Standard Neutron Field Research Articles

Calibration factors for determination of relativistic particle induced fission rates in natU, 235U, 232Th, natPb and 197Au foils

Calibration factors w, for determination of fission rate in metallic foils of natU, 235U, 232Th, natPb and 197Au were determined for foils in contact with synthetic mica track detectors. Proton-induced fission at proton energies of 0.7 GeV and 1.5 GeV were used. Using our experimental results as well as those of the other authors, w for different foil–mica systems were determined. Two methods were used to calculate w, relative to the calibration factor for uranium–mica system, which has been obtained in a standard neutron field of energy 14.7 MeV. One of these methods requires the knowledge of the mean range of the fission fragments in the foils of interest and other method needs information on the values of the fission cross-sections at the required energies as well as the density of the tracks recorded in the track detectors in contact with the foil surfaces. The obtained w-values were compared with Monte Carlo calculations and good agreements were found. It is shown that a calibration factor obtained at low energy neutron induced fissions in uranium isotopes deviates only by less than 10% from those obtained at relativistic proton induced fissions.

Calibration of neutron-sensitive devices

The calibration of a neutron-sensitive device can range from a simple calibration factor at a single energy or energy distribution to a full response characterization over the entire energy range to which the device is sensitive. As the responses of neutron-sensitive devices and the fluence-to-dose-equivalent conversion coefficients can vary with neutron energy and incident angle, both simulation and experiments in standard neutron fields are required. Although several ISO standards present calibration principles in general and detailed discussion on many specific areas, there are certain omissions and limitations that this paper intends to highlight, along with some new recommendations derived from the recent literature, mainly focused on the effective centre, corrections for geometry and neutron scattering, as well as the problem of calibrating in terms of personal dose equivalent.

Monitoring the ambient neutron radiation around a high energy electron/positron collider using PADC dosemeters

At the INFN Frascati National Laboratories (INFN-LNF), the DA Φ NE accelerator complex for fundamental and applied physics is operating since 1997. The routine radiation protection monitoring of the ambient neutron radiation around the accelerator is performed by means of a homemade dosimetry service, based on chemically etched PADC detectors. The LNF dosemeter is based on a four-element configuration, in which the main characteristic is a new procedure to reduce the fraction of “false positives”, i.e. background dosemeters inferred as doses due to spurious high track densities. The calibration of the dosemeters has been performed with a standard 241Am-Be neutron field. In addition, field-specific calibration factors have been derived, relying on the measured neutron spectra in selected workplaces around the accelerator. This work shows the results of the first two months monitoring period and discusses the suitability of the PADC for the neutron dosimetry around particle accelerators. The implications in the accuracy of the measurement, due to the realistic calibration and the new four elements configuration, are also addressed.

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.

Characterisation of kilo electron volt neutron fluence standard with the 45Sc(p,n)45Ti reaction at NMIJ

We are developing a national standard of a monoenergetic kilo electron volt neutron field with the (45)Sc(p,n)(45)Ti resonance reaction. A wide resonance yields 27.4 keV neutrons at 0 degrees with respect to the proton beam. The proton energy was precisely determined in the measurement of the relative neutron yield as a function of the proton energy from the threshold energy to 2.942 MeV. Absolute measurement of the monoenergetic neutron fluence was performed using a (3)He proportional counter. Relative measurement was also carried out using a Bonner sphere calibrated at our 144 keV standard neutron field. Calibration factors were obtained between the count of a neutron monitor and the neutron fluence. A silicon-surface barrier detector with a (6)LiF foil converter was also being developed for the neutron fluence measurement. Successful results were obtained in the tests in the 144 keV standard neutron field.

Spent Fuel Monitoring with Silicon Carbide Semiconductor Neutron/Gamma Detectors

Abstract Silicon carbide semiconductor radiation detectors have been demonstrated for neutron and gamma-ray monitoring of spent nuclear fuel. Neutrons and gamma rays were monitored simultaneously over a 2050-h period, resulting in a gamma dose of over 6000 Gy to the SiC detector. No changes in the neutron and gamma-ray sensitivity were observed as a result of the gamma-ray exposure. After the spent-fuel measurements, the absolute neutron sensitivity was determined in a standard neutron field, and the degree of gamma-ray and neutron spectrum overlap were determined through exposures in intense gamma ray fields. No overlap was observed for gamma-ray dose rates up to 100 G/h.

Neutron and Photon Dosimetry at the LR-0 Reactor Using Paired Detectors

Abstract Paired thermoluminescence dosimeters and ionization chambers were used for characterizing mixed (neutron and photon) radiation fields at the LR-0, a zero power model of VVER reactors. Detector responses were investigated in standard photon and thermal neutron fields. A good agreement was obtained among the doses measured with the two methods. Furthermore, the measured photon and neutron attenuation values are consistent with those from evaluated, independent data.

The ΣΣ secondary intermediate-energy standard neutron field development at the Romania TRIGA Reactor

The ΣΣ intermediate-energy reference spectrum irradiation facility, operated until 1998 at the VVR-S Reactor from Bucharest, was put into operation at the Romania TRIGA Reactor. This paper presents the experimental devices developed for the ΣΣ system operating at the Steady State Reactor (SSR)-TRIGA core, including the monitoring system, and preliminary neutron characterization results. These results show that the spectrum characteristics are slowly deviating from those for the recommended ΣΣ spectrum. The high-energy neutron tail is slowly increasing, compared with the recommended ΣΣ spectrum and other similar facilities.

Criteria for the selection of dosimetry cross sections

This paper defines a process for selecting dosimetry-quality cross sections. The recommended cross-section evaluation depends on screening high-quality evaluations with quantified uncertainties, down-selecting based on comparison to experiments in standard neutron fields, and consistency checking in reference neutron fields. This procedure is illustrated for the /sup 23/Na(n,/spl gamma/)/sup 24/Na reaction.

High-Intensity Neutron Fields at the Center of the Metallic Core of Fast Reactors as the Primary Standard Neutron Fields

It is proposed to use high-intensity neutron fields with an average neutron energy of 0.7–1.5 MeV at the center of the metallic core of fast reactors as the primary standard neutron fields along with the known standard fields of fission neutrons with average energies of 1.93 and 2.13 MeV, respectively, from 235U and 252Cf nuclei, which are used in the world practice of taking neutron measurements. It is also proposed to create standard fields of nuclear fission neutrons with an average energy of ≥2 MeV by the bombardment of 235U converters with a beam of fast monoenergetic neutrons with energies of 2.5 and 14 MeV and 50-MeV protons.

Integral Test on Activation Cross Section of Tag Gas Nuclides Using Fast Neutron Spectrum Fields

Activation cross sections of xenon and krypton isotopes were evaluated as tag gases to identify capsule rupture in liquid metal cooled fast reactors. The accuracy of the activation cross sections was investigated with several tag gas samples irradiated in the standard neutron field of JOYO and YAYOI reactors. Comparing the measured radio activities and calculated values using YAYOI neutron fluence and activation cross section processed from the JENDL-3.2 cross section library yielded C/E values of approximately 0.86 to 2.6 for tag gas nuclides. The discrepancy between calculation and measurement appears due to the cross section uncertainty. This study confirmed the present accuracy of tag gas activation cross sections.

Standard neutron fields and researches on advanced nuclear measurement technique

The present status of standard neutron fields in the YAYOI reactor are reviewed. The research topics using the standard neutron fields are also introduced on the development of advanced nuclear measurement technique in application of the latest technology on lasers, optical fibers and digital signal processing.

Fast neutron source reactor, YAYOI

The characteristics of the fast neutron source reactor, YAYOI of the University of Tokyo are described. The results of major researches are summarized. Those are the studies of fast neutron shielding and neutron transport, development of standard neutron field, advanced neutron detection, measurement of decay heat, development of epithermal neutron columns for boron neutron capture therapy, on-line tritium recovery from fusion materials, fast neutron radiolysis of water, fast neutron radiography and behavior of actinides. Some of the off-pile studies are described. Those are the conceptual design study of advanced power reactors, electromagneto-mechanics, fluid dynamics of a free surface in liquid metal fast breeder reactors and development of computational techniques for nuclear engineering.

Development of methods for monitoring the quality and certification of standard samples of fissile materials

A description is given of methods for monitoring the isotopic composition of standard samples of plutonium using a standard neutron field from a BIR-2 nuclear reactor, and of a method for certifying standard samples of uranium using x-ray fluorescence.

A neutron irradiation facility for damage studies

An intermediate energy standard neutron field provided by the ITN-ΣΣ facility of IAP Bucharest Magurele has been calibrated for hardness studies. The facility is well standardized in neutron fluence and the neutron spectrum is close to the albedo neutron spectrum at high energy accelerator experiments. Special TLDG dosimetrical measurement showed that the gamma fluence is much smaller than for other similar facilities.

Standard neutron field at the center of the active zone of a BIR-2 reactor

The objective of this study was to examine several aspects of research on the BIR-2 fast pulsed reactors in order to determine the feasibility of using the neutron field at the center of the active zone for examination of the cross sections of nuclear reactions that are used for neutron spectrometry. Cross-sections for the neutron spectrum were calculated and presented, as were results for the basic cross-sections, standard fields for nuclear dosimetry, and tests of selected fission detectors. When compared with experimental data, excellent agreement was obtained.

Comparison of Measured and Calculated Bonner Sphere Responses for 24 and 144 keV Incident Neutron Energies

Calculations and experimental calibrations in standard neutron fields are needed in order to describe adequately the response function of a Bonner sphere. Measurements of response functions in reactor filtered neutron beams are intercompared with calculations, and the influence of the moderator's mass density is studied in calculations. The 3He content of the proportional counter used as the central detector has been experimentally determined and roughly confirmed by calculations. The calculational model used was found to be too simple for simulating small Bonner spheres.

Advances in high fluence solid state track recorder neutron dosimetry

Neutron dosimeters based on measurements of neutron-induced fission rates with solid state track recorders (SSTRs) provide a versatile and accurate method for reactor dosimetry applications. The high fluence upper limit of the range of the SSTR neutron dosimetry method is determined by track counting limitations at very high track densities. The recent development of techniques for the preparation and calibration of ultra low-mass fissionable deposits has extended the high-fluence limit to about 5×10 18 neutrons/cm 2 . To date, more than 900 SSTR neutron dosimeters have been exposed in high neutron fluence applications at commercial nuclear power reactors. At the same time, parallel efforts have been carried out to further extend the range of the SSTR neutron dosimetry technique, to improve the precision of the present methods, and to verify the high accuracy of the method through controlled exposures in standard neutron fields. The present status of high fluence SSTR neutron dosimetry techniques as well as recent improvements in the method will be discussed.

Uniformity in dosimetry protocols for therapeutic applications of fast neutron beams.

In this supplement to both the American and the European protocols for clinical neutron dosimetry, new recommendations are given with respect to the basic physical parameters and experimental techniques employed. For neutron dosimetry, the air kerma or exposure calibration in a photon beam is the most suitable method for the calibration of tissue-equivalent ionization chambers until calibration in a standard neutron field becomes available. More recent data are recommended for the physical parameters required for the photon calibration as well as for the measurements in the neutron beam. Water is recommended as the reference phantom material due to its similarity in absorption and scattering properties to muscle. The resulting overall change in absorbed dose calculated according to this supplement, compared with the original protocols, will be smaller than about +/- 2% due to differences in the basic physical parameters. An additional change of several percent occurs at depth in a phantom as a result of the difference between water and the muscle-equivalent liquid formerly recommended as the reference phantom material.

Verification Experiment of Threshold Detector107Ag(n, n′)107mAg on “YAYOI” Standard Neutron Field

The validity of the 107Ag(n, n′)107mAg reaction for reactor neutron dosimetry has been examined through an irradiation experiment using the standard neutron field in the “YAYOI”, a fast neutron source reactor. The test foil used was 99.0% 107Ag enriched silver, which was irradiated along with five reference foils already well evaluated. The reaction rate of 107Ag was determined by K X-ray counting with a high-purity Ge detector. The cross section for the 107Ag(n, n′) 107mAg reaction was compiled from the excitation cross section data of 107Ag published in ENDF/B-IV, in combination with the decay scheme of 107mAg listed in the “Table of Isotopes“. Using the neutron spectrum which had previously been precisely measured, it was found that the compiled activity of 107mAg was underestimated by 12.8±5%. This error is attributed to significant underestimation involved in the cross section of the 107Ag(n, n′)107mAg reaction.