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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 469 records · Page 26
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

A Study on Korea Experience in Regulating the Transport of Low and Intermediate-level Radioactive Wastes - 17182

In the Republic of Korea, the transport of Low and Intermediate-Level Radioactive Wastes (LILW) from nuclear power plants (NPP), nuclear research reactor, and radioactive isotope storage facility to Gyeongju LILW disposal facility has increased since the operation license of facility was approved in December 2014. As a result, there has been an increase in public anxiety and concerns over damage or social disorder with LILW transport. So the necessity of strengthened safety confirmation on LILW has been constantly raised. In August 2016, the Enforcement Regulations of the Nuclear Safety Act were amended in relation to the reporting and inspection of LILW transport. The first section of this paper describes the domestic transport regulations including the recently revised contents. The second section shows that the status of domestic and inspection of LILW transport were introduced and transportation regulatory experience was shared in Korea. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Chemical Compatibility of Conventional Geo-synthetic Clay Liners to Aggressive Low-Level Radioactive Waste Leachate - 17296

representative of leachates in low-level radioactive waste (LLW) disposal facilities affected the hydraulic conductivity of a conventional sodium-bentonite (Na-B) geo-synthetic clay liner (GCL) used in composite liner systems. One leachate was highly alkaline (H-pH) and the other had a low abundance of monovalent cations relative to divalent cations (L-RMD). Tests were also conducted with a typical radioactive synthetic leachate (RSL) and with deionized (DI) water as control. Na-B from the GCL had a swell index of 24 mL/2 g in RSL, 22 mL/2 g in H-pH leachate, and 16 mL/2 g in L-RMD leachate, whereas the swell index was 36 mL/2 g in DI water. Hydraulic conductivity tests were conducted in flexible-wall permeameters using the falling headwater with an effective confining stress at 20 kPa. Hydraulic conductivity of Na-B GCL was low when permeated with RSL (1.8 x 10-10 m/s) and H-pH leachate (2.4 x 10{sup -10} m/s), but higher than the hydraulic conductivity to DI water (2.5 x10{sup -11} m/s). The GCL was two orders of magnitude more permeable when permeated with L-RMD leachate (7.9 x 10{sup -8} m/s). The higher hydraulic conductivity of the GCL to the leachates is attributed to reduced swelling of the Na-B in leachates with higher ionic strength solution (H-pH, RSL) or a predominance of divalent cations (L-RMD). Increasing the effective confining stress from 20 to 450 kPa reduced the hydraulic conductivity approximately two orders of magnitude for the L-RMD and H-pH leachates. (authors)

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DOE OSTI2017

Comparison of U.S. Special Permit/Exemption/Authorization Processes for Licensing Type B Radioactive Material Packages - 17513

In the United States, there are three regulators (agencies) associated with licensing Type B radioactive material packages for shipping. They are the U.S. Department of Energy (DOE), the United States Nuclear Regulatory Commission (NRC), and the United States Department of Transportation (DOT). Each regulator has requirements and guidance which implement United States Code of Federal Regulations 10 CFR 71 - Packaging and Transportation of Radioactive Material and 49 CFR 171 through 180 - Hazardous Materials Regulations. Periodically, select packages may not clearly fit within the requirements or guidance. In such cases, the planned shipper can petition the applicable agency for a special permit/exemption/authorization approving the shipment with the specified deviations from the requirements or guidance. If equivalent safety is successfully demonstrated to the applicable regulator, such authorization may be obtained. This paper compares the: 1) DOE exemption process; 2) the NRC special package authorization process; and, 3) the DOT special permit process. (authors)

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DOE OSTI2015

An experimental program to characterize confined explosions of hydrogen/oxygen mixtures in the context of radioactive material transport between CEA nuclear facilities - 15125

CEA, the French Alternative Energies and Atomic Energy Commission, carries out a variety of nuclear research programs in many research reactors and nuclear facilities, decommissioning and dismantling operations that require multiple transports of different type of radioactive material such as radioactive waste or experimental fuel rods using a wide range of transportation packages. Packages are designed, constructed and operated to meet safety and regulatory requirements and must be approved by the French Safety Authority. Application for package approval (e.g. type B package) requires a complete and thorough safety demonstration in which flammable gas generation like radiolysis has become a major issue in the last few years. To face these challenges, the CEA unit in charge of transport package construction and transport operations (CEA/STMR, Cadarache Center, France), has developed a many-sided approach based on one hand on a better knowledge of both hydrogen gas generation and oxygen consumption in waste materials, and on the other hand, on design of transport packages or waste containers capable of withstanding dynamic pressure loads due to an inner hydrogen/air mixture explosion. This paper describes the experimental program that CEA/STMR has decided to implement in order to better characterize confined space explosions of hydrogen/oxygen mixtures. It presents how pressure profiles were experimentally measured in various pressure and temperature conditions which are representative of normal and accident transport conditions. It describes how different experimental set-ups were designed and improved progressively so as to obtain reliable and reproducible detonation conditions, investigate the influence of water vapor and characterize pressure profiles (peak pressure, residual pressure, peak duration). This experimental data will be used to qualify computational models and tools that are used to calculate the structural response of waste containers or fuel canisters, and finally demonstrate the safety of transport packages following an inner hydrogen explosion. (authors)

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DOE OSTI2015

Tracking radioactivity content in the spent fuel pool in PWRS - 15150

Over the last decade, significant progress has been made in managing radioisotopes generated in the PWR primary systems. These include limiting the use of cobalt containing materials, along with improved fuel performance and zinc injection to passivate the corrosion layers to limit the accumulation of Co-58 and Co-60 on primary coolant surfaces. Attention here is directed to the spent fuel pool where accumulation of spent fuel and activated material represents a growing source of mobile radioactivity supplemented with periodic exchanges of water from the reactor cooling system. Characterization of this source is increasingly important to effective management of disposal options. This paper examines the radioactive resin source term as determined from routine sampling of the coolant and fuel pool liquids. Targeted resin stream sampling and radiochemical analyses conducted at Diablo Canyon serve to inform and provide context for this examination. This examination is supplemented with computer modelling using the DW James Consulting computer program, 3R-SCAN. At most operating plants daily reactor coolant samples are counted using gamma spectroscopy and recorded. These samples serve to track fuel performance and maintain water chemistry. A secondary use for this data is to maintain a day to day tracking of releases of radioisotopes important to waste disposal. Given the coolant concentrations, plant operating history, and corroborative mathematical models and sampling, accumulation of these radioisotopes in coolant cleanup system ion-exchangers can be predicted. Fuel pool sampling is less frequent, but at the same time, things change more slowly there. The fuel pool contains about ten times the volume of the reactor coolant but its cleanup system process rate is comparable to reactor coolant letdown, i.e. 5.047 - 8.833 liters per second [lps] (80-140 gallons per minute [gpm]). A spent fuel pool clean-up demineralizer flow rate is typically about 5.047 lps (80 gpm). Concentrations of longer lived radioisotopes in the spent fuel pool can be comparable to those in reactor coolant. Releases of corrosion products and fission products from the stored fuel occur at a low rate but given the overall inventory can be significant. Recognized scaling factor ratios, particularly those to Co-60, are skewed by the aging of the material in the pool. An additional consideration stems from refueling operations themselves. Prior to refueling, it is common practice to conduct a forced oxidation of the corrosion films within the reactor. This spikes the activity within the reactor coolant. Most of this activity is removed by the reactor coolant purification system prior to removing the reactor head. This paper examines the accumulation of activity in the fuel pool purification ion exchange resins and how it impacts the overall classification of those ion exchange resins. The following discussion focuses on the transport of nickel and cobalt isotopes particularly as it relates to determination of Ni-63 which has become increasingly prominent in disposal classification. (authors)

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DOE OSTI2015

The future of RADCALC- the radioactive material and waste transportation software program - 15154

This paper provides a thorough review of RADCALC, the U.S. Department of Energy (DOE)-developed software tool to assist DOE sites in making radioactive material and waste packaging and transportation determinations in compliance with applicable Federal regulations and DOE Orders and Standards. There are more than 400 registered users of the RADCALC tool across the DOE complex. DOE has also authorized a limited number of users from certain universities and foreign countries. DOE periodically upgrades the software consistent with changing domestic and international requirements, as well as needs unique to DOE. Due to recent changes in regulatory requirements by the Department of Transportation (DOT), it is necessary that the current version of RADCALC should be updated in a timely manner. Prior to investing funds in software upgrades, DOE decided to evaluate the nature of changes required and level of operational support and looked for alternatives for meeting sites' operational needs. DOE formed a joint review team consisting of experts from Energy Facility Contractors Group (EFCOG) sites, commercial waste treatment facilities and the National Laboratories. This team evaluated the RADCALC software, commercially available software packages, and internal DOE site-developed programs against an established set of criteria. The team also reviewed and prioritized regulatory changes and user-recommended enhancements. Based on extensive review, the team recommended that DOE continue to support RADCALC noting that RADCALC is a cost-effective, technically robust tool that can perform complicated radioactive material shipping determinations. The tool promotes accuracy, consistency, and completeness between shipments by a single shipper, between shippers at a given site, and between sites. The isotopic coverage, decay algorithm and the range of calculations are without compare either within DOE or in industry in a tool of this kind. The team also recommended that DOE should actively promote RADCALC use throughout the DOE complex and consider cooperative arrangements with the Department of Defense or other government entities that may have similar needs. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Setting an optimal procedure for transferring intermediate radioactive effluents: contribution of laboratory experiments - 15234

The INB35 nuclear facility at the CEA Saclay center is in charge of collecting and treating radioactive effluents generated from site's research activities. The first step of the process consists of an evaporation. The concentrate is an IL radioactive suspension with high salt concentration and suspended matter content. Until the 1990's, the resulting effluents were bitumen coated and since 2012 they have been cemented. During the 1990's, concentrates were stored in 7 tanks, named MA50X with X from 1 to 7. With today's changes to safety regulations, the effluents must now be transferred before tank dismantling. According to the site records, the compositions of the liquid wastes MA501, MA502, MA503, MA504, MA506 and MA507 are similar. From 2013 to 2014, effluents contained in MA502, MA503, MA504 tanks have been transferred. Several difficulties have been encountered: - The tanks were historically different, especially manhole tank access - Since the 1990's, effluent composition has changed due to water evaporation. The salt concentration has increased and insoluble elements have formed a deposit in the bottom of the tanks resulting in high composition heterogeneity - In some tanks, a wall breeze-blade exists dividing the tank into two zones, which complicates the interventions. Major challenges have been met and will be described in the presentation. - To obtain an effluent which is as homogeneous as possible, the duration of mixing has been optimized. - To pump the liquid, it has been necessary to treat the effluent with nitric acid. This step has been optimized thanks to experiments in laboratory and hydrodynamic modelling. - A mobile workshop has been designed in order to perform the effluent transfer. It is set up directly above the tanks and can be moved from one tank to another on rails. Laboratory experiments have been carried out to characterize the suspension and sludge in order to understand its chemical behavior. - The treatment of the effluent has been studied in the laboratory and is now mastered. - The analytical strategy has been optimized. The close collaboration between the analytical laboratory and the dismantling project was a real advantage. Thanks to this, the duration of the last transfer on MA 504 was reduced by half, compared to the first operations. Transfer operations will be described and the differences and similarities between laboratory and in-situ observations will be pointed out. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Design and testing of a sensitive heat-flow calorimeter for measuring radioactive waste in 55-gallon (US) drums - 15334

A sensitive large volume calorimeter for measuring the thermal power generated by the radioactive decay of heat producing materials in 55-gallon drums has been developed and tested. The technology is applicable to the measurement of all radionuclides that decay by alpha or beta decay such as waste or product material containing plutonium, tritium and americium. It is also relevant to the measurement of drums where gamma rays from radioactive decay are captured locally in attenuating materials and the resulting rate of thermal energy deposition (sample thermal power) is of sufficient magnitude for a calorimetry measurement. The initial application of the drum calorimeter is for measuring and characterizing plutonium-bearing waste drums. The calorimeter design has optimised the two characteristics of high thermal sensitivity and reduced measurement times. In order to achieve these characteristics for large volume drums, the heat-flow method of operation was chosen in a single measurement chamber configuration. Using a drum-lifting device, the waste drum is positioned on a sliding platform and subsequently moved by the operator inside an octagonal inner chamber, which acts as the heat sink. The heat-sink chamber has significant thermal inertia and remains effectively at a constant temperature throughout the duration of a drum measurement. The octagonal inner measurement chamber is constructed inside a multi-layer external thermal enclosure, which provides both thermal insulation and significant thermal inertia. The external thermal enclosure isolates the measurement process from the effects of changes in the external ambient temperature. During a measurement, heat flows from the drum being measured to the heat sink. This heat-flow is measured by a series of thermopile sensor assemblies, which are positioned between the waste drum and the heat sink. A high sensitivity of greater than 200 μV/mW (microvolts per milliwatt) has been achieved using multiple close-coupled thermopile sensors in each assembly. Reduced measurement times are achieved by positioning the thermopile sensor assemblies close to the drum surface in order to reduce the thermal transport delay. The instrument incorporates a calibrated precision power supply, which requires annual recalibration and is traceable to national standards. It is used to power a calibration heater that has been built into a calibration and test drum. The calibration and test drum is used both to calibrate the instrument and provide periodic confirmation of correct operation. Measurements of calorimeter performance are reported for a range of drum thermal powers. Long-term temperature drift measurements are used to provide estimates of the minimum detectable drum thermal power and hence the minimum detectable activity for both plutonium and tritium. (authors)

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DOE OSTI2015

Ecological risk assessment software for radioactive waste management facilities - 15464

Environmental Risk Assessment (ERA) is a systematic process which can be used to identify, quantify and characterize the risk posed by contaminants and physical stressors in the environment on biological receptors, including human and non-human biota. Human receptors can be addressed through a Human Health Risk Assessment and non-human biota can be addressed through an Ecological Risk Assessment (EcoRA). In 2012 the Canadian Standards Association (CSA) issued a CSA standard N288.6 Environmental Risk Assessment at Class I Nuclear Facilities and Uranium Mines and Mills. The standard applies to radioactive waste management facilities, as well as nuclear power plants. The ERA can also be used retrospectively, to estimate the effect that a contaminant or stressor has already had on the environment; and can be used to assist in determining possible remedial or risk management options to return the quality of the affected environment to a predetermined state. While industry-standard tools for estimating radiological doses to humans are readily available, there is a need for a tool to implement the CSA N288.6 standard for Ecological Risk Assessment. To address this need, AMEC developed a computer code called -Assessment of Impact of Contaminants on Ecological Receptors (AICER)-. The code was developed to be consistent with the CSA standard N288.6, by assessing the impact of radiological contamination on non-human biota for normal operations. The code can also be used to screen for the potential effects of conventional contamination on the environment. In addition, the code is also applicable for abnormal events and accidents with the appropriate modification of some parameters used in the code. The AICER code was qualified in accordance with the requirements of the Canadian quality standard CSA N286.7 -Quality Assurance of Analytical, Scientific, and Design Computer Programs for Nuclear Power Plants-. The functionality of the code is illustrated with screen shots in the paper, following the introduction of the theoretical model for the dose calculation. An inter-comparison with RESRAD-BIOTA and ERICA Assessment tool at the conceptual level was conducted. In addition, the practical application of the code for the assessment of the impacts related to an existing radioactive waste management facility is provided. In particular, analysis was conducted to evaluate radiological doses to non-human biota resulting from a fire accident scenario involving an existing waste storage and processing facility. Analysis included all feasible waterborne and airborne exposure pathways. Doses were evaluated for a range of species representing valued ecological components at various trophic levels and compared with benchmark values for indicator species. (authors)

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DOE OSTI2018

Issues Affecting the Assessment of Impacts of Disposal of Radioactive and Hazardous Waste - 18177

BIOPROTA is an international collaborative forum that seeks to address the key uncertainties in long term assessments of radionuclide and other contaminant releases into the environment arising from solid radioactive waste disposal. (see www.bioprota.org). Two workshops have been organized through the BIOPROTA forum to consider the non-radiological post-disposal impacts of radioactive waste disposal. In these workshops, it was noted that, ideally, the assessment of radionuclides and hazardous materials should be such that consistent assumptions and criteria are used in risk evaluation and that in cases where this is not possible, a clear understanding of the reasons why needs to be provided. Accordingly, a detailed review was organized through BIOPROTA to provide information to support development of a consensus on how to address this issue. The study suggests that there is a continuing need for some common measure of hazard that supports identification of risk management priorities for mixed hazardous waste. Without it, effective optimization of management of these mixed hazardous wastes will be restricted. (authors)

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DOE OSTI2019

Recovery of Buried Discrete Radioactive Objects by Channel-Specific Analysis of Gamma Scan Data - 19110

During radiological clean-up activities, gamma scan survey data collected from across an industrial site revealed spatially varying concentrations in radioactivity related to naturally occurring soil constituents or dissimilar fill materials that may have been placed at the site. A three-step process was developed to assess the data in order to identify radiation anomalies that most likely could represent small, buried discrete radioactive objects (ROs). The three-step process consists of: (1) define the radiological composition of the site isotopically, (2) spatially parse the data into more homogeneous subgroupings based on the site's radiological composition, and (3) perform analysis of the data subgroupings to identify statistical and spatial population outliers (i.e., radiation anomalies) that otherwise may be masked by the overall variation in the data. Using this process, locations containing buried low-specific activity ROs or higher activity ROs whose radiation is attenuated by soil can be identified and investigated. (authors)

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DOE OSTI2019

Revisit of Phase Separation of Borosilicate Glass for Vitrification of High Level Radioactive Waste - 19166

Phase separation of glasses for vitrification of high level radioactive waste has been experimentally studied to know uncertainty of the phase separation conditions; temperature and glass compositions as well as to propose possibility of recycling of heavy metals on a longer time scale. Our current results at 700 and 750 deg. C for ternary borosilicate glass system showed that the phase separation region at 700 deg. C spreads along with a direction of increasing Na{sub 2}O composition when compared with the data in literature. On the other hand, at the glass composition employed in the reprocessing plant in Japan, no segregation was observed. Concerning the application of phase separation phenomena, we had previously reported recovery of almost all elements from the vitrified borosilicate glass containing simulated fission products through phase separation of the borosilicate glass and this suggests that we could propose that after vitrification of high-level radioactive waste heavy metals may be retrieved as energy or materials resources when they are required to utilize on a longer time scale. From this standpoint, phase separation of lead borosilicate glass was also studied to seek economical and versatile recycling. (authors)

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DOE OSTI2019

Overview of Novel Canadian Approaches for Decontamination of Solid Radioactive and Mixed Waste - 19476

The availability of decontamination-for-release technology options provide a path forward to economically decontaminate radioactive metallic waste and enable its disposal as clean scrap metal, to decontaminate tools/equipment for unrestricted refurbishment or reuse, and to carry out inspection of in-service components. Projects involving management of solid mixed waste pose complex technical challenges that require the development and application of novel decontamination processes. The objective of this paper is to present project case studies that provide an overview of Kinectrics' experience with the decontamination of solid radioactive and mixed waste. Novel approaches to decontamination, recent developments, key results and lessons learned will also be discussed. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Optimization of Energy Absorbing Materials for Radioactive Packaging

Most applications of honeycomb materials are for lightweight structural members. This area of honeycomb material behavior has been extensively researched. In packaging design, honeycomb materials are used for impact energy absorption. Data concerning the deformation of stainless steel honeycomb structures is not widely available. It is being proposed to use honeycomb stainless steel as an impact energy absorber in a new radioactive material shipping package. Honeycomb materials can be useful for energy absorption applications because the geometry causes the structure to have a constant compressive force during buckling. This Finite Element Analysis is used to predict the compressive buckling strength of various honeycomb structures in order to design an energy absorber for a specific package weight and geometry. The model simulates a drop test where a 50 pound plate lands flat on a honeycomb stainless steel structure from a height of 30 feet. Some actual drop tests have already been performed on real specimens. Most of the resulting deformations of these tests include the bottom of the honeycomb structure buckling while the rest of the structure was either slightly deformed or straight. The deformation of the simulated honeycomb specimen matched the deformation from the actual drop tests. This is one verification that the model is working properly. It is evident that after about 2.5% strain the strength is a relatively linear value, and eventually flattens out to about 875 psi. The crush strength measured from the actual drop tests for the same density honeycomb was just over 800 psi. This is further verification that the model is working correctly. The results gathered from the FEA model successfully mirror the results from a real drop test. The deformation looks realistic, and the compressive strength found is consistent with the empirically measured value. Future work for this model includes modeling different density honeycomb structures which have not been tested before. This will allow for optimization of honeycomb stainless steel use for impact absorption in radioactive material shipping packages.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Non-destructive Material Characterization of Radioactive Waste Packages with QUANTOM{sup R} - 20130

During the last decades, the nuclear and non-nuclear industry has produced a considerable amount of low and intermediate level radioactive wastes. In many countries, it is foreseen to dispose of these waste packages safely in a final repository under strict waste acceptance requirements (e.g. the radiological and material characterization) defined by national licensing and supervisory authorities. The material characterization of waste packages can be performed on the basis of existing documentation or, if the documentation is insufficient, on further destructive or non-destructive analysis. Non-destructive methods are to be preferred to minimize radiation exposures of operating personnel as well as costs. We offer here an innovative non-destructive technology called QUANTOM{sup R} (Quantitative Analysis of Toxic and nontoxic Materials) based on prompt and delayed gamma neutron activation analysis (P and DGNAA). This technology is able to identify, quantify and thus verify the amount of hazardous and non-hazardous substances in 200-l radioactive drums, which is required for a final disposal characterization. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Improvement of Quantification in Non-Destructive Characterization of Radioactive Waste Packages -Introduction to the Ideas of a Currently Started Project - 20197

For the characterization of radioactive waste packages (e.g. 200-L drums) with inhomogeneous activity and density distributions preferably non-destructive methods like segmented gamma scanning and transmission measurements are applied. While the first is used for identification of the radioactive inventory, both are necessary for a reliable activity quantification. Although at nuclear facilities, segmented gamma scanner are state-of-the-art and are mostly equipped with a vertically moveable transmission source the combining of this data and further a-priori information improving accuracy in data estimation giving realistic, reliable and traceable uncertainty estimations are improvable. This challenge is tackled by a R and D project funded by the German Federal Ministry of Education and Research (BMBF). The boundary conditions for this project are (i) taking into account the typical instrumentation available at nuclear facilities, (ii) minimize the overall measurement time and (iii) combining data from emission and transmission measurements by Bayesian methods. For the R and D project, being still in its initial stage, the basic ideas, the data used for evaluation and the result of a preliminary investigation demonstrating the proof of principle are presented. Additionally, some first aspects on spatial resolution in emission and transmission measurements are addressed. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Decision-making Framework to Support the End-of-Life Management of High-Activity Radioactive Sources

Sealed radioactive sources are used for an extremely wide range of purposes, including as radiation sources in various nuclear facilities, universities, hospitals, biomedical and industrial applications, as well as for geological prospecting and exploration. The proper management of these sealed radioactive sources once they have reached their end-of-life is still a global challenge. Very few field-tested Mobile Hot Cells (MHC) exist in the world to support the end-of-life management of disused high-activity radioactive sources. One of the critical challenges that a source recovery team encounters on each site is to find adequate space, and access to that space to complete the staging and recapture process with a MHC; this process is tedious, uncertain, and time-consuming. Therefore, it is essential to understand the planning and management of the MHC setup and takedown. Currently, there isn't an single capable planning tool that will provide an effective medium to plan, inform, educate, and communicate knowledge about the environment where the MHC will be deployed. We present a novel approach for the planning and management of the deployment of the MHC to several different sites that relies on technology. Furthermore, this planning tool will support collaboration, promote information sharing, and foster better decision making processes, resulting in a more efficient, and reliable setup and source recovery with the MHC.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2012

Intraluminal Radioactive Stent Compared with Covered Stent Alone for the Treatment of Malignant Esophageal Stricture

Objective: This study was designed to compare the clinical effectiveness of intraluminal radioactive stent loaded with iodine-125 seeds implantation versus covered stent alone insertion in patients with malignant esophageal stricture. Methods: We studied two groups of patients with malignant esophageal stricture. Group A comprised 28 patients (19 men and 9 women) who underwent intraluminal radioactive stent loaded with iodine-125 seeds implantation and were followed prospectively. Group B comprised 30 patients (18 men and 12 women) who had previously received covered stent alone insertion; these patients were evaluated retrospectively. There was no crossover between the two groups during follow-up. Informed consent was obtained from each patient, and our institutional review board approved the study. The dysphagia score, overall survival rates, complication rates, and reintervention rates were compared in the two groups. Results: There were no significant differences between the two groups in terms of baseline characteristics. Stent placement was technically successful and well tolerated in all patients. The dysphagia score was improved in both groups after stent placement. The median survival was significantly longer in group A than in group B: 11 versus 4.9 months, respectively (P < 0.001). The complications of chest pain, esophageal reflux, and stent migration was more frequent in group B, but this difference did not reach statistical significance. There was no statistical difference in reintervention between two groups. Conclusions: Intraluminal radioactive stent loaded with iodine-125 seeds implantation was a feasible and practical management in treating malignant esophageal stricture and was superior to covered stent alone insertion, as measured by survival.

62 RADIOLOGY AND NUCLEAR MEDICINE↗