Pu-Be Neutron Source Research Articles

A novel method of quantitative assay of PuBe neutron sources by neutron coincidence technique

Using a simple model for PuBe neutron sources, in which the neutrons were assumed to be randomly emitted from the centre of a spherical source, and induce neutron-producing secondary reactions in the material, calculation shows that the ratio of the real coincidence count rate R to the total count rate T, measured by neutron coincidence technique, is proportional to the Pu content of the sample. Therefore, measuring the ratio R/ T can be considered a method for determining the Pu content of PuBe neutron sources. The effect of the moderator of a neutron coincidence collar on the count rate of doubles was investigated experimentally using uranium samples of different enrichment (0.7–2.7% 235U). An important conclusion from the data of the experiments is that the influence of the neutrons backscattered from the moderator for the secondary reaction rate can be neglected during measuring PuBe neutron sources. This result leads to an empirical formula for determining the Pu content of PuBe neutron sources, which is independent of the type of the detector. The accuracy of the method is about 15%.

Quantitative non-destructive assay of PuBe neutron sources

PuBe neutron sources were assayed, using a combination of high resolution γ-spectrometry (HRGS) and neutron correlation technique. In a previous publication [J. Bagi, C. Tam Nguyen, L. Lakosi, Nucl. Instr. and Meth. B 222 (2004) 242] a passive neutron well-counter was reported with 3He tubes embedded in a polyamide (TERRAMID) moderator (lined inside with Cd) surrounding the sources to be measured. Gross and coincidence neutron counting was performed, and the Pu content of the sources was found out from isotope analysis and by adopting specific (α, n) reaction yields of the Pu isotopes and 241Am in Be, based on supplier’s information and literature data. The method was further developed and refined. Evaluation algorithm was more precisely worked out. The contribution of secondary (correlated) neutrons to the total neutron output was derived from the coincidence (doubles) count rate and taken into account in assessing the Pu content. A new evaluation of former results was performed. Assay was extended to other PuBe sources, and new results were added. In order to attain higher detection efficiency, a more efficient moderator was also applied, with and without Cd shielding around the assay chamber. Calibration seems possible using neutron measurements only (without γ-spectrometry), based on a correlation between the Pu amount and the coincidence-to-total ratio. It is expected that the method could be used for Pu accountancy and safeguards verification as well as identification and assay of seized, found, or not documented PuBe neutron sources.

Development of a microstrip-based neutron detector

A gas-filled microstrip detector for thermal neutrons has been built and successfully tested in our laboratory. The detector has an active area of 20 mm x 15 mm and consists of alternate anodes and cathodes of widths 12 mum and 300 mum respectively. The anode to cathode gap is 150mum and the pitch is 612 mum. A high resistance, meandering type horizontal strip connects the anodes at one end and aids in position sensing by charge division method. The detector is tested with gas mixtures He-3+Kr (1: 2) and He-3+CF4 (2: 1)at pressure of 3 atmospheres and using a Pu-Be neutron source. The pulse height spectrum shows energy resolution of similar to8% (FWHM) for the 764 keV peak at anode voltage of 525 V for 3 He+Kr and similar to15% at anode voltage of 800 V for He-3+CF4. Gas gains up to 6.3 x 10(3)and 3.6 x 10(3) are obtained respectively with these gas mixtures. The overall efficiency of the detector along the sensitive length is tested by exposing the active area to neutrons and recording the position spectrum. The detector shows fairly uniform efficiency (similar to45%) over the active length.

Study of neutron-induced ionization in helium and argon chamber gases

We present studies of helium- and argon-filled ionization chambers exposed to intense neutron fluxes from PuBe neutron sources ( E n=1– 10 MeV) . The sources emit about 10 8 neutrons per second. The number of ion pairs in the chamber gas volume per incident neutron is derived. While limited in precision because of a large gamma ray background from the PuBe sources, our results are consistent with the expectation that the neutrons interact purely elastically in the chamber gas.

Prompt-gamma neutron activation analysis facility for in vivo body composition studies in small animals.

The design, calibration, dosimetry and performance evaluation of a prompt-gamma neutron activation analysis facility for in vivo body composition studies in small animals (i.e. rats or rabbits) is discussed. The system design was guided by Monte Carlo transport calculations using MCNP-4C code. A system was built and performance evaluation was made using a 185-GBq Pu-Be neutron source. Prompt-gamma rays produced by neutron capture reactions were detected by a combination of a NaI(Tl) scintillation and a HPGe semiconductor detectors. Nitrogen and chlorine were quantified by analysis of the 10.83-MeV and 6.11-MeV peaks, respectively. Appropriate corrections for the animal body size were determined. The facility described allows the in vivo determination of protein and extracellular space in sets of experimental animals.

Non-invasive measurement of water content in lung phantom by neutrons: latest improvements to the technique.

The novel technique for measuring the water content in lungs based on the slowing down of fast neutrons has been improved making possible its clinical use on patients. This goal has been reached by performing computer simulations that closely mimic the experimental setup and by acquiring a new set of measurements obtained with a plutonium-beryllium (Pu-Be) neutron source provided of an activity suitable to clinical requirements. Finally, accurate radiation dose measurements have been carried out to ensure the acceptability of the technique.

Large size LSO:Ce and YSO:Ce scintillators for 50 MeV range /spl gamma/-ray detector

LSO:Ce and LYSO:Ce scintillators are candidates for the pre-shower detector for the search for /spl mu//sup +//spl rarr/e/sup +/ /spl gamma/ decay with the branching-ratio sensitivity of 10/sup -14/ at PSI (Villigen, Switzerland). The detection involves 52.8 MeV /spl gamma/-rays. To study the response of the crystals to high energy /spl gamma/-rays we have chosen a Pu-Be neutron source which gives mono-energetic 4.439 MeV /spl gamma/-rays due to the /sup 9/Be(/spl alpha/,n) reaction. We have studied the response of a /spl phi/60/spl times/10 mm YSO:0.2%Ce crystal and a half disk of /spl phi/60/spl times/10 mm LSO:0.2%Ce. Despite the large background of /sup 176/Lu, the high energy /spl gamma/-rays can be detected efficiently. We have also studied the response to lower energy /spl gamma/-rays (from /sup 137/Cs and /sup 60/Co sources) and found the conditions of optimum work. The YSO crystal detects 4.439 MeV /spl gamma/-rays primarily by double 511 keV escapes that form a distinct peak in the energy spectrum.

Neutron Dose Measurement at the Brookhaven National Laboratory Whole-Body Activation Facility

The neutron dose at the Brookhaven whole-body activation facility was measured under various operational conditions. A bomab phantom was used in the measurements. The measurements were done with a pair of a tissue-equivalent (TE) and a graphite ionisation chambers. The TE chamber is used for neutron dose measurement and the graphite for gamma dose subtraction. The effect of a polyethylene (PE) moderator was investigated as well as the effect of the source-body distance. In addition to the regular phantom. the dose to an obese phantom was also measured. The facility is made of 14 PuBe neutron sources, ∼50 Ci each. The patient stays in the facility for 300 s, lying on a 2 cm thick polyethylene (PE) bed, covered with a 2 cm thick PE moderator. The neutron dose at the patient's chest under such conditions is 267 mrem. The gamma dose is 12 mrem. When the PE moderator is removed, the dose increases by about 20%. Thinner moderator (∼1 cm) also increases the dose: it also increases as the patient's skin gets closer to the sources. The increase is close to 1/r 2 . A 2.5 cm change in distance at 30 cm increases the dose by 22%. The dose to the patient's back was found to be 10% higher due to neutron scattering from the facility floor and walls. In addition to the dose, the neutron and gamma spectra were measured with and without the phantom and moderator. Based on the measured neutron spectrum, an effective QF=13 was found. Neutron moderation by hydrogen was observed, the high energy neutron peaks are lower when more hydrogen is present in the surrounding material. In the gamma spectra it was found, as expected, that the hydrogen in the PE moderator and more so in the phantom, increases the intensity of the 2.2 MeV gamma rays.

Exploratory studies on neutron radiography with a small neutron source using a nuclear scintillation imaging technique

Neutron radiography based on mobile neutron sources need optimum utilization of available neutron fluxes which are usually lower compared to those available from reactors. For optimum utilization of such low flux devices, a sensitive neutron imaging technique is required. Such a neutron imaging system based on a Li 6F-ZnS scintillator screen has been developed using a pair of image intensifier tubes and a charge coupled device. This detector system has been employed to study the feasibility of neutron radiography using low neutron fluences. The main feature of this imaging system is its ability to detect individual neutron scintillation events with a higher degree of spatial resolution. In order to test the efficiency of this imaging system, a small scale moderator-collimator assembly was designed using a Pu-Be neutron source of strength ≈2 × 10 7 n/s. Details of this imaging system and results of some exploratory experiments for low fluence neutron imaging are presented in this paper.

High sensitivity neutron imaging system for neutron radiography with a small neutron source

Although neutron radiography is a very useful NDT technique. its widespread use has been limited due to the lack of a mobile neutron radiography system. One of the major requirements in designing such a system is the development of a low fluence neutron imaging technique. In recent years several groups in France, Japan and the USA have been working on such an imaging system. We report in this paper the development of such a neutron imaging system based on a Li 6F-ZnS scintillator screen using a pair of image intensifier tubes and a charge coupled device. This detector system has been employed to study the feasibility of neutron radiography using low neutron fluences. The main feature of this imaging system is its ability to detect individual neutron scintillation events with a ligh degree of spatial resolution. Details of this imaging system and the preliminary results obtained using a Pu-Be neutron source of strength ≈ 2 × 10 7 n/s are reported in this paper.

Determination of water in zeolites by a neutron reflection method

Water content in zeolites has been determined by an improved neutron reflection method using a Pu-Be neutron source of 106 n·s−1 intensity and a BF3 counter. It was found that the water content of different types of zeolites collected in Hungary varies between 9 and 12 wt.%. The standard deviation of the determination does not exceed 0.5 wt.%. The matrix effect on the sensitivity and accuracy was also studied. An approximate relation is given between the count rate and the neutron physical parameters of the samples.

Experiences with CR-39 SSNTD in monitoring neutrons emitted from spent reactor fuel

Abstract Neutron emission from spent fuel assemblies was studied using Hungarian made CR-39 for monitoring the transuranium content as the burnup parameter. To detect thermal and fast neutrons Boron loaded and polyethylene radiators were applied. It was recognized that the high gamma backgroung in the spent fuel ponds causes sometimes a drastic change in the bulk etch rate and in the particle detection sensitivity. An empirical formula was deduced to estimate the change of the bulk etch rate: V B ʒ  V Bo = f ∗ t ∗ B T − 1 2 , where factor f depends on the etching conditions, t is the exposure time, B is the burnup parameter and T is the cooling time after removing the fuel from the reactor core. In addition, Co-60 and LINAC irradiation facilities combined with Pu-Be neutron sources were used to study the effect of the high gamma dose on the bulk etch rate and proton and alpha sensitivity of the Hungarian CR-39.

Gadolinium-neutron-activation determination with a Pu-Be source

A nondestructive neutron activation method for determining gadolinium content for reactor construction materials was developed. The method uses a Pu-Be neutron source capable of giving 10/sup 8/ neutrons per second and the neutron reaction with a /sup 160/Gd target. To determine the flux attenuation, induced-activity distributions were measured along the radius with artificial compacts of Al/sub 2/O/sub 3/ + Gd/sub 2/O/sub 3/ specimens with varying gadolinium contents. The specimens were irradiated in unscreened and screened containers. The ratios of the unfiltered and filtered activities were not more than 1.06 +/- 0.04. The dependence of the gamma-ray absorption coefficient on gadolinium content and the effect of gadolinium content on the count rate due to /sup 161/Gd were determined. The nondestructive neutron-activation determination of gadolinium was possible for gadolinium concentrations where the radial induced-activity distribution was constant. The method for calculating the gamma-ray absorption coefficient was simple and reliable for measurement geometry close to 4pi. Neutron activation results agreed with chemical measurement within the error limits.

Fast Neutron TL Dosimetry by Knock-on Protons: Energy Deposition in Thin Dosemeter Layers

Thin layer dosemeters are suitable for rapid laser heating and for the detection of knock-on protons produced in hydrogenous materials by collisions with fast neutrons. In an attempt to establish the optimum design parameters for a fast neutron dosemeter based on such a proton radiator, we calculate the proton conversion efficiencies and the resulting normalised TL signals for CaSO4:Dy layers of various thicknesses. On the basis of the known neutron energy spectra of 252Cf, moderated 252Cf, and Pu-Be neutron sources, the generation of protons in polyethylene foils, placed in close contact with the TLD layer, and the energy deposited by these protons in the CaSO4:Dy layer are computed. The results are compared with experiments performed in mixed neutron-gamma fields with 9 mm2 spots of this TL phosphor on borosilicate glass substrates heated with a computer-controlled CO2 laser beam of uniform intensity profile.

Fast Neutron TL Dosimetry by Knock-on Protons: Energy Deposition in Thin Dosemeter Layers

Abstract Thin layer dosemeters are suitable for rapid laser heating and for the detection of knock-on protons produced in hydrogenous materials by collisions with fast neutrons. In an attempt to establish the optimum design parameters for a fast neutron dosemeter based on such a proton radiator, we calculate the proton conversion efficiencies and the resulting normalised TL signals for CaSO4:Dy layers of various thicknesses. On the basis of the known neutron energy spectra of 252Cf, moderated 252Cf, and Pu-Be neutron sources, the generation of protons in polyethylene foils, placed in close contact with the TLD layer, and the energy deposited by these protons in the CaSO4:Dy layer are computed. The results are compared with experiments performed in mixed neutron-gamma fields with 9 mm2 spots of this TL phosphor on borosilicate glass substrates heated with a computer-controlled CO2 laser beam of uniform intensity profile.

Neutron field measurements at the Aladdin Synchrotron Light Source.

The neutron field near the inflector of the 800-MeV electron storage ring was studied. Photon-induced neutrons are produced by 100-MeV electrons bombarding the inflector during injection into the synchrotron ring. Neutrons were measured with moderating detectors made of 15 X 15 X 20 cm Lucite blocks and Au activation foils. Detector response was established with a Pu-Be neutron source and a 25.4-cm polyethylene sphere and Au foil detector. The neutron yield was 0.97 +/- 0.14 X 10(12) kJ-1. For 1.38 W of electron pulse power, the dose equivalent rate 1 m aside and 1 m above the inflector was 4.35 +/- 0.47 mu Sv s-1 and 3.13 +/- 0.23 mu Sv s-1, respectively. A measured dose equivalent transmission curve for polyethylene yielded an attenuation coefficient of 15.7 m-1.

Originalarbeiten Aufbau einer Wärmeflußmeßeinrichtung zur Bestimmung der thermischen Leistung einer PuBe-Neutronenquelle

The layout of a simple heat-flow calorimeter for the determination of the thermal power of PuBe-neutron sources is described. Using isotopic composition dates for the radioactive material in the sources obtained from γ-spectromctric measurements the mass of 239Pu contained in the PuBe-neutron sources can be determined non-destructively.

Plutonium metallurgy in India

The embryo of plutonium metallurgy in India, as well as in Asia, was formed nearly 25 years back in Trombay. Thereafter, in the intervening 25 years, significant amount of research and development work have been carried out atbarc on this man-made fissile element and indigenisation has been achieved in the fabrication, characterisation and property evaluation of plutonium metal, delta-stabilised alloys, Al-Pu fuels, Pu-Be neutron sources, (UPu)O2 fuels for thermal and fast reactors and (UPu)C and (UPu)N advanced fuels for liquid metal-cooled fast breeder reactors (lmfbr).This paper summarises the development of plutonium metallurgy in India highlighting the safety aspects, our achievements and potential of plutonium for generation of nuclear electricity in the coming decades.

Experimental tests of the Bonner sphere spectrometer using filtered neutron beams

The operation of a Bonner Sphere neutron detector system has been tested using several unique neutron sources. Filtered neutron beams at beamport F at the University of Missouri Research Reactor (MURR) were used as a source of known quasi-monoenergetic neutrons for precise energy spectra analysis and calibration. A PuBe neutron source was used for absolute flux magnitude and spectral verification. Two computer codes, SWIFT and Least Squares Unfolding Techniques (LSUT), were used to unfold the experimental data. Several operational problems were encountered during these tests. First, many of the measurements involved neutron beam measurements in which the beams had a smaller diameter than the moderating spheres. This caused partial illumination of the spheres for which correction factors had to be developed. A partial illumination correction factor has been proposed and tested to account for this problem. Second, reactor core gamma-ray contamination in the neutron beams was of sufficient magnitude to interfere with some measurements. Gamma-ray background subtraction techniques using a multi-channel analyzer were used to alleviate this problem. After correcting for gamma-ray background and applying partial illumination correction factors, unfolded neutron spectra from the unfolding codes gave good results for most neutron sources. In particular the SWIFT results were quite good, exceeding expectations in terms of energy resolution and spectral accuracy.

Determination of aluminum in bulk iron ore samples by neutron activation analysis

A fast neutron activation technique has been developed for the determination of SiO/sub 2/ in 30-kg samples of iron ore of -6-mm particle size. Only a single irradiation with a Pu-Be neutron source is necessary, if particular care is taken to minimize the thermal neutron flux in the sample. By including the sample weight and the epithermal neutron flux in the regression equation, an accuracy of better than +-0.34 percent SiO/sub 2/ (95 percent confidence level) is achieved for silica concentrations ranging from 1 to 10 percent in the presence of low amounts of Al/sub 2/O/sub 3/ (up to 6 percent). Samples must be dried to <1 percent free moisture before analysis, if the best accuracy is required.