Mixed Neutron-gamma Radiation Research Articles

СЛИЧЕНИЕ ДОЗИМЕТРИЧЕСКИХ СИСТЕМ ФОТОННОГО И НЕЙТРОННОГО ИЗЛУЧЕНИЙ, ИСПОЛЬЗУЕМЫХ В ОРГАНИЗАЦИЯХ ГОСКОРПОРАЦИИ «РОСАТОМ» ДЛЯ КОНТРОЛЯ В СИТУАЦИИ ПЛАНИРУЕМОГО ОБЛУЧЕНИЯ

Purpose: Using the example of reviewing and discussing the results of comparison tests of dosimetric systems of gamma and neutron radiation used in organizations of the State Corporation “Rosatom”, to assess the current state of reliability of monitoring the planned exposure of workers in fields of mixed gamma-neutron radiation using the considered dosimetric systems in order to produce recommendations for corrective actions to ensure a unified approach to conduct individual dosimetric control of external exposure. Results: All measuring instruments for individual dose equivalent of photon and neutron radiation, presented in comparison tests, comply with up-to-date requirements for individual dosimetric control systems. All measuring instruments confirmed their measuring capabilities, showed satisfactory quality of measurement results and the absence of a systematic bias in the measurement results. Analysis of the results of measuring the individual dose equivalent of neutron radiation showed that problems affecting the quality of the results obtained were identified in the considered instruments of measuring personal dose equivalent. The following factors might be the sources of problems: lack of knowledge about the real characteristics of radiation fields (spectral characteristics, radiation direction, etc.) at workplaces; insufficient research of the method used for measuring neutron radiation under real conditions (technical and metrological characteristics and features of the individual dosimeters used); failure to take into account the weighing coefficients for neutrons of various energies when measuring instruments are calibrated and when real measurements are conducted. Conclusion: It is necessary to organize and conduct investigations of the metrological characteristics of the measuring instrument that are used under conditions typical for a specific radiation object. After finishing these experimental studies, it is recommended to test the methodology with an analysis of the compliance of the accuracy indicators with the requirements of the relevant guidelines. In order to solve the problem of a lack of knowledge about the real characteristics of radiation fields, radiation safety services of organizations are recommended to organize and conduct research aimed at studying such characteristics using radiometric and spectrometric methods, experimental modeling of the process of personnel exposure using anthropomorphic phantoms and determining correction factors for the individual dosimeters used.

Modelling the response of CLLBC(Ce) and TLYC(Ce) SiPM-based radiation detectors in mixed radiation fields with Geant4

CLLBC(Ce) and TLYC(Ce) are novel scintillation materials capable of measuring mixed gamma ray and neutron radiation fields that have gained significant interest in the areas of space and nuclear safety/security science. To date Geant4, the world’s most popular Monte Carlo radiation modelling toolkit, has yet to be effectively used to simulate the full response of these materials when coupled to near ultra-violet Silicon PhotoMultipliers (SiPMs). In this work an experimentally validated Geant4 application has been developed with optimised material composition, optical data tables, and physics transport settings that is able to accurately simulate the response of CLLBC(Ce) and TLYC(Ce) SiPM-based radiation detectors under both gamma ray and neutron irradiation. Experimental benchmarking for five different radioactive sources (Co-60, Cs-137, Eu-152, Am-241, and Cf-252) illustrated that this developed Geant4 application was able to reproduce the position and structure of all major spectral features (full energy gamma ray photo-/neutron capture peaks, X-ray escape photopeaks, Compton edge, Compton backscatter peaks, and Compton plateau) to high level of accuracy.

A new Multi Leaf Collimator for the fission neutron therapy facility at FRM II

As a part of the general upgrade program of the fission neutron therapy facility MEDAPP at the research reactor FRM II, a newly designed, electronically operated Multi Leaf Collimator (MLC) was introduced. The design of the new collimator was optimized in order to increase dose output and to allow patient positioning closer to the collimator. The new collimator allows divergent and convergent beam shaping in order to maximize the effective area of the fission source used for irradiation. In comparison to the old MLC, the radiation dose for the staff is reduced by the motorized operation of the new MLC. The advanced beam shaping abilities of the new collimator are investigated by a comparison with depth dose curves and beam profiles of the old MLC. For the separation of neutron and gamma dose deposition in water, the two chamber dosimetry method recommended for mixed field dosimetry was applied. A re-calibration of the set of ionization chambers used for dosimetry in mixed neutron-gamma radiation fields at FRM II was performed. The calibration was done in a reference neutron field from 3H(p,n)3He-reactions in a low scattering environment. An overall increased dose output as intended by the design of the new MLC was shown. Also, improved dosimetric characteristics of the treatment field due to the optimized beam shaping abilities of the new MLC were demonstrated. In the calibration measurements of the magnesium ionization chamber used in the two chamber method, an unexpected increase of the neutron sensitivity by a factor of 8 was measured. The new MLC will enable more conformal dose delivery to the target volume in patients and therefore help to reduce healthy tissue damage by the application of neutron radiation.

Neutron spectrometry and dosimetry using a multi-shell Single Sphere Neutron Spectrometer with thermo-luminescent and optically stimulated luminescent detectors

A portable neutron spectrometer with concentric multi-shell moderating assembly and TL, OSL detectors has been validated with standard neutron reference fields to generate the neutron spectra and estimate the ambient dose equivalents, H*(10). Further, the system has been used in the mixed neutron–gamma radiation fields at accelerator facility. The neutron fluence and the ambient dose equivalent estimates were verified using the standard reference neutron fields, viz. 241Am-Be, 252Cf, D-T neutron sources. The system was then employed in the accelerator radiation environment to measure the neutron spectral yield from the interactions of 15 and 20 MeV protons with a thick natural Ta target. The results were compared with those obtained from FLUKA simulations. A conventional neutron dose equivalent meter was also used to verify the ambient dose equivalent estimated from the spectrometric measurements. The reproducibility of the neutron spectra and dose values were found to be satisfactory, thereby encouraging the implementation of the portable single sphere neutron spectrometer in complex radiation environment around particle accelerators. Present studies demonstrate the measurements with the in-house built spectrometer for neutron energies up to 20 MeV.

Shielding Capability Research on Composite Base Materials in Hybrid Neutron-Gamma Mixed Radiation Fields

The 16N monitoring system operates in a mixed neutron-gamma radiation field and is subject to high background radiation, thus triggering instability in the 16N monitoring system measurement data. Due to its property of actual physical process simulation, the Monte Carlo method was adopted to establish the model of the 16N monitoring system and design a structure-functionally integrated shield to realize neutron-gamma mixed radiation shielding. First, the optimal shielding layer with a thickness of 4 cm was determined in this working environment, which had a significant shielding effect on the background radiation and improved the measurement of the characteristic energy spectrum and the shielding effect on neutrons was better than gamma shielding with the increase in the shield thickness. Then, functional fillers such as B, Gd, W, and Pb were added to the matrix to compare the shielding rates of three matrix materials of polyethylene, epoxy resin, and 6061 aluminum alloy at 1 MeV neutron and gamma energy. The shielding performance of epoxy resin as the matrix material was better than that of the aluminum alloy and polyethylene, and the shielding rate of boron-containing epoxy resin was 44.8%. The γ-ray mass attenuation coefficients of lead and tungsten in the three matrix materials were simulated to determine the best material for the gamma shielding performance. Finally, the optimal materials for neutron shielding and gamma shielding were combined, and the shielding performance of single-layer shielding and double-layer shielding in mixed radiation field was compared. The optimal shielding material-boron-containing epoxy resin was determined as the shielding layer of the 16N monitoring system to realize the integration of structure and function, which provides a theoretical basis for the selection of shielding materials in a special working environment.

Use of the chemical Fricke dosimeter and its modifications for dosimetry of gamma neutron radiation of a pulsed reactor

The paper investigates the characteristics of the chemical Fricke dosimeter (with the standard composition (D1), without NaCl addition to the solution (D2), without NaCl but with a tenfold increased concentration of Fe2+ (D3)) under continuous and pulsed irradiation with an ultra-high dose rate of the BARS-6 reactor with unshielded metallic cores. The dosimeter radiosensitivity had a linear dependence on the gamma neutron radiation dose in a range of 25 to 750 Gy and was respectively 1.96 ± 0.05 μGy–1 (D1), 2.04 ± 0.05 μGy–1 (D2), and 2.08 ± 0.5 μGy–1 (D3) in the continuous irradiation mode, and 1.24 ± 0.05 μGy–1, 2.00 ± 0.05 μGy–1, and 1.94 ± 0.05 μGy–1 in the pulsed irradiation mode. This makes ≈ 60% of their sensitivity to the 60Со gamma radiation (3.40 ± 0.02 μGy–1), and 36%, 1.6 times as less, for a standard Fricke dosimeter irradiated in the pulsed mode. The experimental value of the radiation chemical yield, Gn(Fe3+), for all solution modifications and both irradiation modes varied slightly and was 0.84 ± 0.11 μM/J on the average, except for the standard solution in the pulsed mode (0.66 ± 0.07 μM/J). The neutron doses determined by chemical and activation dosimeters coincided within the error limits, but the chemical dosimeter readings were systematically higher, by about 20%. Therefore, in the fission spectrum neutron dose rate range of 0.4 to 7×108 Gy/min, there is no dose rate effect both in the standard Fricke dosimeter version (without NaCl) and in the modified version, which makes it possible to use modified Fricke dosimeters to assess the physical and dosimetry characteristics of mixed gamma neutron radiation beams.

Пероральный термолюминесцентный дозиметр нейтронного излучения для условий аварийного облучения

An oral dosimeter of mixed gamma-neutron radiation for emergency exposure conditions has been developed. The energy dependence of the neutron radiation dosimeter sensitivity is close to the energy dependence of the specific effective dose per unit flux density. For neutron fields containing a significant contribution of fast neutrons, the uncertainty of the dosimeter readings is no more than 25% for the anteroposterior radiation geometry and no more than 35% for the rotation geometry. In neutron fields with a predominance of particles with thermal and intermediate energies, the dosimeter overestimates the effective radiation dose by 2.5 times for the anteroposterior geometry and 3.3 times for the rotation geometry. A staging experiment was carried out, which included placing individual dosimeters inside a canister simulating the torso of a standard adult in a neutron radiation field. The conditionally true values of the effective dose were obtained using the energy and angular distribution of the neutron radiation flux density. Differences in the dosimeter readings and the conditionally true value of the effective dose do not exceed 2.

Certification of a specialized source of gamma radiation spatially combined with modeling reference field of neutron radiation

The results of metrological studies of the field of the experimental channel of the research nuclear reactor OR (experimental base of the NRC “Kurchatov Institute”) with a capacity of 0.1–300 kW are presented. The empirical dependence of the contribution of the absorbed dose of gamma radiation during the raising and lowering reactor power in the total absorbed dose during the irradiation in the reactor and absorbed dose rate from a power reactor OR in the steady state. Based on these results, the absorbed dose and absorbed dose rate of gamma radiation can be pre-calculated. According to metrological studies results, the experimental channel of the reactor OR was certified in terms of the absorbed dose rate of gamma radiation as a specialized source of gamma radiation spatially combined with the modeling reference field of neutrons. The specialized source is designed for calibration of special-purpose measuring instruments and testing for radiation resistance of products operating in conditions of mixed gamma-neutron radiation of nuclear reactors.

Improving Safety of Shielding Chambers Equipped with Viewing Windows

The article describes and proves the possibility of increasing the protection against neutron radiation for personnel by means of immersion liquid without major changes in the design of the equipment used and without reducing the safety of the radiation object. Studies of the components properties of the protection glass and the immersion liquid have shown that the properties of the immersion liquid reduce the dose rate of neutron radiation significantly compared to the protection glass, which is used for the main protection against ionizing radiation when working with uranium-plutonium fuels. The use of two-layer protection in viewing windows allows reducing the dose rate of mixed gamma-neutron radiation to values that do not exceed the safe radiation dose for personnel when working with an ionizing radiation source, which in this case is 50 kg of uranium-plutonium fuel

A mm3 Fiber-Coupled Scintillator for In-Core Thermal Neutron Detection in CROCUS

This article presents the development, full-characterization, and in-core testing of a miniature neutron detector for the investigation of highly localized in-core neutronics effects in the zero-power reactor CROCUS, operated at the Ecole Polytechnique federale de Lausanne (EPFL), Switzerland. A ZnS: 6LiF screen, mixing inorganic scintillator and neutron converter, with a surface of 1 mm2 and a thickness of 0.2 mm, was coupled with a silicon photomultiplier (SiPM) via a 10-m optical fiber. Analog readout electronics were adopted in this prototype version to process the SiPM output signal. A first campaign was carried out to assess the detection system capabilities in terms of neutron detection and counting in mixed neutron-gamma radiation fields. The miniature detector was thoroughly characterized with a Pu-Be source installed in the CARROUSEL facility and was subsequently tested inside a control rod guide tube of the CROCUS reactor in different reactor conditions, i.e., at shutdown, startup, and critical. The detector shows a linear response with a reactor power increase up to 6.5 W (i.e., around $10 {^{\mathrm{ 8}}}\,\,\text {cm}^{-2}\cdot \text {s}^{-1} $ total neutron flux) and excellent neutron counting capabilities, when compared to other localized detection systems available in CROCUS, such as the miniature fission chambers and single-crystal vapor deposited (sCVD) diamond detectors. In addition, the exposure of the miniature detector to a 60Co source with an activity of ~250 GBq, combined with the computation of the gamma flux in CROCUS with the Serpent 2 Monte Carlo code, confirmed the gamma insensitivity of the system in a mixed neutron-gamma field.

Effects of neutron–gamma radiation on the free radical contents in epoxy resin: upconversion luminescence and structural stabilization

The purpose of this work is to study the effect of mixed neutron–gamma radiation (MNGR) on structural, optical and paramagnetic properties on epoxy resin nanostructure thin films [ER]NSTF. These films were prepared and irradiated in the nuclear reactor with dose range between 100 and 900 Gy. After exposure to MNGR, FT-IR results demonstrated the broad ν(OH) and ν(C=O) band to increase with increase in radiation dose, while the ν(C–H) band decreased with increase in radiation dose. This is due to the formation of chain stretch, bending and scission/cross-linking through MNGR that lead to a decrease in the crystallinity region contained within the [ER]NSTF. XRD data confirm these results and show that the crystalline phase was destroyed by MNGR irradiation. PL is used to determine the fundamental transition and defects transitions in [ER]NSTF at 300 K. It was found that MNGR leads to significant enhancement of the luminescence properties. Samples of [ER]NSTF irradiated with 600 Gy dose presented the highest response. Therefore, EPR results confirmed that MNGR induces the paramagnetic center’s formation of the [ER]NSTF whose concentration varies differently.

Design, fabrication and modeling of an AmBe neutron irradiator for TLD screening for neutron dose measurement in mixed radiation fields

TLDs dosimeters are frequently presented as a viable choice for dosimetric studies when dealing with mixed neutron-gamma radiation fields. However, this choice is not without some drawbacks, because not only TLD response is highly dependent on particle type but also on neutron energy spectrum. Therefore, a correct screening and calibration of the dosimeter are required, and a simple shift from gamma screening methodology for mixed field is not suitable. This paper presents the design, fabrication and tests of an irradiator for TLD screening for neutron dose measurement using an AmBe source and polyethylene as moderator material. The design of the irradiator was conducted through Monte Carlo simulations using the MCNP5 code. The experimental validation and tests were performed using Indium activation foils and TLD 600 dosimeters. The manufactured irradiator demonstrated to be suitable for TLD screening under neutron source radiation field, offering very good homogeneity conditions in the radiation field so to guarantee same radiation dose delivered to the TLDs.

Investigation of thermal neutron detection capability of a CdZnTe detector in a mixed gamma-neutron radiation field

Abstract The aim of this study was to investigate the thermal neutron measurement capability of a CdZnTe detector irradiated in a mixed gamma-neutron radiation field. A CdZnTe detector was irradiated in one of the irradiation tubes of a 241Am-Be source unit to determine the sensitivity factors of the detector in terms of peak count rate (counts per second [cps]) per neutron flux (in square centimeters per second) [cps/neutron·cm−2·s−1]. The CdZnTe detector was covered in a 1-mm-thick cadmium (Cd) cylindrical box to completely absorb incoming thermal neutrons via 113Cd(n,γ) capture reactions. To achieve, this Cd-covered CdZnTe detector was placed in a well-thermalized neutron field (f-ratio = 50.9 ± 1.3) in the irradiation tube of the 241Am-Be neutron source. The gamma-ray spectra were acquired, and the most intense gamma-ray peak at 558 keV (0.74 γ/n) was evaluated to estimate the thermal neutron flux. The epithermal component was also estimated from the bare CdZnTe detector irradiation because the epithermal neutron cutoff energy is about 0.55 eV at the 1-mm-thick Cd filter. A high-density polyethylene moderating cylinder box can also be fitted into the Cd filter box to enhance thermal sensitivity because of moderation of the epithermal neutron component. Neutron detection sensitivity was determined from the measured count rates from the 558 keV photopeak, using the measured neutron fluxes at different irradiation positions. The results indicate that the CdZnTe detector can serve as a neutron detector in mixed gamma-neutron radiation fields, such as reactors, neutron generators, linear accelerators, and isotopic neutron sources. New thermal neutron filters, such as Gd and Tb foils, can be tested instead of the Cd filter due to its serious gamma-shielding effect.

LiCaAlF6 scintillators in neutron and gamma radiation fields

Intentionally doped LiCaAlF6 (LiCAF) single crystals are prospective scintillators, especially for thermal neutron detection through the 6Li(n,t)4He nuclear reaction. Four different LiCAF scintillator samples were tested in various neutron and gamma fields. Two of the tested samples were LiCAF:Eu and LiCAF:Eu,Na single crystals, and another two samples were made of LiCAF:Eu micro crystals dispersed in transparent rubber, with different rubber dimensions. All LiCAF samples contain lithium enriched to[Formula: see text]Li. A plutonium–beryllium source was used as a neutron source. The neutron spectrum was modified by moderator and filter to get different ratios between thermal, epithermal and fast neutron fluence rates. The MCNP code was used for calculations of the fluence rates for different configurations. Radionuclides [Formula: see text]Cs and [Formula: see text]Co were applied as gamma radiation sources. The light signal from the scintillator was evaluated with a photomultiplier and a multichannel analyzer. The purpose of this work was to study the characteristics of LiCAF scintillators, especially the ability to discriminate signals from neutron and gamma radiation, which is the basic scintillator condition for neutron detection in mixed neutron-gamma radiation fields. Generally, the discrimination can be done by the pulse height and/or the pulse shape of the evaluated signals. Both methods can be used for a LiCAF scintillator. However, only the pulse height discrimination method is discussed in this paper. The possibility of fast neutron detection with LiCAF scintillators was also tested.

DOSE-DEPENDENT ANALYSIS OF ACUTE MEDICAL EFFECTS OF MIXED NEUTRON-GAMMA RADIATION FROM SELECTED SEVERE 235U OR 239Pu CRITICALITY ACCIDENTS IN USSR, UNITED STATES, AND ARGENTINA

Eight of the most severe cases of acute radiation disease (ARS) known to have occurred in humans (as the result of criticality accidents) had survival times less than 120 h (herein defined as "early death"). These accidents were analyzed and are discussed with respect to the specific accident scenarios and the resulting accident-specific, mixed neutron-gamma radiation clinical dose distributions. This analysis concludes that the cardiovascular system appears to be the most critical organ system failure for causing "early death" following approximate total body, mixed gamma-neutron radiation doses greater than 40-50 Gy. The clinical data also suggest that there was definite chest dose dependence in the resulting survival times for these eight workers, who unfortunately suffered profound radiation injury and unusual clinical effects from such high dose radiation exposures. In addition, "toxemic syndrome" is correlated with the irradiation of large volumes of soft tissues. Doses to the hands or legs greater than 80-100 Gy or radiation lung injury also play significant but secondary roles in causing "early death" in accidents delivering chest doses greater than 50 Gy.

Neutron-Gamma Pulse Shape Discrimination with a NE-213 Liquid Scintillator by Using Digital Signal Processing Combined with Similarity Method

Neutron-Gamma Pulse Shape Discrimination with a NE-213 Liquid Scintillator by Using Digital Signal Processing Combined with Similarity Method. Measurement of mixed neutron-gamma radiation is difficult because a neuclear detector is usually sensitive to both radiations. A new attempt of neutron-gamma pulse shape discrimination for a NE-213 liquid scintillator is presented by using digital signal processing combined with an off-line similarity method. The output pulse shapes are digitized with a high speed digital oscilloscope. The n-γ discrimination is done by calculating the index of each pulse shape, which is determined by the similarity method, and then fusing it with its corresponding pulse height. Preliminary results demonstrate good separation of neutron and gamma-ray signals from a NE-213 scintillator with a simple digital system. The results were better than those with a conventional rise time method. Figure of Merit is used to determine the quality of discrimination. The figure of merit of the discrimination using digital signal processing combined with of line similarity method are 1.9; 1.7; 1.1; 1.1; and 0.8 ; on the other hand by using conventional method the rise time are 0.9; 0.9; 0.9; 0.7; and 0.4 for the equivalent electron energy of 800 ; 278 ; 139 ; 69 ; and 30 keV.

Dosimetry in mixed gamma–neutron radiation fields and energy compensation filters for CaF 2:Tm TL detectors

Strong fields of thermal and epithermal neutrons as used for Boron Neutron Capture Therapy (BNCT) show a relatively small gamma dose. However, for accurate dose estimation, both the neutron as well as the gamma component has to be considered. Different approaches have been studied with various thermoluminescence detectors (TLDs). As a result, CaF 2:Tm detectors (TLD-300) are attributed to be most insensitive to thermal neutrons but not tissue-equivalent. In this paper we focus on the measurement of the gamma dose with TLD-300 in mixed irradiation fields. To overcome the disadvantage of being not tissue-equivalent, a study on energy compensation filters was performed. The paper comprises also calibrations of TLD-300 using energy compensation filters at the PTB-reference field for thermal neutrons, and first test experiments at the TRIGA Mainz including simulations.

A Novel Detector for Measuring Gamma-Ray Fluxes in a Mixed Pulsed Neutron-Gamma Radiation Fields

An ICI (insulator-conductor-insulator) gamma-ray detector designed for measurement of fast rising, high intensity pulsed gamma-ray fluxes has been developed. Like vacuum Compton diode (VCD) and dielectric Compton diode (DCD) devices, the ICI detector operates by utilization of the Compton effect. It has a very fast time response (rise time is less than 1 ns), a large linearity and a wide dynamic range. It is designed to measure intense and rapidly changing gamma-ray fluxes in highly mixed neutron-gamma radiation fields. Its sensitivity to 1.25 MeV gamma rays is about 1.8 x 10-20, for the 50 mm collimation diameter, which is relatively high compared with VCD and DCD devices. The simulated gamma-ray sensitivity agrees well with the experimental value. The gamma-to-neutron sensitivity ratio is about three orders of magnitudes for neutron energies below 14.1 MeV and 1.25 MeV gamma rays. Compared to existing VCD and DCD devices, the ICI detector requires no vacuum and power supply in operation, and also minimally attenuates a collimated gamma-ray beam due to its overall thickness of 3 mm in beam path.

Detection system built from commercial integrated circuits for real-time measurement of radiation dose and quality using the variance method

A small, specialised amplifier using commercial integrated circuits (ICs) was developed to measure radiation dose and quality in real time using a microdosimetric ion chamber and the variance method. The charges from a microdosimetric ion chamber, operated in the current mode, were repeatedly collected for a fixed period of time for 20 cycles of 100 integrations, and processed by this specialised amplifier to produce signal pulse heights between 0 and 10 V. These signals were recorded by a multi-channel analyser coupled to a computer. FORTRAN programs were written to calculate the dose and dose variance. The dose variance produced in the ion chamber is a microdosimetric measure of radiation quality. Benchmark measurements of different brands of ICs were conducted. Results demonstrate that this specialised amplifier is capable of distinguishing differences of radiation quality in various high-dose-rate radiation fields including X rays, gamma rays and mixed neutron-gamma radiation from the research reactor at Texas A&M University Nuclear Science Center.

Intercomparison on Measurements of the Quantity Personal Dose Equivalent Hp(d) in Mixed (Neutron-Gamma) Fields--an IAEA Project

Within its occupational radiation protection programme, the International Atomic Energy Agency (IAEA) initiated and funded an international intercomparison exercise of personal dosemeters to determine the quantity personal dose equivalent in mixed neutron-photon radiation. The objectives of the intercomparison are to assess the capabilities of the dosimetry services in measuring the quantity Hp(10) in mixed neutron-gamma fields; to assist IAEA member states in achieving sufficiently accurate dosimetry; and, if necessary, to provide guidelines for improvements (not simply a test of the performance of the existing dosimetry service). The intercomparison is directed to passive dosemeters to determine, in mixed neutron-gamma radiation fields, either these two components separately or the total personal dose equivalent. The intercomparison consists of two phases: Phase I--Type-test intercomparison: irradiation in selected calibration fields and results used to improve dosimetric procedures of participating laboratories, where needed. Phase II--Simulated workplace field intercomparison: irradiation in radiation fields similar to those in workplaces as a final check of performance. The exercise revealed clear deficiencies in the methodology used by several laboratories and necessitated a detailed analysis of the existing discrepancies. This papers summaries the finding and conclusions for radiation fields similar to those found in nuclear industry.