Photon Dosemeters Research Articles

RESPONSE OF CURRENT AREA DOSEMETERS TO NEW OPERATIONAL QUANTITIES.

New operational quantities for external radiation exposure presented by the ICRU brought changes, which can affect the use of current dosemeters. Dosemeters calibrated in the old quantities should be tested, and the results obtained should be analyzed with respect to the new quantities. The difference between the old and new quantities depends primarily on photon energy. This work concerns instruments for area monitoring, which were exposed to reference photon and neutron spectra of mean energies from 65 to 1250keV and from 0.02 to 10MeV, respectively. As expected, it was revealed that the current photon dosemeters overestimate the new quantity ambient dose. For the measurements within the energy interval chosen, it seems to be acceptable to implement a correction factor to optimize the response. For the purposes of measurements of photons of lower energies, further research would be needed. To adapt the response of the neutron dosemeter tested, recalibration or redesign can be considered, but no fundamental changes seem to be necessary for the investigated spectra.

Peripheral Organ Equivalent Dose Estimation Procedure in Proton Therapy.

The aim of this work is to present a reproducible methodology for the evaluation of total equivalent doses in organs during proton therapy facilities. The methodology is based on measuring the dose equivalent in representative locations inside an anthropomorphic phantom where photon and neutron dosimeters were inserted. The Monte Carlo simulation was needed for obtaining neutron energy distribution inside the phantom. The methodology was implemented for a head irradiation case in the passive proton beam of iThemba Labs (South Africa). Thermoluminescent dosimeter (TLD)-600 and TLD-700 pairs were used as dosimeters inside the phantom and GEANT code for simulations. In addition, Bonner sphere spectrometry was performed inside the treatment room to obtain the neutron spectra, some relevant neutron dosimetric quantities per treatment Gy, and a percentual distribution of neutron fluence and ambient dose equivalent in four energy groups, at two locations. The neutron spectrum at one of those locations was also simulated so that a reasonable agreement between simulation and measurement allowed a validation of the simulation. Results showed that the total out-of-field dose equivalent inside the phantom ranged from 1.4 to 0.28 mSv/Gy, mainly due to the neutron contribution and with a small contribution from photons, 10% on average. The order of magnitude of the equivalent dose in organs was similar, displaying a slow reduction in values as the organ is farther from the target volume. These values were in agreement with those found by other authors in other passive beam facilities under similar irradiation and measurement conditions.

Organ Dose Reconstruction Applicable for a Japanese Nuclear Worker Cohort: J-EPISODE.

An evaluation of cancer risk based on organ-absorbed dose is underway for the Japanese Epidemiological Study on Low-Dose Radiation Effects (J-EPISODE), which has analyzed health effects in association with radiation exposure evaluated with the personal dose equivalent Hp(10). Although the concept of effective dose and its operational definition of Hp(10) are widely used for radiological protection purposes, effective dose is not recommended for epidemiological evaluation. Organ-absorbed dose was instead adopted for the IARC 15-Country Collaborative study (15-Country study), the International Nuclear Workers Study (INWORKS), the Mayak worker study, and the Life Span Study (LSS) of atomic bomb survivors. The reconstruction method in J-EPISODE followed in principle the approach adopted in the 15-Country Study. As part of the approach of J-EPISODE, a conversion factor from photon dosimeter reading to air kerma was developed using dosimeter response data, which were measured by the experiment using an anthropomorphic phantom, and it was confirmed that the 15-Country study’s assumption of photon energy and geometry distribution in a work environment applied to Japanese nuclear workers. This article focuses on a method for reconstructing the conversion factor from photon dosimeter reading to organ-absorbed photon dose for a Japanese nuclear worker cohort. The model for estimating the conversion factor was defined under the assumption of a lognormal distribution from three concerned bias factors: (1) a dosimeter reading per air kerma, i.e., dosimeter response; (2) an organ-absorbed dose per air kerma; and (3) a factor relating to the differences in dose concepts and calibration practices between the roentgen dosimeter era and the present. Dosimeter response data were cited from the companion paper. Data on organ-absorbed photon dose per air kerma were estimated using a voxel phantom with the average Japanese adult male height and weight. The bias factor for the recorded dose in the roentgen era was defined, considering the backscatter radiation from the human body. The estimated values of organ-absorbed photon dose per air kerma were almost the same as those in ICRP Publication 116, revealing that the effect of differences in body size was almost negligible. The conversion factors from dosimeter reading to organ-absorbed dose were estimated by period (the roentgen era or from then), nuclear facility type (nuclear power plant or other), dosimeter type, and tissue or organ. The estimated conversion factors ranged from 0.7 to 0.9 (Gy Sv−1). The estimated cumulative organ-absorbed photon dose for the participants of J-EPISODE demonstrated that organ-absorbed dose values were approximately 0.8 times the recorded doses if neglecting dose-unit differences. J-EPISODE reconstructed an organ-absorbed dose conversion factor and will evaluate the risk of cancer mortality and morbidity using the organ-absorbed dose in the future.

Preparedness intercomparison of passive H*(10) area photon dosemeters in 2017/2018 (IC2017prep)

This intercomparison serves to investigate the long-term behaviour of passive H*(10) dosemeters which may be used in the aftermath of a radiological or nuclear event. In routine operation, such dosemeters are generally used to monitor installations like nuclear power plants and accelerators. Such dosemeters are used in the radiation field of the natural ambient radiation, including terrestrial and secondary cosmic radiation. From October 2017 to April 2018, photon dosemeters of 38 dosimetry systems were exposed to ionising radiation at three dosimetric reference sites which are operated by the Physikalisch-Technische Bundesanstalt (PTB). In addition to measurements which were carried out under natural conditions, a number of dosemeters was also irradiated artificially under two angles in PTB's photon fields. 34 measuring bodies and institutions which are mainly involved in ambient radiation monitoring in Europe took part in this intercomparison in which the response of the dosemeters to terrestrial and also to secondary cosmic radiation was determined. As a result of this intercomparison, some sources of uncertainty and some errors were found. The intercomparison revealed the typical precision that has to be expected when long-term dose measurements are carried out in the natural environment. The successful participation in this intercomparison was documented by certificates that were issued to the participants. The short designation of this intercomparison is “IC2017prep”.

Response of photon dosimeters and survey instruments to new operational quantities proposed by ICRU RC26

Operational quantities in radiation protection are defined as conservative estimates of the protection quantities, which are generally not accessible to measurement. Radiation protection dosimeters and monitors are calibrated in the operational quantities. An ICRU Report Committee works on a proposal to replace the operational quantities for external radiation ambient dose equivalent, H*(10) and personal dose equivalent Hp(d, α) by quantities which are defined in close relation to the protection quantities. In this note, the response of dosimeters and survey instruments for photon radiation to these new operational quantities is evaluated. Survey instruments, having a low-energy cut-off at 50 keV can simply be recalibrated in the new quantity ambient dose and function as usual. Personal dosimeters will show an overreponse to photons with energy below 50 keV and either the dosimeter holder or the evaluation algorithm must be adapted to the new quantity.

Radiation Exposure to Surgeon and Assistant During Flat Panel Mini C-Arm Fluoroscopy in Hand and Wrist Surgical Procedures.

During mini C-arm fluoroscopy, both the patient and surgical team are exposed to scatter radiation. The objective of this study was to quantify body, thyroid, and hand radiation exposure to surgeon and assistant during intraoperative use of flat panel mini C-arm fluoroscopy in hand and wrist surgical procedures. Over 5 months, the surgeon's and assistant's radiation exposure was recorded during all osseous hand and wrist surgical procedures. Whole-body and thyroid radiation exposure were measured with 2 types of dosimeters: a photon thermoluminescence detector and a RaySafe i2 real-time dosimeter. Ring dosimeters were used to quantify hand radiation exposure. Mini C-arm fluoroscopy was used in 94 surgical procedures. Total fluoroscopy time was 1,996 seconds and varied between surgical procedures (range, 1-152 seconds; median, 11 seconds). No thermoluminescence detector photon dosimeter exceeded the threshold limit of 0.1 mSv. The RaySafe i2 real-time dosimeters recorded a cumulated dose of 0.029 mSv for the body and 0.012 mSv for the thyroid position of the surgeon. The assistant received a cumulated dose of 0.011 mSv for the body and 0.011 mSv for the thyroid position. The ring dosimeters showed a cumulated dosage of 1.28 mSv for the surgeon and 0.20 mSv for the assistant. Our results show that the surgeon's and assistant's body, thyroid, and hands were exposed to acceptable levels of scatter radiation during intraoperative use of the flat panel mini C-arm. The surgeon received the highest radiation exposure: 2.9% of the yearly radiation limits for the body, 0.05% for the thyroid position, and 2.56% for the hands. The assistant was exposed to less scatter radiation: 1.1% for the body, 0.04% for the thyroid, and 0.4% for the hands. This study quantified radiation levels to which surgeon and assistant are exposed during mini C-arm fluoroscopy in hand and wrist surgical procedures.

Feasibility study on determining the conventional true value of gamma-ray air kerma in a minitype reference radiation

A minitype reference radiation (MRR) with dimensions of only 1 m × 1 m × 1 m has been developed for the in situ calibration of photon dosimeters. The present work conducts a feasibility study on determining the conventional true value of gamma-ray air kerma at the point of test in the MRR. Owing to its smaller dimensions, the scattered gamma-rays in the MRR are expected to induce a non-negligible interference with the radiation field compared with conditions in the standard reference radiation stipulated by ISO4037-1 or GB/T12162.1. A gamma-ray spectrometer was employed to obtain the spectra of scattered gamma-rays within the MRR, and the feature components of the spectra were extracted by principal component analysis to characterize the interference of a dosimeter probe in the radiation field. A prediction model of the CAK at the point of test was built by least squares support vector machine based on the feature component data obtained from nine sample dosimeters under five different dose rates. The mean prediction error of the CAK prediction model was within ±4.5%, and the maximum prediction error was about ±10%.

The HMGU combined neutron and photon dosemeter

Recently, two independent electronic dosimeters for individual monitoring have been developed at the Helmholtz Center Munich, Germany, one for photons and one for neutrons. Calibration of these dosimeters was performed using mono-energetic neutrons (energy range: 144 keV–14.8 MeV) and photons (12 keV–1.3 MeV), and the resulting calibration curves were compared to Monte Carlo simulations using the GEANT4 tool kit. In the present paper, an integrated dosimeter prototype is described which combines both dosimeters and therefore allows measurement of total dose in mixed neutron/photon radiation fields. It is demonstrated that the neutron channel of the combined dosimeter is insensitive to photons, while the photon channel is insensitive to neutrons. The prototype was tested in the field of an Am-Be neutron source which also includes a photon component. When the Am-Be neutron source was shielded by 9 mm lead, the combined dosimeter indicated a dose rate of 122.6 ± 1.0 μSv/h from neutrons and 7.3 ± 0.1 μSv/h from photons. In contrast, when the Am-Be neutron source was not shielded by lead, the combined dosimeter indicated a dose rate of 125.8 ± 1.0 μSv/h from neutrons and 65.9 ± 0.1 μSv/h from photons indicating a significant contribution from photons to total dose. In a second experiment when the Am-Be neutron source was closer to the combined dosimeter and a 137Cs source was added, a neutron dose rate of 421 ± 20 μSv/h and a photon dose rate of 849 ± 6 μSv/h was obtained. It is concluded that this instrument can be used in mixed photon-neutron fields with doses and dose rates typical for radiation protection purposes.

USE OF A SIMPLE THERMALISED NEUTRON FIELD FOR QUALITY ACCEPTANCE OF WHOLE BODY TLDS.

The individual monitoring service of Public Health England (PHE) uses Harshaw™ whole-body and extremity thermoluminescent dosemeters (TLDs) with high-sensitivity lithium fluoride LiF:Mg,Cu,P, together with Harshaw 8800™ automated readers. The neutron-insensitive, (6)Li-depleted variety of TLD material is used by PHE because the service provides separate neutron and photon dosemeters. The neutron dosemeters are not sensitive to photons and vice versa Since insensitivity to neutrons is a supply requirement for TLDs, there is a need to test every new (annual) consignment for this. Because it is thermal neutrons that produce a response in (6)Li TLDs, a thermal field is needed. To this end, PHE has adopted the simple approach of sandwiching the TLDs between two ISO water-filled slab phantoms. In this arrangement, the fast neutrons from an Am-Be source are effectively thermalised. Details of the method are given, together with the results of supporting MCNP calculations and some typical results.

Establishing local workplace field correction factors for neutron personal dosemeters.

The present personal neutron dosemeters still need local correction factors to be able to provide accuracy comparable with photon dosemeters. Characterisation of the local neutron field is an indispensable part of neutron dosimetry to obtain such correction factors. It is often overlooked that besides characterisation in the neutron energy also the directional distribution of neutrons plays a crucial part in this characterisation. The authors have done such characterisation in the energy and angle for four workplace fields in Paks NPP. For this a relatively simple approximation method was used using the Nprobe for the energy distribution and measurements on the six sides of the slab phantom with personal dosemeters for the directional distribution. This allowed one to estimate a reference neutron Hp(10) rate and to compare it with the response of several neutron personal dosemeters.

Intercomparisons for individual monitoring services in Europe: Organization, experience and results

The European Radiation Dosimetry Group (EURADOS) has set-up a worldwide unique program of self-sustainable international intercomparisons for individual monitoring services (IMS). Starting in 2008, with three intercomparisons for whole body photon dosimeters every 2 nd year, one intercomparison in 2009 for photon and beta extremity dosimeters and in 2012, an intercomparison for neutron whole body dosimeters, the intercomparisons became fundamental for the harmonization of radiation protection and individual monitoring in Europe. With the intercomparison results, the participants can show compliance within their quality management system, compare their results with those from other participants and develop plans for improvement of their system. The costs of the exercises are covered by the participant's fees. For the whole body photon intercomparisons a the number of participating systems increased from 62 in 2008 to 88 in 2012, showing the big demand for such exercises. Some of the system types like film dosimeters showed decreasing deviations of their results in the general and subsequently increased their quality in the dosimetric performance. The calibration of the systems showed to be an important factor for the performance of all system types. The intercomparison for extremity dosimeters for photon and beta radiation revealed large deviations for quite a few of the participating systems. The intercomparison actions are planned to be continued on a regular basis; the next whole body photon intercomparison is scheduled for 2014. International intercomparisons for IMS for external radiation will bring in improvement and harmonization in individual monitoring.

Development of a graphite probe calorimeter for absolute clinical dosimetry

The aim of this work is to present the numerical design optimization, construction, and experimental proof of concept of a graphite probe calorimeter (GPC) conceived for dose measurement in the clinical environment (U.S. provisional patent 61∕652,540). A finite element method (FEM) based numerical heat transfer study was conducted using a commercial software package to explore the feasibility of the GPC and to optimize the shape, dimensions, and materials used in its design. A functioning prototype was constructed inhouse and used to perform dose to water measurements under a 6 MV photon beam at 400 and 1000 MU∕min, in a thermally insulated water phantom. Heat loss correction factors were determined using FEM analysis while the radiation field perturbation and the graphite to water absorbed dose conversion factors were calculated using Monte Carlo simulations. The difference in the average measured dose to water for the 400 and 1000 MU∕min runs using the TG-51 protocol and the GPC was 0.2% and 1.2%, respectively. Heat loss correction factors ranged from 1.001 to 1.002, while the product of the perturbation and dose conversion factors was calculated to be 1.130. The combined relative uncertainty was estimated to be 1.4%, with the largest contributors being the specific heat capacity of the graphite (type B, 0.8%) and the reproducibility, defined as the standard deviation of the mean measured dose (type A, 0.6%). By establishing the feasibility of using the GPC as a practical clinical absolute photon dosimeter, this work lays the foundation for further device enhancements, including the development of an isothermal mode of operation and an overall miniaturization, making it potentially suitable for use in small and composite radiation fields. It is anticipated that, through the incorporation of isothermal stabilization provided by temperature controllers, a subpercent overall uncertainty will be achieved.

Sci-Thur PM: YIS - 09: Development of a graphite probe calorimeter for absolute clinical dosimetry: Numerical design optimization, prototyping and experimental proof-of-concept.

In this work, the feasibility of absolute dose to water measurements using a small-scale graphite probe calorimeter (GPC) in a clinical environment is established. A numerical design optimization study was conducted by simulating the heat transfer in the GPC resulting from irradiation using a finite element method software package. The choice of device shape, dimensions and materials was made to minimize the heat loss in the sensitive volume of the GPC. The resulting design, which incorporates a novel aerogel-based thermal insulator, was built in-house. Absorbed dose to water measurements were made under standard conditions in a 6 MV 1000 MU/min photon beam and subsequently compared against TG-51 derived values. The average measured dose to water was 95.7 ±1.4 cGy/100 MU, as compared to an expected value of 96.6 cGy/100 MU. The Monte Carlo-calculated graphite to water dose conversion factor was 1.099, while the derived heat loss correction factors varied between 1.005 and 1.013. The most significant sources of uncertainty were the repeatability (type A, 1.4%) and thermistor calibration (type B, 2.1%). The contribution of these factors to the overall uncertainty is expected to decrease significantly upon the implementation of active thermal stabilization provided by a temperature controller and direct electrical calibration, respectively. This work demonstrates the feasibility of using the GPC as a practical clinical absolute photon dosimeter and will serve as the basis for a miniaturized version applicable to small and composite fields.

TH-E-BRB-07: Developing a Graphite Probe Calorimeter for Accurate Clinical Dosimetry

Purpose: To optimally design, construct and validate an innovative graphite probe calorimeter (GPC), comparable in size to a 0.6 cm3 cylindrical ionization chamber, for use as an absolute clinical dosimeter. Methods: A numerical finite element method (FEM) based heat transfer study was conducted using COMSOL Multiphysics® to explore the feasibility of the GPC and to optimize its design in terms of shape, dimensions and materials. Through the use of a novel aerogel-based material as the thermal insulation between the graphite bodies, the portability and robustness of the GPC design is made suitable for routine clinical use. A functioning prototype was constructed in-house and used to perform dose to water measurements under a 6 MV stereotactic radiosurgery (1000 MU/min) photon beam from a Novalis Tx, under reference conditions. Heat loss correction factors were determined using FEM analysis and graphite to water absorbed dose conversion factors were calculated using Monte Carlo techniques. Results: The dose to water measurements agreed well with the TG-51 derived values, with an average measured dose output of 95.7 (s = 1.4) cGy/100 MU, compared to an expected output of 96.6 cGy/100 MU. Heat loss correction factors ranged between 1.005 and 1.013, while the graphite to water dose conversion factor was 1.099. The combined relative uncertainty was estimated at 2.7%, with the largest contributions being the repeatability (type A, 1.4%) and thermistor calibration (type B, 2.1%). Conclusions: This proof of concept demonstrates the feasibility of using the GPC as a practical clinical absolute photon dosimeter and lays the foundation for a miniaturized version suitable for small and composite field dosimetry. It is expected that through active thermal stabilization provided by a temperature controller and electrical calibration, the uncertainty due to repeatability and calibration will be decreased, such that the overall uncertainty is better than 1%. This work has been supported in part by Grant RGPIN 298191 of the Natual Sciences and Engineering Research Council of Canada (NSERC). J.R. is a recipient of a NSERC M.Sc. scholarship.

SU-E-T-132: Investigation of Photon and Proton Overlapping Fields in PRESAGE- Dosimeters.

To evaluate the effects of overlapping dose volumes for varying field arrangements in two formulations of PRESAGE®: one intended for, and irradiated with, proton beams and the other photon beams. For each treatment modality (photon, proton), three overlapping field setups were performed. These included a stationary dosimeter irradiated over six fractions, a dosimeter shifted laterally to the field to deliver a dose plateau in two fractions, and a dosimeter rotated on its axis to deliver a two-field (for protons) and four-field (for photons) box treatment overlapping in the center of the dosimeter. All subsequent fractions were given within ten minutes and never less than one minute apart. Two cylindrical PRESAGE® dosimeters approximately 7.5 cm in length by 7.5 cm in diameter were irradiated for each setup. The dosimeters were paired, with one dosimeter given total dose by a single fraction while the other followed one of the overlapping field setups. The dosimeters were analyzed using an optical CT scanner and exported to the CERR environment where the doses were compared between paired dosimeters. Dose profile comparisons showed relative dose agreement between paired dosimeters within 5% along the SOBP region of the proton formulation. In the case of the fractionated proton irradiation, there was an over-response while other setups resulted in under-responses. Dose agreement between the photon dosimeter treated with six fractions showed a dose under-response within 11% and never less than 5%. Future measurements will include the remaining field setups. The proton formulation of PRESAGE® showed good dose agreement between single and multiple field irradiations. While the photon formulation had slightly less agreement, additional field setup comparisons may show improved results. These results will aid future measurements of overlapping field treatment plans delivered to PRESAGE® for treatment verification for proton and photon 3D dosimetry.

A new design of dosemeter for assessment of effective dose in anisotropic photon fields

Abstract Conventional personal dosemeters designed for measurement of H p (10) of photon exposure can approximate effective dose E (quantity of interest in radiation protection) only for situations when irradiation occurs from the front of a human body. It is known that for rear irradiation with lower energies dosemeter readout may significantly underestimate E . The aim of this study was to determine angular and energy dependences of E and estimate respective conversion coefficient from photon dosemeter readout to E for given (left chest) location of a dosemeter and based on such dependences theoretically elaborate a design of a dosemeter with uniform response in terms of E . Monte Carlo simulation was performed for different energies (50, 80, 100, 150 and 200 keV) and incidence angles (with 30° increments both for polar and azimuth angles) for a number of plausible dosemeter designs. It was demonstrated that properly designed copper filter reduces variation of conversion coefficient by a factor of four, thus allowing more certain evaluation of effective dose of occupational exposure than dosemeters of conventional design may do.

Comparison of the Dosimetric Characteristics of High Atomic-numbered Inorganic Scintillators as a Fiber Optic-based Photon Dosimeter

Comparison of the Dosimetric Characteristics of High Atomic-numbered Inorganic Scintillators as a Fiber Optic-based Photon Dosimeter

Measuring cosmic-ray exposure in aircraft using real-time personal dosemeters

Aircrew exposure to radiation was measured on several long-haul flights using two small commercial electronic personal dosemeters: one was a photon dosemeter, the NRF20; the other was a neutron dosemeter, the NRY21-both manufactured by Fuji Electric Systems Co. Ltd. for radiation protection at nuclear facilities. Non-neutron doses were measured using the photon dosemeter, and neutron doses were measured using the neutron dosemeter. The measured non-neutron doses at commercial aviation altitudes agree with the EPCARD (European Program Package for the Calculation of Aviation Route Doses) dose calculation within a difference of 8 %. However, the recorded neutron doses were 5-15 times larger than the EPCARD calculation. These over-measurements are dependent on cut-off rigidities.

Neutron measurements around a TN85-type storage cask with high-active waste

Abstract Several types of casks have been deposited in the German interim storage facility for spent fuel assemblies and vitrified high-active waste (HAW) at Gorleben since 1995, most of them of the CASTOR® type. In 2008 a delivery of 11 TN85-type casks arrived. They belong to the Transnuclear/Areva cask family and, compared to the flasks of the German (GNS) CASTOR® type, they differ in the neutron shielding design. Generally, radiation exposure of personnel during transportation and storage of casks containing spent fuel and vitrified waste is caused by mixed photon/neutron fields. Frequently, especially at casks for vitrified waste from reprocessing, neutrons are the major component of radiation exposure. Spectrometric and dosimetric investigations were made around a cask of the TN85-type. Neutron fluence spectra and reference values of the ambient dose equivalent H*(10) were measured by means of a Bonner sphere spectrometer (BSS) at several locations on the cask surface and in its environment. Moreover, commercial area dosemeters, LB6411 neutron monitors and conventional AD 6-type photon dosemeters were used. In addition, the responses of two electronic personal dosemeters for mixed fields (EPD-N2, DMC 2000GN) and a TLD albedo dosemeter were investigated. The neutron spectra obtained from the BSS are presented and compared with former measurements at CASTOR® type casks. The relative responses of the LB6411 survey meter and the individual dosemeters are discussed. The LB6411 monitor indicates H*(10) around the TN85 cask with tolerable measuring uncertainties. The personal dosemeters provide acceptable results for photons but overestimate the neutron dose considerably.

A BOUNDING ESTIMATE OF NEUTRON DOSE BASED ON MEASURED PHOTON DOSE AROUND SINGLE PASS REACTORS AT THE HANFORD SITE

Neutron and photon radiation survey records have been used to evaluate and develop a neutron to photon (NP) ratio to reconstruct neutron doses to workers around Hanford's single pass reactors that operated from 1945 to 1972. A total of 5,773 paired neutron and photon measurements extracted from 57 boxes of survey records were used in the development of the NP ratio. The development of the NP ratio enables the use of the recorded dose from an individual's photon dosimeter badge to be used to estimate the unmonitored neutron dose. The Pearson rank correlation between the neutron and photon measurements was 0.71. The NP ratio best fit a lognormal distribution with a geometric mean (GM) of 0.8, a geometric standard deviation (GSD) of 2.95, and the upper 95 th % of this distribution was 4.75. An estimate of the neutron dose based on this NP ratio is considered bounding due to evidence that up to 70% of the total photon exposure received by workers around the single pass reactors occurs during shutdown maintenance and refueling activities when there is no significant neutron exposure. Thus when this NP ratio is applied to the total measured photon dose from an individual film badge dosimeter, the resulting neutron dose is considered bounded.