Neutron measurements at nuclear power reactors [55]

Staff from the Pacific Northwest National Laboratory (operated by Battelle Memorial Institute), have performed neutron measurements at a number of commercial nuclear power plants in the United States. Neutron radiation fields at light water reactor (LWR) power plants are typically characterized by low-energy distributions due to the presence of large amounts of scattering material such as water and concrete. These low-energy distributions make it difficult to accurately monitor personnel exposures, since most survey meters and dosimeters are calibrated to higher-energy fields such as those produced by bare or D₂O-moderated ²⁵²Cf sources. Commercial plants typically use thermoluminescent dosimeters in an albedo configuration for personnel dosimetry and survey meters based on a thermal-neutron detector inside a cylindrical or spherical moderator for dose rate assessment, so their methods of routine monitoring are highly dependent on the energy of the neutron fields.Battelle has participated in neutron assessments at a number of LWR facilities to characterize neutron radiation fields and to evaluate the responses of plant dosimeters and survey instruments. In these studies, the tissue equivalent proportional counter was used for measuring neutron dose and dose equivalent rates, and multisphere spectrometers were used to measure energy distributions. The use of these instruments in LWR work locations is usually difficult because of extreme environmental conditions such as high temperatures.These studies have confirmed the presence of low-energy neutron fields in most work locations. The studies have also found that albedo dosimeters used at power plants typically overrespond significantly when using a calibration based on californium exposures. Survey instruments also respond highly in typical LWR environments.     

Characterization of neutron reference fields at US Department of Energy calibration fields

The Health Physics Measurements Group at the Los Alamos National Laboratory (LANL) has initiated a study of neutron reference fields at selected US Department of Energy (DOE) calibration facilities. To date, field characterisation has been completed at five facilities. These fields are traceable to the National Institute for Standards and Technology (NIST) through either a primary calibration of the source emission rate or through the use of a secondary standard. However, neutron spectral variation is caused by factors such as room return, scatter from positioning tables and fixtures, source anisotropy and spectral degradation due to source rabbits and guide tubes. Perturbations from the ideal isotropic point source field may impact the accuracy of instrument calibrations. In particular, the thermal neutron component of the spectrum, while contributing only a small fraction of the conventionally true dose, can contribute a significant fraction of a dosemeter's response with the result that the calibration becomes facility-specific. A protocol has been developed to characterise neutron fields that relies primarily on spectral measurements with the Bubble Technology Industries (BTI) rotating neutron spectrometer (ROSPEC) and the LANL Bonner sphere spectrometer. The ROSPEC measurements were supplemented at several sites by the BTI Simple Scintillation Spectrometer probe, which is designed to extend the ROSPEC upper energy range from 5 to 15 MeV. In addition, measurements were performed with several rem meters and neutron dosemeters. Detailed simulations were performed using the LANL MCNPX Monte Carlo code to calculate the magnitude of source anisotropy and scatter factors.     

Establishment of a simple neutron calibration field using a moderated 252Cf source: Part II. Experimental characterization of simple neutron calibration field

This paper describes the development and experimental standardization of neutron fields simply arranged for detector calibrations used for radiation control and environmental measurement. These fields are the following: (1) bare ²⁵²Cf fission field, (2) iron-moderated ²⁵²Cf field, (3) carbone-moderated Cf field, and (4) polyethylene-moderated ²⁵²Cf field. These fields are most suitable for calibrating the detectors used in and around nuclear and radiation facilities, since the fields are designed to simulate the typical neutron fields in and around the facilities.The direct neutron components of these fields have been standardized by the following two methods: (1) calculation by the ANISN code, and (2) measurements with and without a shadow shield by detectors standardized in the national standard field at the Electrotechnical Laboratory (ETL). The neutron emission rates of the ²⁵²Cf source have been calibrated also at ETL. We have standardized only direct components because of their independence of room size and peripheral structures. The standardized values are energy spectra and dose equivalent rates of the direct neutron components; the accuracies have also been evaluated to be 20% below 100 keV, 15% at 1 MeV, and 50% above 5 MeV. These fields including room scattered components have also been characterized especially to calibrate neutron detectors having sensitivity to low energy room scattered neutrons, because of large errors caused by shadow shield subtraction.     

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.     

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.     

Neutron and gamma spectra measurements and calculations in benchmark spherical iron assemblies with 252Cf neutron source in the centre

The neutron and gamma spectra measurements have been made for benchmark iron spherical assemblies with the diameter of 30, 50 and 100 cm. The ²⁵²Cf neutron sources with different emissions were placed into the centre of iron spheres. In the first stage of the project, independent laboratories took part in the leakage spectra measurements. The proton recoil method was used with stilbene crystals and hydrogen proportional counters. The working range of spectrometers for neutrons is in energy range from 0.01 to 16 MeV, and for gamma from 0.40 to 12 MeV. Some adequate calculations have been carried out. The authors propose to carefully analyse the leakage mixed neutron and gamma spectrum from iron sphere of diameter 50 cm and then adopt that field as standard.     

NIST Calibration of a Neutron Spectrometer ROSPEC

A neutron spectrometer was acquired for use in the measurement of National Institute of Standards and Technology neutron fields. The spectrometer included options for the measurement of low and high energy neutrons, for a total measurement range from 0.01 eV up to 17 MeV. The spectrometer was evaluated in calibration fields and was used to determine the neutron spectrum of an Americium-Beryllium neutron source. The calibration fields used included bare and moderated ²⁵²Cf, monoenergetic neutron fields of 2.5 MeV and 14 MeV, and a thermal-neutron beam. Using the calibration values determined in this exercise, the spectrometer gives a good approximation of the neutron spectrum, and excellent values for neutron fluence, for all NIST calibration fields. The spectrometer also measured an Americium-Beryllium neutron field in a NIST exposure facility and determined the field quite well. The spectrometer measured scattering effects in neutron spectra which previously could be determined only by calculation or integral measurements.     

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.     

Modification of ROSPEC to cover neutrons from thermal to 18 MeV

Rotating Spectrometer (ROSPEC) is a neutron spectrometer designed to measure neutron energy distributions, and provide accurate neutron dosimetry. It is a completely self-contained unit and measures neutron energy via recoiling protons in gas proportional counters. Each of the four original gas counters is dedicated to a particular neutron energy range dictated by sensitivity to gamma rays at the low energy end of the spectrum and by proton collisions with the counter walls at the high energy end. Introduced originally in 1992, ROSPEC has a proven operational record with a program of continued upgrades. The operating range of the original ROSPEC spans 50 keV-4.5 MeV. The range of the ROSPEC has now been extended down to include epithermal and thermal neutrons by adding two 2 in. (3)He counters. Also, an optional simple scintillation spectrometer was designed to extend the upper limit of ROSPEC up to 18 MeV.     

Characteristics of the simulated workplace neutron fields using a 252Cf source surrounded with cylindrical moderators

The authors established the simulated workplace neutron fields using a 252Cf source surrounded with cylindrical moderators at the Japan Nuclear Cycle Development Institute (JNC), Tokai Works. The moderators are annular cylinders made of polymethyl methacrylate and steel. The neutron energy spectrum at the reference calibration point was evaluated from the calculations by MCNP-4B and the measurements by the Bonner multisphere spectrometer and the hydrogen-filled proportional counters. The calculated neutron spectra were in good agreements with the measured ones. These fields can provide the realistic neutron spectra similar to those encountered around the glove-boxes of the fabrication process of MOX (PuO2-UO2 mixed oxide) fuel.     

Characteristics of polyethylene-moderated 252Cf neutron sources

Polyethylene-moderated ²⁵²Cf neutron sources were designed to produce neutron reference fields’ spectra that simulate the spectra observed in the workplaces within nuclear reactors and accelerators. The paper describes the neutron sources and fields. Neutron spectra were calculated by the Monte Carlo method and compared with experimental data.     

Catalogue to select the initial guess spectrum during unfolding

A new method to select the initial guess spectrum is presented. Neutron spectra unfolded from Bonner sphere data are dependent on the initial guess spectrum used in the unfolding code. The method is based on a catalogue of detector count rates calculated from a set of reported neutron spectra. The spectra of three isotopic neutron sources ²⁵²Cf, ²³⁹PuBe and ²⁵²Cf/D₂O, were measured to test the method. The unfolding was carried out using the three initial guess options included in the BUNKIUT code. Neutron spectra were also calculated using MCNP code. Unfolded spectra were compared with those calculated; in all the cases our method gives the best results.     

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%.     

Study of a gold-foil-based multisphere neutron spectrometer

Multisphere neutron spectrometers with active thermal neutron detectors cannot be used in high-intensity radiation fields due to pulse pile-up and dead-time effects. Thus, a multisphere spectrometer using a passive detection system, specifically gold foils, has been investigated in this work. The responses of a gold-foil-based Bonner sphere neutron spectrometer were studied for two different gold-foil holder designs; an aluminium-polyethylene holder and a polyethylene holder. The responses of the two designs were calculated for four incident neutron beam directions, namely, parallel, perpendicular and at +/-45 degrees relative to the flat surface of the foil. It was found that the use of polyethylene holder resulted in a more isotropic response to neutrons for the four incident directions considered. The computed responses were verified by measuring the neutron spectrum of a 252Cf source with known strength.     

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.     

The clinical practice of 252Cf neutron brachytherapy devices on malignant tumor in China

The 252Cf neutron brachytherapy devices were designed and manufactured after solving those problems: the shield difficulties for 252Cf neutron; the method and facilities for 252Cf neutron source deliver; the dose calculation and measurement for 252Cf neutron. From 1999-2009, more than twenty 252Cf neutron brachytherapy devices had been installed in China. The Chinese radiation oncologists implemented the 252Cf neutron brachytherapy on more than 5 000 different kinds of cancer patients, including cervix cancer; endometrium cancer and achieved ideal results. For special presentation, they had breakthrough on radical treatment for low lying rectal adenocarcinoma with sphincter preservation and melanoma patients after the modification of 252Cf neutron intracavitary applicator and dose calibration.     

Investigations of shield effect and type of soil on landmine detection

This paper investigates the possibility of optimizing the performance of the neutron backscattering method in landmine detection by designing a suitable shield around a ²⁵²Cf neutron source to reduce the background due to soil and the neutrons emitted from the source that hit the detector directly. A series of Monte Carlo simulations were performed to improve the source shield thickness and to study the elastically backscattered (EBS) ²⁵²Cf neutrons from the buried explosive material TNT in the soil; the optimal configuration was examined against different soil types and source heights. The results obtained in terms of performance of the relative (EBS) neutrons confirmed that the proposed source shield has significantly improved the signal to background ratio. Higher signal-to-background ratio was observed using ²⁵²Cf neutron source as compared to Pu-Be source.     

Development of a light-weight portable neutron survey meter

A light-weight portable neutron survey meter was developed using a mixed organic gas counter for dose management at nuclear power plants and accelerator facilities. This survey meter, NSN31041, is ~2 kg in weight and W160×H250×L300 mm(3) in size, which is capable of measuring neutron ambient dose equivalent rate from thermal to 15 MeV neutrons. The neutron energy response of the survey meter is evaluated using continuous energy neutron sources of (252)Cf, (241)Am-Be, thermal neutrons generated from a graphite pile loading a (252)Cf source, concrete-moderated neutrons of (241)Am-Be source and D(2)O-moderated neutrons of (252)Cf source. The measured response data show very good agreement with neutron ambient dose equivalent within a 50 % deviation.     

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.     

Calculated neutron air kerma strength conversion factors for a generically encapsulated Cf-252 brachytherapy source

The ²⁵²Cf neutron air kerma strength conversion factor (S_KN/m_Cf) is a parameter needed to convert the radionuclide mass (μg) provided by Oak Ridge National Laboratory into neutron air kerma strength required by modern clinical brachytherapy dosimetry formalisms indicated by Task Group No. 43 of the American Association of Physicists in Medicine (AAPM). The impact of currently used or proposed encapsulating materials for ²⁵²Cf brachytherapy sources (Pt/Ir-10%, 316L stainless steel, nitinol, and Zircaloy-2) on S_KN/m_Cf was calculated and results were fit to linear equations. Only for substantial encapsulation thicknesses, did S_KN/m_Cf decrease, while the impact of source encapsulation composition is increasingly negligible as Z increases. These findings are explained on the basis of the non-relativistic kinematics governing the majority of ²⁵²Cf neutron interactions. Neutron kerma and energy spectra results calculated herein using MCNP were compared with results of Colvett et al. and Rivard et al.