Standardisation of water-moderated 241Am-Be neutron source using De Pangher neutron long counter: experimental and Monte Carlo modelling

A convenient neutron source is made for calibration of neutron survey instruments and personal dosimeters that are used in various nuclear installations such as fuel reprocessing, waste management, fuel fabrication and oil and well logging facilities, etc. This source consists of a bare (241)Am-Be neutron source placed at the centre of a 15-cm radius stainless steel spherical shell filled with distilled water. This paper describes the standardisation of the source at Bhabha Atomic Research Centre, using De Pangher neutron long counter both experimentally and using the Monte Carlo simulation. The ratio of neutron yield of water moderated to the bare (241)Am-Be neutron source was found to be 0.573. From the simulation, the neutron-fluence-weighted average energy of water-moderated (241)Am-Be source (fluence-weighted average energy of 2.25 MeV, dose-weighted average energy of 3.55 MeV) was found to be nearly the same as that of a (252)Cf source (fluence-weighted average energy of 2.1 MeV, dose-weighted average energy of 2.3 MeV). This source can be used for calibration in addition to (252)Cf, to study the variation in response of neutron monitoring instruments.     

Compliance of electronic personal neutron dosemeters with the new International Standard IEC 61526

The recommendations and test requests for the dose equivalent response of personal neutron dosemeters formulated by the new International Standard IEC 61526 are summarised. In particular, IEC 61526 allows the use of broad fields if dosemeters do not fulfil the hard requirements using monoenergetic neutrons. Some broad fields which can work as a replacement field using ISO sources ((252)Cf, (252)Cf (D(2)O mod.), (241)Am-Be) and simulated workplace fields (CANEL and SIGMA) are described. This work shows the results of recent measurements of the personal dose equivalent response for the dosemeters Thermo Electron EPD-N2, Aloka PDM-313 and the prototype dosemeter PTB DOS-2002, and discusses their compliance with respect to the new IEC 61526 standard.     

French comparison exercise with the rotating neutron spectrometer, 'ROSPEC'

The French laboratories in charge of 'neutron' dosimetry using the spectrometer 'ROSPEC', formed a working group in 2001. The participants began to study the behaviour of the instrument with a comparison exercise in broad energy neutron fields recommended by the International Organisation for Standardisation (ISO) and available at the LMDN in Cadarache. The complete version of the ROSPEC is made up of six spherical proportional counters fixed to a rotating platform. These counters cover different energy ranges which overlap each other to provide a link between the detectors, within the energy range from thermal neutrons to 4.5 MeV. The irradiation configurations chosen were ISO standard sources (252Cf, (252Cf+D2O)(/Cd), 241Am-Be) and the SIGMA facility. The results show that the 'thermal and epithermal' neutron fluence was widely overestimated by the spectrometer in all configurations.     

Evaluation of the characteristics of the neutron reference field using D2O-moderated 252Cf source

The ambient/personal dose equivalent per fluence for D(2)O moderated (252)Cf neutron source was determined by measurement. An appropriate subtraction of the scattered neutrons is required for the accurate measurement of direct neutrons. A cubic shadow object was used for the subtraction of the scattered neutrons from the surroundings. The scattered neutrons to be subtracted vary with the position of the shadow object due to the large volume of the source. Using the Monte Carlo code MCNP-4C, the optimum positions of the shadow object were surveyed for subtracting the scattered neutrons. The energy spectra of direct neutrons were measured in the optimum position. The dosimetric parameters for the D(2)O moderated (252)Cf neutron source were reasonable, taking into account the uncertainties of the parameters.     

Results of field trials using the NPL simulated reactor neutron field facility

The NPL simulated reactor neutron field facility provides neutron spectra similar to those found in the environs of UK gas-cooled reactors. Neutrons are generated by irradiating a thick lithium-alloy target with monoenergetic protons between 2.5 and 3.5 MeV (depending on the desired spectrum), and then moderated by a 40-cm diameter sphere of heavy water. This represents an extremely soft workplace field, with a mean neutron energy of 25 keV and, more significantly, a mean fluence to ambient dose equivalent conversion coefficient of the order of 20 pSv cm(2), approximately 20 times lower than those of the ISO standard calibration sources (252)Cf and (241)Am-Be. Results of field trials are presented, including readings from neutron spectrometers, personal dosimeters (active and passive) and neutron area survey meters, and issues with beam monitoring are discussed.     

Experimental comparison of 241Am-Be neutron fluence energy distributions

(241)Am-Be(alpha,n) neutron sources provide one of the most commonly used neutron fields for routine calibration of neutron sensitive devices. The neutron energy distribution of the IRSN standard (241)Am-Be source was measured in the energy region above 1.65 MeV using a BC501A proton-recoil liquid scintillator. The experimental data were compared to the ISO-recommended neutron energy distribution for an (241)Am-Be source. Some differences in shape were observed, with large variations mainly within the energy interval 3-6 MeV and around 8 MeV. Within the framework of a collaboration between three national metrological institutes (PTB, Germany; NPL, UK and LNE-IRSN, France), the neutron energy distributions of (241)Am-Be sources at each laboratory have been compared. The IRSN-BC501A proton-recoil scintillator was used to measure all the sources. The results show different energy distributions a priori influenced by the origin of the source, i.e. the manufacturing process. The maximum deviation observed for the integral dose equivalent, in the measured BC501A energy range, is within the 4% uncertainty recommended by ISO standard 8529-2 to allow for variations of the neutron spectrum among different (241)Am-Be sources. However, knowledge of the energy distribution of an (241)Am-Be source provides a way to reduce the uncertainty in the dose equivalent rate delivered by such a source.     

Evaluation of spectrum measurement devices for operational use

Several neutron spectrometers manufactured by Bubble Technology Industries (BTI) were tested and evaluated in a variety of neutron fields. Findings and conclusions are presented for the following BTI instruments: a modification of the Rotational Spectrometer (ROSPEC) that includes a thermal and epithermal capability, the Simple Scintillation Spectrometer that is used in conjunction with the ROSPEC to extend its high-energy range, and the MICROSPEC N-Probe which is capable of providing a crude spectrum over the energy range from thermal to 18 MeV. The main objective of these measurements was to determine the accuracy of both the energy spectrum and dose equivalent information generated by these devices. In addition, the dose response of the Wide-Energy Neutron Detection Instrument (WENDI-II) was measured in all neutron fields relative to a bare ²⁵²Cf calibration. The performance of the WENDI-II rem meter was compared to the dose information generated by the neutron spectrometers. The instruments were irradiated to bare ²⁵²Cf and ²⁴¹AmBe sources, and in a series of moderated ²⁵²Cf fields using a standard D₂O sphere and a set of polyethylene spheres. The measured spectra were benchmarked with a set of detailed Monte Carlo calculations with the same energy bin structure as that of the instruments under test. These calculations allowed an absolute comparison to be made with the measurements on a bin by bin basis. The simulations included the effects of room return and source anisotropy.     

Radionuclide neutron sources in calibration laboratory--neutron and gamma doses and their changes in time

The calibration laboratory, having standard neutron fields of radionuclide sources, should perform regular measurements of fields' parameters in order to check their stability and to get knowledge of any changes. Usually, accompanying gamma radiation is not of serious concern, but some personal dosemeters, old neutron dose equivalent meters with scintillation detectors and the dose meters of mixed radiation require the determination of this component. In the Laboratory of Radiation Protection Measurements in the Institute of Atomic Energy, Poland, the fields of radionuclide neutron sources (252)Cf, (241)Am-Be and (239)Pu-Be were examined for nearly 20 y. A number of detectors and methods have been applied for the determination of neutron ambient dose equivalent rate and for the determination of neutron and gamma dose components. This paper presents the recent results of measurements of gamma and neutron dose and dose equivalent, compared with the results accumulated in nearly 20 y.     

Calculations of anisotropy factors for radionuclide neutron sources due to scattering from source encapsulation and support structures

A model has been developed for calculating the angular neutron fluence distributions for radionuclide neutron sources that are heavily encapsulated or surrounded by source support structures as a source holder and a source movement system. These structures may cause an anisotropic neutron fluence distribution. This should be taken into account in the neutron-measuring instruments calibration procedure. The calculations were made for two types of widely used neutron sources, (241)Am-Be and (252)Cf, by combining an in-house code simulating the (9)Be(alpha,n) reactions and the Monte Carlo code MCNP-4C. As a result, anisotropy factors in the direction perpendicular to the source capsule axis for bare neutron sources were evaluated to be 1.012, 1.030 and 1.039 for (252)Cf in a standard Amersham X1 capsule, (241)Am-Be in a X3 capsule and (241)Am-Be in a X4 capsule, respectively. These values are in reasonable agreement with the published data. If the support structures are included in the MCNP simulation, the anisotropy factors for these neutron sources increase by approximately 10%.     

Response of Harshaw neutron thermoluminescence dosemeters in terms of the revised ICRP/ICRU recommendations

To monitor workers for external neutron radiation dose, the Y-12 National Security Complex utilises the thermoluminescence dosemeters (TLDs) manufactured by Harshaw. At Y-12, the majority of external dose to workers is due to low-energy photon and/or beta particles emitted from uranium and its progeny. However, some neutron dose is expected since neutrons are produced from (alpha,n) reactions in various compounds found at the plant, including UF4 and UF6. Neutron sources, such as 252Cf, are also used throughout the complex. The Harshaw neutron dosemeter consists of two gamma-sensitive elements (7Li) and two neutron-sensitive elements enriched in 6Li with various shielding/filter materials placed around each of them. In this work, the energy response of the dosemeter to neutrons has been calculated using the Monte Carlo transport code MCNP Version 4-C and, these results are compared with the measured response of the dosemeter to unmoderated and D2O-moderated 252Cf neutrons. The response of the dosemeter has also been determined in terms of the personal absorbed dose and personal dose equivalent as a function of neutron energy based on the recommendations of the ICRP Publication 60 and ICRU Report 49. The energy response of the dosemeter characteristics can be used to generate spectral conversion coefficients for routine neutron absorbed dose and dose equivalent calculations.     

Neutron calibration facility with radioactive neutron sources at KRISS

For the purpose of radiation protection, the reference neutron field for calibration of neutron monitors was constructed using radioactive neutron sources-bare-(252)Cf, D(2)O-moderated (252)Cf and (241)Am-Be(alpha,n)-at Korea Research Institute of Standards and Science (KRISS). The well-specified neutron source with its emission rate and the anisotropy was installed at the centre of the neutron irradiation room, which is 6.6 x 7.6 x 6.3 m(3) in size. The neutron spectra of each source was measured using the Bonner sphere spectrometer (KRISS-BSS). Calculations using MCNP5 with realistic geometry and materials in the neutron irradiation room were performed. The calculations and measurements were found to be in good agreement, showing that the neutron calibration facility at KRISS is well established.     

Shielding design studies for a neutron irradiator system based on a 252Cf source.

This study aims to investigate a shielding design against neutrons and gamma rays from a source of 252Cf, using Monte Carlo simulation. The shielding materials studied were borated polyethylene, borated-lead polyethylene and stainless steel. The Monte Carlo code MCNP4B was used to design shielding for 252Cf based neutron irradiator systems. By normalising the dose equivalent rate values presented to the neutron production rate of the source, the resulting calculations are independent of the intensity of the actual 252Cf source. The results show that the total dose equivalent rates were reduced significantly by the shielding system optimisation.     

Assessment of neutron dosemeters around standard sources and nuclear fissile objects

In order to evaluate the neutron doses around nuclear fissile objects, a comparative study has been made on several neutron dosemeters: bubble dosemeters, etched-track detectors (CR-39) and 3He-filled proportional counters used as dose-rate meters. The measurements were made on the ambient and the personal dose equivalents H*(10) and Hp(10). Results showed that several bubble dosemeters should have been used due to a low reproducibility in the measurements. A strong correlation with the neutron energy was also found, with about a 30% underestimation of Hp(10) for neutrons from the PuBe source, and about a 9% overestimation for neutrons from the 252Cf source. Measurements of the nuclear fissile objects were made using the CR-39 and the dose-rate meters. The CR-39 led to an underestimation of 30% with respect to the neutron dose-rate meter measurements. In addition, the MCNP calculation code was used in the different configurations.     

Shielding evaluation of a medical linear accelerator vault in preparation for installing a high-dose rate 252Cf remote afterloader

In support of the effort to begin high-dose rate 252Cf brachytherapy treatments at Tufts-New England Medical Center, the shielding capabilities of a clinical accelerator vault against the neutron and photon emissions from a 1.124 mg 252Cf source were examined. Outside the clinical accelerator vault, the fast neutron dose equivalent rate was below the lower limit of detection of a CR-39 etched track detector and below 0.14 +/- 0.02 muSv h(-1) with a proportional counter, which is consistent, within the uncertainties, with natural background. The photon dose equivalent rate was also measured to be below background levels (0.1 muSv h(-1)) using an ionisation chamber and an optically stimulated luminescence dosemeter. A Monte Carlo simulation of neutron transport through the accelerator vault was performed to validate measured values and determine the thermal-energy to low-energy neutron component. Monte Carlo results showed that the dose equivalent rate from fast neutrons was reduced by a factor of 100,000 after attenuation through the vault wall, and the thermal-energy neutron dose equivalent rate would be an additional factor of 1000 below that of the fast neutrons. Based on these findings, the shielding installed in this facility is sufficient for the use of at least 5.0 mg of 252Cf.     

Standard fields of old neutron sources--parameters and traceability

Standard neutron fields of isotope neutron sources, established in the Institute of Atomic Energy, Poland, have been examined for nearly 20 y. The neutron dose equivalent and gamma dose rates were measured, applying various techniques. The neutron emission of standard neutron sources of 241Am-Be and 252Cf has been originally determined in primary standard laboratory, then checked, making use of transfer instrument calibrated in primary laboratory. The growth of neutron emission of 239Pu-Be source, used for routine calibrations, has been demonstrated. The total uncertainty of determined parameters has been discussed. The periodically repeated checks of neutron fields of standard sources with the use of transfer instrument, calibrated in primary laboratory, should provide the traceability to primary laboratory and the fields could be officially recognised.     

Response components of LiF:Mg,Ti around a moderated Am-Be neutron source

The responses of TLD-1010, TLD-700 and TLD-600 thermoluminescence dosemeters to the radiation field inside a water tank enclosing an isotopic 241Am-Be neutron source are analysed. Separate contributions coming from thermal neutrons, neutrons with energies above thermal and gamma rays to the total response of the three types of TLD are obtained. This is accomplished by assuming that the gamma responses for materials with different 6Li enrichments are identical and that the neutron response of TLD-700 is negligible compared to TLD-100 and TLD-600. The last assumption is tested by Monte Carlo simulations of the neutron energy spectrum at the points where the TLDs are irradiated.     

Characteristics of the neutron field of the facility at DIN-UPM

A new source facility (241Am-Be) has been installed in a bunker-type room of large dimensions. To characterise the neutron fields in the facility, detailed calculations have been made with MCNP-4C, showing the different components of the neutron radiation reaching the reference points (direct, inscattered, backscattered). The contribution from neutrons scattered in the walls to the total ambient dose equivalent remains reasonably low (<10%) in the reference points. Additionally, spectra measurements have been performed with a Bonner spheres spectrometer with a 6LiI(Eu) scintillator (0.4 phi x 0.4 cm2), UTA4 response matrix and BUNKIUT unfolding code. The calculated and experimentally obtained spectra are compared, with small differences found in the epithermal and thermal region, attributable to the concrete composition used in the calculations. The H*(10) rate has been determined from the spectra, and then compared to the reading of an active dosemeter (LB6411), with differences found lower than 8%.     

Calibration and testing of a TLD dosemeter for area monitoring

The response of a TLD-600/TLD-700 area dosemeter has been characterized in neutron fields around the 590 MeV cyclotron ring at the Paul Scherrer Institute (PSI). The dosemeter is based on a cylindrical paraffin moderator with three of each type of TLD chip at the centre, and is intended to use for area monitoring around accelerator facilities. The dosemeter is calibrated in terms of ambient dose equivalent using a non-moderated 252Cf neutron source. The ambient dose equivalent response has been tested in five locations where the neutron fields and dose rates have been well characterized by Bonner sphere spectrometer and active neutron monitor measurements. The different spectrum shapes and dose rates in the five locations permit the comparison of the behavior of the active and passive dosemeters in these neutron fields.     

Development and testing of an active area neutron dosemeter

This paper describes the design, development and testing of an active area neutron dosemeter (AAND). The classic moderator and central detector is retained but in AAND this arrangement is augmented by small thermal neutron detectors positioned within the moderating body. The outputs from these detectors are combined using an appropriately weighted linear superposition to fit both the ambient dose equivalent and the radiation weighting factor. Experimental verifications of both the modelled detector energy reponses and the overall AAND response are given. In the relatively soft D2O moderated 252Cf spectra, the AAND determined both the H*(10) and mean radiation weighting factor to better than +10%.     

Measured and calculated angular responses of panasonic UD-809 thermoluminescence dosemeters to neutrons

Multi-element thermoluminescence dosemeters (TLD), such as the Panasonic UD-809, are used in personal dosimetry. The Panasonic UD-809 dosemeter consists of one gamma sensitive and three neutron sensitive TLD elements with different filter materials. In this work, the neutron energy responses (the number of (n,alpha) reactions per neutron) of the neutron-sensitive TLD elements of the Panasonic UD-809 dosemeter were calculated using the MCNP Monte Carlo transport code. Experiments were performed in a calibration geometry with an unmoderated 252Cf neutron source. These measurements were made with the dosemeter placed on the centre front face of a polymethylmethacrylate (PMMA) slab phantom. The phantom was rotated in the horizontal plane from -90 to +90 degrees, in 15 degree increments. Good agreement between calculated and measured element responses was observed. The angular dependency of personal dose equivalent was also calculated for parallel beams of 252Cf neutrons and compared to the TLD element angular responses.