Criticality Accident Dosimetry Research Articles

A review of criticality dosimetry at the Y-12 National Security Complex and practical importance of dose accuracy in emergency response

A nuclear criticality results in the emission of both neutron and gamma radiation and can produce doses to personnel near the event that exceed 0.1 Gy (10 rad). The primary purpose of nuclear accident dosimetry is to rapidly identify affected personnel in need of prompt medical treatment and to reassure personnel who have been only minimally exposed. While accurate dosimetry is desired, it must be recognized that dose determinations made from whole-body dosimeters or simple triage methods are very rough estimates and contain significant uncertainties. Even when accounting for factors like varying neutron energy spectra, mean photon energies, body orientation within the radiation field, and transient effects on dosimeter response, etc., the end value is a dosimetric quantity defined for very specific radiological conditions and determined within a simple phantom usually at a single depth. Of more importance is the biological response to the radiation, which will vary by person and can be affected by the individual's radiation sensitivity, age, gender, mass, and underlying health conditions. The overall biological, person-specific response to a given dose cannot be precisely determined except by patient symptom observation and individual biological dosimetry (e.g. chromosome analysis, lymphocyte ratios, etc.). This work describes and discusses the criticality accident dosimetry program at the Y-12 National Security Complex, a United States Department of Energy National Nuclear Security Administration facility. The primary goals of the Y-12 accident dosimetry program are, among others, the rapid identification of significantly exposed persons, prompt routing of exposed workers for medical evaluation and treatment, and the ultimate processing of dosimeters to assign doses to personnel.

A micro-gap, air-filled ionisation chamber as a detector for criticality accident dosimetry

A micro-gap air-filled ionisation chamber was designed for criticality dosimetry. The special feature of the chamber is its very small gap between electrodes of only 0.3 mm. This prevents ion recombination at high dose rates and minimises the influence of gas on secondary particles spectrum. The electrodes are made of polypropylene because of higher content of hydrogen in this material, when compared with soft tissue. The difference between neutron and gamma sensitivity in such chamber becomes practically negligible. The chamber's envelope contains two specially connected capacitors, one for polarising the electrodes and the other for collecting the ionisation charge.

Measurement of Neutron Dose under Criticality Accident Conditions at TRACY Using TLDs

Applicability of a neutron dose equivalent monitor with thermoluminescence dosimeter (TLD monitor) as an area dosimeter for criticality accidents was studied through measurements at Transient Experiment Critical Facility (TRACY). The TLD monitor is composed of two TLD badges and a cubical polyethylene case, and has a response similar to the values of the conversion coefficients for the ambient dose equivalent in the energy range from 0.1 to 10MeV. TRACY was operated with and without a water reflector when irradiating the TLD monitors. The neutron doses measured with the TLD monitors were converted into tissue kerma using dose conversion factors calculated with MCNP5. Factors to correct for the difference between responses of the TLD monitor to the spontaneous fission spectrum of a 252Cf calibration source and to spectra of TRACY were also calculated with MCNP5 and applied to the tissue kerma. The values of tissue kerma were proportional to the integrated power of TRACY. The measured tissue kerma ranged from 0.034 to 16 Gy, which covers the range from 0.1 to 10Gy specified as important in criticality accident dosimetry by the IAEA. The presented method of measurement also satisfies the time limit on dose determination.

Measurement of Neutron Dose under Criticality Accident Conditions at TRACY Using TLDs

Applicability of a neutron dose equivalent monitor with thermoluminescence dosimeter (TLD monitor) as an area dosimeter for criticality accidents was studied through measurements at Transient Experiment Critical Facility (TRACY). The TLD monitor is composed of two TLD badges and a cubical polyethylene case, and has a response similar to the values of the conversion coefficients for the ambient dose equivalent in the energy range from 0.1 to 10MeV. TRACY was operated with and without a water reflector when irradiating the TLD monitors. The neutron doses measured with the TLD monitors were converted into tissue kerma using dose conversion factors calculated with MCNP5. Factors to correct for the difference between responses of the TLD monitor to the spontaneous fission spectrum of a 252Cf calibration source and to spectra of TRACY were also calculated with MCNP5 and applied to the tissue kerma. The values of tissue kerma were proportional to the integrated power of TRACY. The measured tissue kerma ranged from 0.034 to 16 Gy, which covers the range from 0.1 to 10Gy specified as important in criticality accident dosimetry by the IAEA. The presented method of measurement also satisfies the time limit on dose determination.

TL detectors for gamma ray dose measurements in criticality accidents

Determination of gamma ray dose in mixed neutron+gamma ray fields is still a demanding task. Dosemeters used for gamma ray dosimetry are usually in some extent sensitive to neutrons and their response variations depend on neutron energy i.e., on neutron spectra. Besides, it is necessary to take into account the energy dependence of dosemeter responses to gamma rays. In this work, several types of thermoluminescent detectors (TLD) placed in different holders used for gamma ray dose determination in the mixed fields were examined. Dosemeters were from three different institutions: Ruder Bosković Institute (RBI), Croatia, JoZef Stefan Institute (JSI), Slovenia and Autoridad Regulatoria Nuclear (ARN), Argentina. All dosemeters were irradiated during the International Intercomparison of Criticality Accident Dosimetry Systems at the SILENE Reactor, Valduc, June 2002. Three accidental scenarios were reproduced and in each irradiation the dosemeters were exposed placed on the front of phantom and 'free in air'. Following types of TLDs were used: 7LiF (TLD-700), CaF2:Mn and Al2O3:Mg,Y-all from RBI; CaF2:Mn from JSI and 7LiF (TLD-700) from ARN. Reported doses were compared with the reference values as well as with the values obtained from the results of all participants. The results show satisfactory agreement with other dosimetry systems used in the Intercomparison. The influence of different types of holders and applied corrections of dosemeters' readings are discussed.

Assessment of Human Body Surface and Internal Dose Estimations in Criticality Accidents Based on Experimental and Computational Simulations

For a study on the applicability of a personal dosimetry method to criticality accident dosimetry, an assessment of the human body surface and internal dose estimations was performed by experimental and computational simulations. The experimental simulation was carried out in a criticality accident situation created at the TRACY facility. The neutron and gamma-ray absorbed doses in muscle tissue were separately estimated by a dosimeter set of an alanine dosimeter and a thermoluminescence dosimeter made of enriched lithium tetra borate with a phantom. The computational simulation was conducted with a Monte Carlo code taking account of dose components of neutrons, prompt gamma-rays and delayed gamma-rays. Prior to the assessment of the body surface and internal dose estimations, the computational simulation was ascertained to be valid by comparison between the calculated dose distributions in the phantom and the measured ones. The assessment of the body surface and internal dose estimations based on the experimental and computational simulations confirmed that the personal dosimetry using the dosimeter set provided a first estimation of the body surface and internal doses with precision enough for emergency medical care.

Criticality dosimetry using a sulfur disk and a priori neutron spectral knowledge

This paper proposes the design of a new approach to criticality accident dosimetry, which uses a priori knowledge of the neutron spectra for criticality accident conditions, and depends upon accurate detector spectral response characterization and environmental modeling. The sulfur disk was selected as a potential neutron detector, for neutrons of higher energies. Several sulfur disks were exposed to a critical Godiva experiment, the spectral response function determined using Monte Carlo methods, and the activity determined using knowledge of the known criticality spectrum. The new method, possibly with two or more different detectors to measure other energy ranges, holds promise as a practical approach to neutron criticality dosimetry.

Criticality Accident Dosimetry: An International Intercomparison at the Silene Reactor

criticality accident dosimetry, specialised techniques, which differ markedly from those used in routine radiological protection, are used to measure separately the gamma ray and neutron components of dose. In order to improve these techniques and to evaluate the performance of their dosimetry systems, fifteen laboratories participated in an international intercomparison at the SILENE reactor, Valduc, France. The physical techniques used in criticality dosimetry are first described, for personnel and installed dosemeters, and for the initial sorting of those exposed. The SILENE reactor is briefly described together with a review of the dosimetry and spectrometry techniques used to characterise the radiation field at a reference position and to study the uniformity of the field on an arc around the source. The experimental arrangements of the participants' dosemeters for the intercomparison irradiations (in free-air and on phantoms) are given, followed by a summary of the doses measured by the participants. From an overall analysis of this intercomparison, it is concluded that the participants' results are in good agreement with the reference measurements for both the neutron and gamma ray components of dose.

Criticality accident dosimetry with ESR spectroscopy

The suitability of the ESR alanine and sugar detectors for criticality accident dosimetry was experimentally investigated during an intercomparison of dosimetry techniques. Tests were performed irradiating detectors both free-in-air and on-phantom during controlled criticality excursions at the SILENE reactor in Valduc, France. Several grays of absorbed dose were imparted in neutron gamma-ray fields of various relative intensities and spectral distributions. Analysed results confirmed the potential of these systems which can immediately provide an acute dose assessment with an average underestimate of 30% in the various fields. This performance allows for the screening of severely exposed individuals and meets the IAEA recommendations on the early estimate of accident absorbed doses.

Spectrophotometric measurement of track density in CR39 for high neutron dose determinations

Abstract Due to track overlapping, visual counting and automatic counting are limited to low track densities. Conventional reading techniques cannot be used for high dose measurement. An attempt has been made to develop a track density determination method suitable for high neutron dose irradiation (ex : criticality accident dosimetry). We have found that the measurement of the optical transmission of a monochromatic light through CR39 detector is a possible method. Different kinds of CR39 (produced in Bristol, Italy, or by American Acrylics) were studied. They were irradiated in the field of the SILENE reactor and then chemically etched. The results obtained show that : • - it is possible to perform measurements from about 0.2 Gy to 100 Gy, • - a linear relation between the optical density and neutron kerma can be obtained by suitably adjusting chemical etching conditions.

Nuclear Accident Dosimetry Intercomparison Studies

Twenty-two nuclear accident dosimetry intercomparison studies utilizing the fast-pulse Health Physics Research Reactor at the Oak Ridge National Laboratory have been conducted since 1965. These studies have provided a total of 62 different organizations a forum for discussion of criticality accident dosimetry, an opportunity to test their neutron and gamma-ray dosimetry systems under a variety of simulated criticality accident conditions, and the experience of comparing results with reference dose values as well as with the measured results obtained by others making measurements under identical conditions. Sixty-nine nuclear accidents (27 with unmoderated neutron energy spectra and 42 with eight different shielded spectra) have been simulated in the studies. Neutron doses were in the 0.2-8.5 Gy range and gamma doses in the 0.1-2.0 Gy range. A total of 2,289 dose measurements (1,311 neutron, 978 gamma) were made during the intercomparisons. The primary methods of neutron dosimetry were activation foils, thermoluminescent dosimeters, and blood sodium activation. The main methods of gamma dose measurement were thermoluminescent dosimeters, radiophotoluminescent glass, and film. About 68% of the neutron measurements met the accuracy guidelines (+/- 25%) and about 52% of the gamma measurements met the accuracy criterion (+/- 20%) for accident dosimetry.

Performance Characteristics of a High-Level Solid-State Personnel Dosimetry System in Pulsed Radiation Environments

The Health Physics Research Reactor at Oak Ridge National Laboratory was used to evaluate performance characteristics of a high-level, solid-state, personnel dosimetry system developed for the US Army for the rapid measurement of total (gamma plus neutron) radiation doses to soldiers in locations where tactical nuclear weapons could be used. The gamma sensitive element of the dosemeter consists of a silver-activated phosphate glass and the neutron detector is a silicon diode. Areas considered in this evaluation included pre-irradiation dose indication; accuracy and precision of gamma, neutron, and total dose measurements; time dependence; and temperature effects. Study results indicate that for doses above about 0.4 Gy, the system can provide rapid and accurate dose estimates within accuracy limits specified by several international regulatory and standards agencies for criticality accident dosimetry. However, theoretical and observed lower limits of detection for this system are somewhat higher than the 0.25 Gy minimum dose of interest for general accident dosimetry.