Urine Bioassay Research Articles

Simultaneous Determination of Transuranium Radionuclides in Urine by Tandem Quadrupole ICP-MS/MS with Mass-Shift Mode Combined with Chemical Separation.

The urine bioassay method for transuranium nuclides (237Np, 239,240,241Pu, 241Am, and 244Cm) is needed to quickly assess the potential internal contamination in emergency situations. However, in the case that the analysis of multiple radionuclides is required in the same sample, time-consuming/tedious sequential analytical procedures using multiple chromatographic separation resins would have to be employed for the separation of every single radionuclide. In this work, a rapid method for the simultaneous determination of transuranium nuclides in urine was developed by using triple quadrupole inductively coupled plasma mass spectrometry (ICP-MS/MS) combined with a single DGA resin column. The chemical behaviors of Np/Pu and Am/Cm on the DGA resin were consistent in 8-10 mol/L HNO3 and 0.005-0.02 mol/L NaNO2 when 242Pu and 243Am were selected as tracers for Np/Pu and Am/Cm yield monitoring. Based on their different reaction rates with O2, 237Np, 239,240,241Pu, 241Am, and 244Cm in the same solution were simultaneously measured by ICP-MS/MS in the same run. The elimination efficiency of 238U+ tailing (7.43 × 10-9), 238U1H16O2+/238U16O2+ (8.11 × 10-8) and cross contamination of 241Pu and 241Am (<1%) were achieved using 10.0 mL/min He-0.3 mL/min O2 even if the eluate was directly measured without any evaporation. The detection limits of transuranium nuclides were at the femtogram level, demonstrating the feasibility of ICP-MS/MS for simultaneous transuranic radionuclides urinalysis. The developed method was validated by analyzing the spiked urine samples.

Chelation Modeling of a Plutonium-238 Inhalation Incident Treated with Delayed DTPA.

This work describes an analysis, using a previously established chelation model, of the bioassay data collected from a worker who received delayed chelation therapy following a plutonium-238 inhalation. The details of the case have already been described in two publications. The individual was treated with Ca-DTPA via multiple intravenous injections and then nebulizations beginning several months after the intake and continuing for four years. The exact date and circumstances of the intake are unknown. However, interviews with the worker suggested that the intake occurred via inhalation of a soluble plutonium compound. The worker provided daily urine and fecal bioassay samples throughout the chelation treatment protocol, including samples collected before, during, and after the administration of Ca-DTPA. Unlike the previous two publications presenting this case, the current analysis explicitly models the combined biokinetics of the plutonium-DTPA chelate. Using the previously established chelation model, it was possible to fit the data through optimizing only the intake (day and magnitude), solubility, and absorbed fraction of nebulized Ca-DTPA. This work supports the hypothesis that the efficacy of the delayed chelation treatment observed in this case results mainly from chelation of cell-internalized plutonium by Ca-DTPA (intracellular chelation). It also demonstrates the validity of the previously established chelation model. As the bioassay data were modified to ensure data anonymization, the calculation of the "true" committed effective dose was not possible. However, the treatment-induced dose inhibition (in percentage) was calculated.

Sequential separation of actinides and strontium in urine samples using TRU and Sr-spec resin in tandem

Abstract This paper describes the sequential separation of actinides such as U, Pu, Am, and Sr in urine samples. Urine bioassay is one of the preferred methods for assessing low level of internal contamination in radiation workers. Any special monitoring case necessitates the completion of sample analysis as soon as possible. Hence, in this work, efforts have been made to reduce the overall analysis time for the estimation of actinides and strontium in urine samples. In the described method, TRU and Sr-spec resins are used in tandem for the separation of actinides and strontium. The proposed method is very convenient to handle, rapid, and robust when compared with ion exchange and other extraction chromatography work. Radiochemical recoveries of 232U, 242Pu, 243Am, and 90Sr obtained are in the range of 66%–89%, 63%–99%, 63%–96%, and 61%–79% with an average and standard deviation of 77.2% ±7.2, 79.8% ±11.4, 80.9% ±11.2, and 71.8% ±5.9, respectively.

Gamma Radioactivity Detection Limits and Associated Radionuclide Intakes Study in Artificial Human Urine Using Sodium-iodide and High-purity Germanium Detectors.

The performance of several gamma detectors was investigated for emergency urine bioassay screening of two radionuclides of concern: 131 I and 137 Cs. Unspiked artificial urine samples were measured for 10 min each on four different gamma detectors: 80% relative efficiency high-purity Ge detector in standard shielding, 102% low-background high-purity Ge detector equipped with top muon shield, 78% high-purity Ge well detector in standard shielding, and 4″ × 4″ NaI well detector in standard shielding. The measured gamma spectra were analyzed in two ways: (1) for the 364-keV peak region of 131 I and 662-keV peak region of 137 Cs and (2) for the total counts in the full energy spectrum (50-2,048 keV). The results were analyzed using the principles of signal detection theory according to the Currie's formalism extended by a complete uncertainty propagation. This enabled calculation of the detection capability in terms of detection limit (Bq L -1 ) of urine, the latter referred to as minimum detectable activity. The NaI well detector had the lowest minimum detectable activities for total spectra, whereas the high-purity Ge well detector had the lowest peak minimum detectable activity values. Minimum detectable inhalation and ingestion intakes from urine bioassay were calculated from the minimum detectable activity values for urine collection 1 d, 1 wk, and 1 mo past the initial intake. The calculated intakes were compared with annual limits on intake. The results are interpreted with respect to a large-scale radiological emergency response.

URINARY EXCRETION OF PLUTONIUM IN MAYAK WORKERS DURING AND AFTER CA-DTPA ADMINISTRATION.

Chelation therapy is sometimes used after potential exposures to plutonium to increase urinary excretion of the radionuclide to improve the accuracy of bioassay measurements. The purpose of this report is to describe the enhancement of urinary excretion of plutonium during and after the administration of the trisodium salt of calcium diethylenetriaminepentaacetate (Ca-DTPA) daily for 3d to a group of male and female plutonium workers from the Mayak Production Association in Ozyorsk, Russia. One-hundred and two cases (18 females and 84 males) were selected where urinary contents of plutonium, prior to chelation, exceeded the detection threshold. Daily urine collections were obtained during the 3d of Ca-DTPA treatments. In addition, 58 of these cases had urine bioassays at 1-45d after chelation. The daily enhancement over baseline values excretion of plutonium was found to be 50.4×/1.4 (geometric mean and geometric standard deviation); 58.9×/1.2; 72.9×/1.4 in the first, second and third days of Ca-DTPA administration. The mean enhancement for the 3-d period was 60.1×/1.7. The rate of plutonium excretion from 1 to 45d after chelation decreased with a half-period of 3.9d and the chelation enhancement factor (Кenh-i) is described by the function Кenh-i=(0.79±0.24)+(42.9±1.2)×e-(0.18±0.01)×day.

Plutonium in Manhattan Project workers: Using autopsy data to evaluate organ content and dose estimates based on urine bioassay with implications for radiation epidemiology.

PurposeRadiation dose estimates in epidemiology typically rely on intake predictions based on urine bioassay measurements. The purpose of this article is to compare the conventional dosimetric estimates for radiation epidemiology with the estimates based on additional post-mortem tissue radiochemical analysis results.MethodsThe comparison was performed on a unique group of 11 former Manhattan Project nuclear workers, who worked with plutonium in the 1940s, and voluntarily donated their bodies to the United States Transuranium and Uranium Registries.ResultsPost-mortem organ activities were predicted using different sets of urine data and compared to measured activities. Use of urinalysis data collected during the exposure periods overestimated the systemic (liver+skeleton) deposition of 239Pu by 155±134%, while the average bias from using post-exposure urinalyses was –4±50%. Committed effective doses estimated using early urine data differed from the best estimate by, on average, 196±193%; inclusion of follow-up urine measurements in analyses decreased the mean bias to 0.6±36.3%. Cumulative absorbed doses for the liver, red marrow, bone surface, and brain were calculated for the actual commitment period.ConclusionOn average, post-exposure urine bioassay results were in good agreement with post-mortem tissue analyses and were more reliable than results of urine bioassays collected during the exposure.

Sexually transmitted infections and prior antibiotic use as important causes for negative urine cultures among adults presenting with urinary tract infection symptoms to primary care clinics in Zimbabwe: a cross-sectional study

ObjectiveUrinary tract infections (UTIs) are common in primary care. The yield of urine cultures in patients with UTI symptoms can be considerably different between high-income and low-income settings. This study...

Maximum Possible Doses for a Cohort of Individuals with Intakes Possibly Containing a Component of a Ceramic-type Material.

Recently, a glovebox breach led to the potential exposure of 15 Los Alamos National Laboratory employees to 238Pu. Given what is known about the material involved in the incident, the possibility of an intake with a ceramic-type component must be considered. Incidents in which intakes of ceramic solubility-type material is suspected represent a challenge for internal dose assessment via urine bioassay because even relatively large doses cannot be detected in urine until many months after the intake. Ideally, in these situations fecal samples should be collected to assess the intake, but in this case fecal sampling was delayed. This paper presents a method to calculate the maximum possible doses for all individuals involved in an incident using only early time-decreasing urine bioassay measurements.

Response to a Skin Puncture Contaminated with 238Pu at Los Alamos National Laboratory.

The three principal pathways for intakes of plutonium are ingestion, inhalation, and contaminated wounds. In August 2018, a glovebox worker at Los Alamos National Laboratory (LANL) sustained a puncture from a thread of a braided steel cable contaminated with Pu. The puncture produced no pain, no blood, and little or no visible mark. As a result, the potential for a contaminated wound was not immediately recognized, and a wound count was not conducted until elevated urine bioassay results were received 12 d after the incident. This paper discusses the circumstances of the incident, along with the medical response and dose assessment, and a discussion of the risks and benefits of the medical interventions.

Multiplexed Protein Biomarker Detection with Microfluidic Electrochemical Immunoarrays.

Electrochemistry is a multidisciplinary field encompassing the study of analytes in solution for detection and quantification. For the medical field, this brings opportunities to the clinical practice of disease detection through measurements of disease biomarkers. Specifically, panels of biomarkers offer an important future option that can enable physicians' access to blood, saliva, or urine bioassays for screening diseases, as well as monitoring the progression and response to therapy. Here, we describe the simultaneous detection of eight protein cancer biomarkers in a 30-min assay by a microfluidic electrochemical immunoarray.

Long-term Retention of Plutonium in the Respiratory Tracts of Two Acutely-exposed Workers: Estimation of Bound Fraction.

Inhalation of plutonium is a significant contributor of occupational doses in plutonium production, nuclear fuel reprocessing, and cleanup operations. Accurate assessment of the residence time of plutonium in the lungs is important to properly characterize dose and, consequently, the risk from inhalation of plutonium aerosols. This paper discusses the long-term retention of plutonium in different parts of the respiratory tract of two workers who donated their bodies to the US Transuranium and Uranium Registries. The post-mortem tissue radiochemical analysis results, along with the urine bioassay data, were interpreted using Markov Chain Monte Carlo and the latest biokinetic models presented in the Occupational Intakes of Radionuclides series of ICRP publications. The materials inhaled by both workers were found to have solubility between that of plutonium nitrates and oxides. The long-term solubility was also confirmed by comparison of the activity concentration in the lungs and the thoracic lymph nodes. The data from the two individuals can be explained by assuming a bound fraction (fraction of plutonium deposited in the respiratory tract that becomes bound to lung tissue after dissolution) of 1% and 4%, respectively, without having to significantly alter the particle clearance parameters. Effects of different assumptions about the bound fraction on radiation doses to different target regions was also investigated. For inhalation of soluble materials, an assumption of fb of 1%, compared to the ICRP default of 0.2%, increases the dose to the most sensitive target region of the respiratory tract by 258% and that to the total lung by 116%. Some possible alternate methods of explaining higher-than-expected long-term retention of plutonium in the upper respiratory tract of these individuals-such as physical sequestration of material into the scar tissues and possible uptake by lungs-are also briefly discussed.

The effect of quench agent on urine bioassay for various radionuclides using QuantulusTM1220 and Tri-CarbTM3110

Following a radiological or nuclear incident, the National Response Plan has given the Department of Health and Human Services/Centers for Disease Control and Prevention the responsibility for assessing population’s contamination with radionuclides. In the public health response to the incident, valuable information could be obtained in a timely and accurate manner by using liquid scintillation counting techniques to determine who has been contaminated above background for alpha and beta emitting radionuclides. The calibration plays a major role in this process therefore, knowing the effect of quench agents on calibration is essential.

Use of gamma cameras for individual monitoring of photon emitting radionuclides based on urine samples in emergency situations

IntroductionIn nuclear and radiological accidents there is a risk of radioactive contamination and accidental intake of radionuclides. Therefore, the individual monitoring of internal exposures is essential for the prompt response to an emergency situation. Individual monitoring allows classifying and evaluating the accident, preventing the increase of contaminated areas and number of affected people and assessing the efficacy of the remediation actions and the decontamination procedures.This study evaluated the use of gamma cameras for individual monitoring of photon-emitting radionuclides incorporated into the human body based on urine samples to apply as a prompt response in emergency situations. MethodsFor this, a gamma camera available in a military public hospital located in the city of Rio the Janeiro was calibrated using standard liquid sources of 57Co, 22Na, 137Cs and 54Mn supplied by the National Laboratory of Metrology (LMNRI) of the Institute of Radiation Protection and Dosimetry (IRD), all in a 1 L bottle geometry, simulating an urine bioassay. “Efficiency vs Energy” curves at 5, 10, 15 and 20 cm source-to-detector distances were obtained. Based on the “Efficiency vs Energy” curves Minimum Detectable Activities, Minimum Detectable Intakes and Minimum detectable committed effective dose were calculated for 106Ru, 103Ru, 131I, 134Cs and 137Cs at the four source-to-detector distances. ResultsThe results showed that the gamma camera presents Minimum detectable committed effective dose bellow 1 mSv for 103Ru, 131I, 134Cs and137Cs. For 106Ru the Minimum detectable committed effective dose calculated ranged from 1.84 to 3.40 mSv, which is acceptable in emergency situations. ConclusionsIt has been concluded that the equipment evaluated in this work shows enough sensitivity for use as screening method in accident situations involving intakes of such radionuclides.

Nuclear medicine: workplace monitoring and internal occupational exposure during a ventilation/perfusion single-photon emission tomography.

The administration of 99mTc-HDP to diagnose pulmonary thromboembolisms leads to the presence of 99mTc in the environment of a nuclear medicine department, which could pose a potential risk of internal contamination to medical staff. Therefore, air samples from the administration room, gamma camera room and corridor of such a department were taken for the purpose of performing a workplace monitoring program of the medical centre under study, with maximum activity values of 640 ± 30kBq/m3, 1.5 ± 0.1kBq/m3 and 54 ± 3kBq/m3, respectively, being obtained. These results correspond to committed effective doses received by exposed employees, via inhalation, when one ventilation/perfusion single-photon emission tomography study was performed, of 0.7μSv, 0.004μSv and 0.2μSv, respectively. As inhalation is the employees' main exposure pathway to radio-aerosols, the internal dose of the nuclear medicine department's medical staff was also evaluated via urine bioassay measurements. Nuclear medicine nurses showed the highest 99mTc activity in 24-h urine samples (2100 ± 130Bq/day), resulting in a committed effective dose of 21μSv for each diagnostic study performed. Even so, the performance of ventilation/perfusion diagnostic studies did not constitute a substantial radiological risk since the annual dose limit for exposed employees was not exceeded.

Updated mortality analysis of the Mallinckrodt uranium processing workers, 1942–2012

Purpose Mallinckrodt Chemical Works (MCW) was the earliest uranium processing facility in the United States, and in 1942 produced the uranium oxide used for the first sustained and controlled nuclear fission chain-reaction at the University of Chicago. A second follow-up through 2012 was conducted of 2514 White male workers employed 1942–1966 at the MCW for dose-response analyses for selected causes of death. Materials and methods Organ/tissue-specific dose reconstruction included both external (12,686 MCW film badge records, 210 other facility film badge records, and 31,297 occupational chest x-rays) and internal sources of uranium and radium (39,451 urine bioassays, 2341 breath radon measurements, and 6846 ambient radon measurements). Dust measurements from pitchblende facilitated quantitative risk estimates for non-radiogenic effects on the lung and kidney. Vital status was determined from multiple sources including the National Death Index and the Social Security Administration. Cox regression models were used for dose response analyses. Results Vital status was determined for 99% of the workers, of whom 75% had died. The mean lung dose from all sources of external and internal radiation combined was 69.9 mGy (maximum 885 mGy; percent workers >100 mGy, 10%) and there was no evidence for a dose response for lung cancer (Hazard Ratio (HR) of 0.95 (95% CI = 0.81–1.12) at 100 mGy). A significant association with radiation was found for kidney cancer (HR of 1.73 (95% CI = 1.04–2.79) at 100 mGy) and suggested for nonmalignant kidney diseases (HR of 1.30 (95% CI = 0.96–1.76) at 100 mGy). A non-radiation etiology could not be discounted, however, because of the possible renal toxicities of uranium, a heavy metal, and silica, a component of pitchblende dust. Non-significant HRs at 100 mGy for other sites of a priori interest were 0.36 (0.06–2.03) for leukemia other than CLL, 0.68 (0.17–2.77) for liver cancer, and 1.23 (0.79–1.90) for non-Hodgkin lymphoma. The HR at 100 mGy was 1.09 (0.99–1.20) for ischemic heart disease. An association was seen between dust and combined malignant and non-malignant lung disease, HR at 10 mgm−3year−1 of 1.01 (1.00–1.02). Conclusions A positive radiation dose response was observed for malignant and non-malignant kidney disease, and a negative dose response for malignant and non-malignant lung disease. Cumulative measures of dust were significantly associated with malignant and non-malignant lung disease and suggested for malignant and non-malignant kidney disease. Small numbers preclude definitive interpretations which will await the combination with similar studies of early uranium processing workers.

Validation of a system of models for plutonium decorporation therapy.

A recently proposed system of models for plutonium decorporation (SPD) was developed using data from an individual occupationally exposed to plutonium via a wound [from United States Transuranium and Uranium Registries (USTUR) Case 0212]. The present study evaluated the SPD using chelation treatment data, urine measurements, and post-mortem plutonium activities in the skeleton and liver from USTUR Case 0269. This individual was occupationally exposed to moderately soluble plutonium via inhalation and extensively treated with chelating agents. The SPD was linked to the International Commission on Radiological Protection (ICRP) Publication 66 Human Respiratory Tract Model (HRTM) and the ICRP Publication 30 Gastrointestinal Tract model to evaluate the goodness-of-fit to the urinary excretion data and the predictions of post-mortem plutonium retention in the skeleton and liver. The goodness-of-fit was also evaluated when the SPD was linked to the ICRP Publication 130 HRTM and the ICRP Publication 100 Human Alimentary Tract Model. The present study showed that the proposed SPD was useful for fitting the entire, chelation-affected and non-affected, urine bioassay data, and for predicting the post-mortem plutonium retention in the skeleton and liver at time of death, 38.5years after the accident. The results of this work are consistent with the conclusion that Ca-EDTA is less effective than Ca-DTPA for enhancing urinary excretion of plutonium.

Improved Modeling of Plutonium-DTPA Decorporation.

Individuals with significant intakes of plutonium (Pu) are typically treated with chelating agents, such as the trisodium salt form of calcium diethylenetriaminepentaacetate (CaNa3-DTPA, referred to hereafter as Ca-DTPA). Currently, there is no recommended approach for simultaneously modeling plutonium biokinetics during and after chelation therapy. In this study, an improved modeling system for plutonium decorporation was developed. The system comprises three individual model structures describing, separately, the distinct biokinetic behaviors of systemic plutonium, intravenously injected Ca-DTPA and in vivo-formed Pu-DTPA chelate. The system was linked to ICRP Publication 100, "Human Alimentary Tract Model for Radiological Protection" and NCRP Report 156, Development of a Biokinetic Model for Radionuclide-Contaminated Wounds and Procedures for Their Assessment, Dosimetry and Treatment." Urine bioassay and chelation treatment data from an occupationally-exposed individual were used for model development. Chelation was assumed to occur in the blood, soft tissues, liver and skeleton. The coordinated network for radiation dosimetry approach to decorporation modeling was applied using a chelation constant describing the secondorder, time-dependent kinetics of the in vivo chelation reaction. When using the proposed system of models for plutonium decorporation, a significant improvement of the goodness-of-fit to the urinary excretion data was observed and more accurate predictions of postmortem plutonium retention in the skeleton, liver and wound site were achieved.

Dosimetry is Key to Good Epidemiology: Workers at Mallinckrodt Chemical Works had Seven Different Source Exposures.

Mallinckrodt Chemical Works was the earliest uranium processing facility in the Manhattan Project, beginning in 1942. Even then, concern existed about possible health effects resulting from exposure to radiation and pitchblende dust. This concern was well founded as the facility processed Belgian Congo pitchblende ore that was up to 60% pure uranium with high U content and up to 100 mg of radium per ton. Workers were exposed to external gamma radiation plus internal radiation from inhalation and ingestion of pitchblende dust (uranium, radium, and silica). Multiple sources of exposure were available for organ dose reconstruction to a degree unique for an epidemiologic study. Personal film badge measures available from 1945 captured external exposures. Additional external exposure included 15,518 occupational medical x-rays and 210 radiation exposure records from other facilities outside of Mallinckrodt employment. Organ dose calculations considered organ-specific coefficients that account for photon energy and job-specific orientation of workers to the radiation source during processing. Intakes of uranium and radium were based on 39,451 uranium urine bioassays and 2,341 breath radon measurements, and International Commission on Radiological Protection (ICRP) Publication 68 biokinetic models were used to estimate organ-specific radiation absorbed dose. Estimates of exposure to airborne radon and its short-lived progeny were based on radon measurements in work areas where radium-containing materials were handled or stored, together with estimated exposure times in these areas based on job titles. Dose estimates for radon and its short-lived progeny were based on models and methods recently recommended in ICRP Publication 137. This comprehensive dosimetric approach follows methods outlined by the National Council on Radiation Protection Scientific Committee 6-9 for the Million Worker Study. Annual doses were calculated for six organs: lung, brain, heart, kidney, colon and red bone marrow. Evaluation and adjustment for individual cumulative measures of pitchblende dust inhalation were made for lung and kidney diseases.

A Case Study of the NCRP 156 Wound Model of Embedded DU Using Data From Urine Uranium Concentrations of Wounded Veterans.

Depleted uranium (DU) munitions were initially used by the United States (U.S.) military during the first Persian Gulf War in 1991 in order to penetrate heavily armored vehicles. However, as a result of friendly fire, several U.S. military personnel received intakes from DU munitions. One of the ongoing concerns for these wounded veterans is the potential long-term exposure received from DU embedded fragments. The United States Army Institute of Public Health (AIPH) is the first laboratory that analyzes the urine bioassays from Army Soldiers that are injured with DU fragments. The United States Air Force School of Aerospace Medicine also evaluates bioassays from DU injuries. The urine bioassay data collected by AIPH was evaluated using the NCRP 156 wound model coefficients for the DU-Wafer, Fragment, and Particle models. The maximum likelihood method was used in the Integrated Modules for Bioassay Analysis (IMBA-PPAE) to calculate the estimates of intake and tissue doses. Evaluating the three models for wound retention, the DU-Wafer and Fragment model yielded a credible fit to the bioassay data. Comparing the two models, the DU-Wafer model fits the data better than the Fragment model when comparing their autocorrelation coefficient and chi-squared values of (P 1.73 × 10, c 4.83 × 10), (P 2.01 × 10, c 1.09), respectively. This evaluation supports the validity of both the DU-wafer model as well as the default fragmentation model proposed by NCRP 156.

Frequent Community Use of Antibiotics among a Low-Economic Status Population in Manila, the Philippines: A Prospective Assessment Using a Urine Antibiotic Bioassay.

The widespread unregulated use of antibiotics without medical consultation contributes to the burden of antibiotic resistance in Southeast Asian countries. This study investigated antibiotic use before hospital consultation. In a prospective observational study from February 2, 2015, to July 2, 2015, we enrolled febrile patients attending the emergency room in San Lazaro Hospital, Manila, the Philippines. A urine sample was collected and a bioassay was used to detect antibiotic activity in urine using Bacillus stearothermophilus (ATCC7953), Escherichia coli (ATCC25922), and Streptococcus pyogenes (ATCC19615). Patients or caregivers reported their medication history, clinical information, and socioeconomic status. During the study period, 410 patients were enrolled. The median (interquartile range) age was 14 (7-23) years and 158 (39%) reported prior antibiotic use, predominantly a beta-lactam antibiotic. A total of 164 (40%, 95% confidence interval [CI]: 35-45) patients were urine bioassay positive with any of three organisms. The Bacillus assay was the most sensitive, detecting 162 (99%, 95% CI: 96-100) cases. Among bioassay positive patients, dengue (N = 91, 55%, 95% CI: 48-63) was the most frequent diagnosis, followed by other viral infections, including measles, rubella, and mumps (N = 17, 10%, 95% CI: 6-16). Patients with a positive bioassay were significantly more likely to be from the lowest-income group (adjusted odds ratio [AOR]: 1.7; 95% CI: 1.1-2.6) and required hospital admission (AOR: 2.1; 95% CI: 1.3-3.5). Unnecessary antibiotic use for febrile illnesses before hospital consultation is common in a low-income, highly populated urban community in Manila. Education targeting this group should be implemented to reduce unnecessary antibiotic use.