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All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Measurement Quality Objective Considerations for Residual Alpha Surface Radioactivity Measurements at Remediation Sites - 16522

Surveys for residual alpha surface radioactivity on materials, equipment, and other non-porous surfaces are most-commonly performed using the traditional 'best available' equipment. The best available equipment for measuring alpha surface radioactivity includes portable gas-flow proportional or scintillation detectors with active surface areas typically ranging from 50 to 800 square centimeters (cm{sup 2}). The measurement equipment is offered by numerous service providers and instrument vendors applying basic detection principles that date back to the early-Atomic Age. The detector face is held as close to the surface as possible and scanned at a typical rate of 2 to 5 centimeters per second (cm/s). Any audible or visual indication of elevated alpha activity would result in a more careful assessment of the possible 'hot spot'. Modern surface contamination surveys continue to employ these best available technologies while endeavoring to achieve ever more restrictive measurement quality objectives and screening or release criteria. This paper considers the possibly unforeseen consequences to sensitivity and precision as scanning speeds are reduced to meet constraining measurement quality objectives. Sensitivity calculations are conducted to demonstrate how application of real-world efficiencies and the most common survey approaches are inadequate to meet sensitivity objectives based on American National Standard N13.12-2013, Surface and Volume Radioactivity Standards for Clearance screening criteria. Alternate data collection and assessment methodologies are offered to enhance the defensibility of collected data with stakeholders. Limitations of available or applied alpha survey technologies should be considered when establishing survey programs at project sites where alpha-emitting radionuclides are expected to contribute to residual surface radioactivity. These include but are not limited to nuclear fuel-cycle sites, military/weapons/energy research and production facilities, and 19.-20. Century-era commercial/ industrial/research/medical facilities where radioactive materials were handled. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Development and Delivery of National Strategic Approaches to Radioactive Waste Management in the UK - 16595

The UK has limited capacity for the management of radioactive wastes. Such capacity is needed to enable nuclear and non-nuclear sectors to continue to make responsible use of radioactive substances (e.g. for health care provision) as well as to deliver key objectives such as the decommissioning and clean-up of the civil nuclear legacy. A revision in UK LLW policy in 2007 led to the development of a range of national LLW strategies, each focusing on a particular radioactive waste sector. For various reasons each of the strategies was developed in a different way. The success of these strategies to-date includes the development of a range of new routes for LLW management, greater transparency and understanding of the challenges facing LLW management across the UK, and recognition of priorities for national work, e.g., to address issues such as the management of radioactive asbestos waste and the treatment of metallic radioactive wastes. The work also revealed common issues across each of the waste producing sectors, and the opportunity to improve future waste management through further integration. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Radioactive Test Platform Demonstrating the Direct Feed Low-Activity Waste Process at Hanford - 18628

The Hanford Site is implementing a phased process for treating liquid legacy radioactive waste. The process is called the Direct Feed Low-Activity Waste process (DFLAW). The DFLAW includes a filtration and ion-exchange unit in the Low-Activity Waste Pretreatment System (LAWPS) that removes solids and cesium. This pretreat feed is then sent to a vitrification facility where glass-forming solids are added and the waste is turned into glass in a melter. The melter off-gas liquid condensate is evaporated in the Effluent Management Facility (EMF), with the concentrate recycled back into the feed. A laboratory-scale test platform has been built to test various aspects of the DFLAW process using samples of actual radioactive waste. The current platform represents four of the major unit operations: filtration, ion-exchange, vitrification, and off-gas condensation. A lab scale EMF is being designed to supplement the current platform. The platform is expandable to add any or all unit operations. The radioactive test platform has three missions. The first is simply to demonstrate the DFLAW process with real waste. A second objective is to address flowsheet gaps and opportunities, some of which require testing with real waste. Some gaps and opportunities include addressing the fate of mercury in the flowsheet, evaluating salt and aluminum precipitation, and proving new glass formulations with real waste. A third objective of the radioactive test platform is to demonstrate or evaluate future operations and alternative flowsheets. Some of the future operations evaluated will be the life expectancy of the resin in real waste, resin sampling, evaluation of how waste affects melter throughput, and grouting various primary and secondary waste streams. The radioactive test platform is therefore a crucial tool for verifying and optimizing the DFLAW flowsheet at Hanford with real waste. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Environmental evaluation of radioactivity levels and associated radiation hazards in groundwater around the WIPP site

Groundwater may contain radioactive substances which can be dangerous to human health. Concentrations of natural radionuclides polonium (Po), thorium (Th), uranium (U), and radium (Ra) isotopes were measured in groundwater samples collected from different locations in the vicinity of the Waste Isolation Pilot Plant (WIPP) site in Carlsbad, New Mexico. The average values of gross activity concentrations of 210 Po, 228 Th, 238 U, 234 U, 226 Ra and 228 Ra isotopes were determined to be 1.62 Bq L -1 in shallow groundwater and 5.88 Bq L -1 in deep groundwater, respectively. The total radioactivity in deep groundwater was higher than that in shallow groundwater, and most of the radioactivity in the water is from 226 Ra. Furthermore, the effective doses for ingestion of natural radionuclides were about 0.333 mSv y–1 for shallow groundwater and about 1.338 mSv y –1 for deep groundwater samples, which are higher than the World Health Organization (WHO, 2017) guideline level (0.1 mSv y –1 ) for drinking water. Ra dominated the total ingestion dose, contributing 93.06 % and 75.40 % of the total effective doses to the deep and shallow groundwater, respectively. The ingrowth and decay of natural radionuclides suggested that 228 Ra/ 226 Ra ratio can be a useful indicator of the source of radioactive contamination. The radioactivity data obtained from the investigated groundwater samples can be used to establish a baseline for radioactivity levels in groundwater around the WIPP site.

63 RADIATION, THERMAL, AND OTHER ENVIRON. POLLUTAN↗
DOE OSTI2023

Review of particle deposition to and removal from clothing, skin, and hair after a radioactive airborne dispersal event

Explosive Radiological Dispersal Devices (RDD) – aka dirty bombs – are seen as a credible method to carry out a radiological terror attack. After exploding a radioactive source, the radionuclide-laden plume will be blown downwind of ground zero, with particles falling out and potentially depositing on people caught in and under the cloud. Some of these people may not show any sign of radiation sickness and therefore not realize they have been contaminated and may take the radioactive particulate with them on their daily activities, thus spreading the radioactive particulate outside the initially contaminated area. This paper reviews the scientific literature to better understand the rate at which particulate deposits on and is removed from the different “surfaces” of a person, i.e., hair, skin, and clothing. Prior research indicates that: 1) particle deposition is usually higher on skin than on hair and clothing; 2) particle deposition is greater for a person with higher skin moisture, 3) stronger wind increases the deposition flux onto a person, and 4) the fraction of particulate deposited on the hair, skin, and clothing respectively depends on the length of the hair, assuming all the hair surface is available for deposition. The studies taken into consideration show that the largest uncertainty in particulate deposition onto a person is due to clothing type because of the different possible weave arrangements and tightness which translate into differences in actual surface area and surface roughness. A factor of 2-to-20 variation in deposition rate was found. Removal of the particulate from the contaminated person may be due to wind, a person's movement, and/or contact transfer, i.e., by touching a different clean surface. Experiments show that the majority of the particulate is resuspended within 2–6 h mostly depending on the intensity of physical activity. The largest uncertainty in particulate removal from skin depends on the skin moisture, transfer rate of single-contact, and how many objects/people a person touches per hour. No data for hair were found for particle removal and resuspension. The studies considered did not utilize radionuclides directly; however, data on adhesion of radioactive vs. their non-radioactive counterpart have shown that the uncertainty due to the radioactivity of the particles is lower than that due to other factors. In conclusion, an idealized scenario involving a single building in the path of the cloud showed the impact of building-influenced flow on the cloud transport path and mixing, which affects the radiological dose the downwind population is exposed to and consequently the health effects.

63 RADIATION, THERMAL, AND OTHER ENVIRON. POLLUTAN↗
DOE OSTI2020

Impact of boundary layer simulation on predicting radioactive pollutant dispersion: a case study for HANARO Research Reactor using the WRF-MMIF-CALPUFF modeling system

Wind plays an important role in cases of unexpected radioactive pollutant dispersion, deciding distribution and concentration of the leaked substance. The accurate prediction of wind has been challenging in numerical weather prediction models, especially near the surface because of the complex interaction between turbulent flow and topographic effect. As such, in this study, we investigated the characteristics of atmospheric dispersion of radioactive material (i.e. 137 Cs) according to the simulated boundary layer around the HANARO research nuclear reactor in Korea using the Weather Research and Forecasting (WRF)-Mesoscale Model Interface (MMIF)-California Puff (CALPUFF) model system. We examined the impacts of orographic drag on wind field, stability calculation methods, and planetary boundary layer parameterizations on the dispersion of radioactive material under a radioactive leaking scenario. We found that inclusion of the orographic drag effect in the WRF model improved the wind prediction most significantly over the complex terrain area, leading the model system to estimate the radioactive concentration near the reactor more conservatively. We also emphasized the importance of the stability calculation method and employing the skillful boundary layer parameterization to ensure more accurate low atmospheric conditions, in order to simulate more feasible spatial distribution of the radioactive dispersion in leaking scenarios.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2020

Opportunities in the Synthesis and Design of Radioactive Thin Films and Nanoparticles

Studies of radioactive isotopes at the liquid-solid or gas-solid interface are enabling a detailed mechanistic understanding the effects of radioactive decay on physical, biological, and chemical systems. In recent years, there has been a burgeoning interest in using radioactive isotopes for both imaging and therapeutic purposes by attaching them to the surface of colloidal nanoparticles. By merging the field of nanomedicine with the more mature field of internal radiation therapy, researchers are discovering new ways to diagnose and treat cancer. In this perspective, we discuss state-of-the-art radioactive thin films as applied to both well-defined surfaces and more complex nanoparticles. Here, we highlight the design considerations that are unique to radioactive films, which originate from the damaging and potentially self-destructive emissions produced during radioactive decay and highlight future opportunities in the largely underexplored area between radioisotope chemistry and nanoscience.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2018

Radioactivity Calculator for Research and Test Reactor Neutron-Activated Materials

The Idaho National Laboratory (INL) Nuclear Science User Facility (NSUF) supports nuclear energy-relevant research into the effects of radiation on nuclear fuels and structural materials. The NSUF provides access, at no cost to the researcher, to the Advanced Test Reactor (ATR), the High-Flux Isotope Reactor (HFIR), and other university and national laboratory research and test reactors for radiation effects experiments. Post-irradiation examination of materials poses some difficulties when studying neutron-irradiated materials due to the radioactivity levels of the materials. Inside a test reactor, under high neutron flux, stable materials can activate (capture a neutron) and become unstable or radioactive. These newly created radioactive materials decay, based on their respective half-lives, and emit energetic radiation. Nuclides with short half-lives are able to decay to sufficiently low radiation levels prior to the sample being removed from the reactor. However, nuclides with longer half-lives present a challenge since they must be handled with extreme caution and stored until their radioactivity levels decrease. If immediate examination is required, the examination may take place in a hot cell, which can dramatically increase the cost of the project. The NSUF has access to a variety of hot cells and characterization capabilities at national laboratories, universities, and industrial sites across the United States. In order to estimate the post-irradiation radioactivity levels of materials early in the experiment cycle, we created a radioactivity calculator, hosted on the NSUF web site (nsuf.inl.gov), which will consider three stages of neutron activations and decays for sample materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Real breakthrough in detection of radioactive sources by portal monitors with plastic detectors and New Advanced Source Identification Algorithm (ASIA-New)

Radiation Portal Monitors (RPM) with plastic detectors represent the main instruments used for primary border (customs) radiation control. RPM are widely used because they are simple, reliable, relatively inexpensive and have a high sensitivity. However, experience using the RPM in various countries has revealed the systems have some grave shortcomings. There is a dramatic decrease of the probability of detection of radioactive sources under high suppression of the natural gamma background (radiation control of heavy cargoes, containers and, especially, trains). NORM (Naturally Occurring Radioactive Material) existing in objects under control trigger the so-called 'nuisance alarms', requiring a secondary inspection for source verification. At a number of sites, the rate of such alarms is so high it significantly complicates the work of customs and border officers. This paper presents a brief description of new variant of algorithm ASIA-New (New Advanced Source Identification Algorithm), which was developed by the Rapiscan company. It also demonstrates results of different tests and the capability of a new system to overcome the shortcomings stated above. New electronics and ASIA-New enables RPM to detect radioactive sources under a high background suppression (tested at 15-30%) and to verify the detected NORM (KCl) and the artificial isotopes (Co- 57, Ba-133 and other). New variant of ASIA is based on physical principles, a phenomenological approach and analysis of some important parameter changes during the vehicle passage through the monitor control area. Thanks to this capability main advantage of new system is that this system can be easily installed into any RPM with plastic detectors. Taking into account that more than 4000 RPM has been installed worldwide their upgrading by ASIA-New may significantly increase probability of detection and verification of radioactive sources even masked by NORM. This algorithm was tested for 1,395 passages of different transports (cars, trucks and trailers) without radioactive sources. It also was tested for 4,015 passages of these transports with radioactive sources of different activity (Co-57, Ba-133, Cs-137, Co-60, Ra-226, Th-232) and these sources masked by NORM (K-40) as well. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2017

Assessment on Transport of Radioactive Species from Nuclear Facilities - 17079

It is generally recognized that radioactive nuclides are most hazardous fission products that could be released in the event of reactor accidents. The reactor coolant system of a pressurized water reactor (PWR) nuclear plant exhibits radioactivity from various sources. These radiation sources consist of the fission products from fuel element and activated products from corrosion products. The complex infrastructure of the processes does not allow detailed measurement of activity until after the reactor is shutdown. All of the components that require dismantling will be radioactive which was caused by neutron activation during operation, contamination by corrosion products, contamination by damaged by fuel rods and leaks by equipment. Due mainly to the various operation modes and their aging, there are the increasing needs to carry out repair works that are not sufficiently provided for by the original design. Many difficulties are involved in these tasks, namely, high radiation doses, cramped areas, hard to reach components, coupled with stringent quality conditions. The need to simulate the behavior of corrosion products in a nuclear power plant environment has attracted many researchers' attention. It has performed the in-depth reviews of the various theories and physics associated with the corrosion products transportation, the laboratory loop testing, and the field measurements in operating plants. Before we start the decommissioning of nuclear power plants, workers, the public and the environment shall be properly protected from radiation hazards. For this purpose, a safety assessment should be carried out relating to the potential hazards that may occur during the decommissioning. The radiological hazards involved in the decommissioning activities also should be identified, and be taken into account in the safety assessment. The radiation conditions of the primary circuit are determined by the activity of corrosion and fission products that have accumulated on the internal surfaces of the components. Therefore the estimation of contamination on the components is very important and, may be the first step in the reactor operation and decommissioning of components. The predominant radionuclide in the PWR primary coolant is Co-60, although Co-58 contributes significantly to radioactivity level at the early stage of a reactor operation. Since Co-60 is produced from the in-core Co-58, which is the source of cobalt in a reactor coolant system, many researchers have investigated its transmutation in the core, its transport around the system, and its subsequent deposition on surfaces. The major parameter influencing the behavior of radioactivity, such as Co-60 in the coolant, is the corrosion product oxides. The objective of this work is to analyze the corrosion product activity transported in the primary circuit by using a multi-region model. This model, which is incorporated with the mass and activity transport between the dissolved corrosion products in the coolant and the surface, describes the specific activity of corrosion products in coolant and on the surfaces according to the operation modes as well as decommissioning stage. In order to evaluate the analysis of the corrosion products in an integral type reactor, a multi-region mass balance model is applied. The specific activities of the various species on the surface of core and steam generator throughout the effective full power hours, 15,000 hours, are simulated for the various modes of operations. As a result, the specific activities of corrosion products from the surface of primary circuit were shown as very low compared to those of other commercial nuclear power plants. In this simulated reactor, a significant reduction of specific activities from the activated corrosion products is obtained due to the introduction of titanium alloy to the steam generator tube and the avoidance of boric acid induced corrosion. From this result we conclude that we are able to predict the level of radioactive nuclides generated by corrosion products in a primary circuit according to the various modes of operations by using multi-region model. In this point of view it is concluded that this multi-region model can be used as an engineering tool for analysis of the corrosion products within the primary coolant for the nuclear power plants. It is also considered that it is possible to extend to a calculation tool for estimating the radiological hazards involved in the decommissioning activities in terms of safety assessment. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2016

Study on Gaseous and Liquid Radioactive Effluents in AP1000 Plants

During normal operation, the gaseous and liquid radioactive effluents will be released in nuclear plants. It is the base of effluents management and environment influence evaluation to analyze the anticipated annual activity releases. In China, both the annual release of radioactive liquid and gaseous effluents limit value (for single plant and whole-site) and the liquid concentration of discharge canal are introduced in GB6249-2011. This standard is mandatory and the limit values must be met. Because the limit values are very strict, it is meaningful to determine a suitable and logical method of liquid and gaseous effluents in China. In this paper, according to the operation actuality and related rules of law in China, the analysis method of nuclides except H-3 and C-14 in the gaseous and liquid effluents of AP1000 plant is introduced. The method and results can be reference for effluents analysis for AP1000 plants. According to the revised method, the gaseous and liquid radioactive effluents for Sanmen 1 and 2 plants are calculated. The results and the comparison with the limit value in GB6249-2011 are given. The liquid concentration of discharge canal is 868.4 Bq/l, which is less than the limit value 1000 Bq/L in GB6249-2011. It can be seen that according to the revised method, the gaseous and liquid radioactive effluents for Sanmen 1 and 2 plants can meet the limit value in GB6249-2011. This revised method considers all the source, the detailed purification process and release pathways for gaseous and liquid radioactive effluents. The effluents calculation in the revised method can be a possible reference to the analysis of the gaseous and liquid radioactive effluents in China, though there may be some detailed optimization and improvement in future. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

A Low-Tech, Low-Budget Storage Solution for High Level Radioactive Sources

The need for safe, secure, and economical storage of radioactive material becomes increasingly important as beneficial uses of radioactive material expand (increases inventory), as political instability rises (increases threat), and as final disposal and treatment facilities are delayed (increases inventory and storage duration). Several vendor-produced storage casks are available for this purpose but are often costly — due to the required design, analyses, and licensing costs. Thus the relatively high costs of currently accepted storage solutions may inhibit substantial improvements in safety and security that might otherwise be achieved. This is particularly true in areas of the world where the economic and/or the regulatory infrastructure may not provide the means and/or the justification for such an expense. This paper considers a relatively low-cost, low-technology radioactive material storage solution. The basic concept consists of a simple shielded storage container that can be fabricated locally using a steel pipe and a corrugated steel culvert as forms enclosing a concrete annulus. Benefits of such a system include 1) a low-tech solution that utilizes materials and skills available virtually anywhere in the world, 2) a readily scalable design that easily adapts to specific needs such as the geometry and radioactivity of the source term material), 3) flexible placement allows for free-standing above-ground or in-ground (i.e., below grade or bermed) installation, 4) the ability for future relocation without direct handling of sources, and 5) a long operational lifetime . ‘Le mieux est l’ennemi du bien’ (translated: The best is the enemy of good) applies to the management of radioactive materials – particularly where the economic and/or regulatory justification for additional investment is lacking. Development of a low-cost alternative that considerably enhances safety and security may lead to a greater overall risk reduction than insisting on solutions that remain economically and/or politically ‘out of reach’.

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

Sixth National Report for the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management

The United States of America (U.S.) ratified the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management (Joint Convention) on April 9, 2003. The Joint Convention establishes an international peer review process among Contracting Parties and provides incentives for nations to take appropriate steps to bring their spent fuel and radioactive waste management activities into compliance with general safety standards and practices. An extensive U.S. legal and regulatory structure ensures the safety of spent fuel and radioactive waste management. This report describes radioactive waste management in the U.S. in both the commercial and government sectors, and provides annexes with information on spent fuel and waste management facilities, inventories, and ongoing decommissioning projects. It also provides detailed information on spent fuel and radioactive waste management safety, as well as transboundary movements (imports/exports) and disused sealed sources, as required by the Joint Convention.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Radioactive Waste Water Treatment for Fukushima Daiichi Nuclear Power Plant - 16081

The Great East Japan Earthquake of 11 March 2011 caused serious damage to the Fukushima Daiichi Nuclear Power Station (NPS) of the Tokyo Electric Power Company, Inc.. Toshiba and Shaw/AVANTech installed a Simplified Active Water Retrieve and Recovery System (SARRY{sup TM}) to remove radioactive cesium (Cs) from the contaminated water. Toshiba, in partnership with EnergySolutions, also developed and installed the Multi Radionuclide Removal System (MRRSTM), which can remove radioactive nuclides being treated, except tritium, to a level less than the concentration limit in water. As contaminated water treatment has progressed, the composition of the contaminated water has changed. Both salinity and radioactive Cs concentration in the contaminated water have been decreasing. Thus, a new adsorbent that is suitable for the present water chemistry conditions is required. In order to provide an appropriate contaminated-water treatment system, Toshiba has been examining the removal of not only radioactive Cs but also radioactive strontium (Sr) by SARRY{sup TM}. SARRY{sup TM} has been operated for removing Cs by using a Cs adsorbent. Instead of a Cs adsorbent, a new media that has high selectivity for both Cs and Sr should be installed. This paper describes the results obtained when using the newly developed adsorbent for removing Cs and Sr at once. As the Cs and Sr adsorbent, we have selected crystalline silico-titanate (CST). The distribution coefficient (K{sub d}) of CST was determined by the batch method. The lifetime of the adsorbent was evaluated by a column examination. The selected CST had a K{sub d} value of over 10{sup 4} ml/g for both Cs and Sr. The highest-selectivity CST had a lifetime (defined as the period for the outlet concentration to decrease to one tenth of the inlet level) of more than 125000 BV (bed volume) for Cs and more than 18000 BV for Sr, when the test was conducted at 10 BV/h. Calculation code was also developed for estimating the relation between the BV and the decontamination factor (DF) using an adsorption rate model for diffusion and adsorption. The calculation results of the Cs and Sr adsorption behavior agreed well with the experimental results. To evaluate the replacement cycle of adsorption tower in actual equipment, the radiation exposure during the replacement operation and the required Cs and Sr removal performance were used as limiting conditions for the estimation. The results of the evaluation showed that the selected CST has sufficient performance for practical usage. Based on this finding, Toshiba proposed loading the CST in SARRY{sup TM}, and SARRY{sup TM} loaded with this CST started removing Cs and Sr in December 2014 resulting in decreased contaminated water. It has been continuously working since then. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Regulatory Treatment of Non-Core Sources of Radioactivity for Advanced Reactor Designs

The recent resurgence in advance (non-light water) reactor development has been paralleled by the development of risk-informed performance-based (RIPB) licensing pathways. Specifically, the creation of the RIPB Licensing Modernization Project (LMP) approach and subsequent endorsement by the U.S. Nuclear Regulatory Commission (NRC) now provides advanced reactor vendors with a defined RIPB method to develop an affirmative safety case for licensing. In addition, the Technology Inclusive Content of Applications Project (TICAP) has published guidance on developing a license application based on the LMP approach. To support the utilization of risk information as part of advanced reactor design and licensing efforts, the American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) Joint Committee on Nuclear Risk Management (JCNRM) has developed a probabilistic risk assessment (PRA) standard for advanced reactors. The standard, which was formerly approved by the American National Standards Institute (ANSI) in 2021 and recently endorsed by the NRC in trial use Regulatory Guide (RG) 1.247, is an integral standard, covering from initiating events to offsite consequence. A major feature of the standard is that it permits the inclusion of any source of radioactivity material at the site within the plant PRA. Therefore, non-core sources of radioactivity, such as fuel storage, fuel processing, and purification systems, can be included within a single comprehensive plant PRA. For those advanced reactor vendors utilizing a RIPB licensing approach, there is an opportunity to include the non-core sources of radioactivity within the RIPB framework for licensing decision-making, such as the categorization of events, classification of structures, systems, and components (SSCs), and evaluation of the adequacy of defense-in-depth (DID). For advanced reactor designs that contain multiple non-core sources of radioactivity, or for monolithic plant sites that include associated fuel facilities, this approach could potentially simplify licensing applications through the use of a single, uniform, and consistent decision-making framework across all radioactive sources at the site. In addition, a RIPB approach could provide additional insights regarding plant behavior, flexibility regarding licensing decision-making, and potentially allow the use of risk information as part of the plant oversight process. Risk-informing these aspects of advanced reactor licensing would also be consistent with the NRC’s risk policy statement. However, there is diverse regulation and guidance regarding the licensing of non-core sources of radioactivity and generally limited experience using RIPB approaches for the evaluation as part of licensing.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Study of two different radioactive sources for prostate brachytherapy treatment

In this study we evaluated two radioactive sources for brachytherapy treatments. Our main goal was to quantify the absorbed doses on organs and tissues of an adult male patient, submitted to a brachytherapy treatment with two radioactive sources. We evaluated a {sup 192}Ir and a {sup 125}I radioactive sources. The {sup 192}Ir radioactive source is a cylinder with 0.09 cm in diameter and 0.415 cm long. The {sup 125}I radioactive source is also a cylinder, with 0.08 cm in diameter and 0.45 cm long. To evaluate the absorbed dose distribution on the prostate, and other organs and tissues of an adult man, a male virtual anthropomorphic phantom MASH, coupled in the radiation transport code MCNPX 2.7.0, was employed.We simulated 75, 90 and 102 radioactive sources of {sup 125}I and one of {sup 192}Ir, inside the prostate, as normally used in these treatments, and each treatment was simulated separately. As this phantom was developed in a supine position, the displacement of the internal organs of the chest, compression of the lungs and reduction of the sagittal diameter were all taken into account. For the {sup 192}Ir, the higher doses values were obtained for the prostate and surrounding organs, as the colon, gonads and bladder. Considering the {sup 125}I sources, with photons with lower energies, the doses to organs that are far from the prostate were lower. All values for the dose rates are in agreement with those recommended for brachytherapy treatments. Besides that, the new seeds evaluated in this work present usefulness as a new tool in prostate brachytherapy treatments, and the methodology employed in this work may be applied for other radiation sources, or treatments. (authors)

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2017

1st Quarter Transportation Report FY2017: Waste Shipments To and From the Nevada National Security Site (NNSS), Radioactive Waste Management Complex

This report satisfies the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Field Office (NNSA/NFO) commitment to prepare a quarterly summary report of waste shipments to the Nevada National Security Site (NNSS) Radioactive Waste Management Complex (RWMC) at Area 5. This report summarizes the 1st quarter of fiscal year (FY) 2017 low-level radioactive waste (LLW), mixed low-level radioactive waste (MLLW) and classified non-radioactive (CNR) shipments. There were no shipments sent for offsite treatment from a NNSS facility and returned to the NNSS this quarter of FY2017.

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

2nd Quarter Transportation Report FY2015: Radioactive Waste Shipments to and from the Nevada National Security Site

This report satisfies the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Field Office (NNSA/NFO) commitment to prepare a quarterly summary report of radioactive waste shipments to and from the Nevada National Security Site (NNSS) Radioactive Waste Management Complex (RWMC) at Area 5. There were no shipments sent for offsite treatment and returned to the NNSS this quarter. This report summarizes the 2nd quarter of Fiscal Year (FY) 2015 low-level radioactive waste (LLW) and mixed low-level radioactive waste (MLLW) shipments.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗