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Cost and Performance Report for the ASTD Reuse of Concrete Within DOE from D&D Projects

Description: This cost and performance report describes the Accelerated Site Technology Deployment project that developed the Protocol for Development of Authorized Release Limits for Concrete at U.S. DOE Sites, which identifies the steps for obtaining approval to reuse concrete from Deactivation and Decommissioning of facilities. This protocol compares the risk and cost of various disposition paths for the concrete and follows the authorized release approach described in the DOE's draft handbook, Controlling Release for Reuse or Recycle of Property Containing Residual Radioactive Material. This approach provides for the development of authorized release limits through a series of prescribed steps before approval for release is granted. A case study was also completed on a previously decommissioned facility.
Date: September 1, 2000
Creator: Kamboj, S.; Arnish, J.; Chen, S. Y.; Phillips, Ann Marie; Meservey, Richard Harlan & Tripp, Julia Lynn
Partner: UNT Libraries Government Documents Department

Volumetric Radioactivity Viewed as Surface Radioactivity for Free Release Assessment Purposes

Description: As a part of the SRS Beneficial Reuse Program, stainless steel radioactive scrap metal is melted, pour into ingots, and roll into sheets. The sheets are then fabricated into boxes and barrels for beneficial reuse. The melting activity is a partial decontamination process. Certain isotopes separate from the melted steel, while others stay in solution. Cobalt-60 is the primary constituent, which remains in solution, and becomes the major contributor to the volumetric radioactivity of the finished products (boxes and barrels). There is currently no ``de minimis`` free release level for volumetrically radioactive material. However, under certain circumstances, pathway analysis can be used (and have been used) to free release volumetrically radioactive material. This paper presents an analysis using empirical data derived from over sixty ``melts``, to demonstrate that the implied surface radioactivity for specific beneficial reuse products is within free release limit. The approach can be applied to other recycled metal products.
Date: July 8, 1998
Creator: Boettinger, W.L.
Partner: UNT Libraries Government Documents Department

Advanced Materials Laboratory hazards assessment document

Description: The Department of Energy Order 55OO.3A requires facility-specific hazards assessments be prepared, maintained, and used for emergency planning purposes. This hazards assessment document describes the chemical and radiological hazards associated with the AML. The entire inventory was screened according to the potential airborne impact to onsite and offsite individuals. The air dispersion model, ALOHA, estimated pollutant concentrations downwind from the source of a release, taking into consideration the toxicological and physical characteristics of the release site, the atmospheric conditions, and the circumstances of the release. The greatest distance at which a postulated facility event will produce consequences exceeding the Early Severe Health Effects threshold is 23 meters. The highest emergency classification is a General Emergency. The Emergency Planning Zone is a nominal area that conforms to DOE boundaries and physical/jurisdictional boundaries such as fence lines and streets.
Date: October 1, 1995
Creator: Barnett, B. & Banda, Z.
Partner: UNT Libraries Government Documents Department

Summary of Surface Swipe Sampling for Beryllium on Lead Bricks and Shielding

Description: Approximately 25,000 lbs of lead bricks at Site 300 were assessed by the Site 300 Industrial Hygienis tand Health Physicist for potential contamination of beryllium and radiation for reuse. These lead bricks and shielding had been used as shielding material during explosives tests that included beryllium and depleted uranium. Based on surface swipe sampling that was performed between July 26 and October 11, 2010, specifically for beryllium, the use of a spray encapsulant was found to be an effective means to limit removable surface contamination to levels below the DOE release limit for beryllium, which is 0.2 mcg/100 cm{sup 2}. All the surface swipe sampling data for beryllium and a timeline of when the samples were collected (and a brief description) are presented in this report. On December 15, 2010, the lead bricks and shielding were surveyed with an ion chamber and indicated dose rates less than 0.05 mrem per hour on contact. This represents a dose rate consistent with natural background. An additional suevey was performed on February 8, 2011, using a GM survey instrument to estimate total activity on the lead bricks and shielding, confirming safe levels of radioactivity. The vendor is licensed to possess and work with radioactive material.
Date: August 3, 2011
Creator: Paik, S Y & Barron, D A
Partner: UNT Libraries Government Documents Department

Radiological Dose Calculations for Fusion Facilities

Description: This report summarizes the results and rationale for radiological dose calculations for the maximally exposed individual during fusion accident conditions. Early doses per unit activity (Sieverts per TeraBecquerel) are given for 535 magnetic fusion isotopes of interest for several release scenarios. These data can be used for accident assessment calculations to determine if the accident consequences exceed Nuclear Regulatory Commission and Department of Energy evaluation guides. A generalized yearly dose estimate for routine releases, based on 1 Terabecquerel unit releases per radionuclide, has also been performed using averaged site parameters and assumed populations. These routine release data are useful for assessing designs against US Environmental Protection Agency yearly release limits.
Date: April 1, 2003
Creator: Abbott, Michael L.; Cadwallader, Lee C. & Petti, David A.
Partner: UNT Libraries Government Documents Department

Closure Report for Corrective Action Unit 116: Area 25 Test Cell C Facility, Nevada National Security Site, Nevada

Description: This Closure Report (CR) presents information supporting closure of Corrective Action Unit (CAU) 116, Area 25 Test Cell C Facility. This CR complies with the requirements of the Federal Facility Agreement and Consent Order (FFACO) that was agreed to by the State of Nevada; the U.S. Department of Energy (DOE), Environmental Management; the U.S. Department of Defense; and DOE, Legacy Management (FFACO, 1996 [as amended March 2010]). CAU 116 consists of the following two Corrective Action Sites (CASs), located in Area 25 of the Nevada National Security Site: (1) CAS 25-23-20, Nuclear Furnace Piping and (2) CAS 25-41-05, Test Cell C Facility. CAS 25-41-05 consisted of Building 3210 and the attached concrete shield wall. CAS 25-23-20 consisted of the nuclear furnace piping and tanks. Closure activities began in January 2007 and were completed in August 2011. Activities were conducted according to Revision 1 of the Streamlined Approach for Environmental Restoration Plan for CAU 116 (U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office [NNSA/NSO], 2008). This CR provides documentation supporting the completed corrective actions and provides data confirming that closure objectives for CAU 116 were met. Site characterization data and process knowledge indicated that surface areas were radiologically contaminated above release limits and that regulated and/or hazardous wastes were present in the facility.
Date: September 29, 2011
Creator: National Security Technologies, LLC
Partner: UNT Libraries Government Documents Department

Lessons Learned from Characterization, Performance Assessment, and EPA Regulatory Review of the 1996 Actinide Source Term for the Waste Isolation Pilot Plant

Description: The Waste Isolation Pilot Plant (WIPP) is a US Department of Energy (DOE) facility for the permanent disposal of transuranic waste from defense activities. In 1996, the DOE submitted the Title 40 CFR Part 191 Compliance Certification Application for the Waste Isolation Pilot Plant (CCA) to the US Environmental Protection Agency (EPA). The CCA included a probabilistic performance assessment (PA) conducted by Sandia National Laboratories to establish compliance with the quantitative release limits defined in 40 CFR 191.13. An experimental program to collect data relevant to the actinide source term began around 1989, which eventually supported the 1996 CCA PA actinide source term model. The actinide source term provided an estimate of mobile dissolved and colloidal Pu, Am, U, Th, and Np concentrations in their stable oxidation states, and accounted for effects of uncertainty in the chemistry of brines in waste disposal areas. The experimental program and the actinide source term included in the CCA PA underwent EPA review lasting more than 1 year. Experiments were initially conducted to develop data relevant to the wide range of potential future conditions in waste disposal areas. Interim, preliminary performance assessments and actinide source term models provided insight allowing refinement of experiments and models. Expert peer review provided additional feedback and confidence in the evolving experimental program. By 1995, the chemical database and PA predictions of WIPP performance were considered reliable enough to support the decision to add an MgO backfill to waste rooms to control chemical conditions and reduce uncertainty in actinide concentrations, especially for Pu and Am. Important lessons learned through the characterization, PA modeling, and regulatory review of the actinide source term are (1) experimental characterization and PA should evolve together, with neither activity completely dominating the other, (2) the understanding of physical processes required to develop conceptual models is ...
Date: March 22, 1999
Creator: Larson, K.W.; Moore, R.C.; Nowak, E.J.; Papenguth, H.W. & Jow, H.
Partner: UNT Libraries Government Documents Department

Cold Vacuum Drying facility design basis accident analysis documentation

Description: This document provides the detailed accident analysis to support HNF-3553, Annex B, Spent Nuclear Fuel Project Final Safety Analysis Report (FSAR), ''Cold Vacuum Drying Facility Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the FSAR. The calculations in this document address the design basis accidents (DBAs) selected for analysis in HNF-3553, ''Spent Nuclear Fuel Project Final Safety Analysis Report'', Annex B, ''Cold Vacuum Drying Facility Final Safety Analysis Report.'' The objective is to determine the quantity of radioactive particulate available for release at any point during processing at the Cold Vacuum Drying Facility (CVDF) and to use that quantity to determine the amount of radioactive material released during the DBAs. The radioactive material released is used to determine dose consequences to receptors at four locations, and the dose consequences are compared with the appropriate evaluation guidelines and release limits to ascertain the need for preventive and mitigative controls.
Date: August 8, 2000
Creator: CROWE, R.D.
Partner: UNT Libraries Government Documents Department

Release process for non-real property containing residual radioactive material

Description: It is DOE`s objective to operate its facilities and to conduct its activities so that radiation exposures to members of the public are maintained within acceptable limits and exposures to residual radioactive materials are controlled. To accomplish this, DOE has adopted Order DOE 5400.51 `Radiation Protection of the Public and the Environment`, and will be promulgating IO CR Part 834 to codify and clarify the requirements of DOE 5400.5. Under both DOE 5400.5 and 10 CR Part 834, radioactively contaminated DOE property is prohibited from release unless specific actions have been completed prior to the release. This paper outlines a ten-step process that, if followed, will assist DOE Operations and contractor personnel in ensuring that the required actions established by Order DOE 5400.5 and 10 CR Part 834 have been appropriately completed prior to the release for reuse or recycle of non-real property (e.g., office furniture, computers, hand tools, machinery, vehicles and scrap metal). Following the process will assist in ensuring that radiological doses to the public from the released materials will meet applicable regulatory standards and be as low as reasonably achievable (ALARA).
Date: February 1, 1997
Creator: Ranek, N.L.; Chen, S.Y.; Kamboj, S.; Hensley, J.; Burns, D.; Fleming, R. et al.
Partner: UNT Libraries Government Documents Department

Predicted radionuclide release from reactor-related unenclosed solid objects dumped in the Sea of Japan and the Pacific Ocean, east coast of Kamchatka

Description: Between 1978 and 1991 reactor-related solid radioactive waste was dumped by the former Soviet Union as unenclosed objects in the Pacific Ocean, east coast of Kamchatka, and the Sea of Japan. This paper presented estimates for the current (1994) inventory of activation and corrosion products contained in the reactor-related unenclosed solid objects. In addition, simple models derived for prediction of radionuclide release from marine reactors dumped in the Kara Sea are applied to certain of the dumped objects to provide estimates of radionuclide release to the Pacific Ocean, east coast of Kamchatka, and Sea of Japan environments. For the Pacific Ocean, east coast of Kamchatka, total release rates start below 0.01 GBq yr{sup -1} and over 1,000 years, fall to 100 Bq yr{sup -1}. In the Sea of Japan, the total release rate starts just above 1 GBq yr{sup - 1}, dropping off to a level less than 0.1 GBq yr{sup -1}, extending past the year 4,000.
Date: June 1, 1996
Creator: Mount, M.E.; Lynn, N.M. & Warden, J.M.
Partner: UNT Libraries Government Documents Department

Preliminary evaluation of predicted peak release rates from the engineered barrier system for a potential repository at Yucca Mountain, Nevada

Description: Any potential repository for the ultimate disposal of the nation`s high-level radioactive wastes is subject to meeting post-closure regulatory requirements as specified by the NRC. Three NRC sub-system performance measures are relevant to the evaluation of the Yucca Mountain site and possible engineered barriers. These performance requirements are specified in 10 CFR 60. These include the substantially complete containment requirement, the engineered barrier system (EBS) release requirement, and the pre-waste emplacement groundwater travel time requirement. The present paper documents an initial evaluation of the peak EBS release rates. A number of key factors significantly impact the maximum release rate from the engineered barrier system. The authors have conducted four simulations to approximate the effects of delaying and spreading out the failure distribution that are based on different thermal loads and criteria for the initiation of aqueous corrosion. Using an assumed outer barrier of 10 cm and an inner barrier of 0.95 cm and the Stahl model for aqueous pitting corrosion, they have analyzed the EBS release rates for thermal loads of 28.5, 57 and 83 kW/Ac using temperature as the corrosion limiting factor and at 57 kW/Ac for saturation limiting the initiation of corrosion. The later had the earliest failures and the most rapid failure rates observed in the TSPA-1993 analyses so provides the upper bound on the release rates.
Date: December 31, 1995
Creator: Andrews, R.W.; McNeish, J.A. & Lee, J.H.
Partner: UNT Libraries Government Documents Department

WIPP Case Study - Compliance Monitoring, Passive Institutional Controls, and Record Keeping

Description: The WIPP Case Study describes the compliance monitoring program, record keeping requirements, and passive institutional controls that are used to help ensure the Waste Isolation Pilot Plant (WIPP) will safety contain radioactive waste and indicate dangers and location of the wastes. The radioactive components in the waste are regulated by the U.S. Environmental Protection Agency (EPA) while the hazardous components in the waste are regulated by the New Mexico Environment Department (NMED). This paper addresses monitoring relating to radionuclide containment performance, passive institutional controls, and record keeping over a 10,000-year time frame. Monitoring relating to the hazardous components and the associated regulator are not addressed in this paper. The WIPP containment performance is mandated by release limits set by regulation. Regulations also require the radioactive waste containment performance of the WIPP to be predicted by a ''Performance Assessment.'' The EPA did not base the acceptance of the WIPP solely on predicted containment but included additional assurance measures. One such assurance measure is monitoring, which may be defined as the on-going measurement of conditions in and around the repository. This case study describes the evolution of the WIPP monitoring program as the WIPP project progressed through the planning, site characterization, regulatory promulgation, and eventual operational stages that spanned a period of over 25 years. Included are discussions of the regulatory requirements for monitoring, selection of monitoring parameters, trigger values used to identify unexpected conditions, assessment of monitoring data against the trigger values, and plans for post-closure monitoring. The United EPA established the requirements for Passive Institutional Controls (PICs) for disposal sites. The requirements state the a disposal site must be designated by the most permanent markers, records, and other passive institutional controls practicable to indicate the dangers of the wastes and their location. The PIC Task Force assessed the effectiveness of ...
Date: July 1, 2002
Creator: Wagner, Stephen W.; Beauheim, Richard L.; Pfeifle, Tom W.; Bethel, Amy; Sosa-Yates, Grace Ann; Williams, Cecelia V. et al.
Partner: UNT Libraries Government Documents Department

Authorized Limits for the Release of a 25 Ton Locomotive, Serial Number 21547, at the Area 25 Engine Maintenance, Assembly, and Disassembly Facility, Nevada Test Site, Nevada

Description: This document contains process knowledge and radiological data and analysis to support approval for release of the 25-ton locomotive, Serial Number 21547, at the Area 25 Engine Maintenance, Assembly, and Disassembly (EMAD) Facility, located on the Nevada Test Site (NTS). The 25-ton locomotive is a small, one-of-a-kind locomotive used to move railcars in support of the Nuclear Engine for Rocket Vehicle Application project. This locomotive was identified as having significant historical value by the Nevada State Railroad Museum in Boulder City, Nevada, where it will be used as a display piece. A substantial effort to characterize the radiological conditions of the locomotive was undertaken by the NTS Management and Operations Contractor, National Security Technologies, LLC (NSTec). During this characterization process, seven small areas on the locomotive had contamination levels that exceeded the NTS release criteria (limits consistent with U.S. Department of Energy [DOE] Order DOE O 5400.5, “Radiation Protection of the Public and the Environment”). The decision was made to perform radiological decontamination of these known accessible impacted areas to further the release process. On February 9, 2010, NSTec personnel completed decontamination of these seven areas to within the NTS release criteria. Although all accessible areas of the locomotive had been successfully decontaminated to within NTS release criteria, it was plausible that inaccessible areas of the locomotive (i.e., those areas on the locomotive where it was not possible to perform radiological surveys) could potentially have contamination above unrestricted release limits. To access the majority of these inaccessible areas, the locomotive would have to be disassembled. A complete disassembly for a full radiological survey could have permanently destroyed parts and would have ruined the historical value of the locomotive. Complete disassembly would also add an unreasonable financial burden for the contractor. A decision was reached between the NTS regulator and NSTec, ...
Date: April 8, 2010
Creator: Frenette, Jeremy Gwin and Douglas
Partner: UNT Libraries Government Documents Department

Verification Survey of Rooms 113, 114, and 208 of the Inhalation Toxicology Laboratory, Lovelace Respiratory Research Institute, Albuquerque, NM

Description: The objectives of the verification survey were to confirm that accessible surfaces of the three laboratories meet the DOE’s established criteria for residual contamination. Drain pipes and ductwork were not included within the survey scope.
Date: June 25, 2008
Creator: Vitkus, T. J.
Partner: UNT Libraries Government Documents Department

Supplemental Release Limits for the Directed Reuse of Steel in Road Barriers and Lead in Shielding Products by the Department of Energy

Description: The DOE National Center of Excellence for Metals Recycle (NMR) proposes to define and implement a complex-wide directed reuse strategy for surplus radiologically impacted lead (Pb) and steel as part of the U.S. Department of Energy's commitment to the safe and cost-effective recycle or reuse of excess materials and equipment across the DOE complex. NMR will, under this proposal, act on behalf of the DOE Office of Environmental Management, Office of Technical Program Integration (specifically EM-22), as the Department's clearinghouse for DOE surplus lead, steel and products created from these materials by developing and maintaining a cost-effective commercially-based contaminated lead and steel recycle program. It is NMR's intention, through this directed reuse strategy, to mitigate the adverse environmental and economic consequences of managing surplus lead and steel as a waste within the complex. This approach promotes the safe and cost-effective reuse of scrap metals in support of the Department's goals of resource utilization, energy conservation, pollution prevention and waste minimization. This report discusses recommendations for supplemental radiological release limits for the directed reuse of contaminated lead and steel by the DOE within the nuclear industry. The limits were originally selected from the American National Standards Institute and Health Physics Society standard N13.12 titled ''Surface and Volume Radioactivity Standards for Clearance'' (Health Physics Society, 1999) but were subsequently modified as a result of application-specific issues. Both the health and measurement implications from the adoption and use of the limits for directed reuse scenarios are discussed within this report.
Date: April 7, 2006
Creator: Coleman, RL
Partner: UNT Libraries Government Documents Department

Supplemental Release Limits for the Directed Reuse of Lead in Shielding Products by the Department of Energy

Description: The DOE National Center of Excellence for Metals Recycle (NMR) proposes to define and implement a complex-wide directed reuse strategy for surplus radiologically impacted lead (Pb) as part of the U.S. Department of Energy's commitment to the safe and cost-effective recycle or reuse of excess materials and equipment across the DOE complex. NMR will, under this proposal, act on behalf of the DOE Office of Environmental Management, Office of Technical Program Integration (specifically EM-22), as the Department's clearinghouse for DOE surplus lead and lead products by developing and maintaining a cost-effective commercially-based contaminated lead recycle program. It is NMR's intention, through this directed reuse strategy, to mitigate the adverse environmental and economic consequences of managing surplus lead as a waste within the complex. This approach would promote the safe and cost-effective reuse of DOE's scrap and surplus lead in support of the Department's goals of resource utilization, energy conservation, pollution prevention and waste minimization. This report discusses recommendations for supplemental radiological limits for the directed reuse of contaminated lead and lead products by the DOE within the nuclear industry. The limits were selected--with slight modification--from the recently published American National Standards Institute and Health Physics Society standard N13.12 titled Surface and Volume Radioactivity Standards for Clearance (ANSI/HPS 1999) and are being submitted for formal approval by the DOE. Health and measurement implications from the adoption and use of the limits for directed reuse scenarios are discussed within this report.
Date: August 22, 2001
Creator: Coleman, R.L.
Partner: UNT Libraries Government Documents Department

Releases from exotic waste packages from partitioning and transmutation

Description: Partitioning the actinides in spent nuclear fuel and transmuting them in actinide-burning liquid-metal reactors has been proposed as a potential method of reducing the public risks from geologic disposal of nuclear waste. To quantify the benefits for waste disposal of actinide burning, we calculate the release rates of key radionuclides from waste packages resulting from actinide burning, and compare them with release rates from LWR spent fuel destined for disposal at the potential repository at Yucca Mountain. The wet-drip water-contact mode has been used. Analytic methods and parameter values are very similar to those used for assessing Yucca Mountain as a potential repository. Once released, the transport characteristics of radionuclides will be largely determined by site geology. For the most important nuclides such as I-129 and {Tc}-99, which are undiminished by actinide-burning reactors, it is not surprising that actinide burning offers little reduction in releases. For important actinides such as Np-237 and Pu isotopes, which are reduced in inventory, the releases are not reduced because the release rates are proportional to solubility, rather than inventory.
Date: September 1, 1991
Creator: Lee, W.W.L. & Choi, J.S.
Partner: UNT Libraries Government Documents Department

P2Pro(RSM) : a computerized management tool for implementing DOE's authorized release process for radioactive scrap metals.

Description: Within the next few decades, several hundred thousand tons of metal and several million cubic meters of concrete are expected to be removed from nuclear facilities across the US Department of Energy (DOE) complex as a result of decontamination and decommissioning (D&D) activities. These materials, together with large quantities of tools, equipment, and other items that are commonly recovered from site cleanup or D&D activities, constitute non-real properties that warrant consideration for release from regulatory control for reuse or recycle, as permitted and practiced under current DOE policy. The provisions for implementing this policy are contained in the Draft Handbook for Controlling Release for Reuse or Recycle of Non-Real Property Containing Residual Radioactive Material published by DOE in 1997 and distributed to DOE Field Offices for interim use and implementation. This manual describes a computer management tool, P2Pro(RSM), that implements the first 5 steps of the 10-step process stipulated by the Handbook. P2Pro(RSM) combines an easy-to-use Windows interface with a comprehensive database to facilitate the development of authorized release limits for non-real property.
Date: July 22, 1999
Creator: Arnish, J.; Chen, S. Y.; Kamboj, S. & Nieves, L.
Partner: UNT Libraries Government Documents Department

Gaseous release of carbon-14: Why the high level waste regulations should be changed

Description: The high-level nuclear waste regulations pertaining to gaseous release of carbon-14 from a repository should be changed to allow greater release, for several reasons. Some of them are as follows. First, the total amount of carbon-14 that would be placed in a repository is small compared to that produced naturally in the atmosphere by cosmic rays. Second, the dose that would result to an individual from total release of repository carbon-14 would be very small compared to that from natural radiation sources and would be well below the ``Below Regulatory Concern`` criterion. Third, the limits on gaseous carbon-14 release from a repository have been set unreasonably low compared to the limits set for carbon-14 release from other fuel cycle facilities. Fourth, the additional cost for waste packages to attempt to meet the regulations for carbon-14 release would likely be of the order of a billion dollars or more, too high to be justified by the small reduction in dose that might result. 32 refs.
Date: April 1, 1991
Creator: Van Konynenburg, R.A.
Partner: UNT Libraries Government Documents Department

Approach toward development of release standards for D and D cleanup.

Description: The release of materials containing residual radioactivity from a controlled environment in decontamination and decommissioning (D&D) activities has been problematic. The primary impediment to such a release is the lack of a suitable framework within which release standards can be developed. The concept of clearance for radioactive materials was recently introduced by the International Atomic Energy Agency (IAEA) (l). This concept is being evaluated by the international regulatory communities as a basis for setting standards for releasing from control solid materials containing residual radioactivity. Accordingly, both the IAEA (2) and the European Commission (EC) (3) have recently proposed clearance standards. In the US, the Nuclear Regulatory Commission (USNRC) has just begun its rule-making process on clearance. The term ''clearance'' was introduced as a regulatory process for releasing radioactive materials posing negligible risks. A trivial risk level has been determined to be a 10{sup {minus}6} to 10{sup {minus}7} annual risk to an exposed individual, and a population risk of no more than 0.1 for an annual practice. Under these strict constraints, exposure scenarios would be developed to estimate potential doses to affected individuals. Such scenarios may account for processing, disposal, and product end-use of materials. This paper discusses these scenarios and also describes the technical basis for deriving release levels under the suggested risk for dose constraints.
Date: January 27, 1999
Creator: Chen, S. Y.
Partner: UNT Libraries Government Documents Department

Recycling Of Uranium- And Plutonium-Contaminated Metals From Decommissioning Of The Hanau Fuel Fabrication Plant

Description: Decommissioning of a nuclear facility comprises not only actual dismantling but also, above all, management of the resulting residual materials and waste. Siemens Decommissioning Projects (DP) in Hanau has been involved in this task since 1995 when the decision was taken to decommission and dismantle the Hanau Fuel Fabrication Plant. Due to the decommissioning, large amounts of contaminated steel scrap have to be managed. The contamination of this metal scrap can be found almost exclusively in the form of surface contamination. Various decontamination technologies are involved, as there are blasting and wiping. Often these methods are not sufficient to meet the free release limits. In these cases, SIEMENS has decided to melt the scrap at Siempelkamp's melting plant. The plant is licensed according to the German Radiation Protection Ordinance Section 7 (issue of 20.07.2001). The furnace is a medium frequency induction type with a load capacity of 3.2 t and a throughput of 2 t/h for steel melting. For safety reasons, the furnace is widely operated by remote handling. A highly efficient filter system of cyclone, bag filter and HEPA-filter in two lines retains the dust and aerosol activity from the off-gas system. The slag is solidified at the surface of the melt and gripped before pouring the liquid iron into a chill. Since 1989, in total 15,000 t have been molten in the plant, 2,000 t of them having been contaminated steel scrap from the decommissioning of fuel fabrication plants. Decontamination factors could be achieved between 80 and 100 by the high affinity of the uranium to the slag former. The activity is transferred to the slag up to nearly 100 %. Samples taken from metal, slag and dust are analyzed by gamma measurements of the 186 keV line of U235 and the 1001 keV line of Pa234m for ...
Date: February 26, 2003
Creator: Kluth, T.; Quade, U. & Lederbrink, F. W.
Partner: UNT Libraries Government Documents Department

ASTM STANDARD GUIDE FOR EVALUATING DISPOSAL OPTIONS FOR REUSE OF CONCRETE FROM NUCLEAR FACILITY DECOMMISSIONING

Description: Within the nuclear industry, many contaminated facilities that require decommissioning contain huge volumes of concrete. This concrete is generally disposed of as low-level waste at a high cost. Much of the concrete is lightly contaminated and could be reused as roadbed, fill material, or aggregate for new concrete, thus saving millions of dollars. However, because of the possibility of volumetric contamination and the lack of a method to evaluate the risks and costs of reusing concrete, reuse is rarely considered. To address this problem, Argonne National Laboratory-East (ANL-E) and the Idaho National Engineering and Environmental Laboratory teamed to write a ''concrete protocol'' to help evaluate the ramifications of reusing concrete within the U.S. Department of Energy (DOE). This document, titled the Protocol for Development of Authorized Release Limits for Concrete at U.S. Department of Energy Site (1) is based on ANL-E's previously developed scrap metal recycle protocols; on the 10-step method outlined in DOE's draft handbook, Controlling Release for Reuse or Recycle of Property Containing Residual Radioactive Material (2); and on DOE Order 4500.5, Radiation Protection of the Public and the Environment (3). The DOE concrete protocol was the basis for the ASTM Standard Guide for Evaluating Disposal Options for Concrete from Nuclear Facility Decommissioning, which was written to make the information available to a wider audience outside DOE. The resulting ASTM Standard Guide is a more concise version that can be used by the nuclear industry worldwide to evaluate the risks and costs of reusing concrete from nuclear facility decommissioning. The bulk of the ASTM Standard Guide focuses on evaluating the dose and cost for each disposal option. The user calculates these from the detailed formulas and tabulated data provided, then compares the dose and cost for each disposal option to select the best option that meets regulatory requirements. ...
Date: February 27, 2003
Creator: Phillips, Ann Marie & Meservey, Richard H.
Partner: UNT Libraries Government Documents Department

ASSET RECOVERY OF HAZARDOUS MATERIALS BENEFICIAL REUSE OF RADIOLOGICALLY ENCUMBERED LEAD STOCKS

Description: Underutilized and surplus lead stocks and leaded components are a common legacy environmental problem across much of the Department of Energy (DOE) Complex. While seeking to dispose of these items through its Environmental Management Program, DOE operational programs continue to pursue contemporary mission requirements such as managing and/or storing radioactive isotopes that require lead materials for shielding. This paradox was identified in late 1999 when DOE's policies for managing scrap metal were assessed. In January 2000, the Secretary of Energy directed the National Center of Excellence for Materials Recycle (NMR) to develop and implement a comprehensive lead reuse program for all of DOE. Fluor Hanford, contractor for DOE Richland Operations, subsequently contacted NMR to pilot lead reclamation and reuse at the Hanford Site. This relationship resulted in the development of a beneficial reuse pathway for lead reclaimed from spent fuel transport railcars being stored at Hanford. The 1.3 million pounds of lead in the railcars is considered radiologically encumbered due to its prior use. Further, the material was considered a mixed Resource Conservation and Recovery Act (RCRA) low-level radioactive waste that would require expensive storage or macro encapsulation to meet land disposal restrictions prior to burial. Working closely with Flour Hanford and the Office of Air, Water, and Radiation (EH-412), NMR developed a directed reuse pathway for this and other radiologically encumbered lead. When derived supplemental release limits were used, the lead recovered from these railcars became eligible for reuse in shielding products to support DOE and commercial nuclear industry operations. Using this disposition pathway has saved Hanford one third of the cost of disposing of the lead and the cost of acquiring additional lead for nuclear shielding applications. Furthermore, the environmental costs associated with mining and producing new lead for shielding products a nd stewardship of the waste was ...
Date: February 27, 2003
Creator: Lloyd, E.R. & Meehan, R.W.
Partner: UNT Libraries Government Documents Department