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DOE OSTI2012

Radioactive Waste Characterization Strategies; Comparisons Between AK/PK, Dose to Curie Modeling, Gamma Spectroscopy, and Laboratory Analysis Methods- 12194

In the coming fiscal years of potentially declining budgets, Department of Energy facilities such as the Los Alamos National Laboratory (LANL) will be looking to reduce the cost of radioactive waste characterization, management, and disposal processes. At the core of this cost reduction process will be choosing the most cost effective, efficient, and accurate methods of radioactive waste characterization. Central to every radioactive waste management program is an effective and accurate waste characterization program. Choosing between methods can determine what is classified as low level radioactive waste (LLRW), transuranic waste (TRU), waste that can be disposed of under an Authorized Release Limit (ARL), industrial waste, and waste that can be disposed of in municipal landfills. The cost benefits of an accurate radioactive waste characterization program cannot be overstated. In addition, inaccurate radioactive waste characterization of radioactive waste can result in the incorrect classification of radioactive waste leading to higher disposal costs, Department of Transportation (DOT) violations, Notice of Violations (NOVs) from Federal and State regulatory agencies, waste rejection from disposal facilities, loss of operational capabilities, and loss of disposal options. Any one of these events could result in the program that mischaracterized the waste losing its ability to perform it primary operational mission. Generators that produce radioactive waste have four characterization strategies at their disposal: - Acceptable Knowledge/Process Knowledge (AK/PK); - Indirect characterization using a software application or other dose to curie methodologies; - Non-Destructive Analysis (NDA) tools such as gamma spectroscopy; - Direct sampling (e.g. grab samples or Surface Contaminated Object smears) and laboratory analytical; Each method has specific advantages and disadvantages. This paper will evaluate each method detailing those advantages and disadvantages including; - Cost benefit analysis (basic materials costs, overall program operations costs, man-hours per sample analyzed, etc.); - Radiation Exposure As Low As Reasonably Achievable (ALARA) program considerations; - Industrial Health and Safety risks; - Overall Analytical Confidence Level. The concepts in this paper apply to any organization with significant radioactive waste characterization and management activities working to within budget constraints and seeking to optimize their waste characterization strategies while reducing analytical costs. (authors)

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

International Implementation of IAEA's Borehole Disposal Concept for Sealed Radioactive Sources - 18545

With funding from the Government of Canada, and support from the U.S. Nuclear Regulatory Commission's (NRC's) Office of International Program and many others, the International Atomic Energy Agency (IAEA) is supporting the field deployment of the Borehole Disposal Concept (BDC) for secure disposal of radioactive sources in Ghana, the Philippines and Malaysia. This is a first-of-its-kind implementation. Though small in volume, disused sealed radioactive sources can be intensively radioactive and create a significant safety and security liability for the majority of the 168 Member States in the IAEA. Millions of sealed radioactive sources have been manufactured since 1901, and it is estimated that hundreds of thousands are now unwanted. The irony is that some of the countries with the smallest inventories have the greatest difficulties managing their radioactive wastes. The BDC was originally conceptualized in an IAEA-study undertaken by the South African Nuclear Energy Corporation (NECSA) for countries with small inventories of disused sealed radioactive sources. There are multiple, passive engineered and natural barriers in the BDC design, including an inner stainless-steel capsule (which holds the wastes), an inner cement containment barrier, an outer stainless steel container and outer backfill cement, as well as the surrounding geosphere. In common subsurface geochemical environments, these barriers will isolate the wastes from the biosphere, provide effective containment and ensure long-term safety. To greatly limit the likelihood of inadvertent human intrusion (safety), and deliberate human intrusion (security), these waste packages will be disposed of in a narrow diameter borehole (26 cm) at depths greater than 30 m. This type of borehole disposal creates an exceptionally small footprint, removes the wastes from normal human surficial activities, and isolates the waste packages from near-surface processes. Over the past 20 years, the IAEA, NECSA and many others have advanced the BDC on the shoulders of the IAEA's broader work supporting safe and secure disposal of radioactive wastes and management of sealed radioactive sources. With recent funding and support from many others, the IAEA has assisted Ghana, Malaysia and the Philippines in assessing their inventories and conditioning their disused sealed radioactive sources, in drilling investigation boreholes and characterizing their proposed sites, and in designing their disposal systems. These countries plan to place the uppermost waste package at depths greater than 100 m. While activities in the Philippines remain focused on-site characterization, Ghana is in the latter stages of developing their Safety Case for operational and 'post-closure' safety and Malaysia has completed their Safety Case and submitted a license application to construct and operate a BDC facility. Though this paper identifies the support provided to Member States by the IAEA, it is the Ghana Atomic Energy Commission and the Malaysian Nuclear Agency that are responsible for the safety and security of the disposals and that have been responsible for carrying-out their disposal programs. Once the BDC is licensed by the relevant national regulatory bodies, these first-of-a-kind disposal approaches will provide a template for other countries to safely dispose of their disused sealed radioactive sources - permanently eliminating their safety and security liabilities. (authors)

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

Transportation of Low-Level Radioactive Waste from KAERI to the Final Repository - 18179

The Korea Atomic Energy Research Institute (KAERI) collects the radioactive wastes generated during the operation of nuclear facilities and R and D processes. Radioactive wastes are treated and repackaged according to the national radioactive waste management policy, and then transferred to the final disposal site at the Korea Radioactive Waste Agency (KORAD). After 2015, the wastes delivered to the final disposal site have consisted of spent filters and combustible wastes such as clothes and gloves generated during R and D or facility operation processes at KAERI. The wastes generated by each facility are collected at the radioactive waste treatment facility, then treated and repackaged in 200 L drums according to their generation characteristics, i.e. each drum contains waste of the same type from the same facility, and with the same date of generation. Then, a radionuclide inventory is obtained for the repackaged drum by using a direct measurement method such as the drum assay system, radiochemical analysis through random sampling or an indirect measurement method such as applying a scaling factor. When performing sampling of the radioactive waste drums, the waste samples with a high contamination are selected in a conservative approach as representative. The radioactive waste drums are inspected and checked for acceptability by the Quality Management Division of KAERI, and a preliminary inspection is carried out on-site by KORAD. At this time, the acceptability of the radioactive waste drums is assessed through visual inspection, the appropriateness of their nuclide analysis value is confirmed, and a survey or smear test of sample drums is performed. Drums that meet the acceptance criteria are transferred to KORAD after a radiation environmental impact assessment of the transportation. Since 2015, 1,084 low-level operational radioactive wastes have been transferred from KAERI to the final disposal site, and further preparation work is now being made for the transportation of additional drums. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

Radioactive Demonstration of Final Mineralized Waste Forms for Hanford Waste Treatment Plant Secondary Waste (WTP-SW) by Fluidized Bed Steam Reforming (FBSR) using the Bench Scale Reformer Platform

The U.S. Department of Energy's Office of River Protection (ORP) is responsible for the retrieval, treatment, immobilization, and disposal of Hanford's tank waste. Currently there are approximately 56 million gallons of highly radioactive mixed wastes awaiting treatment. A key aspect of the River Protection Project (RPP) cleanup mission is to construct and operate the Waste Treatment and Immobilization Plant (WTP). The WTP will separate the tank waste into high-level and low-activity waste (LAW) fractions, both of which will subsequently be vitrified. The projected throughput capacity of the WTP LAW Vitrification Facility is insufficient to complete the RPP mission in the time frame required by the Hanford Federal Facility Agreement and Consent Order, also known as the Tri-Party Agreement (TPA), i.e. December 31, 2047. Therefore, Supplemental Treatment is required both to meet the TPA treatment requirements as well as to more cost effectively complete the tank waste treatment mission. In addition, the WTP LAW vitrification facility off-gas condensate known as WTP Secondary Waste (WTP-SW) will be generated and enriched in volatile components such as 137 Cs, 129 I, 99 Tc, Cl, F, and SO 4 that volatilize at the vitrification temperature of 1150°C in the absence of a continuous cold cap (that could minimize volatilization). The current waste disposal path for the WTP-SW is to process it through the Effluent Treatment Facility (ETF). Fluidized Bed Steam Reforming (FBSR) is being considered for immobilization of the ETF concentrate that would be generated by processing the WTP-SW. The focus of this current report is the WTP-SW. FBSR offers a moderate temperature (700-750°C) continuous method by which WTP-SW wastes can be processed irrespective of whether they contain organics, nitrates, sulfates/sulfides, chlorides, fluorides, volatile radionuclides or other aqueous components. The FBSR technology can process these wastes into a crystalline ceramic (mineral) waste form. The mineral waste form that is produced by co-processing waste with kaolin clay in an FBSR process has been shown to be as durable as LAW glass. Monolithing of the granular FBSR product is being investigated to prevent dispersion during transport or burial/storage, but is not necessary for performance. A Benchscale Steam Reformer (BSR) was designed and constructed at the SRNL to treat actual radioactive wastes to confirm the findings of the non-radioactive FBSR pilot scale tests and to qualify the waste form for applications at Hanford. BSR testing with WTP SW waste surrogates and associated analytical analyses and tests of granular products (GP) and monoliths began in the Fall of 2009, and then was continued from the Fall of 2010 through the Spring of 2011. Radioactive testing commenced in 2010 with a demonstration of Hanford's WTP-SW where Savannah River Site (SRS) High Level Waste (HLW) secondary waste from the Defense Waste Processing Facility (DWPF) was shimmed with a mixture of 125/129 I and 99 Tc to chemically resemble WTP-SW. Prior to these radioactive feed tests, non-radioactive simulants were also processed. Ninety six grams of radioactive granular product were made for testing and comparison to the non-radioactive pilot scale tests. The same mineral phases were found in the radioactive and non-radioactive testing.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Current Situation for Management of Disused Sealed Radioactive Sources in Japan - 13025

As for the Sealed Radioactive Source currently used in Japan, many of them are imported from overseas. The U.S., Canada, Germany, the Netherlands, Belgium and Czech Republic are the main exporting States. Many of disused sealed radioactive sources are being returned to exporting States. The sealed radioactive sources which cannot be returned to exporting States are appropriately kept in the domestic storage facility. So, there are not main problem on the long term management of disused sealed radioactive sources in Japan. However, there are some difficulties on repatriate. One is reservation of a means of transport. The sea mail which conveys radioactive sources owing to reduction of movement of international cargo is decreasing in number. And there is a denial of shipment. Other one is that the manufacturer has already resigned from the work and cannot return disused sealed radioactive sources, or a manufacturer cannot specify and disused sources cannot be returned. The disused sealed radioactive source which cannot be repatriated is a little in term of radioactivity. As for the establishment of national measure of final disposal facility for disused sealed radioactive sources, in Japan, it is not yet installed with difficulty. Since there are many countries for which installation of a final disposal facility for disused sealed radioactive sources is difficult, the source manufacture country should respond positively to return the source which was manufactured and sold in the past. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2016

Development of In-Situation Radioactivity Inspection System for Food Materials

The impacts on the ocean of releases of radionuclides from the Fukushima nuclear power plants remain unclear. Korean people worry about radioactive contamination of Japanese and Korean sea products. Radioactive contamination of processed foodstuffs, livestock, sea products, farm products imported from Japan and fishes caught in coastal oceans of Korea has become an important social issue. Recently, there are also needs of inspection system for monitoring of public meals such like school feedings of kindergarten, elementary/middle/high school and university. Radioactivity inspections of food materials have been executed manually with portable measuring instruments or at labs using their samples. But, it's very important the radioactivity inspection should be performed at the field of feeding facilities. To protect the health of people from radioactivity contained in Japanese marine products imported to Korea, a system to inspect radioactivity in real time has been developed. The system is to measure the radioactivity level of farm and marine products and public meals continuously and automatically at inspection sites of an agency checking radiation of imported foodstuffs or school. The performance test of NaI(Tl) detector was examined using the quantified Cs-137 source 10 Bq/kg, 30 Bq/kg, 50 Bq/kg, 300 Bq/kg and 900 Bq/kg at Korea Research Institute of Standards and Science(KRISS). Source was put in a box and radioactivity was measured in the middle of the box. First of all, Background was measured and calibration was performed using the Cs-137 300 Bq and Ba-133 3700 Bq for 60 minutes. After setting the background and calibration data, tests of the Cs-137 10, 30, 50, 300, 900 Bq/kg were executed for 10 minutes. A peak-shaped graph is shown at the location of 662 KeV (80∼100 channel). This is due to the gamma radiation energy with 662 KeV emitted by the Cs-137. Because through KRISS examination, the MDA (minimum detectable activity) is about 23 Bq/kg, it is reasonable Cs-137 10 Bq/kg cannot be measured. And in cases of Cs-137 300 Bq/kg and 900 Bq/kg, there are some peak movement on the graph. It can be adjusted by controlling gain or high voltage. The examination was identified through the performance test (Cs-137 10, 30, 50, 300, 900 Bq/kg) at KRISS. In order to increase the measurement efficiency of Cs-137 and the system performance, database will be built up based on radioactivity measurement results from 10 to 300 Bq/kg with the unit 10. Thus, Cs-137 concentration of food materials can be identified below the maximum permissible concentration of 100 Bq/kg on a basis of the spectrum characteristic of Cs-137. Also, anxiety about radioactive contamination of foods after the Fukushima nuclear accident will be eased and trust of the radioactive inspection is expected definitely. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

Cement Solidification of Residual Metal Oxide from Radioactive Waste

A lot of uranium contaminated soil and concrete waste is generated from the dismantlement of uranium conversion facilities. There are several disposal methods for radioactive materials such as regulation exemption, decontamination and long-term storage. It is necessary for us to achieve the permanent disposal of radioactive waste. The methods for solidifying radioactive waste include cement solidification, asphalt solidification, and polymer solidification. Cement solidification has several advantages such as well-known materials and technologies, various applications, and reasonable cost. On the contrary, it has certain disadvantages such as a large volume expansion of solidification for radioactive waste. Because cement solidification has excellent structural strength and a shielding effect, it is widely used for the permanent disposal of radioactive waste. Cement solidification operates as a barrier against the leaching of nuclides during radioactive waste disposal. To evaluate the criteria of KORAD (Korea Radioactive Waste Agency), integrity tests on the solidification of radioactive waste need to be performed. In this study, to inspect the criteria of KORAD, we carried out the cement solidification of residual metal oxide from radioactive waste. An integrity inspection such as the pressure strength, inundation, and leaching test on cement solidification was performed. We also analyzed the experimental results according to the criteria of KORAD. The calculation of the leachability index on the primary and secondary cement solidification for leaching detection of nuclides is carried out. The integrity of primary and secondary cement solidification is evaluated. (authors)

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

Development of the Mobile Systems for Conditioning of Disused Sealed Radioactive Sources in Serbia - 20105

Sealed radioactive sources (SRS) are being used worldwide in the field of medicine, agriculture, industry and research. They can be found in mobile as well as stationary devices. SRS contains radioactive material that is (a) permanently sealed in a capsule or (b) closely bounded and in a solid form. The capsule or material of an SRS should be strong enough to maintain leak tightness under the conditions of use and purpose for which the source was designed, also in case of accidents. In this case only emitted radiation is utilized. Firstly, the hazard from external radiation has to be considered, but the possibility of contamination due to fracture of the capsule should not be disregarded. The radioactive sources are composed of the radiating isotope contained in the filling medium, the single or double isotope holder that partially or totally surrounds the filling medium, the outer cover that contains the parts mentioned above and the capsule closed airtightly by welding or using some other method. The capsule must be tested for leakage periodically. If the SRS is no longer needed (e.g. replaced by a different technique) or it becomes useless for the intended application (e.g. the activity becomes too weak, the equipment containing the source works poorly or becomes obsolete, the source is damaged or leaking) it is considered disused. Disused sealed radioactive sources (DSRS) are typically conditioned and disposed if a facility is available. If the disposal option is not available, conditioned DSRS should be stored under proper conditions. In some cases, the radionuclide(s) in DSRS can be recovered/recycled or the DSRS can be repurposed for other applications. Conditioning of DSRS ensures containment of the radioactive material, provides confinement for leaking sources, provides sufficient radiation shielding, reduces storage/disposal volume by allowing consolidation of multiple sources into a single storage/disposal container, facilitates transport operations and contributes to safety and security as well. Typically, conditioning technologies are deployed either as permanently installed stationary systems in centralized or mobile on-site waste processing facilities, or in a mobile configuration. Centralized stationary facilities provide a single processing location for multiple users that requires transport of the waste to the facility. On the other hand, mobile systems may be provided for the selection and application of the optimum technology for a specific waste stream (such as DSRS) by bringing the process to the point where the waste is generated. In addition, mobile systems could offer additional flexibility by sharing equipment among multiple waste generating sites for processing campaigns that vary in duration, from very short periods to several years. The term 'mobile processing system' refers to any radioactive waste processing system or component which is designed to be transportable and which is not considered permanently installed. Two mobile system for conditioning of disused sealed radioactive sources are developed in the Public Company Nuclear Facilities of Serbia. Development of these mobile systems was supported by SRB9005 national project via Technical Cooperation of the IAEA. The first mobile system, built inside the 20 feet ISO container, will be used for conditioning of DSRS category 3 to 5. The second mobile system, built inside the 7 m long vehicle (Iveco Daily Van), will be used for dismantling of ionizing smoke detectors mostly with Am-241 sources. Designs of the mobile systems were defined in cooperation with two companies from Belgium (Belgoprocess and Leniko) and a Croatian company Ekoteh as well as with the support of the IAEA experts. The generic safety assessment and operational procedures for the mobile systems are developed. Based on safety assessment the acceptance criteria and operational limits and conditions are established. Operational procedures include: (a) equipment and material requirements, (b) assembling procedure of the mobile unit, (c) procedure for acceptance of devices for dismantling and conditioning, (d) dismantling procedure for devices to recover the DSRS, (e) characterization of DSRS, (f) encapsulation procedure of DSRS, (g) disassembling procedure of the mobile unit, and (h) keeping records, identification and traceability. In addition, radiation safety, health safety, security and emergency preparedness plans are prepared. The generic safety assessment and operational procedures could be updated with site specific requirements, DSRS inventory, and different needs for future customers. Developed mobile systems could be used in all situations when it is feasible to perform conditioning of DSRS on the spot in the county and worldwide. Development of these mobile units was just the first step to create the Reference Center for Radioactive Waste Treatment and Disused Radioactive Sources Conditioning for Small Facilities which can become a regional training center in the future, and/or as a tool for comprehensive national search and secure programmes. In the next phases development of e-learning platforms and blended learning packages as well as application for the IAEA Qualified Technical Centre (QTC) for the management of DSRS is foreseen. (authors)

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

Conceptual Design of Decontamination System for Radioactive Contaminated Water with Microalgae

In 2011, Fukushima nuclear power plant accident in Japan caused contamination of lakes and rivers due to radioactive materials released. As time passes, the radioactivity of radioactive contaminated water has been lowered, but there is still some level of radioactive contamination. At present, decontamination of radioactive contaminated water uses a method of removing radioactive materials in polluted water by using resin or filter. This conventional decontamination method can lower the radioactive concentration of contaminated water to a certain level. However, it is difficult to reduce the amount of contaminated water to a level lower than the very low level due to technical and economic reasons. In this study, we propose a method to decontaminate water contaminated with radioactive materials using microalgae. In addition, a system for the treatment of radioactive contaminated water using microalgae has been conceptually designed.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

Research and Education Campus Facilities Radioactive Waste Management Basis and DOE Manual 435.1-1 Compliance Tables

U.S. Department of Energy Order 435.1, 'Radioactive Waste Management,' along with its associated manual and guidance, requires development and maintenance of a radioactive waste management basis for each radioactive waste management facility, operation, and activity. This document presents a radioactive waste management basis for Idaho National Laboratory Research and Education Campus facilities that manage radioactive waste. The radioactive waste management basis for a facility comprises existing laboratory-wide and facility-specific documents. Department of Energy Manual 435.1-1, 'Radioactive Waste Management Manual,' facility compliance tables also are presented for the facilities. The tables serve as a tool to develop the radioactive waste management basis.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

Materials and Security Consolidation Complex Facilities Radioactive Waste Management Basis and DOE Manual 435.1-1 Compliance Tables

Department of Energy Order 435.1, 'Radioactive Waste Management,' along with its associated manual and guidance, requires development and maintenance of a radioactive waste management basis for each radioactive waste management facility, operation, and activity. This document presents a radioactive waste management basis for Idaho National Laboratory's Materials and Security Consolidation Center facilities that manage radioactive waste. The radioactive waste management basis for a facility comprises existing laboratory-wide and facility-specific documents. Department of Energy Manual 435.1-1, 'Radioactive Waste Management Manual,' facility compliance tables also are presented for the facilities. The tables serve as a tool for developing the radioactive waste management basis.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

Advanced Test Reactor Complex Facilities Radioactive Waste Management Basis and DOE Manual 435.1-1 Compliance Tables

U.S. Department of Energy Order 435.1, 'Radioactive Waste Management,' along with its associated manual and guidance, requires development and maintenance of a radioactive waste management basis for each radioactive waste management facility, operation, and activity. This document presents a radioactive waste management basis for Idaho National Laboratory's Advanced Test Reactor Complex facilities that manage radioactive waste. The radioactive waste management basis for a facility comprises existing laboratory-wide and facility-specific documents. U.S. Department of Energy Manual 435.1-1, 'Radioactive Waste Management Manual,' facility compliance tables also are presented for the facilities. The tables serve as a tool to develop the radioactive waste management basis.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

Materials and Fuels Complex Facilities Radioactive Waste Management Basis and DOE Manual 435.1-1 Compliance Tables

Department of Energy Order 435.1, 'Radioactive Waste Management,' along with its associated manual and guidance, requires development and maintenance of a radioactive waste management basis for each radioactive waste management facility, operation, and activity. This document presents a radioactive waste management basis for Idaho National Laboratory's Materials and Fuels Complex facilities that manage radioactive waste. The radioactive waste management basis for a facility comprises existing laboratory-wide and facility-specific documents. Department of Energy Manual 435.1-1, 'Radioactive Waste Management Manual,' facility compliance tables also are presented for the facilities. The tables serve as a tool for developing the radioactive waste management basis.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

Central Facilities Area Facilities Radioactive Waste Management Basis and DOE Manual 435.1-1 Compliance Tables

Department of Energy Order 435.1, 'Radioactive Waste Management,' along with its associated manual and guidance, requires development and maintenance of a radioactive waste management basis for each radioactive waste management facility, operation, and activity. This document presents a radioactive waste management basis for Idaho National Laboratory's Central Facilities Area facilities that manage radioactive waste. The radioactive waste management basis for a facility comprises existing laboratory-wide and facilityspecific documents. Department of Energy Manual 435.1-1, 'Radioactive Waste Management Manual,' facility compliance tables also are presented for the facilities. The tables serve as a tool for developing the radioactive waste management basis.

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

Design and Development of an Automated Instrument for Measuring Radioactivity in an Excavator Bucket - 16246

Increasing interest and emphasis is being given to the presence of radioactive contamination in the environment and the methods by which this contamination can be remediated. Radioactive contamination may arise from industrial process including the operation of nuclear facilities. It may also arise from the enhancement of naturally occurring radioactive material by industrial processes. An area of particular concern is the remediation of soil and rubble with radioactive contamination. Traditional methods of soil remediation involve sampling followed by the excavation of areas where contamination is found. The soil removal process is normally very inefficient and can involve multiple handling of excessive quantities of soil and ultimately an over estimate of the quantity of soil with radioactive contamination. In order to both minimize material handling and to reduce the final volume of contaminated material, which must be handled for disposal, a sensitive instrument has been developed and tested which is designed to measure the radioactive content of an excavator bucket. The instrument detector is positioned automatically just above the open top of the bucket and the measurement employs a high efficiency sodium iodide crystal which has a graded shadow shield to reduce the effect of background radioactivity. The measurement efficiency is enhanced as the detector is located in close proximity to the soil in the bucket and the thick steel shell of the bucket acts as a shield to further reduce the effect of background from the ground below. The method has obvious advantages over methods, which attempt to measure the potentially radioactive soil through the shielding provided by the thick steel bucket shell. As the first stage of the bucket measurement the excavator operator fills the bucket and locates it in a designated position with the open top of the bucket facing upwards, close to the operator cab. A measurement technician, located outside the danger area of the excavator boom, begins the measurement using a tablet computer with a wireless connection to the instrument. The shielded detector is moved from a storage position above the operator cab to the measurement position close to the open surface of the excavator bucket. Once the measurement is completed in typically one to two minutes, the detector is repositioned in its storage location and the measurement data is transmitted to the tablet computer. At this point the measurement technician makes the decision based on the measurement results as to how to disposition the bucket contents. Typically contaminated material will be transferred to a vehicle for off-site disposal. Alternatively, and this is the great advantage of this technique, uncontaminated material may be returned directly to the site of the excavation without the need for further handling. The development, detailed design and preliminary testing of the instrument are described. (authors)

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

The UK's National Policy and Programme for the Management of Spent Fuel and Radioactive Waste - 16574

This paper provides a broad overview of the United Kingdom's (UK's) national policies and national programme for the management of spent fuel and radioactive waste. The aim of the paper is to ensure a clear understanding of the approach taken in the UK so that opportunities for learning lessons and joint working, where they exist, can be identified and to ensure a general understanding of the interdependencies between the various approaches taken in the UK. With a legacy that dates back to the late 1940's/early 1950's the UK have a range of complex and difficult issues to resolve that have evolved over time. What was considered to be acceptable in the past is not thought of in the same way today. We therefore need to be forward thinking, innovative and above all ready to adapt our practices and policies to address previously unknown issues and complications. Due to the wide range of types of radioactive wastes and materials to be managed in the UK we have needed to establish a set of high level principles to cover all radioactive waste types while at the same time ensure that specific approaches for specific issues are developed within that framework. In the UK we have also had to consider the issue of integration to ensure that an approach that addresses one issue does not inadvertently cause a problem elsewhere - for example: in disposing of the high volumes of very low level waste we needed an approach to ensure that other low level waste has a disposal route. Financing of decommissioning and future waste management is clearly one of the biggest challenges to be addressed. As mentioned above ensuring that the legacy left for future generation is not one of a burden - we need to ensure that measures are in place to fund and manage long lived radioactive wastes. Finally, it is clear that not all radioactive wastes are the same and therefore different approaches are needed for managing the different types of material. This requires a proportionate and prioritised approach so while the same ultimate goal is applied (i.e.: the safe and responsible management of radioactive waste) some radioactive wastes are more equal than others and need to be treated accordingly. In the UK we believe that we have a broad programme in place that allows the right balance to be struck in managing all our radioactive waste and spent fuel in a proportionate, safe and responsible manner.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Induced Radioactivity Analysis for the NSRL Linac in China Using Monte Carlo Simulations and Gamma Spectroscopy - Paper 141

The Synchrotron Radiation Facility of the National Synchrotron Radiation Laboratory (NSRL) located in Hefei, China consists of a 200-MeV electron linear accelerator and an 800-MeV electronic storage ring. Constructed in 1989, the 200-MeV Electron linac is one of the earliest high-energy electron linear accelerators in China. The NSRL is a large-scale experiment device, having been operated more than twenty years, so it is hard to give a historical recount about the injection of NSRL Linac. Lower-energy electrons emitted from the accelerator electron gun are progressively accelerated to about 200 MeV in five acceleration tubes, and then are injected into the electron storage ring, reaching the final energy of 800 MeV. The linear accelerator is a traveling wave linear accelerator, consisting of pre-injection, four 6-m acceleration segment, some beam sensing elements, and focusing inserts. The electrons are accelerated in the acceleration tubes to 32 MeV, 74 MeV, 116 MeV, 158 MeV, respectively, and the electronic energy reaches 200 MeV at the end of the acceleration trip. A scraper is used for beam collimation and is located at the end of each acceleration tube, with a length that increases with the energy to provide sufficient collimation. The NSRL Linac was retired in 2012 due to a plan to upgrade the facility. The decommissioning produced many real challenges related to the methods to handle the large size of the device and the time required to store structural materials until the radioactivity is reduced in the waste repository. These problems should be solved as soon as possible in order for the upgrade to proceed smoothly. At present, no prior decommissioning experience existed in China on such a high-energy electron linear accelerator. Therefore, the decommission project provided an opportunity to study and evaluate the challenges, in particular, involving the induced radioactivity for the purposes of radiation protection. For a well-shielded accelerator facility, the induced radioactivity is an important source of radiation for people who perform the decommissioning. In addition, information on the level of induced radioactivity is needed for disposing of the retired accelerator materials. Therefore it is significant to predict such radioactivity level accurately. However, the induced radioactivity study for such type high-energy electron linear accelerator is very seldom. Particle accelerators that operate at energy greater than 10 MeV will produce induced radioactivity due to well-known photonuclear interactions. However, the assessment of such as induced radioactivity of large-scale installations is a time-consuming process. Experimental methods are time-consuming and require special radiation safety measures to protection the investigators. On the other hand, empirical formula cannot yield sufficiently accurate results. Therefore, Monte Carlo simulation method is the primary means to study such problems. In the decommissioning process of the NSRL Linac, the induced radioactivity assessment will be based on Monte Carlo simulation method as the main auxiliary means to expand a series of studies. (authors)

43 PARTICLE ACCELERATORS↗
DOE OSTI2012

The EU Approach for Responsible and Safe Management of Spent Fuel and Radioactive Waste - 12118

In July 2011 legislation on responsible and safe management of spent fuel and radioactive waste was adopted in the European Union (EU). It aims at ensuring a high level of safety, avoiding undue burdens on future generations and enhancing transparency. EU Member States are responsible for the management of their spent fuel and/or radioactive waste. Each Member State remains free to define its fuel cycle policy. The spent fuel can be regarded either as a valuable resource that may be reprocessed or as radioactive waste that is destined for direct disposal. Whatever option is chosen, the disposal of high level waste, separated at reprocessing, or of spent fuel regarded as waste should be considered. The storage of radioactive waste, including long-term storage, is an interim solution, but not an alternative to disposal. To this end, each Member State has to establish, maintain and implement national policy, framework and programme for management of spent fuel and/or radioactive waste in the long term. Member States will invite international peer reviews to ensure that high safety standards are achieved. The EU approach is anchored in internationally endorsed principles and requirements of the IAEA safety standards and the Joint Convention and in this context makes them legally binding and enforceable in the EU. The EU approach of regulating the management of spent fuel and radioactive waste is anchored in the competence of the national regulatory authorities and in the internationally endorsed principles and requirements of the IAEA Safety Standards and the Joint Convention. Member States have to report to the Commission on the implementation of Directive 2011/70/Euratom for the first time by 23 August 2015, and every 3 years thereafter, taking advantage of the review and reporting under the Joint Convention. On the basis of the Member States' reports, the Commission will submit to the European Parliament and the Council a report on progress made and an inventory of radioactive waste and spent fuel present in the EU territory and the future prospects. Directive 2011/70/Euratom is a logical next step after the Council Directive 2009/71/Euratom on the nuclear safety of nuclear installations. The EU is the first major regional actor providing a binding legal framework on nuclear safety and on responsible and safe management of spent fuel and radioactive waste, and thus is a real model to progress spent fuel and waste management in a safe and responsible manner. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗