Neutron dosimetry using activation of thermoluminescent CaSO4

Sulfur activation in calcium sulfate doped with dysprosium (CaSO₄:Dy) thermoluminescent powder, which is bound in pure sulfur, has been used to measure the fast neutron dose at the tangential beam port of a Triga Mark III reactor. After a post-irradiation time of 3 d, the dosimeters were annealed at 600°C for 30 min in order to erase all the thermoluminescence acquired during the irradiation. The dosimeters were then stored to allow self-irradiation by betas from ³²P produced by sulfur activation. The thermoluminescent signal accumulated during a post-irradiation time of 20 d due to a neutron fluence of 2.2 × 10¹¹ n/cm² was equivalent to an absorbed dose of 10 mGy of ⁶⁰Co gamma rays. The thermoluminescence as a function of fast neutron dose fitted to a straight line on a log-log scale from 1 Gy to 10⁴Gy.     

Development of a technique using MCNPX code for determination of nitrogen content of explosive materials using prompt gamma neutron activation analysis method

Nuclear-based explosive detection methods can detect explosives by identifying their elemental components, especially nitrogen. Thermal neutron capture reactions have been used for detecting prompt gamma 10.8 MeV following radioactive neutron capture by ¹⁴N nuclei. We aimed to study the feasibility of using field-portable prompt gamma neutron activation analysis (PGNAA) along with improved nuclear equipment to detect and identify explosives, illicit substances or landmines. A ²⁵²Cf radio-isotopic source was embedded in a cylinder made of high-density polyethylene (HDPE) and the cylinder was then placed in another cylindrical container filled with water. Measurements were performed on high nitrogen content compounds such as melamine (C₃H₆N₆). Melamine powder in a HDPE bottle was placed underneath the vessel containing water and the neutron source. Gamma rays were detected using two NaI(Tl) crystals. The results were simulated with MCNP4c code calculations. The theoretical calculations and experimental measurements were in good agreement indicating that this method can be used for detection of explosives and illicit drugs.     

Beryllium neutron activation detector for pulsed DD fusion sources

A compact fast neutron detector based on beryllium activation has been developed to perform accurate neutron fluence measurements on pulsed DD fusion sources. It is especially well suited to moderate repetition-rate (<0.2 Hz) devices, such as the plasma focus or Z-pinch. The detector comprises a beryllium metal sheet sandwiched between two large-area xenon-filled proportional counters. A methodology for calculating the absolute response function of the detector using a “first principles” approach is described. This calibration methodology is based on the ⁹Be(n,α)⁶He cross-section, energy calibration of the proportional counters, and numerical simulations of neutron interactions and beta-particle paths using MCNP5. The response function R(E_n) is determined over the neutron energy range 2-4 MeV. The count rate capability of the detector has been studied and the corrections required for high neutron fluence measurements are discussed. For pulsed DD neutron fluencies >3×10⁴ cm⁻², the statistical uncertainty in the fluence measurement is better than 1%. A small plasma focus device has been employed as a pulsed neutron source to test two of these new detectors, and their responses are found to be practically identical. Also the level of interfering activation is found to be sufficiently low as to be negligible.     

High-sensitivity activation neutron detector

We demonstrate that neutron fluences of nearly 1000 cm^&#x2212;2 can be measured when the fonctions of an activation detector and a device for recording induced activity are combined in one object i.e., a NaI(Tl) scintillation crystal. The beta radiation of the product from the¹²⁷I(n, &#x03B3;)¹²⁸ reaction is detected with close to 100% efficiency by the crystal. The detector can be used in work with low-intensity sources of neutron fields such as neutron generators, radiosotope sources, and TOKAMAK devices.Translated from Izmeritel'naya Tekhnika, No. 2, p. 56, February, 1995.     

Status of the cyric neutron TOF facilities upgrade

We describe progress in the Tohoku neutron time-of-flight (TOF) facilities improved to facilitate high resolution measurements for neutrons in the range 10 ≤ E_n ≲ 60 MeV. Efforts have been concentrated on completing the neutron detection system consisting of twelve neutron detectors, in which 23 1 of NE213 liquid scintillator are encapsulated, and a CAMAC-based data acquisition system. A systematic study of the neutron detection efficiency has been performed by Monte Carlo calculation for monochromatic neutrons with E_n ≲ 34 MeV. Results have been tested by counting neutrons from the ⁷Li(p, n)⁷Be reaction and comparing the yield with the absolute neutron fluence determined by activation.     

Dose measuring system for boron neutron capture therapy

Dose measuring systems for boron neutron capture therapy (BNCT) of brain tumors are presented. The systems are a real-time monitoring system, an integral measuring system and a ¹⁰B concentration measuring system. The real-time monitoring with a small PN junction silicon detector made it possible to simultaneously measure the thermal neutron flux and the gamma dose rate in a patient during neutron therapy. Another monitoring of dose equivalents of thermal neutrons and gamma rays was performed with a BGO scintillation detector connected to an optical fiber. The accurate neutron fluence and gamma dose were determined with the integral measurements of the foil activation method and thermoluminescent dosimeters (TLDs) after irradiation. Kerma doses of thermal neutrons and gamma-rays were also measured with the TLD at the same time. Preliminary measurements of ¹⁰B concentration in tissue and blood of a patient were carried out by prompt gamma-ray spectroscopy.     

Neutron-irradiation effects on track etching and optical characteristics of CR-39 (DOP) nuclear track detector

The effects of thermal neutron-irradiation in the neutron fluence range of (1.77–7.08) 10¹¹ n/cm², on the etching and optical characteristics of diethylene glycol bis allyl carbonate (dioctyl phthalate doped), CR-39 (DOP) nuclear track detector have been studied using etching and UV–visible spectroscopic techniques. The bulk etch rates determined at different fluences were found to increase with an increase in neutron fluence up to 3.54 × 10¹¹ n/cm², and then decrease at higher neutron fluence. The optical absorption spectra in the wavelength range of 200–800 nm were also recorded for the unirradiated and neutron irradiated samples in the above fluence range. The optical energy gaps (E _g) were determined by the shift in optical absorption edges as observed by UV–visible spectra of the neutron irradiated sample, using Tauc’s expression. The UV–visible spectra results were further supported by determining the activation energies for bulk etching.     

Prospective Large-Scale Transmutation of Fission Iodine

The effective neutron capture cross section of ¹²⁹I was determined at irradiation of 11.3±0.55 mg of KI (83.5% ¹²⁹I and 16.5% ¹²⁷I) for 6 h in a V10 channel of the beryllium reflector of VVR-M water-cooled water-moderated reactor (PINF) at average power of 12 MW. The neutron spectrum was measured by the activation procedure. For this purpose a series of thermal, resonance, and threshold detectors and a weighed portion of iodine were irradiated. The neutron spectrum was restored from the -ray spectra of these samples by the best fit procedure. The neutron capture cross section of ¹²⁹I, averaged over the neutron spectrum in the beryllium reflector of VVR-M reactor, was 17.8±3.2 b. The activities of products formed from a model ¹²⁷I portion of iodine were irradiated. The neutron spectrum was restored from the -ray spectra of these samplesby the best fit procedure. The neutron capture cross section of ¹²⁹I, averaged over the neutron spectrum in theberyllium reflector of VVR-M reactor, was 17.8±3.2 b. The activities of products formed from a model 127Isample irradiated in a loop of VVR-M reactor under steady-state conditions were calculated.     

Dosimetry of fast neutron beams using CaSO4:Dy (TLD-900) pellets

This paper describes the use of commercially avialable CaSO₄:Dy (TLD-900) pellets for the measurement of absorbed doses of fast neutrons and gamma rays in mixed fields with one single detector. The gamma ray absorbed doses could be estimated by recording the thermoluminiscence (TL) induced during the neutron beam irradiations, whereas the fast neutron absorbed doses were measured by employing a post-irradiation TL accumulation due to activation of sulphur by the threshold nuclear reaction ³²S(n, p)³²P in CaSO₄:Dy.     

A CdTe detector with a Gd converter for thermal neutron detection

A CdTe detector with a Gd converter has been developed and investigated as a neutron detector for neutron imaging. The fabricated Gd/CdTe detector with the 25 μm thick Gd was designed on the basis of simulation results of thermal neutron detection efficiency and spatial resolution. The energy resolution of the Gd/CdTe detector is less than 4 keV, which is enough to discriminate neutron capture gamma rays from background gamma emission. The Gd/CdTe detector shows the detection of neutron capture gamma ray emission in the ¹⁵⁵Gd(n, γ)¹⁵⁶Gd, ¹⁵⁷Gd(n, γ)¹⁵⁸Gd and ¹¹³Cd(n, γ)¹¹⁴Cd reactions and characteristic X-ray emissions due to conversion-electrons generated inside the Gd film. The observed efficient thermal neutron detection with the Gd/CdTe detector shows its promise in neutron radiography application.     

Absolute neutron emission rate measurement of a 252Cf source by the manganous sulfate bath method

A standard program for the absolute neutron emission rate measurement of a ²⁵²Cf source by the manganous sulfate bath method has been completed for the standardization of neutron source measurements. The MnSO₄ circulation system enables the ⁵⁶Mn activity to be measured during activation of the bath by the neutron source. The neutron capture cross section ratio of hydrogen to manganese has been determined to be 0.02506 by varying the manganese concentration in the MnSO₄ solution. From the graphical analysis of a straight line fitted to the six sets of data, the observed neutron emission rate was 2.544 × 10⁷ n/s for the ²⁵²Cf source at the reference date, January 1, 1987.     

Yields of activation products in 226Ra irradiation in the high-flux SM reactor

A target containing 3.16 mg of ²²⁶RaCO₃ was irradiated in the neutron trap of the SM reactor for 25 effective days at a thermal neutron flux of 1.5 × 10¹⁵ cm^?2 s^?1. After the irradiation and storage for 17 days, the irradiated material was dissolved and the Ra activation products were chemically isolated. The yields of ²²⁹Th, ²³⁰Th, ²²⁷Ac, and ²²⁸Th were determined by ??- and ??-ray spectroscopy and by mass spectrometry. Formation of significant amounts of ²²⁸Ra, not predicted by the calculations, was revealed. It was suggested that the short-lived isotope ²²⁷Ra has a high neutron capture cross section [??_eff(²²⁷Ra) ?? 1.5 × 10³ b].     

Simulation of corrosion processes in the closed system steel-water coolant

Methodology based on principles of neutron activation analysis is developed for simulation and study of steel corrosion in water coolants at thermal and nuclear power plants. Radionuclide generation directly in the steel samples at thermal neutron fluence not exceeding 10¹⁹ n cm^?2 and the corresponding calculated exposure time ensures identity of the chemical behavior of the radionuclides formed in the irradiated samples and of their stable isotopes and, at the same time, does not involve irreversible changes in the steel matrix structure. Owing to high sensitivity of methods based on activation analysis, the corrosion rate of pearlite steels in the initial stages of their contact with deionized (desalinated) water was determined reliably. The methods developed also allow monitoring of mass exchange processes in the steel-aqueous medium systems, simulating intracircuit processes. An additional advantage of the methods developed is the possibility of simultaneous monitoring of the behavior of not only iron (by ⁵⁹Fe and ⁵⁵Fe radionuclides), but also alloying elements (by ⁵¹Cr, ⁵⁴Mn, ⁶⁰Co, etc.).     

Nondestructive evaluation of jades by means of 14 MeV neutron activation

Fourteen MeV neutron activation of jades was used to test the authenticity of jades: to ascertain whether they are genuine jadeite and whether their greenish colour is genuine, both of which are important criteria for the high market value of jadeite. For given activation and measurement conditions of the jades, the gamma-ray spectrum was observed to change as a function of the type of the jades; the counting ratio of the photopeaks from the two most prominent constituent elements varies greatly as a function of the type of the jades. The ratio of counts for the Compton edges of the 1.779 MeV gamma-ray of²⁸Al, from²⁸Si(n,p)²⁸Al reaction, and the 1.434 MeV gamma-ray of⁵²V, from⁵²Cr(n,p)⁵²V reaction, varied as a function of the type of the jades and also as a function of the color. These results can be applied for a fast and nondestructive evaluation of jades.     

New storage phosphors for thermal neutron detection

The main requirements on neutron detectors based on storage phosphors (SPs) are formulated. It is shown that commercial gadolinium-containing neutron SPs (Gd-SPs) based on mixtures of BaFBr:Eu²⁺ with Gd₂O₃ (a neutron converter) do not possess optimum neutron characteristics because of a high sensitivity to gamma radiation. A comparative study of photostimulated luminescence (PSL) kinetics was performed for samples of halogen-containing borates Sr₂B₅O₉Br(Cl) and Ca₂B₅O₉Br(Cl) activated with Ce³⁺ ions and of the commercial Gd-SPs upon neutron and gamma irradiation. The results suggest that the PSL characteristics of the above borates with 100% content of ¹⁰B must be similar to those of Gd-SPs. Use of the halogen borates as SPs offers an advantage to Gd-SPs, because the former neutron detectors are less sensitive to gamma radiation.     

Characterisation of an accelerator-based neutron source for BNCT versus beam energy

Neutron capture in ¹⁰B produces energetic alpha particles that have a high linear energy transfer in tissue. This results in higher cell killing and a higher relative biological effectiveness compared to photons. Using suitably designed boron compounds which preferentially localize in cancerous cells instead of healthy tissues, boron neutron capture therapy (BNCT) has the potential of providing a higher tumor cure rate within minimal toxicity to normal tissues. This clinical approach requires a thermal neutron source, generally a nuclear reactor, with a fluence rate sufficient to deliver tumorcidal doses within a reasonable treatment time (minutes). Thermal neutrons do not penetrate deeply in tissue, therefore BNCT is limited to lesions which are either superficial or otherwise accessible. In this work, we investigate the feasibility of an accelerator-based thermal neutron source for the BNCT of skin melanomas. The source was designed via MCNP Monte Carlo simulations of the thermalization of a fast neutron beam, generated by 7 MeV deuterons impinging on a thick target of beryllium. The neutron field was characterized at several deuteron energies (3.0–6.5 MeV) in an experimental structure installed at the Van De Graaff accelerator of the Laboratori Nazionali di Legnaro, in Italy. Thermal and epithermal neutron fluences were measured with activation techniques and fast neutron spectra were determined with superheated drop detectors (SDD). These neutron spectrometry and dosimetry studies indicated that the fast neutron dose is unacceptably high in the current design. Modifications to the current design to overcome this problem are presented.     

Two types of multi-moderator neutron spectrometers: Gamma-ray insensitive type and high-efficiency type

Two types of multi-moderator neutron spectrometers were developed; one is a gamma-ray insensitive type, and the other is a high-efficiency type. An indium activation detector is loaded in the former spectrometer, which can measure the photon-dominant pulsed neutron field such as in the primary photon beam of a high-energy medical electron accelerator. The latter, in which a ³He counter is loaded, is so sensitive that it can measure leakage neutrons from a well shielded facility or even the skyshine neutrons. The response functions of the spectrometers were measured by thermal and mono-energetic neutron standard fields, and were also calculated by the one-dimensional discrete ordinates transport code, ANISN. The measured and calculated responses showed generally good agreement. A benchmark measurement of ²⁵²Cf fission neutrons by using these two spectrometers agreed well with the calculated spectrum. The spectrometers were used in the measurements of neutrons produced by a medical electron accelerator and of skyshine neutrons from an intense 14 MeV neutron source facility.     

A multiwire chamber system for measurements of neutron spectra

A new type of detector for measuring neutron flux and energy over a wide range of angles and energies is being developed. Measurements of neutron elastic and inelastic scattering as well as neutron energy continua are possible. Time-of-flight is not used for measuring outgoing neutron energy, and so for continuum measurements this system has some distinct advantages over conventional neutron detectors. Neutron energy measurement is carried out by measuring the energy and angle of the recoil proton produced by the neutron in a CH₂ converter. Spectra from ⁷Li(p, n)⁷Be at 62 MeV and ⁴⁰Ca(n, n′χ) at 65 MeV are presented.     

Use of neutron activation analysis for evaluating the efficiency of water treatment to remove heavy metal ions by zone freezing

The efficiency of zone freezing for potable water treatment to remove inorganic impurities was examined. The content of impurities in model solutions before and after zone freezing was evaluated by neutron activation analysis with the formation of ⁵⁶Mn, ^116m In, and ¹⁹⁸Au radionuclides. The zone freezing procedure is the most efficient at low ice front velocities (no greater than 0.15 cm h⁻¹).     

Neutron sensitivity of vanadium self powered neutron detectors

The neutron sensitivity of vanadium self powered neutron detectors (Type 5503-B), manufactured by Studsvik, Sweden, has been measured. Using the calculational model developed by Warren, the neutron sensitivity of these detectors has been computed taking into account additional factors viz. flux depression caused by the detector and the interaction of ⁵²V gamma rays with the emitter, the correction factors due to which have been evaluated to be 0.957 and 1.03 respectively. The evaluation of different parameters that are necessary in this connection and the neutron flux density determination at the position of irradiation of these detectors by gold foil activation, are presented. The measured and calculated values of neutron sensitivity are observed to agree to within 4%.