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All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

WOW Technology's Innovative Radioactive Liquid Waste Treatment - 16128

WOW presents its revolutionary technology and equipment for liquid radioactive waste treatment: outperforming ultimate water decontamination and purification process, enhanced sludge concentration, no secondary waste nor consumables, fully automated, remote controlled and self-decontaminating device. The WOW's technology is based upon a never before observed discovery of fluid dynamics science: the possibility of performing a molecular separation between solute and suspended elements and the solvent. The combination of such a molecular separation process with a standard vacuum evaporation improves the abatement performances by thousands of times, with respect to those of the state of the art vacuum evaporators. In addition to this, no secondary waste is produced during the process, as no filters, membranes, resins or additives are used. WOW equipment, automated and remote controlled, self decontaminates after use and can be designed and constructed either tailored to the application needs or with a modular approach for enhanced transportability and application flexibility. After the preliminary verification by CNR, the Italian National Research Center, WOW Technology's decontamination device was tested c/o LENA, the Laboratory of Applied Nuclear Energy of the University of Pavia, Italy with a simulated solution 6000 times more contaminated than the nuclear reactor's cooling water of Fukushima-Daiichi NPP. In addition to that, WOW's technology was also used in a real case at the Radiochemistry laboratory of the Pavia's University Chemistry department. Both the above mentioned contaminated fluids have been successfully decontaminated without production of additional or secondary waste. WOW's technology has already performed on industrial scale c/o the Nuclear Repository of S.S.M. in Saluggia, Italy: 45000 liters of acid radioactive solution have been successfully decontaminated to a Decontamination Factor (DF) of 335000 for Cs-137 by one single evaporation step and without using any additional tool to control steam. [Performances independently certified by: National Physics Laboratory (UK) and Pavia University Radiochemistry Laboratory, Chemistry Department (Italy)]. The high decontamination factors and the high volume reduction of the sludge make this technology suitable for the ultimate treatment of highly contaminated liquid waste such as the Fukushima-Daiichi nuclear reactor's cooling water. The WOW's technology addresses as well to interventions in both, operating and decommissioning nuclear power plants and any other nuclear and radioactive facility whose produced and/or stored radioactive liquid solutions, including acid liquid solutions, can be successfully treated, the waste volume reduced and no secondary waste produced. (authors)

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

Comparative Study of the Standards and Guides of a Management System Applicable to Radioactive Waste - Paper 18292

A Management System may be defined as a set of standardized and interrelated practices that aim to improve effectiveness and produce better results. This system should be structured in such a way as to allow essential information to flow properly within the organization, including technical, administrative, production and human resources information, among others, within an adopted compliance model. There are, currently, several standards and guidelines for structuring a compliance model related to a management system. A system that covers various aspects of management in a single system is usually called integrated. Only the International Atomic Energy Agency - IAEA has a guide with this approach for radioactive waste: GS-G-3.3, 'The Management System for the Processing, Handling and Storage of Radioactive Waste' and GS-G-3.4, 'The Management System for the Disposal of Radioactive Waste'. The objective of this work is to present a comparison between existing standards, even if not specific for the management of radioactive waste, but which are widely used by companies. In due course, the result of this work will be object of proposal for specific standard to be presented to regulatory agency for this type of activity. This study should be useful for countries where there are no specific guidelines by the regulatory agency, adopting the best standard or guide as a conformity standards, creating and establishing a hybrid structure best suited to the safety and security aspects of the local culture. (authors)

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

Management of Documents and Information on Radioactive Waste - 19455

The assurance of the safety of a repository for low- and intermediate-level radioactive waste, during the 300 years of institutional control after closure, depends on assuming that the responsible institution will be capable to control the documents with information regarding the facility, the disposed of waste and the established measures for counteracting events that could breach the isolation of the radioactive material. Management of documents or information does not represent a major challenge in itself in view of consolidated practices and international standards, such as ISO 15489. However, considering the time scale involved in the disposal of the radioactive wastes, efforts should be made to ensure the preservation and proper access to documents and information. In this matter, the technological aspect of document production can be a complicating factor. In fact, it is common to be unable to access electronic documents produced 20 or 30 years ago, due either to problems related to the media where the information is stored, or to a lack of compatible systems for reading the files. This is just one example of complication when thinking about production, organization, representation, preservation and retrieval of recorded information that is to last for centuries. In this way, this work seeks to present the situation of the management of documents and information on radioactive waste in the scope of the Institute of Energy and Nuclear Research (IPEN) and with national regulation in Brazil. (authors)

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

Risk perception and social representation on radioactive waste: a case study with students of IPEN

Proper management of radioactive waste in teaching and research laboratories involves not only techniques and procedures for the treatment, storage and disposal of these materials but also the responsibility of all those who handle them. Responsibility begins with with methods that generate less waste and with the correct segregation of contaminated materials prior to disposal, In this context, the present study aimed to know, identify, understand and analyze the perception of risk and social representations about radioactive waste and its management by students and researchers of the Nuclear and Energy Research Institute. According to Renn (1992), two of the approaches to risk perception are cultural and sociological, which served as the basis to the present study. Within this analysis, the concept of risk is understood as a phenomenon and, therefore, the phenomena experienced by the individuals are related to their life histories, social relations, religious beliefs and academic formation, elements that interfere in the perception of risk. The method used to collect data was free evocation, with participants receiving the following instructions: 'Please write five words that pop up into your mind when you listen to the words radioactive waste and, sequentially, write a sentence using these words'. All participants received instructions of the voluntary character of the participation and of the confidentiality of the data. A total of 73 students and researchers were interviewed, including scientific initiation, masters degree, doctorate and postdoctoral researchers. 353 words were mentioned, of which 209 were different from each other. After analyzing the sentences, the words were inserted in 12 categories and later analyzed in the software INVOCADO, which provides a diagram with the possible structure of the social representations of the group studied. The analysis of the four-quadrants and their categories, as well as the mentioned words that compose them, are therefore consistent with the universe of the public analyzed. While it is not the totality of students dealing with radioactive materials, in general these individuals understand the importance of technology and that the adoption of safe practices ensures the successful development of research. Future developments: Analyze quantitatively and qualitatively the other questions of the questionnaire. Analyze the data and information obtained from the participant observation, which consisted of a follow-up of three students in their laboratory practices. Propose to the institute a manual of good practices that can be used by the students in their laboratory practices.

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

A radioactive tracer dilution method to determine the mass of molten salt

A new technique for molten salt mass determination, termed radioactive tracer dilution, that uses 22Na as a tracer was validated. It has been a challenging problem to determine the mass of molten salt in irregularly shaped containers, where a highly radioactive, high-temperature molten salt was used to process nuclear spent/used fuel during electrochemical recycling (pyro-processing) or for coolant/fuel salt from molten salt reactors. A radioactive source with known activity is dissolved into the salt. After a complete mixture, a small amount of the salt is sampled and measured in terms of its mass and radioactivity. By finding the ratio of the mass to radioactivity, the unknown salt mass in the original container can be precisely determined.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Ultra-pure nickel for structural components of low-radioactivity instruments

The next generation of nuclear and particle physics rare-event search experiments demand structural materials combining ultra-low levels of radioactive contamination with exceptional mechanical strength. Here, this study evaluates chemical vapor deposition (CVD) nickel as a low radioactive background candidate structural material for such applications. Manufacturer-supplied CVD Ni grown on aluminum substrates was assayed via inductively coupled plasma mass spectrometry (ICP-MS) employing isotope-dilution. These material assays produced measured bulk concentration of 232 Th, 238 U, and nat K at the levels of ~70 ppq, ≲100 ppq, and ~900 ppt, respectively, which is the lowest reported in nickel. Surface-etch profiling uncovered higher concentrations of these contaminants extending ~10 μ m beneath the surface, likely associated with the aluminum growth substrate. Additionally, the CVD Ni underwent tensile testing alongside standard Ni samples. CVD Ni exhibited a planar tensile strength of ~600 MPa, significantly surpassing standard nickel. However, heat treatment was found to reduce the tensile strength to levels comparable to standard Ni, with implications for high-temperature weld joining methods. The results reported are compared to the one other well documented usage of CVD Ni in a low radioactive background physics research experiment and a discussion is provided on how the currently reported results may arise from changes in CVD fabrication or testing process. These results establish CVD Ni as a promising low-radioactivity structural material, while outlining the need for further development in surface cleaning and weld-joining techniques to fully realize its potential in large-scale, low radioactive background rare-event search experiments.

47 OTHER INSTRUMENTATION↗
DOE OSTI2012

Demonstration and development of control mechanism for radioactive sources in Saudi Arabia

Saudi Arabia have no nuclear industry. Nevertheless, many radioactive sources, for different purposes, have been used in the country. There is upswing in the number of companies that recruit nuclear technology in their daily work. The National Center for Radiation Protection (NCRP) takes the full commitment and responsibility for monitoring and regulating the movement of radioactive sources in the country. NCRP issues the licenses for import, export, and use of radioactive sources. It, also, protects the country from any trespassing radiation through a sizable net of early warning and radiation monitoring stations along the borders of Saudi Arabia. This paper talks about the procedures of licensing, importing, exporting of radioactive sources. It, also, sheds light on types of implementing radioactive sources in different practices encompass medicine, industry, research. The NCRP has established an electronic web site to ease the communication with all users in the country. This site is yet in the experimental stage.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2015

Study on effect of geometrical configuration of radioactive source material to the radiation intensity of betavoltaic nuclear battery

Nuclear batteries have strategic applications and very high economic potential. One Important problem in application of nuclear betavoltaic battery is its low efficiency. Current efficiency of betavoltaic nuclear battery reaches only arround 2%. One aspect that can influence the efficiency of betavoltaic nuclear battery is the geometrical configuration of radioactive source. In this study we discuss the effect of geometrical configuration of radioactive source material to the radiation intensity in betavoltaic nuclear battery system. received by the detector. By obtaining the optimum configurations, the optimum usage of radioactive materials can be determined. Various geometrical configurations of radioactive source material are simulated. It is obtained that usage of radioactive source will be optimum for circular configuration.

30 DIRECT ENERGY CONVERSION↗
DOE OSTI2015

SU-E-T-534: Level of Residual Radioactivity of Activated Parts of a Decommissioned Cyclotron

Purpose: CTI cyclotron RDS-111 was used at the Hong Kong Sanatorium and Hospital (HKSH) to produce radiopharmaceuticals and radioactive tracers for diagnostic scans between 1999 and 2007. During the operation, some machine components became radioactive by activation. For the safety of staff, decommissioning took place in 2009, two years after the cyclotron had stopped operation. This study investigates the residual radioactivity and radionuclides found in different cyclotron components in 2014 in compliance with the local regulations in Hong Kong for transfer of radioactive waste. Methods: A representative sample of each part was counted using a high-purity germanium detector (manufacturer: ORTECT) for at least four hours. GammaVision, a multichannel analyzer software, was used to identify the radionuclides found in the cyclotron components, as well as the associated activities. A standard library and a Mariscotti peak search algorithm were used to identify the present radionuclides. Only radionuclides with half-life greater than 180 days were considered. Results: Among the components, the Havar target foil has the highest specific activity ((4.6±0.6)×10{sup 2} Bq/g), with Co-60 being the most prominent ((3.8±0.5)×10{sup 2} Bq/g). The total activity of the target foil, however, is still low due to its small mass of 0.04 g. Radioisotopes Mn-54 (46±6 Bq/g), Na-22 (6.8±0.8 Bq/g), Co-57 (7.3±0.9 Bq/g), and Fe-59 (6.0±0.9 Bq/g) have also been detected in the target foil. The target window holder and the vacuum window register a specific activity of 88.3±0.6 Bq/g and 48.6±0.1 Bq/g, respectively. Other components, such as the collimator, the target tube, the valve body and the beamline, are also found with trace amounts of radionuclides. Conclusion: Even seven years after the cyclotron had stopped operation, some components still exhibited residual radioactivity from activation exceeding the IAEA clearance levels. Special consideration for radiological protection may need to be taken in the disposal of the cyclotron parts.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2012

Radioactive Liquid Waste Treatment Facility Discharges in 2011

This report documents radioactive discharges from the TA50 Radioactive Liquid Waste Treatment Facilities (RLWTF) during calendar 2011. During 2011, three pathways were available for the discharge of treated water to the environment: discharge as water through NPDES Outfall 051 into Mortandad Canyon, evaporation via the TA50 cooling towers, and evaporation using the newly-installed natural-gas effluent evaporator at TA50. Only one of these pathways was used; all treated water (3,352,890 liters) was fed to the effluent evaporator. The quality of treated water was established by collecting a weekly grab sample of water being fed to the effluent evaporator. Forty weekly samples were collected; each was analyzed for gross alpha, gross beta, and tritium. Weekly samples were also composited at the end of each month. These flow-weighted composite samples were then analyzed for 37 radioisotopes: nine alpha-emitting isotopes, 27 beta emitters, and tritium. These monthly analyses were used to estimate the radioactive content of treated water fed to the effluent evaporator. Table 1 summarizes this information. The concentrations and quantities of radioactivity in Table 1 are for treated water fed to the evaporator. Amounts of radioactivity discharged to the environment through the evaporator stack were likely smaller since only entrained materials would exit via the evaporator stack.

03 NATURAL GAS↗
DOE OSTI2012

2011 Radioactive Materials Usage Survey for Unmonitored Point Sources

This report provides the results of the 2011 Radioactive Materials Usage Survey for Unmonitored Point Sources (RMUS), which was updated by the Environmental Protection (ENV) Division's Environmental Stewardship (ES) at Los Alamos National Laboratory (LANL). ES classifies LANL emission sources into one of four Tiers, based on the potential effective dose equivalent (PEDE) calculated for each point source. Detailed descriptions of these tiers are provided in Section 3. The usage survey is conducted annually; in odd-numbered years the survey addresses all monitored and unmonitored point sources and in even-numbered years it addresses all Tier III and various selected other sources. This graded approach was designed to ensure that the appropriate emphasis is placed on point sources that have higher potential emissions to the environment. For calendar year (CY) 2011, ES has divided the usage survey into two distinct reports, one covering the monitored point sources (to be completed later this year) and this report covering all unmonitored point sources. This usage survey includes the following release points: • all unmonitored sources identified in the 2010 usage survey, • any new release points identified through the new project review (NPR) process, and • other release points as designated by the Rad-NESHAP Team Leader. Data for all unmonitored point sources at LANL is stored in the survey files at ES. LANL uses this survey data to help demonstrate compliance with Clean Air Act radioactive air emissions regulations (40 CFR 61, Subpart H). The remainder of this introduction provides a brief description of the information contained in each section. Section 2 of this report describes the methods that were employed for gathering usage survey data and for calculating usage, emissions, and dose for these point sources. It also references the appropriate ES procedures for further information. Section 3 describes the RMUS and explains how the survey results are organized. The RMUS Interview Form with the attached RMUS Process Form(s) provides the radioactive materials survey data by technical area (TA) and building number. The survey data for each release point includes information such as: exhaust stack identification number, room number, radioactive material source type (i.e., potential source or future potential source of air emissions), radionuclide, usage (in curies) and usage basis, physical state (gas, liquid, particulate, solid, or custom), release fraction (from Appendix D to 40 CFR 61, Subpart H), and process descriptions. In addition, the interview form also calculates emissions (in curies), lists mrem/Ci factors, calculates PEDEs, and states the location of the critical receptor for that release point. [The critical receptor is the maximum exposed off-site member of the public, specific to each individual facility.] Each of these data fields is described in this section. The Tier classification of release points, which was first introduced with the 1999 usage survey, is also described in detail in this section. Section 4 includes a brief discussion of the dose estimate methodology, and includes a discussion of several release points of particular interest in the CY 2011 usage survey report. It also includes a table of the calculated PEDEs for each release point at its critical receptor. Section 5 describes ES's approach to Quality Assurance (QA) for the usage survey. Satisfactory completion of the survey requires that team members responsible for Rad-NESHAP (National Emissions Standard for Hazardous Air Pollutants) compliance accurately collect and process several types of information, including radioactive materials usage data, process information, and supporting information. They must also perform and document the QA reviews outlined in Section 5.2.6 (Process Verification and Peer Review) of ES-RN, 'Quality Assurance Project Plan for the Rad-NESHAP Compliance Project' to verify that all information is complete and correct.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2018

Air Dispersion Modeling of Radioactive Releases During Proposed 236Z Rubble Pile Removal Activities

This report is part of the planning process for the removal of the 236Z rubble remaining as of January 1, 2018, at the Plutonium Finishing Plant (PFP) on the Hanford Site. This report documents an analysis considering standard techniques for the removal of demolition rubble. It does not document any decisions about the removal approaches. This document is intended to guide the establishment of a boundary for the application of an air monitoring regime. Atmospheric dispersion modeling using estimated release rates has been conducted to provide information on the location and levels of radioactive contamination that may be expected as the result of rubble removal activities. Hourly meteorological data collected over a 6-year period (2004–2009) were used to examine the effects of wind speed, direction, and stability on projected concentrations of contaminants in air and deposited on nearby surfaces. The source terms modeled in this report were based on the residual transuranic contamination levels that are anticipated to remain in the 236Z rubble following building demolition. The potential airborne concentrations were established using the five-factor formula from DOE-HDBK-3010 (DOE 1994) considering material-at-risk, damage ratio, airborne release fraction, respirable fraction, and leak path factor. Radioactive contamination emissions were calculated by release mechanism and rubble pile area for on-shift and off-shift activities. The emissions from the applicable sources were combined to provide emissions estimates for each day from each cleanup area. The U.S. Environmental Protection Agency’s (EPA’s) AERMOD computer code was used to estimate atmospheric dispersion and deposition of the released radioactive materials in the immediate vicinity of the planned rubble removal activities Both airborne and surface concentrations were modeled with AERMOD. Hourly derived air concentrations (DAC) were modeled for an array of receptors covering the rubble pile site and surrounding area. Peak (95th and 99th percentile) values of time-integrated air concentrations at these receptor points are derived from these hourly values, with modeling results reported as total incremental air concentrations in DAC-hours occurring over the selected time period. Total accumulated deposition amounts were also evaluated with AERMOD using the same array of receptors, with results reported as disintegrations per minute (dpm) per 100 cm2. Each rubble pile was considered in terms of its size, location, and estimated contamination levels. The modeling was based on the assumed sequence of the rubble removal activities. The results in this report are based on cleanup schedules and methods proposed by CHPRC. As of January 1, 2018, there are broadly three large rubble piles (west, footprint, and east) remaining of the 236Z Building process cell. It is assumed that the overall cleanup will require either 1) 30 days over a 6-week period, or 2) 60 days over a 12-week period. Work was assumed to be performed during 10-hour day shifts. The analysis shows that some releases of radioactive material are to be anticipated during the cleanup of the 236Z rubble piles. The modeling results presented here are closely tied to the details of how the rubble removal is to be conducted. The results indicate that for the rubble removal, the radiological exposures from the planned activities are predicted to be well below the designated limits for air and soil exposures. The cleanup of the 236Z main process cell rubble is expected to release some alpha-emitting radionuclides. However, results based on 6 years of hourly climatological data, two representative cleanup periods, realistic mitigating actions, and a bounding source term indicate that concentrations at the fence line of the PFP facilities should remain well below 12 DAC-hours/week. The results presented in this report are a reevaluation of a subset of the results presented in PNNL-20173 Rev. 4.

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

Role of Congress in the High Level Radioactive Waste Odyssey: The Wisdom and Will of the Congress - 13096

Congress has had a dual role with regard to high level radioactive waste, being involved in both its creation and its disposal. A significant amount of time has passed between the creation of the nation's first high level radioactive waste and the present day. The pace of addressing its remediation has been highly irregular. Congress has had to consider the technical, regulatory, and political issues and all have had specific difficulties. It is a true odyssey framed by an imperative and accountability, by a sense of urgency, by an ability or inability to finish the job and by consequences. Congress had set a politically acceptable course by 1982. However, President Obama intervened in the process after he took office in January 2009. Through the efforts of his Administration, by the end of 2012, the US government has no program to dispose of high level radioactive waste and no reasonable prospect of a repository for high level radioactive waste. It is not obvious how the US government program will be reestablished or who will assume responsibility for leadership. The ultimate criteria for judging the consequences are 1) the outcome of the ongoing NRC's Nuclear Waste Confidence Rulemaking and 2) the concomitant permissibility of nuclear energy supplying electricity from operating reactors in the US. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Management of Disused Radioactive Sealed Sources in the Slovak Republic - 12100

After splitting-up the Czechoslovak Federation in 1993, the system of management of institutional radioactive waste, where disused sources represent its significant part, had had to build from beginning, since all corresponding activities had remained in the Czech part of the Federation. The paper presents the development of legislative and institutional framework of the disused radioactive sealed source management, development of the national inventory and development of management practices. According the Governmental decision (1994), the management of disused sealed sources and institutional radioactive waste at whole was based on maximal utilization of facilities inside nuclear facilities, particularly in the NPP A1 (shut down in the past, currently under decommissioning). This approach has been recently changing by Governmental decision (2009) to construct 'non-nuclear facility' - central storage for remained disused sealed sources collected from the places of use, where they were stored in some cases for tens of years. The approaches to siting and construction of this storage facility will be presented, as well as the current approaches to solution of the disused radioactive sources final disposal. Environmental impact assessment process in regard to the given facility/activity is slowly drawing to a close. The final statement of the Ministry of Environment can be expected in January or February 2012, probably recommending option 1 as preferred [6]. According to the Slovak legislation, the final statement has a status of recommendation for ongoing processes leading to the siting license. Very recently, in December 2012, Government of the Slovak republic decided to postpone putting the facility into operation by the end of June, 2014. (author)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2012

Romanian Experience for Enhancing Safety and Security in Transport of Radioactive Material - 12223

The transport of Dangerous Goods-Class no.7 Radioactive Material (RAM), is an important part of the Romanian Radioactive Material Management. The overall aim of this activity is for enhancing operational safety and security measures during the transport of the radioactive materials, in order to ensure the protection of the people and the environment. The paper will present an overall of the safety and security measures recommended and implemented during transportation of RAM in Romania. Some aspects on the potential threat environment will be also approached with special referring to the low level radioactive material (waste) and NORM transportation either by road or by rail. A special attention is given to the assessment and evaluation of the possible radiological consequences due to RAM transportation. The paper is a part of the IAEA's Vienna Scientific Research Contract on the State Management of Nuclear Security Regime (Framework) concluded with the Institute for Nuclear Research, Romania, where the author is the CSI (Chief Scientific Investigator). The transport of RAM in Romania is a very sensible and complex problem taking into consideration the importance and the need of the security and safety for such activities. The Romanian Nuclear Regulatory Body set up strictly regulation and procedures according to the Recommendation of the IAEA Vienna and other international organizations. There were implemented the adequate regulation and procedures in order to keep the environmental impacts and the radiological consequences at the lower possible level and to assure the effectiveness of state nuclear security regime due to possible malicious acts in carrying out these activities including transport and the disposal site at the acceptable international levels. The levels of the estimated doses and risk expectation values for transport and disposal are within the acceptable limits provided by national and international regulations and recommendations but can increase, significantly during potential malicious acts. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Radiation Field of Packages Carrying Spent Co-60 Radioactive Sources - 12437

Among the diverse radioactive sources commonly exploited in medical and industrial applications, Co- 60 is increasingly used as strong gamma emitter. Over time, source manufacturers favored Co-60 as opposed to other gamma emitters because its relatively short half-life (5.27 year) that minimizes issues related to the management of disused sources. Disused Co-60 sources can retain a significant amount of radioactivity (from hundreds of MBq to several GBq) that still poses safety concerns on their handling and transportation. In this context a detailed knowledge of their radiation field would provide the necessary information for taking actions in preventing unnecessary doses to the workers and the population by optimizing transportation procedures and handling operations. We modeled the geometry and the materials constituting a transportation packaging of a spent Co-60 source which had an original maximum activity of a few GBq and was enclosed in a small lead irradiator. Then we applied a Monte Carlo transport code (MCNP5) for tracking down the gamma photons emitted by the source, including the secondary photons resulting by the interaction of the source photons with the surrounding materials. This allowed for the evaluation of the radiation field inside and outside the packaging, and the corresponding equivalent dose useful for checking the compliance with the regulations and the health risk of possible radiation exposure. We found that a typical 60-liters drum carrying a spent Co-60 source, enclosed in its original irradiator, with a residual activity of 300 MBq could already overcome an equivalent dose of 0.2 mSv/h on the drum external surface, which is the maximum equivalent dose at any point of the surface for this packaging as prescribed by local regulations. This condition is even more apparent when the source is slightly displaced with respect to the rotation axis of the drum, an easily occurring condition for sources not properly packaged, generating non-compliant hot-spots on the drum surface. As an example, a displacement of 5 cm translates in an increase of 80% in the dose level on the nearest side of the drum. We also found that the equivalent dose is significantly influenced by the scattered source photons and the secondary photons, whose contribution to the radiation field is mainly determined by the package geometry. The developed model resulted in an important tool for exploring the detail of the radiation field of a spent Co-60 source packaged for transportation allowing to check for compliance with the regulations and to evaluate risks to the workers and the population. It is worth to point out that this modeling approach is completely general and can be applied to a variety of different problems not limited to the transportation of radioactive material. In the example illustrated in this work, a small Co-60 source inside its irradiator and properly packaged in a drum for transport, presents a hot-spot on the drum surface if the irradiator is displaced of just few centimeters with respect to the rotation axis of the drum. A Monte Carlo simulation provided a complete description of the radiation field in the volume surrounding the packaging as a function of photon energy showing that a significant contribution comes from source photons scattered by the package itself. As an example, the Transport Index (TI) of this packaging was calculated, resulting in a value of 0.5. This value is low when compared to the regulations, however suggests that an individual just 1 meter afar from the packaging receives, in a few hours, a dose comparable to that received with a radiograph. This might be the case of the driver of a truck transporting the packaging, in particular when a number of packages are transported at the same time. It is important to notice that the results presented in this work are general, and can be appropriately scaled to represent the radiation field of similar geometries with gamma sources of different activity. However, this quick application of a Monte Carlo simulation provides useful insights in the process of preparing a packaging for transport. This simple model suggests that this approach can be efficient for screening a number of packaging configurations and flagging those potentially not compliant with regulations, that will require a more thorough care in the packaging process. Furthermore the knowledge of the details of the radiation field can be useful for optimizing the geometry of the transport cargo, and becomes essential when a variety of packaging are transported or stored, in order to prevent hazardous situations during their handling, and hence preventing unnecessary doses to the workers and the populations. It is worth to point out that this modeling approach is completely general and can be applied to a variety of different problems not limited to the transportation of radioactive material. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Disposing of High-Level Radioactive Waste in Germany - A Note from the Licensing Authority - 12530

Following the national German consensus on the termination of utilisation of nuclear energy in the summer of 2011, the Federal and Laender Governments have declared their intention to work together on a national consensus on the disposal of radioactive waste as well. Projected in the early 1970's the Federal Government had started exploring the possibility to establish a repository for HLW at the Gorleben site in 1977. However, there is still no repository available in Germany today. The delay results mainly from the national conflict over the suitability of the designated Gorleben site, considerably disrupting German society along the crevice that runs between supporters and opponents of nuclear energy. The Gorleben salt dome is situated in Lower Saxony, the German state that also hosts the infamous Asse mine repository for LLW and ILW and the Konrad repository project designated to receive LLW and ILW as well. With the fourth German project, the Morsleben L/ILW repository only 20 km away across the state border, the state of Lower Saxony carries the main load for the disposal of radioactive waste in Germany. After more than 25 years of exploration and a 10 year moratorium the Gorleben project has now reached a cross-road. Current plans for setting up a new site selection procedure in Germany call for the selection and exploration of up to four alternative sites, depending only on suitable geology. In the meantime the discussion is still open on whether the Gorleben project should be terminated in order to pacify the societal conflict or being kept in the selection process on account of its promising geology. The Lower Saxony Ministry for Environment and Climate Protection proposes to follow a twelve-step-program for finding the appropriate site, including the Gorleben site in the process. With its long history of exploration the site is the benchmark that alternative sites will have to compare with. Following the national consensus of 2011 on the termination of nuclear energy utilisation, it is now the time to reach a national consensus on the disposal of radioactive waste as well. This is a task that the country and society, federal and state governments, political parties and the citizens will have to jointly master within the current generation and within German territory. The basis for the consensus will be a reset to the beginning of this process. It has to start with a new site selection procedure that will take into account and compare up to four alternative sites. This procedure will have to follow the principle of highest possible security. It should be based on a stepwise approach, strictly following scientific criteria. Public confidence in the process and trust can only be achieved by a transparent procedure allowing for the participation of the public and the stakeholders. It is therefore mandatory to consult, both on a national and regional level, all involved parties (public authority, scientist and citizen). The national consensus must also include a decision on the future of the Gorleben exploratory site. The site selection procedure must therefore take this site into account as well. Furthermore, the final decision on safe disposal of German radioactive wastes must be made by sovereign rule by Federal Parliament and Federal Council. (authors)

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

Characterizations of the radioactive waste by the remotely-controlled collimated spectrometric system

Decontamination and decommissioning of the research reactors MR (Testing Reactor) and RFT (Reactor of Physics and Technology) has recently been initiated in the National Research Center (NRC) 'Kurchatov institute', Moscow. In the building, neighboring to the reactor, the storage of HLRW is located. The storage is made of monolithic concrete in which steel cells depth 4 m are located. In cells of storage the HLRW packed into cases are placed. These the radioactive waste are also subject to export on long storage in the specialized organization. For characterization of the radioactive waste in cases the remote-controlled collimated spectrometer system was used. The system consists of a spectrometric collimated gamma-ray detector, a color video camera and a control unit, mounted on a rotator, which are mounted on a tripod with the host computer. For determination of specific activity of radionuclides in cases, it is developed programs of calculation of coefficients of proportionality of specific activity to the corresponding speeds of the account in peaks of full absorption at single specific activity of radionuclides in cases. For determination of these coefficients the mathematical model of spectrometer system based on the Monte-Carlo method was used. Dependences of calibration coefficients for various radionuclides from distance between the detector and a case at various values of the radioactive waste density in cases are given. Measurements of specific activity in cases are taken and are discussed. By results of measurements decisions on the appeal of the radioactive waste being in cases are made. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗