252Cf Neutron Research Articles

A Large Volume (1m3) Spherical Proportional Counter for Atmospheric Neutron Measurements

A large volume (1m3) spherical proportional counter has been developed at CEA/Saclay, for low flux neutron measurements. Neutrons can be measured successfully, with high sensitivity, using 3He gas in the detector. The proton and tritium energy deposition in the drift gaseous volume, from the reaction 3He(n,p)3H, can provide the neutron spectra from thermal neutrons up to several MeV. The α resolution of the detector has been measured using 222Rn gas, in order to have the response of the total active volume of the detector. The three α peaks from 222Rn decay and 222Rn daughters decay have been measured with FWHM less then 2%, with gas mixture Ar 98% and CH4 2% at a pressure of 200 mbar. The neutron sensitivity has been tested inserting in the sphere 100mgr of 3He. The 0.765 MeV peak of p and 3H is well separated form the cosmic ray background after irradiation with 252Cf neutron source. The ground level atmospheric thermal neutrons have been also measured after several hours of exposure. The very low noise, the limited wall effect, the high resolution and the high sensitivity are some of the advantages of the detector.

Enhancing neutron detection capabilities of a water Cherenkov detector

In this work we show the neutron detection capabilities of a water Cherenkov detector (WCD) using pure water, as well as an aqueous solution of sodium chloride as the detection volume. The experiments were performed using a WCD with a single photomultiplier tube (PMT), and a 252Cf neutron source. We show that fast neutrons from a 252Cf source directly impinging the detector can be clearly detected and identified over the natural radiation background. We also present results from numerical simulations that describe the response of the WCD to neutrons of different energies. To do this, a detailed model for the WCD and of the neutron source spectra have been implemented. The implemented simulation code takes into account the interaction processes involved in the detection of neutrons using WCD, and support the experimental evidence introduced in this work. Being both the active volumes analyzed in this work (pure water) and the additive (sodium chloride), cheap, non-toxic and easily accessible materials, the results obtained are of interest for the development of large neutron detectors for different applications. Of particular importance are the detection of special nuclear materials, as well as those applications related with space weather phenomena. We conclude that WCD used as neutron detectors could be a replace or a complementary tool for standard neutron monitors based on 3He.

IMPROVEMENT OF GAMMA-RAY SUBTRACTION PROCEDURE FOR A CURRENT-MODE NEUTRON DETECTOR WITH A PAIR OF 6Li- AND 7Li-GLASS SCINTILLATORS.

A current-mode neutron detector with a pair of 6Li- and 7Li-glass scintillators has been developed to measure high-flux neutrons in a boron neutron capture therapy field. Neutrons are basically measured by subtracting gamma-ray component using current outputs from the 7Li-glass scintillator. In the present study, the difference in the gamma-ray sensitivity between the 6Li- and 7Li-glass scintillators and the neutron sensitivity for the 7Li-glass scintillator due to the 6Li contamination were also considered to improve the gamma-ray subtraction precision. The gamma-ray subtraction procedure was experimentally investigated in thermal neutron fields with 252Cf and 241Am-Be neutron sources, which have different gamma-ray intensities per unit neutron fluence. A linear relation between neutron fluence and current output was obtained for the neutron detector in the two types of thermal neutron fields with different gamma-ray intensities. It was found that the gamma-ray subtraction procedure is useful for current-mode neutron detectors.

Design and dosimetry of a facility to study health effects following exposures to fission neutrons at low dose rates for long durations

Purpose During extended missions into deep space, astronauts will be exposed to a complex radiation field that includes high linear energy transfer (LET) radiation from high energy, heavy ions (HZE particles) at low dose rates of about 0.5 mGy/d for long durations. About 20% of the dose is delivered by ions with LET greater than 10 keV/µm. There are sparse empirical data in any species for carcinogenic effects from whole-body exposures to external sources of mixed or high LET radiation at this level of dose rates. For the induction of solid tumors, acute exposures to HZE ions have been shown to be substantially more effective per unit dose than low LET exposures associated with photons. To determine the health effects of high LET radiation at space-relevant dose rates on experimental animals, we developed a vivarium in which rodents could be irradiated with Californium (252Cf) neutrons for protracted periods of time. Materials and methods The neutron source is a panoramic irradiator containing 252Cf located in a concrete shielded vault with a footprint of 53 m2. The vault can accommodate sufficient caging to simultaneously irradiate 900 mice and 60 rats for durations up to 400 d at a dose rate of 1 mGy/d and is approved for extended animal husbandry. Results The mixed field fluence is a combination of neutrons and photons emitted directly from the source and scattered particles from the concrete walls and floor. Mixed field dosimetry was performed using a miniature GM counter and CaF2:Dy thermoluminescent dosimeters (TLD) for photons and tissue-equivalent proportional counters (TEPC) for neutrons. TEPC data provided macroscopic dose rates as well as measurements of radiation quality based on lineal energy, y, and LET. The instantaneous dose rate from the source decreases with a half-life of 2.6 years. The exposure time is adjusted weekly to yield a total dose 1 mGy/d. The photon contribution is 20% of the total dose. The uncertainty in the delivered dose is estimated to be ±20% taking into account spatial variations in the room and random position of mice in each cage. The dose averaged LET for the charged particle recoil nuclei is 68 keV/µ. Conclusions We have developed a facility to perform high LET studies in mice and rats at space relevant dose rates and career-relevant doses using neutrons emitted from the spontaneous fission of 252Cf.

Fabrication and performance of single crystal diamond neutron and gamma ray detectors

Two high performance single crystal diamond (SCD) radiation detectors for neutrons and gamma ray have been fabricated. One was coated with a 5 μm 6LiF (95% enriched in 6Li) layer in order to detect the neutrons and the other one is without any conversion layer for comparison. The detectors are irradiated with the 241Am-Be and 252Cf neutron sources and 152Eu gamma source to test their performance. Under neutron sources, the neutron signal peak was clearly observed in the detector with the 6LiF conversion layer. Meanwhile, both detectors are responsive to the gamma rays at the low energy. The 152Eu gamma source was specially used to verify the gamma ray detection ability of the fabricated SCD detectors. The results show that the detector’s count rate and the gamma peak center position in the PHS depend on the incident gamma spectra critically. Combined, these results demonstrate that the design and fabrication process of the detectors are feasible.

Conceptual design and first results for a neutron detector with interaction localization capabilities

A new high-precision detector for studying neutrons from beta-delayed neutron emission and direct reaction studies is proposed. The Neutron dEtector with Xn Tracking (NEXT) array is designed to maintain high intrinsic neutron detection efficiency while reducing uncertainties in neutron energy measurements. A single NEXT module is composed of thin segments of plastic scintillator, each optically separated, capable of neutron-gamma discrimination. Each segmented module is coupled to position sensitive photodetectors enabling the high-precision determination of neutron time of arrival and interaction position within the active volume. A design study has been conducted based on simulations and experimental tests leading to the construction of prototype units. First results from measurements using a $^{252}$Cf neutron source and accelerator-produced monoenergetic neutrons are presented.

Study on the Relationship between Seed Absorbed Dose and Seed Composition of 252Cf Neutron Source Irradiated Bean Seed

This study aims to further identify the biological effects of neutron-irradiated plants and provides insights into the mutation breeding of such plants. In this study, the neutron irradiation device designed by our institute was used to analyze the relationship between the seed components in different legume crops and their neutron absorption dose rate, fission gamma absorption dose rate, and induced gamma absorption dose rate. The results show that the effect sizes of the components on the neutron absorbed dose rate are as follows: ash > fat > moisture > carbohydrate > protein. The effect sizes of the components on the absorbed dose rate of fission gamma are as follows: ash > moisture > fat > carbohydrate > protein. There is a positive correlation between fission gamma absorbed dose rate and the weight of ash, water and fat, while a negative correlation with carbohydrate and protein. However, the linear relationship between each component and the absorbed dose rate of induced gamma is not significant, this needs to be identified by further researches. Based on the results of the present study, we conclude that the neutron absorbed dose can be calculated without taking into account the fat composition of bean crop seeds (except for soybean seeds) in the process of mutation breeding induced by radiation. In special cases where the accuracy requirement of the dose rate is not high, it is possible to use protein instead of legume crop seeds for neutron absorption dose calculations.

Performance test of liquid-moderator-based neutron spectrometer for BNCT: Mono-energetic and spontaneous fission spectrum neutron measurements

Abstract Boron neutron capture therapy (BNCT) is an effective cancer therapy that uses neutron irradiation; it is necessary to evaluate the neutron spectrum in the treatment field as precisely as possible. To conduct such evaluations, we developed a new device, called a liquid-moderator-based neutron spectrometer. In the present study, the performance of the developed spectrometer was tested by measuring a mono-energetic neutron field and a fast neutron field generated by a 252Cf neutron source. Finally, we revealed that the response function of the spectrometer could be appropriately calculated and confirmed the capability of the device to measure the neutron energy spectra between 1 keV and 10 MeV.

Application of 252Cf neutron source for precise nuclear data experiments

Purpose of this paper is to provide extensive information helpful for anyone performing any experiment involving 252Cf neutron source and aiming for high precision experiments. The paper summarizes basic characteristics and fields of study using 252Cf neutron source. We show the basic characteristics of our source, precise geometry in MCNP6, isotopic content, distribution of the source in the encapsulation and possible use of encapsulation for 27Al(n,2n)26Al reaction estimation and the way of handling of the 252Cf neutron source in our laboratory. Furthermore, we prove that our source is volumetric, i.e. non-point and can be considered as an isotropic in our experimental settings. Influence of the palladium matrix density on the reaction rates is also investigated. Concerning nuclear data, we are measuring fast neutron leakage spectra in various benchmark sets including spheres of different materials. Our work extends benchmark sets to possible use of cubes, in our case graphite cube of side 30 cm. The neutron spectra are measured by stilbene scintilation detector in the energy range of 1–10 MeV in the steps of 100 keV in the distance of 100 cm from the centre of the cube. There is no significant difference between measurements performed using cube or sphere except lower cost for the cube production. The agreement between calculation using ENDF/B-VII.1 library and experimental data is not satisfactory especially in the regions around 3.5 MeV and 8 MeV. The agreement within 4% of calculation to experiment ratio for neutron leakage fluxes of pure 252Cf source validates our MCNP model for studying reaction rates and leakage fluxes in the region above 1 MeV.

Measurement of fast neutrons with 238U-coated fission chamber

A fission chamber with two layers of 238U fissile materials has been developed for the measurements of fast neutrons. The developed fission chamber has been tested by 60Co gamma source and 252Cf neutron source. The results have indicated that it is very easy to cut off the gamma rays from the neutrons by using a discriminator, when the fission chamber works in the pulse mode. The detector works well, even if the static air is used as the filling gas. To discuss the application of the detector, Monte Carlo simulations have been performed by considering four 238U-coated fission chambers located outside the spallation target for the fast neutron measurement from spallation reaction. The simulation results have indicated that the detector can be used to measure the fast neutron flux and neutron flux ratio of four fission chambers. Furthermore, a 238U-coated fission chamber can be used within a reactor for longer time, in comparison with a 235U-coated fission chamber.

ESTIMATION OF THE ANISOTROPY EMISSION OF LPN/CIEMAT NEUTRON SOURCES: EFFECT OF HEAVY CAPSULE HOLDERS.

In order to estimate the anisotropy emission of 241Am-Be and 252Cf neutron sources from the Spanish Neutron Standards Laboratory (LPN/CIEMAT) detailed models of sources capsules and capsule holders were designed with the MCNPX code. Simulations of the sources inside the capsules without the capsule holders were done to validate the MC model by comparison with experimental results provided by other authors. After that, the capsule holders were incorporated to the simulation. In general, a good agreement has been found between measurements and our calculations. Results show the additional encapsulations have significant influence on anisotropy factors, energy spectra and dose rates.

Investigation of 127I(n,2n)126I and 23Na(n,2n)22Na Reactions Using 252Cf Neutron Source

This work deals with 23Na(n,2n)22Na and 127I(n,2n)126I reactions in the 252Cf spontaneous fission neutron source. 252Cf neutron source with approximate emission of 6·× 108 n/s was employed for the irradiation of sodium iodide. The spectrum-averaged cross sections (SACS) were then inferred from experimentally determined reaction rates and compared with calculations in MCNP6 using various nuclear data libraries. The experimental reaction rates were derived from the net peak areas (NPAs) measured using the high purity germanium spectroscopy. The measured SACS for the 23Na(n,2n)22Na reaction in the 252Cf spectrum was determined as equal to (8.98±0.32)·× 10−6 b. The resulting SACS for the 127I(n,2n)126I reaction in the 252Cf spectrum was derived as (2.044±0.0072)·× 10−3 b. These experimental data can be used for nuclear data libraries validation and to specify high energy tail of the 252Cf neutron spectrum.

Feasibility study on neutron energy spectrum measurement utilizing prompt gamma-rays

Abstract A new approach based on Prompt Gamma-ray Neutron Activation Analysis (PGNAA) was proposed for measuring neutron energy spectra . A system, which consisted of a high purity germanium (HPGe) detector and a multi-layer sample, was used to validate the method. The composite layers of the sample comprised of lead, polymethyl methacrylate, sodium chloride powder and borated polyethylene . Prompt gamma-rays emitted from Pb, H, O, C, Cl and B were selected to determine the incident neutron energy spectrum. The response function for the production of gamma rays from these elements by monoenergetic neutrons was obtained through Monte Carlo simulations. Then the neutron fields of Cf-252 neutron source and D-T neutron generator were investigated with this approach. With the measured prompt gamma-rays and the response matrix, the neutron spectra were obtained through the PSO algorithm. The unfolded results have a good agreement with the simulation spectrum.

Comparative neutron detection efficiency in He-3 proportional counters and liquid scintillators

Abstract We present new results on the neutron detection efficiency of detection systems based on He-3 proportional counters compared to organic liquid scintillators. The experiments were conducted using a well-characterized Cf-252 neutron source, measured bare, and shielded with slabs of polyethylene. The results show that the intrinsic neutron detection efficiency for the bare source is approximately 25% for the He-3 system, and 67% for the liquid scintillator system. When polyethylene sheets were used to shield the source in 2.54 cm increments, the neutron detection efficiency of both systems decreased exponentially to approximately 10% for the case of 25.4 cm. Experimental results were simulated using MCNPX-PoliMi and good agreement was found between measurement and simulation results for all the cases considered here.

Neutron detection capabilities of Water Cherenkov Detectors

Abstract In this work we show the neutron detection capabilities of a water Cherenkov detector (WCD). The experiments presented here were performed using a simple WCD with a single photomultiplier tube (PMT) and a 252 Cf neutron source . We compared the use of pure water and water with non contaminant additive as the detection volume. We show that fast neutrons from the 252Cf source can be detected over the flux of atmospheric particles background. Our first estimation for the neutron detection efficiency is at the level of (19)% for pure water and (44)% for the water with the additive. We also present the simulation of the response of the WCD to neutrons using a simulated 252Cf source. We implemented a detailed model of the WCD and of the neutron source spectra using Geant 4. The results of our simulations show the detailed mechanism for the detection of neutrons using WCD and support the experimental evidences presented. Since both active volumes studied, H 2 O pure and with additive, are cheap, non-toxic and easily accessible materials, the results obtained are of interest for the development of large neutron detectors for different applications. Of special importance are those related with space weather phenomena as well as those for the detection of special nuclear materials. We conclude that WCD used as neutron detectors can be a complementary tool for standard neutron monitors based on 3 He.

Development of a quality assurance process for the SoLid experiment

The SoLid experiment has been designed to search for an oscillation pattern induced by a light sterile neutrino state, utilising the BR2 reactor of SCK⋅CEN, in Belgium. The detector leverages a new hybrid technology, utilising two distinct scintillators in a cubic array, creating a highly segmented detector volume. A combination of 5 cm cubic polyvinyltoluene cells, with 6LiF:ZnS(Ag) sheets on two faces of each cube, facilitate reconstruction of the neutrino signals. Whilst the high granularity provides a powerful toolset to discriminate backgrounds; by itself the segmentation also represents a challenge in terms of homogeneity and calibration, for a consistent detector response. The search for this light sterile neutrino implies a sensitivity to distortions of around \U0001d4aa(10)% in the energy spectrum of reactor ν̄e. Hence, a very good neutron detection efficiency, light yield and homogeneous detector response are critical for data validation. The minimal requirements for the SoLid physics program are a light yield and a neutron detection efficiency larger than 40 PA/MeV/cube and 50% respectively. In order to guarantee these minimal requirements, the collaboration developed a rigorous quality assurance process for all 12800 cubic cells of the detector. To carry out the quality assurance process, an automated calibration system called CALIPSO was designed and constructed. CALIPSO provides precise, automatic placement of radioactive sources in front of each cube of a given detector plane (16×16 cubes). A combination of 22Na, 252Cf and AmBe gamma and neutron sources were used by CALIPSO during the quality assurance process. Initially, the scanning identified defective components allowing for repair during initial construction of the SoLid detector. Secondly, a full analysis of the calibration data revealed initial estimations for the light yield of over 60 PA/MeV and neutron reconstruction efficiency of 68%, validating the SoLid physics requirements.

Model design of a compact delayed gamma-ray moderator system using 252Cf for safeguards verification measurements

Delayed gamma-ray spectroscopy (DGS) is an active-interrogation nondestructive assay technique that can be used to determine the composition of a nuclear material sample by comparing the ratios of fission product gamma-ray peak intensities. However, high-radioactivity nuclear material (HRNM) contains long-lived fission products that can overwhelm a detector and shielding must be used to reduce the count rate, while minimally affecting those gamma rays from short-lived fission products with energy above 3-MeV. To compensate for the signal loss through the shielding, low-energy neutrons are required to induce more fission events from the high thermal cross-section of fissile nuclides. To improve practical safeguards DGS capabilities, we are developing a compact interrogation system to moderate ∼ 2-MeV neutrons that are easier to moderate than 14-MeV neutrons from standard deuterium-tritium generators. This work describes the optimization of an ideal moderator system for a 252Cf neutron point source that results in a neutron fluence of 25.9×10−4 cm−2nsource−1 passing through the sample space with ≳ 70% of those below 1-eV. Modifications for practical fabrication resulted in ⩽20% reductions of the flux compared to the optimized ideal design. Finally, evaluations made of HRNM DGS signals and backgrounds conclude that a 252Cf source intensity of 8.9×107 neutrons per second is required for a single-pass interrogation within this ideal moderator system. However, this can be as low as 3.6×106 neutrons per second using smaller samples that require less shielding.

In-depth evaluation of 252Cf absolute calibration by passive neutron correlation counting method

252Cf neutron sources are particularly convenient, and therefore widely used, to characterize and calibrate many kinds of neutron detectors for both scientific and technological applications. An emerging interest is the use of 252Cf to calibrate neutron correlation counters for the verification and assay of Pu-items as part of the international effort to safeguard nuclear materials. Passive neutron correlation counting (PNCC) methods, combined with knowledge of the fission neutron multiplicity distribution, can be used as an absolute technique to determine the neutron emission rate of 252Cf sources. This means that common equipment used by the safeguards community can, in principle, be used to self-certify, to high accuracy, the strength of physically small 252Cf sealed sources for use in calibration. This is the first step in using 252Cf in place of extensive Pu reference materials for experimental calibration by the safeguards community. The theoretical foundation is already clear but the present work presents the first systematic attempt at demonstrating the practical utility of the concept across a variety of typical instruments. We compare four emission rate estimators for four detector configurations on a common source previously certified by the MnSO4-bath technique. We find that the estimator based on the singles to doubles rate ratio provides consistent and precise results across all of the instruments and that these agree with the MnSO4-bath technique within combined uncertainties. This success suggests that the absolute PNCC method could become a technically defensible and commonly used, method capable of sub 1% accuracy for routine calibration work.

Poor Understanding of Radiation Profiles in Deep Space Causes Inaccurate Findings and Misleading Conclusions.

The radiation environment in deep space, where astronauts are behind the shelter provided by the Earth’s magnetosphere, is a major health concern. Galactic cosmic rays (GCR) and solar particle events (SPE) are two basic sources of space radiation in the solar system. The health risks of exposure to high levels of space radiation can be observed either as acute and delayed effects. Zhang et al. in their recently published paper entitled “γ-H2AX responds to DNA damage induced by long-term exposure to combined low-dose-rate neutron and γ-ray radiation” have addressed the effects of different cumulative radiation doses on peripheral blood cell, subsets of T cells of peripheral blood lymphocytes and DNA damage repair. These researchers exposed animals to low dose rate 60Co-rays at 0.0167 Gy h−1for 2 h/d and 252Cf neutrons at 0.028 mGy h−1for 20 h/d for 15, 30, or 60 consecutive days. They reported that the mRNA of H2AX increased significantly, and showed a positive correlation with dose. Despite strengths, this paper has several shortcomings such as poor definition of low dose radiation as well as space and reactor radiation environments. Another shortcoming of this paper comes from this point that blood cell studies do not represent the biological effects of ionizing radiation on the total body. Moreover, the effects of the human immune system and DNA repair mechanisms are not included in the study. The role of pre-exposures and induction of adaptive response phenomena in decreasing the risk of radiation in deep space missions are also ignored.

Research on design of seed irradiation device based on 252Cf fission neutron source

252Cf neutron source has been widely used in biological effects of neutron radiation, neutron activation analysis and nuclear data acquisition. A seed irradiation device based on 252Cf fission neutron source was designed in this paper to study mutation breeding induced by neutron irradiation and biological effects of radiation. Based on the Monte Carlo method, neutron absorbed dose of pea seed samples in the designed irradiation region was simulated and calculated by MCNP. Results demonstrated that the relative standard deviation of the absorbed dose of pea seed samples in the three-layer partition is controlled lower than 10%, indicating the high dose uniformity in designed irradiation region. The experiment verified that the design of shield in this device meet the irradiation protection requirements.