors comparable to the TN due to its nature. ConclusionThis study presents the optimum thickness for the moderator to convert FN into TN and measure F&TN using TLDs.