Richard H. Olsher

High-energy response of the PRESCILA and WENDI-II neutron rem meters

WENDI-II was designed at the Los Alamos National Laboratory (LANL) specifically as a wide-range rem meter, suitable for applications at particle accelerators, with response extension to 5 GeV. PRESCILA was also designed at LANL, mainly as a lightweight alternative to traditional rem meters, but has shown excellent response characteristics above 20 MeV. This Note summarises measurements performed over a span of 4 y to characterise the high-energy neutron response (>20 MeV) of these meters to several hundred million electron volts. High-energy quasi-monoenergetic beams utilised as part of this study were produced by the cyclotron facilities at the Université Catholique de Louvain (33 and 60 MeV) and the T. Svedberg Laboratory ( 46, 95, 143 and 173 MeV). In addition, measurements were also conducted at the Los Alamos Neutron Science Center, 800 MeV spallation neutron source, in broad energy fields with an average energy of 345 MeV. For the sake of completeness, data collected between 2.5 and 19 MeV in monoenergetic neutron fields at the German Physikalisch-Technische Bundesanstalt (PTB) facility are also included in this study.

PRESCILA: a new, lightweight neutron rem meter.

Abstract— Conventional neutron rem meters currently in use are based on 1960’s technology that relies on a large neutron moderator assembly surrounding a thermal detector to achieve a rem-like response function over a limited energy range. Such rem meters present an ergonomic challenge, being heavy and bulky, and have caused injuries during radiation protection surveys. Another defect of traditional rem meters is a poor high-energy response above 10 MeV, which makes them unsuitable for applications at high-energy accelerator facilities. Proton Recoil Scintillator–Los Alamos (PRESCILA) was developed as a low-weight (2 kg) alternative capable of extended energy response, high sensitivity, and moderate gamma rejection. An array of ZnS(Ag) based scintillators is located inside and around a Lucite light guide, which couples the scintillation light to a sideview bialkali photomultiplier tube. The use of both fast and thermal scintillators allows the energy response function to be optimized for a wide range of operational spectra. The light guide and the borated polyethylene frame provide moderation for the thermal scintillator element. The scintillators represent greatly improved versions of the Hornyak and Stedman designs from the 1950’s, and were developed in collaboration with Eljen Technology. The inherent pulse height advantage of proton recoils over electron tracks in the phosphor grains eliminates the need for pulse shape discrimination and makes it possible to use the PRESCILA probe with standard pulse height discrimination provided by off-the-shelf health physics counters. PRESCILA prototype probes have been extensively tested at both Los Alamos and the German Bureau of Standards, Physikalisch-Technische Bundesanstalt. Test results are presented for energy response, directional dependence, linearity, sensitivity, and gamma rejection. Initial field tests have been conducted at Los Alamos and these results are also given. It is concluded that PRESCILA offers a viable, ergonomically superior, alternative to traditional rem meters that is effective for a wide range of neutron fields. The probe is capable of excellent sensitivity (40 counts per minute per μSv h−1 for 241AmBe) and extended energy response to beyond 20 MeV. Directional response is uniform (±15%) over a wide range of energies. Response linearity has been characterized to over 20 mSv h−1. Gamma rejection is effective in gamma fields up to 2 mSv h−1. The PRESCILA technology has been commercialized and is now offered under license by Ludlum Measurements, Inc.

A real-time pulsed photon dosimeter

Abstract Radiation sources producing short pulses of photon radiation are now widespread. Such sources include electron and proton linear accelerators, betatrons, synchrotrons, and field-emission impulse generators. It is often desirable to measure leakage and skyshine radiation from such sources in real time, on a single-pulse basis as low as 8.7 nGy (1 μR) per pulse. This paper describes the design and performance of a prototype, real-time, pulsed photon dosimeter (PPD) capable of single-pulse dose measurements over the range from 3.5 nGy to 3.5 μGy (0.4 to 400 μR). The PPD may also be operated in a multiple-pulse mode that integrates the dose from a train of radiation pulses over a 3-s period. A pulse repetition rate of up to 300 Hz is accommodated. The design is eminently suitable for packaging as a lightweight, portable, survey meter. The PPD uses a CdWO 4 scintillator optically coupled to a photodiode to generate a charge at the diode output. A pulse amplifier converts the charge to a voltage pulse. A digitizer circuit generates a burst of logic pulses whose number is proportional to the peak value of the voltage pulse. The digitizer output is recorded by a pulse counter and suitably displayed. A prototype PPD was built for testing and evaluation purposes. The performance of the PPD was evaluated with a variety of pulsed photon sources. The dynamic range, energy response, and response to multiple pulses were characterized. The experimental data confirm the viability of the PPD for pulsed photon dosimetry.

Personal dose equivalent conversion coefficients for neutron fluence over the energy range of 20–250 MeV

Monte Carlo simulations were performed to extend existing neutron personal dose equivalent fluence-to-dose conversion coefficients to an energy of 250 MeV. Presently, conversion coefficients, H(p,slab)(10,alpha)/Phi, are given by ICRP-74 and ICRU-57 for a range of angles of radiation incidence (alpha = 0, 15, 30, 45, 60 and 75 degrees ) in the energy range from thermal to 20 MeV. Standard practice has been to base operational dose quantity calculations <20 MeV on the kerma approximation, which assumes that charged particle secondaries are locally deposited, or at least that charged particle equilibrium exists within the tally cell volume. However, with increasing neutron energy the kerma approximation may no longer be valid for some energetic secondaries such as protons. The Los Alamos Monte Carlo radiation transport code MCNPX was used for all absorbed dose calculations. Transport models and collision-based energy deposition tallies were used for neutron energies >20 MeV. Both light and heavy ions (HIs) (carbon, nitrogen and oxygen recoil nuclei) were transported down to a lower energy limit (1 keV for light ions and 5 MeV for HIs). Track energy below the limit was assumed to be locally deposited. For neutron tracks <20 MeV, kerma factors were used to obtain absorbed dose. Results are presented for a discrete set of angles of incidence on an ICRU tissue slab phantom.

A New Neutron Dose Equivalent Meter

The design and performance of a new neutron dose equivalent meter or ``rem meter`` are described. Key design elements include a 6.38-inch diameter polyethylene moderator sphere loaded with Tungsten Carbide powder to 5% tungsten by weight, and a 2-mil thick cadmium foil located at a radius of 5 cm. A detailed response function for the new rem meter (dubbed the Model 652) was derived over the neutron energy range from thermal to 20 MeV using Monte Carlo transport techniques. Compared with the standard Eberline NRD rem mete, the new design is about 1.5 kg lighter, and is 28% more sensitive per unit dose of Cf-252. It is shown that for a large variety of standard and operational neutron spectra, the model 652 rem meter provides a response accuracy of {plus_minus}15%.

A TECHNICAL BASIS FOR THE CONTINUED USE OF EXPOSURE AS AN ACCEPTABLE OPERATIONAL QUANTITY IN RADIATION PROTECTION

Within the tabulated values of the new [to U.S. Department of Energy (DOE)] radiation weighting factors, it can be seen that a doubling of the neutron factor occurs for the 0.1 to 2 MeV neutron energy range. Hence, with the effective replacement of the quality factor by these new radiation weighting factors (for the protection quantities), it has been widely understood that the new changes will most definitely impact neutron dosimetry. However, it is less well understood that the new changes could also affect photon (and beta) dosimetry, i.e., photon reference fields, instrument design, and instrument calibrations. This paper discusses the ramifications, and ultimately concludes that the use of exposure for workplace measurements complies with both current and amended DOE requirements.

A 3He counter version of the Thermo Fisher Scientific NRD neutron rem meter.

Thermo Fisher Scientific’s NRD rem meter has been in production for almost 40 y and is the primary rem meter in use at many U.S. Department of Energy facilities. An upgrade project was initiated at the Los Alamos National Laboratory with the primary goal of increasing the NRD’s neutron sensitivity through the substitution of pressurized 3He gas (4 atmospheres) for the stock counter tube’s BF3 fill gas. Historically, BF3 counters were far less expensive relative to 3He and were usually chosen on the basis of cost. That is no longer the case, with pricing for both types of counters being similar. Test results have shown that the 3He counter version of the NRD exhibits stable operation at a reasonable bias voltage and good gamma rejection. Sensitivity has been increased by about a factor of four with no penalty in cost.

USE OF TRANSPORTABLE RADIATION DETECTION INSTRUMENTS TO ASSESS INTERNAL CONTAMINATION FROM INTAKES OF RADIONUCLIDES PART II: CALIBRATION FACTORS AND ICAT COMPUTER PROGRAM

AbstractThe detonation of a radiological dispersion device or other radiological incidents could result in widespread releases of radioactive materials and intakes of radionuclides by affected individuals. Transportable radiation monitoring instruments could be used to measure radiation from gamma-emitting radionuclides in the body for triaging individuals and assigning priorities to their bioassay samples for in vitro assessments. The present study derived sets of calibration factors for four instruments: the Ludlum Model 44‐2 gamma scintillator, a survey meter containing a 2.54 × 2.54-cm NaI(Tl) crystal; the Captus 3000 thyroid uptake probe, which contains a 5.08 × 5.08‐cm NaI(Tl) crystal; the Transportable Portal Monitor Model TPM-903B, which contains two 3.81 × 7.62 × 182.9‐cm polyvinyltoluene plastic scintillators; and a generic instrument, such as an ionization chamber, that measures exposure rates. The calibration factors enable these instruments to be used for assessing inhaled or ingested intakes of any of four radionuclides: 60Co, 131I, 137Cs, and 192Ir. The derivations used biokinetic models embodied in the DCAL computer software system developed by the Oak Ridge National Laboratory and Monte Carlo simulations using the MCNPX radiation transport code. The three physical instruments were represented by MCNP models that were developed previously. The affected individuals comprised children of five ages who were represented by the revised Oak Ridge National Laboratory pediatric phantoms, and adult men and adult women represented by the Adult Reference Computational Phantoms described in Publication 110 of the International Commission on Radiological Protection. These calibration factors can be used to calculate intakes; the intakes can be converted to committed doses by the use of tabulated dose coefficients. These calibration factors also constitute input data to the ICAT computer program, an interactive Microsoft Windows-based software package that estimates intakes of radionuclides and cumulative and committed effective doses, based on measurements made with these instruments. This program constitutes a convenient tool for assessing intakes and doses without consulting tabulated calibration factors and dose coefficients.

Computer-Based Investigative Techniques: A Comparison of Dose Using the MCNP Code for Optically Stimulated Light Dosimeters

Abstract A computer-based investigative technique, using the Los Alamos Monte Carlo code MCNP5 version 1.51 (Radiation Safety Information Computational Center), was completed to assess the shallow dose equivalent (SDE) reported on the Landauer, Inc.,™ Luxel+ optically stimulated light (OSL) dosimeter. Experimental test irradiations were conducted on 18 OSL dosimeters through various controlled exposures to the source (10 mCi 90Sr). The reported SDE for each test irradiation was compared to the results for SDE calculated using MCNP5. All test irradiation experiments were conducted with the 90Sr source placed in direct contact with the dosimeter with slight placement changes across the dosimeter face. It was found that these slight adjustments caused vast differences in reported doses by Landauer. The SDE determined in a tissue matrix using MCNP5 was studied for two of the dosimeter badge geometries, and it was found that some qualitative agreement exists between the reported and simulated doses in contradiction with the experimental results. Further simulated analysis was not conducted because precise source-dosimeter geometries and the algorithm used by Landauer to analyze its Luxel+ OSL dosimeters were not known. These results indicate that a future study should be conducted with more rigorous simulated benchmarking to verify these results.

Photon energy response of an aluminum oxide TLD environmental dosimeter

Because of aluminum oxides significant advantage in sensitivity over LiF, minimal fading characteristics, and ease of processing, aluminum oxide thermoluminescent dosimeters (TLDs) are being phased in at Los Alamos for environmental monitoring of photon radiation. The new environmental dosimeter design consists of a polyethylene holder, about 0.5 cm thick, loaded with a stack of four aluminum oxide TLD chips, each 1 mm thick and 5 mm in diameter. As part of the initial evaluation of the new design, the photon energy response of the dosimeter was calculated over the range from 10 keV to 1 MeV. Specific goals of the analysis included the determination of individual chip response in the stack, assessment of the response variation due to TLD material (i.e., LiF versus Al/sub 2/O/sub 3/), and the effect of copper filtration in flattening the response.<<ETX>>