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All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

Fall Protection Procedures for Sealing Bulk Waste Shipments by Rail Cars at Formerly Utilized Sites Remedial Action Program (FUSRAP) Sites - 13509

Rail-cars loaded with radioactive materials must be closed and fastened to comply with United States Department of Transportation (DOT) requirements before they shipped. Securing waste shipments in a manner that meets these regulations typically results in the use of a sealable rail-car liner. Workers accessing the tops of the 2.74 m high rail-cars to seal and inspect liners for compliance prior to shipment may be exposed to a fall hazard. Relatively recent revisions to the Fall Protection requirements in the Safety and Health Requirements Manual (EM385-1-1, U.S. Army Corps of Engineers) have necessitated modifications to the fall protection systems previously employed for rail-car loading at Formerly Utilized Sites Remedial Action Program (FUSRAP) sites. In response these projects have developed site-specific procedures to protect workers and maintain compliance with the improved fall protection regulations. (authors)

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

Approaches for Communicating Technical Information - 16012

Projects within the nuclear industry are generally complex and associated with high hazards. To achieve success there is often a necessity to apply innovation or adopt longer term strategies and solutions. Because of these long timescales and the fact that some approaches revolve around allowing natural radioactive decay to take place prior to visible action, the industry is sometimes perceived to be technically unable to make progress with its environmental and waste management obligations. Work within the industry additionally takes on a high profile and invariably creates greater levels of anxiety and mistrust in the eyes of stakeholders than comparable projects from non-nuclear related industries. Many aspects of the industry are still seen as secretive, environmentally detrimental and 'risky' and all components of the nuclear life-cycle are therefore open to external scrutiny. Although communication and engagement has increased greatly over the years, especially in developed countries, there is still room for improvement in this field. Such improvement might relate to the frequency and level of engagement, the subject matter or the actual process itself. Communication with stakeholders is required throughout the various phases of the nuclear lifecycle whether the activity relates to mining/milling of uranium, fuel production, operational plant activities, decommissioning, remediation or waste management. The concerns and aspirations of stakeholders during these different activities will vary but need to be adequately addressed. Some organisations now have a basic engagement program and convince themselves that they are carrying out true engagement. But just ticking a box to say that an engagement program is in place does not provide evidence of its appropriateness or success as on many occasions' organisations might still not understand how to adequately engage and communicate technical information. It is not always the information we choose to communicate but that which might be withheld that is often perceived to be most relevant to the stakeholder community. Communicating technical information isn't as straightforward as it sounds, but still has to be undertaken, otherwise trust, understanding and acceptance of chosen options will not be achieved. I utilise what I call the 'ten principles of engagement' which I believe can help when communicating technical Information in relation to nuclear and radiological matters. These are; 1. Develop Trust. 2. Educating people and providing information in a transparent manner. 3. Develop an engagement process that is truly a two way process. 4. Openly discuss risk perception and risk communication. 5. Never try to trivialise risk. 6. Where possible, simplify the language you use. 7. Put background and naturally occurring radioactivity into context. 8. Explain how radioactive materials are often used in everyday life. 9. Try to demonstrate that you too are a member of the public. 10.Walk the talk - Try to undertake what you wish others to undertake. Further research is being undertaken to determine if Mental Models might assist with the communication of nuclear and radiological information. A mental model is an explanation of someone's thought process about how something works in the real world. The mental model approach aims to investigate the judgment and decision-making processes of both expert and lay groups, thus allowing a comparison and visualisation of the different views to be made. It is undoubtedly worthwhile keeping a watching brief on the outcome of such research as we strive to improve our communication with the general public. The aim of this paper is to highlight some simple principles and approaches that can assist with the communication of technical information. (authors)

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

SU-E-T-388: Estimating the Radioactivity Inventory of a Cyclotron Based Pencil Beam Proton Therapy Facility

Purpose: Parts of the cyclotron and energy degrader are incidentally activated by protons lost during the acceleration and transport of protons for radiation therapy. An understanding of the radioactive material inventory is needed when regulatory requirements are assessed. Methods: First, the tumor dose and volume is used to determine the required energy deposition. For spot scanning, the tumor length along the beam path determines the number of required energy layers. For each energy layer the energy deposition per proton can be calculated from the residual proton range within the tumor. Assuming a typical layer weighting, an effective energy deposition per proton can then be calculated. The total number of required protons and the number of protons per energy layer can then be calculated. For each energy layer, proton losses in the energy degrader are calculated separately since its transmission efficiency, and hence the amount of protons lost, is energy dependent. The degrader efficiency also determines the number of protons requested from the cyclotron. The cyclotron extraction efficiency allows a calculation of the proton losses within the cyclotron. The saturation activity induced in the cyclotron and the degrader is equal to the production rate R for isotopes whose half-life is shorter that the projected cyclotron life time. R can be calculated from the proton loss rate and published production cross sections. Results: About 1/3 of the saturation activity is produced in the cyclotron and 2/3 in the energy degrader. For a projected case mix and a patient load of 1100 fractions per week at 1.8 Gy per fraction a combined activity of 180 mCi was estimated at saturation. Conclusion: Calculations were used to support to application of a radioactive materials license for the possession of 200 mCi of activity for isotopes with atomic numbers ranging from 1-83.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2012

Evaluation of Low-Level Waste Disposal Receipt Data for Los Alamos National Laboratory Technical Area 54, Area G Disposal Facility - Fiscal Year 2011

The Los Alamos National Laboratory (LANL or the Laboratory) generates radioactive waste as a result of various activities. Operational or institutional waste is generated from a wide variety of research and development activities including nuclear weapons development, energy production, and medical research. Environmental restoration (ER), and decontamination and decommissioning (D and D) waste is generated as contaminated sites and facilities at LANL undergo cleanup or remediation. The majority of this waste is low-level radioactive waste (LLW) and is disposed of at the Technical Area 54 (TA-54), Area G disposal facility. U.S. Department of Energy (DOE) Order 435.1 (DOE, 2001) requires that radioactive waste be managed in a manner that protects public health and safety, and the environment. To comply with this order, DOE field sites must prepare and maintain site-specific radiological performance assessments for LLW disposal facilities that accept waste after September 26, 1988. Furthermore, sites are required to conduct composite analyses that account for the cumulative impacts of all waste that has been (or will be) disposed of at the facilities and other sources of radioactive material that may interact with the facilities. Revision 4 of the Area G performance assessment and composite analysis was issued in 2008 (LANL, 2008). These analyses estimate rates of radionuclide release from the waste disposed of at the facility, simulate the movement of radionuclides through the environment, and project potential radiation doses to humans for several on-site and off-site exposure scenarios. The assessments are based on existing site and disposal facility data and on assumptions about future rates and methods of waste disposal. The accuracy of the performance assessment and composite analysis depends upon the validity of the data used and assumptions made in conducting the analyses. If changes in these data and assumptions are significant, they may invalidate or call into question certain aspects of the analyses. For example, if the volumes and activities of waste disposed of during the remainder of the disposal facility's lifetime differ significantly from those projected, the doses projected by the analyses may no longer apply. DOE field sites are required to implement a performance assessment and composite analysis maintenance program. The purpose of this program is to ensure the continued applicability of the analyses through incremental improvement of the level of understanding of the disposal site and facility. Site personnel are required to conduct field and experimental work to reduce the uncertainty in the data and models used in the assessments. Furthermore, they are required to conduct periodic reviews of waste receipts, comparing them to projected waste disposal rates. The radiological inventory for Area G was updated in conjunction with Revision 4 of the performance assessment and composite analysis (Shuman, 2008). That effort used disposal records and other sources of information to estimate the quantities of radioactive waste that have been disposed of at Area G from 1959, the year the facility started receiving waste on a routine basis, through 2007. It also estimated the quantities of LLW that will require disposal from 2008 through 2044, the year in which it is assumed that disposal operations at Area G will cease. This report documents the fourth review of Area G disposal receipts since the inventory was updated and examines information for waste placed in the ground during fiscal years (FY) 2008 through 2011. The primary objective of the disposal receipt review is to ensure that the future waste inventory projections developed for the performance assessment and composite analysis are consistent with the actual types and quantities of waste being disposed of at Area G. Toward this end, the disposal data that are the subject of this review are used to update the future waste inventory projections for the disposal facility. These projections are compared to the future inventory projections that were developed for Revision 4 of the performance assessment and composite analysis. The approach used to characterize the FY 2008 through 2011 waste is generally the same as that used to characterize the inventory for the Revision 4 analyses (Shuman, 2008). This methodology is described in Section 2. The results of the disposal receipt review are presented in Section 3 and discussed in terms of their significance to the Area G analyses.

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

Commercial sealed sources and timely disposal: Challenges and incentives - 15456

It is widely accepted as a radioactive material management best-practice that storage of disused sealed sources should be a temporary measure. Effective end-of-life management, including timely disposition of disused and unwanted sealed sources, is an important way to address national security, public health, and safety concerns presented by long-term storage of disused sealed sources. However, significantly increased disposal access since 2012 has not led commercial sealed source generators in the U.S. to dispose of their disused and unwanted sources as quickly as many had expected. As a result, policy-makers and non-government stakeholders are considering ways to incentivize sealed source generators to make more timely decisions regarding the disposition of their disused and unwanted sources. To help ensure that ongoing efforts to develop sealed source management and disposal incentives are efficient and effective, NNSA/GMS is engaging both public and private sector partners to better understand generator considerations, motivations, and challenges with regard to sealed source management and disposal decisions. The results of these engagements to date indicate potentially effective ways to encourage sealed source waste generators to disposition their sources in a timely fashion. (authors)

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

The Joint Convention, Its Global Impact and U.S Continuing Involvement - 16645

The Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management (hereafter called the Joint Convention) is an international convention, conducted under the auspices of the International Atomic Energy Agency (IAEA). It is one of a suite of international instruments dealing with nuclear safety and physical security, which serve to promote a global culture for the safe use of radioactive materials. Unlike other safety or security conventions, however, the Joint Convention deals exclusively with the end of the nuclear fuel cycle. The Joint Convention focuses on the lessons to be learned from past experience and existing facility operations, which can be implemented in the policies and designs for future operating facilities in the nuclear fuel cycle. The U.S. has participated fully in the review process, as well as setting forth initiatives to sponsor increased ratification. The U.S. holds the current Joint Convention presidency for the Fifth Review Cycle. Because of the important benefits associated with active participation, the U.S. has strongly supported a Regional Conference Initiative outreach program to increase membership. We also provided an expert for each of the conferences to assist in advancing the message to non-Member States, in particular developing nations considering ratification. The Joint Convention process also provides opportunities to identify future areas of bilateral and multilateral technical and regulatory cooperation with other Parties. Furthermore, the Joint Convention is consistent with U.S. foreign policy considerations to support, as a priority, the strengthening of the worldwide safety culture in the use of nuclear energy. (authors)

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

French management model for sustainable transportation of used fuel - 15376

Safe and efficient transportation of radioactive material is key to the success of our industry. Used fuel transports are inherently complex due to the nature of the material and the high visibility. Only a few countries have developed comprehensive used fuel transportation systems that require a wide range of equipment, services and expertise. In the many countries, such as the Unites States, where interim consolidated storage, recycling or geological repository are not available yet, similar systems will have to be implemented. For more than 50 years AREVA TN has safely shipped more than 7,000 used fuel transport casks. The transportation system that was initially developed in the 1970's has been adapted and enhanced over the years to meet more restrictive regulatory requirements and evolving customer needs, and to address public concerns. The numerous 'lessons learned' have offered data and guidance that have allowed for also efficient and consistent improvement over the decades. Based on AREVA TN extensive expertise, this paper will describe the different phases and milestones that need to be met to set up, license and operate a successful used fuel transportation management system. Transportation of spent fuel in France could not be performed effectively and efficiently without strong collaboration with Electricite de France (EDF) who plays a critical role as the shipper. EDF will provide a nuclear operator's perspective on the most critical aspects of a successful used fuel transportation program. (authors)

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

The study of equivalent dose of uranium in long bean (V. U. Sesquipedalis) and the effect on human

In the case of accidental release of Uranium-238 ({sup 238}U) radionuclides in a nuclear facility or in the environment, internal contamination by either acute or chronic exposure has the potential to induce both radiological and chemical toxic effects. A study was conducted to estimate the {sup 238}U radionuclide concentration in the long beans using Induced Coupled Mass Plasma-Spectrometry (ICP-MS). {sup 238}U radionuclide is a naturally occurring radioactive material that can be found in soil and can be transferred to the long bean (Vigna unguiculata subsp. Sesquapedalis) directly or indirectly via water or air. Kidney and liver are the major sites of deposition of {sup 238}U radionuclide. The obtained dose exposed in the liver and kidney is used to assess the safety level for public intake of {sup 238}U radionuclide from the consumption of long beans. The concentration of {sup 238}U radionuclide measured in long bean samples was 0.0226 ± 0.0009 mg/kg. Total activity of {sup 238}U radionuclide was 0.0044 ± 0.0002 Bq/day with the daily intake of 0.3545 ± 0.0143 µg/day and the annual committed effective dose due to ingestion of {sup 238}U radionuclide in long beans was 0.2230 ± 0.0087 µSv/year. The committed equivalent dose of {sup 238}U radionuclide from the assessment in the liver and kidney are 0.4198 ± 0.0165 nSv and 10.9335 ± 0.4288 nSv. The risk of cancer of {sup 238}U radionuclide was determined to be (86.0466 ± 3.3748) × 10-9. Thus, the results concluded that {sup 238}U radionuclide in local long beans was in the permitted level and safe to consume without posing any significant radiological threat to population.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2015

Nevada National Security Site Environmental Report 2014

This Nevada Nuclear Security Site Environmental Report (NNSSER) was prepared to satisfy DOE Order DOE O 231.1B, “Environment, Safety and Health Reporting.” Its purpose is to (1) report compliance status with environmental standards and requirements, (2) present results of environmental monitoring of radiological and nonradiological effluents, (3) report estimated radiological doses to the public from releases of radioactive material, (4) summarize environmental incidents of noncompliance and actions taken in response to them, (5) describe the NNSA/NFO Environmental Management System and characterize its performance, and (6) highlight significant environmental programs and efforts.This NNSSER summarizes data and compliance status for calendar year 2014 at the Nevada National Security Site (NNSS) (formerly the Nevada Test Site) and its two support facilities, the North Las Vegas Facility (NLVF) and the Remote Sensing Laboratory–Nellis (RSL-Nellis). It also addresses environmental restoration (ER) projects conducted at the Tonopah Test Range (TTR) and the Nevada Test and Training Range (NTTR). Through a Memorandum of Agreement, NNSA/NFO is responsible for the oversight of these ER projects, and the Sandia Field Office of NNSA (NNSA/SFO) has oversight of all other TTR and NTTR activities. NNSA/SFO produces the TTR annual environmental report, which is available at http://www.sandia.gov/news/publications/environmental/index.html.Major Site Programs and Facilities NNSA/NFO directs the management and operation of the NNSS and six sites across the nation. The six sites include two in Nevada (NLVF and RSL-Nellis) and four in other states (RSL-Andrews in Maryland, Livermore Operations in California, Los Alamos Operations in New Mexico, and Special Technologies Laboratory in California). Los Alamos, Lawrence Livermore, and Sandia National Laboratories are the principal organizations that sponsor and implement the nuclear weapons programs at the NNSS. National Security Technologies, LLC (NSTec), is the current Management and Operating contractor accountable for the successful execution of work and ensuring that work is performed in compliance with environmental regulations. The six sites all provide support to enhance the NNSS as a location for weapons experimentation and nuclear test readiness. The three major NNSS missions include National Security/Defense, Environmental Management, and Nondefense. The major programs that support these missions are Stockpile Stewardship and Management, Nonproliferation and Counterterrorism, Nuclear Emergency Response, Strategic Partnership Projects, Environmental Restoration, Waste Management, Conservation and Renewable Energy, Other Research and Development, and Infrastructure. The major facilities that support the programs include the U1a Facility, Big Explosives Experimental Facility (BEEF), Device Assembly Facility, Dense Plasma Focus Facility, Joint Actinide Shock Physics Experimental Research Facility, Radiological/Nuclear Countermeasures Test and Evaluation Complex, Nonproliferation Test and Evaluation Complex (NPTEC), Radiological/Nuclear Weapons of Mass Destruction Incident Exercise Site (known as the T-1 Site), Area 5 Radioactive Waste Management Complex (RWMC), and the Area 3 Radioactive Waste Management Site (RWMS).

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2021

REGULATORY TESTING AND POSTTEST ANALYSIS OF THE DPP-3 TYPE B SHIPPING CONTAINER FOR NCT AND HAC TESTS

The safe transportation of radioactive material using appropriately designed and certified transportation packages is important to protect the public and environment. Therefore, a comprehensive and rigorous system of packaging requirements are required. The purpose of this test protocol is to determine the effects of drop testing on package performance in a Type B shipping package (DPP-3). Required physical tests for the DPP-3 package include normal conditions of transport and hypothetical accident conditions test requirements. The preparation, regulatory testing, and post-test evaluation of a Type B shipping package tested to the 10 CFR 71.71 and 10 CFR 71.73 requirements will be presented. Six packages were tested and the responses when subjected to the HAC sequence of free drop, puncture, and thermal tests according to 10 CFR 71.73 were measured. The posttest helium leak test showed a leak rate less than 1 × 10-7 He-cc/sec which means that the containment boundary was maintained after the regulatory testing, therefore maintaining a leak tight environment. To determine the amount of change in the package performance measurements, statistical analyses were conducted using SPSS version 26. The null hypothesis of the testing was that there would not be a change in the measurements from pre to post-testing large enough to find statistical significance. Therefore, a finding of statistical significance indicated too much variability in the package design and was deemed an unfavorable finding. A repeated measures t-test was used to analyze the drum lids and CV torques. A one-way ANOVA was used to analyze height and diameter packages. Results of the repeated measures t-test indicated that all six containment vessel lid torques were statistically significant. However only three out of the six drum lid torques were statistically significant. Therefore, design of the lids caused more variability than necessary when subjected to the following drop tests: 1.) horizontal and cold; 2.) vertical lid down; and 3.) corner with lid up. Results of the one-way ANOVA analysis on height indicated that four out of the six CTUs were statistically significant. Thus, design of the CTUs caused more variability then necessary when subjected to the following drop tests: 1.) horizontal hot; 2.) horizontal cold; 3.) corner lid up; and 4.) corner lid down. Results of the one-way ANOVA analysis on diameter indicated that all six CTUs were statistically significant. Thus, design of the CTUs caused more variability then necessary when subjected to all of the tests. The DPP-3 design has been shown to be an acceptable design for the transport of radioactive materials. However, results of the statistical analysis reveal a possible improvement in package and test design to reduce variability in package performance measurement before and after testing.

Martinez, Oscar↗
DOE OSTI2013

Pressure Build-Up During the Fire Test in Type B(U) Packages Containing Water - 13280

The safety assessment of packages for the transport of radioactive materials with content containing liquids requires special consideration. The main focus is on water as supplementary liquid content in Type B(U) packages. A typical content of a Type B(U) package is ion exchange resin, waste of a nuclear power plant, which is not dried, normally only drained. Besides the saturated ion exchange resin, a small amount of free water can be included in these contents. Compared to the safety assessment of packages with dry content, attention must be paid to some more specific issues. An overview of these issues is provided. The physical and chemical compatibility of the content itself and the content compatibility with the packages materials must be demonstrated for the assessment. Regarding the mechanical resistance the package has to withstand the forces resulting from the freezing liquid. The most interesting point, however, is the pressure build-up inside the package due to vaporization. This could for example be caused by radiolysis of the liquid and must be taken into account for the storage period. If the package is stressed by the total inner pressure, this pressure leads to mechanical loads to the package body, the lid and the lid bolts. Thus, the pressure is the driving force on the gasket system regarding the activity release and a possible loss of tightness. The total pressure in any calculation is the sum of partial pressures of different gases which can be caused by different effects. The pressure build-up inside the package caused by the regulatory thermal test (30 min at 800 deg. C), as part of the cumulative test scenario under accident conditions of transport is discussed primarily. To determine the pressure, the temperature distribution in the content must be calculated for the whole period from beginning of the thermal test until cooling-down. In this case, while calculating the temperature distribution, conduction and radiation as well as evaporation and condensation during the associated process of transport have to be considered. This paper discusses limiting amounts of water inside the cask which could lead to unacceptable pressure and takes into account saturated steam as well as overheated steam. However, the difficulties of assessing casks containing wet content will be discussed. From the authority assessment point of view, drying of the content could be an effective way to avoid the above described pressure build-up and the associated difficulties for the safety assessment. (authors)

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

First Road / River Barge Security Transport of SNF under the New German Transport Security Guidelines - 19102

A transfer of 342 spent fuel assemblies from NPP Obrigheim to NPP Neckarwestheim, both located alongside the river Neckar in the south of Germany, is required to promote the decommissioning process of NPP Obrigheim, which started in 2008 after the NPP had been shut down three years before. The spent fuel storage facility in Neckarwestheim offers sufficient storage space to accept the 342 SNF assemblies, stored in fifteen dry casks type CASTOR{sup R} 440/84 mvK. After several transport feasibility studies had been conducted, Daher Nuclear Technologies GmbH was awarded the commission to plan, prepare and conduct the SNF transport. Since the combined road / river barge transport promised to be the most suitable mode of transportation Daher Nuclear Technologies started to prepare its security concept, a prerequisite to apply for and attain the transport license mandatory for transporting spent nuclear fuel in Germany. In March 2014 the company applied for the transport license, which, after more than three years and numerous amendments to the security concept, was finally granted to Daher Nuclear Technologies on May 16, 2017 by the German Federal Office for the Safety of Nuclear Waste Management (successor organization of the BfS). Simultaneously to the license application phase the transport equipment - amongst others the heavy-duty trucks and trailers, the pusher tugs, the river barge and transport frames - had to be provided and modified to meet the intensified requirements of the New German Transport Security Guidelines Road / Rail as well as several other legal regulations referring to the transport of dangerous goods by road and inland waterways (e.g. ADR - European Agreement concerning the International Carriage of Dangerous Goods by Road and ADN - European Agreement concerning the International Carriage of Dangerous Goods by Inland Waterways). The planning process involved various meetings with representatives of different authorities and administrative bodies incl. accredited experts as well as subcontractors. Tremendous training efforts had to be undertaken to train more than 100 transport participants regarding security measures, health, safety and environmental measures, as well as the proper handling of radioactive material. This also included the training of the security staff, that had to be qualified in compliance with the new Transport Security Guidelines Road and Rail. Finally, in June 2017, when all preparations where complete, all requirements met, and all permits obtained, the first ever security SNF transport on a German river was performed. Three laden CASTOR{sup R} 440/84 mvK transport and storage casks were transported in one batch via heavy-duty truck-trailer from the NPP Obrigheim to the NPP's own harbor, to be rolled on to the river barge, that would carry the three casks upstream the river Neckar to the approx. fifty kilometer distant NPP Neckarwestheim. The journey on the river took about eleven hours, in which several bridges and locks had to be passed. The entire transport was escorted by qualified security staff and auxiliary police forces on waterways and alongside on roads. It attracted nationwide public attention, due to its pioneer status and the fact, that there hadn't been any SNF or HLW shipments in Germany since 2011. After the convoy safely arrived at NPP Neckarwestheim's pier the three laden CASTOR{sup R} 440/84 mvK casks were rolled off the river barge via heavy-duty truck-trailers to be stored in the on-site interim storage facility. This first transport was a success and was followed by four additionally successful security transports in September, October, November and December 2017. The public attention receded significantly during the transport campaign, which ended on December 19, 2017 with the transfer of the last three of fifteen casks to the on-site interim storage facility in Neckarwestheim. With the last cask safely transferred to the interim storage facility Daher Nuclear Technologies successfully completed the transport campaign on schedule, meeting one of EnBW's fundamental objectives to finish the transfer of the 342 SNF assemblies by the end of 2017. (authors)

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

Special Analysis: Disposal Plan for Pit 38 at Technical Area 54, Area G

Los Alamos National Laboratory (LANL) generates radioactive waste as a result of various activities. Operational waste is generated from a wide variety of research and development activities including nuclear weapons development, energy production, and medical research; environmental restoration (ER), and decontamination and decommissioning (D&D) waste is generated as contaminated sites and facilities at LANL undergo cleanup or remediation. The majority of this waste is low-level radioactive waste (LLW) and is disposed of at the Technical Area 54 (TA-54), Area G disposal facility. U.S. Department of Energy (DOE) Order 435.1 (DOE, 2001) requires that radioactive waste be managed in a manner that protects public health and safety, and the environment. To comply with this order, DOE field sites must prepare site-specific radiological performance assessments for LLW disposal facilities that accept waste after September 26, 1988. Furthermore, sites are required to conduct composite analyses that account for the cumulative impacts of all waste that has been (or will be) disposed of at the facilities and other sources of radioactive material that may interact with the facilities. Revision 4 of the Area G performance assessment and composite analysis was issued in 2008 (LANL, 2008). These analyses estimate rates of radionuclide release from the waste disposed of at the facility, simulate the movement of radionuclides through the environment, and project potential radiation doses to humans for several on- and off-site exposure scenarios. The assessments are based on existing site and disposal facility data, and on assumptions about future rates and methods of waste disposal. The Area G disposal facility consists of Material Disposal Area (MDA) G and the Zone 4 expansion area. To date, disposal operations have been confined to MDA G and are scheduled to continue in that region until MDA G undergoes final closure at the end of 2013. Given its impending closure, efforts have been made to utilize the remaining disposal capacity within MDA G to the greatest extent possible. One approach for doing this has been to dispose of low-activity waste from cleanup operations at LANL in the headspace of selected disposal pits. Waste acceptance criteria (WAC) for the material placed in the headspace of pits 15, 37, and 38 have been developed (LANL, 2010) and the impacts of placing waste in the headspace of these units has been evaluated (LANL, 2012a). The efforts to maximize disposal efficiency have taken on renewed importance because of the disposal demands placed on MDA G by the large volumes of waste that are being generated at LANL by cleanup efforts. For example, large quantities of waste were recently generated by the retrieval of waste formerly disposed of at TA-21, MDA B. A portion of this material has been disposed of in the headspace of pit 38 in compliance with the WAC developed for that disposal strategy; a large amount of waste has also been sent to off-site facilities for disposal. Nevertheless, large quantities of MDA B waste remain that require disposal. An extension of pit 38 was proposed to provide the disposal capacity that will be needed to dispose of institutional waste and MDA B waste through 2013. A special analysis was prepared to evaluate the impacts of the pit extension (LANL, 2012b). The analysis concluded that the disposal unit could be extended with modest increases in the exposures projected for the Area G performance assessment and composite analysis, as long as limits were placed on the radionuclide concentrations in the waste that is placed in the headspace of the pit. Based, in part, on the results of the special analysis, the extension of pit 38 was approved and excavation of the additional disposal capacity was started in May 2012. The special analysis presented here uses performance modeling to identify a disposal plan for the placement of waste in pit 38. The modeling uses a refined design of the disposal unit and updated radionuclide inventories to identify a disposal configuration that promotes efficient utilization of the pit and ensures continued compliance with DOE Order 435.1 performance objectives. Section 2 describes the methods used to conduct the analysis; the results of the evaluation are provided in Section 3. The disposal plan for pit 38 is provided in Section 4 and the conclusions of the investigation are provided in Section 5. Throughout the report, pit 38 is used to refer to the entire disposal unit, including the existing pit and the extension that is currently under construction. Where a distinction between the two portions of the pit is necessary, the existing unit is referred to as pit 38 proper and the new portion of the pit as the pit 38 extension or, more simply, the extension.

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

Phyto-screening for Volatile Organic Compounds in a Clay-based Glacial Till at the Niagara Falls Storage Site, New York - 17372

The Niagara Falls Storage Site (NFSS) and associated vicinity properties (VPs) occupy approximately 607 hectares (1,500 acres) of the original 3,035-hectare (7,500-acre) Lake Ontario Ordnance Works (LOOW). The NFSS was used to store radioactive materials, residues and waste from uranium ore processing between 1944 and 1952. The waste-handling practices produced an array of groundwater impacts throughout the site. Although the groundwater at the NFSS is naturally high in total dissolved solids that preclude untreated consumption, the groundwater contaminant levels in some areas of the site are very high and may pose risk to a Construction Worker receptor, as well as limit the options for future site re-use. Volatile organic compound (VOC) impacted soil in the northern portion of the site has produced a groundwater plume with significant concentrations of tetrachloroethene (PCE) and associated daughter compounds. Active groundwater remediation is impractical due to the clayey soils, as evident from adjacent property owners who have unsuccessfully attempted active remediation of their VOC plumes. Consequently, the Buffalo District may evaluate excavation options, with secondary phyto-remediative approaches as passive alternatives to lessen the groundwater impacts during the project's life cycle. A preliminary field-study in the VOC plume area sampled the interior of a Black Locust (Robinia pseudoacacia) tree using a Haglof tree corer. Detectable levels of tetrachloroethene (PCE), trichloroethene (TCE), and cis-1,2,-dichloroethene (cis-DCE) were found in the woody interior of the tree. Subsequent sampling and analyses were performed on proximal trees of varying species in and around the currently delineated VOC plume. The results indicate that poplar and native cottonwood variants preferentially contain the VOCs, whereas collocated hardwood trees produced non-detectable results. Additional sampling of Cottonwoods in the VOC-impacted area promoted further delineations of the impacts. The tree-sampling data can be considered a broad surrogate for soil sampling to optimally delineate the VOC plume without performing additional invasive sampling of the subsurface (e.g., drilling). The results also will be used to determine which native tree species preferentially uptake the VOCs from soil and groundwater. These species, and possibly other hybrid species known for VOC uptake and transpiration, would be propagated to promote interim phyto-extraction of the VOCs during the maintenance period between decision-document completion and likely remedial action under the Formerly Utilized Sites Remedial Action Program (FUSRAP). (authors)

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

Corrosion Testing of Carbon Steel Exposed to Sludge Heel Chemical Cleaning Solutions - 16109

As efforts continue to treat and dispose of millions of gallons of legacy radioactive materials from the production of nuclear weapons, non-compliant waste storage tanks will gradually be emptied of the bulk waste volume leaving heel materials requiring removal prior to tank closure. The waste heel slurries are distributed on the floor of large, million gallon tanks, which frequently contain numerous obstructions that limit the effectiveness of mechanical removal methods. As a result, chemical cleaning methods are needed for the effective removal of the heels as well as chemical scales that are present on the tank walls and other interior surfaces. Two tank cleaning technologies have already been implemented at the Savannah River Site (SRS): Low Temperature Aluminum Dissolution (LTAD) and Bulk Oxalic Acid Cleaning (BOAC). Recent chemical cleaning efforts on SRS Tank 12 were very successful with regard to bulk sludge heel (especially for Al, Fe, and U phases) and beta/gamma radionuclide removal. Although chemical cleaning using these technologies has been shown to be effective, no disposition path has been identified for oxalate added during BOAC, and insoluble oxalate salts are accumulating within the SRS tank farm and waste processing facilities (evaporators, etc.). Extensive sludge washing is also required to remove moderately soluble sodium oxalate salts prior to sludge vitrification in the SRS Defense Waste Processing Facility (DWPF). As a result, oxalate additions to the tank farm need to be minimized by the use of supplementary acids to assist sludge removal in OA or the use of other cleaning reagents or processing strategies. Previous SRNL testing revealed the importance of pH control for BOAC and recommended the use of a supplementary acid (dilute HNO{sub 3}) with OA to minimize oxalate additions. Primary drivers in SRS Tank Closure Performance Assessments, which evaluate the fate and impact of tank sludge residuals on a geological timescale, are the removal of alpha emitting radionuclides present at low concentrations such as Pu, Am, and Np, which are not highly soluble in currently utilized chemical cleaning reagents. Scoping studies conducted at SRNL revealed promising methods to dissolve the actinides within the HLW tank heels. Oxidation of the actinides with permanganate in either strong caustic (10 M NaOH) or dilute acidic (0.2 M HNO{sub 3}) solutions was shown to result in dissolution of oxy/hydroxide phases of these metals in the absence of major sludge phases. The waste tanks at SRS and Hanford are constructed of carbon steel. Carbon steel corrodes rapidly in most acidic and/or oxidizing environments. Previous electrochemical corrosion testing indicated that the corrosion rates in dilute HNO{sub 3} with OA may result in manageable corrosion rates, over short exposure periods. Additional corrosion testing was performed to provide data for the nitric acid/OA blend as well as the acid/permanganate and caustic permanganate based chemical cleaning solutions. Corrosion rates from the passive coupon tests for the nitric acid/oxalic acid blend were significantly greater than those observed previously with the electrochemical testing. This result may be due to simultaneous electrochemical and chemical dissolution of the steel. The corrosion rate data from the coupon tests in the nitric acid/oxalic acid blend were evaluated to determine the degree of potential structural damage. The corrosion rates, although relatively high, would not be expected to cause damage that would reduce the capacity of the tank primary wall if the process is completed within a month. The corrosivity of sodium permanganate in four proposed cleaning solutions, 1 nitric acid solution and 3 sodium hydroxide solutions, was studied by electrochemical methods at room temperature. The corrosion rates were significantly less aggressive than in the nitric acid/oxalic acid blend corrosion tests. Passive coupon tests needs to be performed in these solutions to ensure that chemical dissolution of the metal is not occurring simultaneously. (authors)

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

Derivation of heterogeneous material distributions and their sensitivity to HM-coupled two-phase flow models exemplified with the LASGIT experiment

Abstract Advective gas transport in bentonite, a possible buffer material in repositories for radioactive materials, is difficult to simulate in numerical continuum models, partly due to the complicated microstructure of bentonite. To generate reliable models of repositories nevertheless, spatially distributed heterogeneous material properties can be used to allow localization of gas flow. In this study, a pore-size-dependent stochastic approach of the gas entry pressure is derived from Mercury Intrusion Porosimetry, which is used to replicate measurements from the LASGIT experiment. In addition, three benchmark tests are simulated to investigate the dependence of heterogeneous distributions of material properties on the mesh discretization, the temporal dependence, and the coupling between the processes influenced by the heterogeneous parameters. The numerical modeling results of the LASGIT experiment show that the onset of gas flow into the system and the subsequent increase in pressure and stress can be well reproduced using heterogeneous distributions. Compared to a model with homogeneous material properties, heterogeneous distributions may allow the generation of dilatancy-controlled microfractures—an important feature with regard to the advective gas flow in bentonites. However, it can be observed that the heterogeneous distributions in LASGIT are less significant, as technical gaps or differences in material types could have a greater impact.

Environmental Sciences & Ecology↗
DOE OSTI2024

Safety Basis Compliant Flexible Operations [Capstone Project]

Compliant nuclear operations must be weighed against the radiological risk to the worker, public, and the environment. This project set out to look at one facility at Los Alamos National Laboratory (LANL or the Laboratory) and compare the risk with the operational restrictions associated with the Technical Safety Requirements (TSRs). The hope was to identify controls that could be reduced or simplified while maintaining compliance with 10 CFR 830, Nuclear Safety Management. This would result in increasing facility operational flexibility and efficiency. Controls were identified to be simplified or eliminated based on risk. This identification was initially envisioned from the perspective of reanalysis of the accident and hazard analysis consequences. Additionally, it was envisioned that fire testing could show that the consequences of a fire are not as dramatic as analyzed. Specific facility controls were identified that cause operational inefficiencies. Based on data collection, two controls were identified for further evaluation. The first control was combustible loading within the waste storage buildings, and the second control is related to the seismic power cutoff switch. In evaluating the combustible loading control, the thought is that the fire suppression system (FSS) provides adequate protection and the combustible loading controls do not need elevation to a TSR control. The FSS, along with the Fire Protection Program, would provide protection to prevent a fire from impacting large amounts of stored waste material. The seismic switch was more of a challenge. When reviewing the facility operation, it was determined that the facility does not operate anywhere near the established material-at-risk (MAR – radioactive material) limits. By adjusting these limits, a reduced consequence could be achieved and hence, a significant risk reduction achieved. A new operating mode would be used to achieve this reduction. A reduced MAR limit was successfully used as a test case for snow removal during this project. This project expands on that trial case as related to combustible loading and the seismic power cutoff switch TSRs. This will increase the efficiency of operations with reduced TSR controls while reducing overall facility risk. This is a win-win situation for Triad National Security, LLC (Triad) and the U.S. Department of Energy (DOE). Due to schedule and budgetary constraints, not all envisioned work scope was accomplished. However, it is maintained in this report for future efforts.

61 RADIATION PROTECTION AND DOSIMETRY↗