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At least 379 records · Page 21
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Analytical results from routine DSSHT and SEHT monthly samples

Strip Effluent Hold Tank (SEHT) and Decontaminated Salt Solution Hold Tank (DSSHT) samples from several of the “microbatches” of Integrated Salt Disposition Project (ISDP) Salt Batch (“Macrobatch”) 8B have been analyzed for 238 Pu, 90 Sr, 137 Cs, cations (Inductively Coupled Plasma Emission Spectroscopy - ICPES), and anions (Ion Chromatography Anions - IC-A). The analytical results from the current microbatch samples are similar to those from previous macrobatch samples. The Cs removal continues to be excellent, with decontamination factors (DF) averaging 22,100 (114% RSD). The bulk chemistry of the DSSHT and SEHT samples does not show any signs of unusual behavior, other than lacking the anticipated degree of dilution that is calculated to occur during Modular Caustic-Side Solvent Extraction Unit (MCU) processing.

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

Cesium Ion Exchange with Crystalline Silicotitanate Literature Review

Crystalline silicotinante (CST) is planned for use as the ion exchange media to accomplish the cesium (Cs) decontamination of low-activity waste (LAW) in the current evolution of the Direct Feed Low-Activity Waste flowsheet. CST has yet to be deployed in a production-scale operating waste treatment facility on the Hanford Site. However, CST was used nearly 20 years ago to successfully treat approximately 270,000 gallons of supernatant waste in modular columns at Oak Ridge National Laboratory. As part of the technical maturation of the media for use at Hanford, an assessment of the state-of-the-art was performed by reviewing work in the technical literature. The review considered work from the U.S. Department of Energy (DOE) complex (i.e., national laboratories and government documents), the CST manufacturer, and the open technical literature nationally and internationally. The review focused on information most relevant to the caustic, high-sodium chemical environment typical of Hanford LAW. The topic areas explored in this literature review document include the following: CST structural and physical properties; Isotherm and batch contact data; Mass transport: intraparticle, interparticle, dispersion in the column; Column performance at multiple scales; Hydraulic performance; The effect of process conditions; Gas generation and retention behavior; Post-processing options for spent CST media The reviewed literature demonstrates that CST is well-suited for Cs decontamination of Hanford LAW. However, after consideration of the available literature data in these subject areas, some uncertainties regarding the application of CST for the treatment of Hanford LAW are identified and summarized. These uncertainties are areas where additional information may permit better prediction of CST performance or provide margin to facility operations and safety basis analyses.

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

Next Generation Solvent Performance in the Modular Caustic Side Solvent Extraction Process - 15495

Changes to the Modular Caustic Side Solvent Extraction Unit (MCU) flow-sheet were implemented in the facility. Implementation included changing the scrub and strip chemicals and concentrations, modifying the O/A ratios for the strip, scrub, and extraction contactor banks, and blending the current BoBCalixC6 extractant-based solvent in MCU with clean MaxCalix extractant-based solvent. During the successful demonstration period, the MCU process was subject to rigorous oversight to ensure hydraulic stability and chemical/radionuclide analysis of the key process tanks (caustic wash tank, solvent hold tank, strip effluent hold tank, and decontaminated salt solution hold tank) to evaluate solvent carryover to downstream facilities and the effectiveness of cesium removal from the liquid salt waste. Results indicated the extraction of cesium was significantly more effective with an average Decontamination Factor (DF) of 1,129 (range was 107 to 1,824) and that stripping was effective. The contactor hydraulic performance was stable and satisfactory, as indicated by contactor vibration, contactor rotational speed, and flow stability; all of which remained at or near target values. Furthermore, the Solvent Hold Tank (SHT) level and specific gravity was as expected, indicating that solvent integrity and organic hydraulic stability were maintained. The coalescer performances were in the range of processing results under the BOBCalixC6 flow sheet, indicating negligible adverse impact of NGS deployment. After the Demonstration period, MCU began processing via routine operations. Results to date reiterate the enhanced cesium extraction and stripping capability of the Next Generation Solvent (NGS) flow sheet. This paper presents process performance results of the NGS Demonstration and continued operations of MCU utilizing the blended BobCalixC6-MaxCalix solvent under the NGS flowsheet.

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

EPA and Idaho National Laboratory create first of its scale

Advancing the science and engineering of decontaminating water distribution systems and safely disposing of high-volumes of contaminated water are high priorities for the EPA. To help address these science gaps, EPA homeland security researchers have developed the first-of-its-scale Water Security Test Bed (WSTB). Water is an important component of everyday life. We rely on safe, clean water for our health and the health of the environment. Protecting this vital resource is critical to living safely and securely. The U.S. Environmental Protection Agency is responsible for working with water utilities to protect water systems from contamination and to clean up systems that become contaminated. Purposeful or inadvertent distribution system contamination can result in large amounts of infrastructure and water that must be cleaned. Water distribution systems contamination can be caused by, for example, natural disasters such as Superstorm Sandy or by individuals hoping to cause harm. The WSTB, constructed at the Department of Energy’s (DOE) Idaho National Laboratory, replicates a section of a typical municipal drinking water system including a piping system with roughly 445 feet of pipe and two fire hydrants laid out in an “L” shape using 40-year-old, eight-inch cement mortar lined, ductile iron pipes, excavated after twenty years of use for water conveyance. Researchers built the WSTB above ground for easy access during experiments, and to facilitate fast leak detection. Over the next several years, EPA and partner researchers will conduct experiments using simulants of various biological, chemical, and radioactive materials that simulate high toxic versions of these agents. At this “full” sized system, researchers will demonstrate approaches to contamination detection, sensor and model testing, infrastructure decontamination, water treatment and cyber testing developed at lab and pilot scale. Researchers are also able to verify if laboratory results that were effective in controlled conditions are as effective in full-scale testing. Having the ability to conduct bench scale studies in a real world setting can help address specific research gaps, and help water researchers understand conditions in which real world drinking water systems operate.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2013

Treatment of Irradiated Graphite from French Bugey Reactor - 13424

Beginning in 2009, in order to determine an alternative to direct disposal for decommissioned irradiated graphite from EDF's Bugey NPP, Studsvik and EDF began a test program to determine if graphite decontamination and destruction were practicable using Studsvik's thermal organic reduction (THOR) technology. The testing program focused primarily on the release of C-14, H-3, and Cl-36 and also monitored graphite mass loss. For said testing, a bench-scale steam reformer (BSSR) was constructed with the capability of flowing various compositions of gases at temperatures up to 1300 deg. C over uniformly sized particles of graphite for fixed amounts of time. The BSSR was followed by a condenser, thermal oxidizer, and NaOH bubbler system designed to capture H-3 and C-14. Also, in a separate series of testing, high concentration acid and peroxide solutions were used to soak the graphite and leach out and measure Cl-36. A series of gasification tests were performed to scope gas compositions and temperatures for graphite gasification using steam and oxygen. Results suggested higher temperature steam (1100 deg. C vs. 900 deg. C) yielded a practicable gasification rate but that lower temperature (900 deg. C) gasification was also a practicable treatment alternative if oxygen is fed into the process. A series of decontamination tests were performed to determine the release behavior of and extent to which C-14 and H-3 were released from graphite in a high temperature (900-1300 deg. C), low flow roasting gas environment. In general, testing determined that higher temperatures and longer roasting times were efficacious for releasing H-3 completely and the majority (80%) of C-14. Manipulating oxidizing and reducing gas environments was also found to limit graphite mass loss. A series of soaking tests was performed to measure the amount of Cl-36 in the samples of graphite before and after roasting in the BSSR. Similar to C-14 release, these soaking tests revealed that 70-80% Cl-36 is released during roasting tests. (authors)

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

New Remote Method for Estimation of Contamination Levels of Reactor Equipment - 13175

Projects for decommissioning of shutdown reactors and reactor facilities carried out in several countries, including Russia. In the National Research Centre 'Kurchatov Institute' decontamination and decommissioning of the research reactor MR (Material Testing Reactor) has been initiated. The research reactor MR has a long history and consists of nine loop facilities for experiments with different kinds of fuel. During the operation of main and auxiliary equipment of reactors it was subjected to strong radioactive contamination. The character of this contamination requires individual strategies for the decontamination work. This requires information about the character of the distribution of radioactive contamination of equipment in the premises. A detailed radiation survey of these premises using standard dosimetric equipment is almost impossible because of high levels of radiation and high-density of the equipment that does not allow identifying the most active fragments using standard tools of measurement. The problem can be solved using the method of remote measurements of distribution of radioactivity with help of the collimated gamma-ray detectors. For radiation surveys of the premises of loop installations remotely operated spectrometric collimated system was used [1, 2, 3]. As a result of the work, maps of the distribution of activity and dose rate for surveyed premises were plotted and superimposed on its photo. The new results of measurements in different areas of the reactor and at its loop installations, with emphasis on the radioactive survey of highly-contaminated samples, are presented. (authors)

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

Oak Ridge Cleanup Vision: Moving to the Future by Cleaning Up the Past - 13291

The Oak Ridge Office of Environmental Management (EM) strives to be the leader in the Department of Energy's (DOE's) EM Complex regarding successful and safe project execution and stakeholder interactions that yield positive results. EM's goal has been to become 'Investment Worthy' and, in order to accomplish that important objective, has also had to improve communications both within and outside of the Department. One of our most important missions is to assist the Department in achieving the sustainability goals set forth in Executive Order 13514. In this regard, EM's primary role is to return land to beneficial use and reduce energy impacts and maintenance costs by demolishing unneeded and deteriorating structures and remediating environmental contamination. Recent accomplishments toward meeting these goals include significant progress in the decontamination and demolition of the country's largest facility, the former K-25 Gaseous Diffusion Building, constructed in 1942 to enrich uranium to help end World War II; the disposition of the first phase of Uranium-233 material from the Oak Ridge National Laboratory (ORNL) which involved the transfer of Zero Power Reactor Plates to the National Nuclear Security Administration (NNSA); and a host of other project successes associated with transuranic (TRU) waste processing, hot cell decontamination and demolition, remediation of highly contaminated soils and burial grounds, and removal of mercury from storm sewers and surface waters. With regard to successful stakeholder interactions, recent accomplishments include a new method for collaboration that has renewed EM's working relationship with the regulators, and success in completing an extensive consultation process with over a dozen parties on the historic preservation of the former Oak Ridge Gaseous Diffusion Plant, which is now called the East Tennessee Technology Park (ETTP). Regarding improved communications, EM has successfully revised Program priorities and has received buy-in from the leadership in Headquarters, the regulators, and the community. Issues EM was facing in 2009 are presented. Resulting lessons learned and subsequent changes that the Office has gone through in the past several years in order to improve performance in the safe execution of work, relationships with external stakeholders, and communications both internally and externally are discussed. Results of these efforts are provided as a summary of Program accomplishments, including a strong focus on the future. EM's motto, Moving to the Future by Cleaning up the Past, will be demonstrated through the Program's mission, which includes protecting the region's health and environment; ensuring the continuation of ongoing vital missions being conducted by DOE on the Oak Ridge Reservation; and making clean land available for future use at all three sites, with a near-term focus on Re-industrialization of ETTP. (authors)

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

Environmental remediation following the Fukushima-Daiichi accident

A wide area of Fukushima Prefecture was contaminated with radioactivity released by the Fukushima Daiichi nuclear accident. The decontamination pilot projects conducted by JAEA aimed at demonstrating the applicability of different techniques to rehabilitate affected areas. As most radioactive cesium is concentrated at the top of the soil column and strongly bound to mineral surfaces, there are 3 options left to decrease the gamma dose rate (usually measured 1 m above the ground surface): the stripping of the contaminated topsoil (i.e. direct removal of cesium), the dilution by mixing and the soil profile inversion. The last two options do not generate waste. As the half-distance of {sup 137}Cs gammas in soil is in the order of 5-6 cm (depending on density and water content), the shielding by 50 cm of uncontaminated deep soil would theoretically reduce gamma doses by about 3 orders of magnitude. Which option is employed depends basically on the Cesium concentration in the topsoil, averaged over a 15-cm thickness. The JAEA's decontamination pilot projects focus on soil profile inversion and topsoil stripping. Two different techniques have been tested for the soil profile inversion: one is the reversal tillage by which surface soil of thickness of several tens of cm is reversed by using a tractor plough and the other is the complete interchanging of contaminated topsoil with uncontaminated subsoil by using a back-hoe. Reversal tillage with a tractor plough cost about 30 yen/m{sup 2}, which is an order of magnitude lower than that of topsoil-subsoil interchange (about 300 yen/m{sup 2}). Topsoil stripping is significantly more costly (between 550 yen/m{sup 2} and 690 yen/m{sup 2} according to the equipment used)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Design and Validation Testing of TruckScan to Assay Large Sacks of Fukushima Radioactive Debris on a Truck

As a result of the March 2011 earthquake and resulting tsunami in Japan, there was a serious accident at the Fukushima Dai-ichi Nuclear Power Plant. This accident has contaminated soil and vegetation in a wide area around the plant. Decontamination projects over the last 4 years have resulted in large numbers of 1 cubic meter canvas bags of debris, commonly called Super Sacks [SS]. These are currently stored nearby where they were generated, but starting in 2015, they will be moved to various Interim Storage Facilities [ISF]. Trucks will typically carry 8-20 of these SSs. When the trucks arrive at the ISF they need to be rapidly sorted into groups according to radioactivity level, for efficient subsequent processing. Canberra Industries, Inc. [CI] has designed a new truck monitoring system 'TruckScan' for use at these ISFs. The TruckScan system must measure the entire truck loaded with multiple closely packed SSs, and generate a nuclide specific assay report showing the radioactivity in each individual SS. The Canberra-Japan office, along with Obayashi Corporation have performed validation testing to demonstrate to the regulatory authorities that the proposed technique was sufficiently accurate. These validation tests were conducted at a temporary storage area in Fukushima prefecture. Decontaminated waste of various representative types and of various levels of radioactivity was gathered and mixed to create homogeneous volumes. These volumes were sampled multiple times and assayed with laboratory HPGe detectors to determine the reference concentration of each pile. Multiple SSs were loaded from each pile. Some of the SSs were filled 50% full, others 75% full, and others 100% full, to represent the typical loading configuration of the existing SSs in the field. The content of the SSs are either sand, soil, or vegetation with densities ranging from 0.3 g/cc - 1.6 g/cc. These SSs with known concentrations of Cs-134 and Cs-137 were then loaded onto trucks in a variety of configurations, typical of how they might be on the real trucks. A partial system was installed at the site and used to assay these trucks with the various loading configurations. Whereas the full system will have 8 collimated 3 x 3'' NaI detectors, the test system only had two detectors; therefore the truck was moved and counted 4 times. The data were acquired and analyzed with the Canberra Genie software to determine the peak counts for both Cesium nuclides. That data was then analyzed with a prototype version of a Maximum Entropy algorithm, to determine the individual SS activity. The goal of the validation tests was to demonstrate that the system could detect 1000 Bq/kg in 15 seconds, and to determine how accurately it could quantify individual SSs. The validation tests demonstrated that the product would perform as predicted. The TruckScan results were consistent with the sample assay results [y = 1.0029 x, R{sup 2} = 0.914]. The Total Propagated Uncertainty, including both uncertainties from these tests and others that were estimated but not tested was 16.6% percent. (authors)

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

There Are Safe Limits for Radioactivity - Does WIPP Need to Filter its Exhaust Air? - 17165

Pilot Plant (WIPP) underground repository resulted in the release of americium and plutonium (mostly {sup 241}Am) into the environment and contaminated portions of the repository, primarily along the ventilation path from the location of the incident. The WIPP uses a ventilation system to ensure that underground air is circulated through the repository and to ensure that air quality conditions are safe for the workers. Since the air in the repository exits to the surface through an exhaust shaft, this shaft is the sole potential pathway for airborne radioactivity release from the repository. During normal operations before the 2014 incident, exhaust air was released to the environment unfiltered, with an automatic switch to filtration when airborne radioactivity was detected immediately down-stream from the active waste disposal operation. However, since the radiation release event, the WIPP ventilation system has been, and will likely remain in what is called 'filtration mode,' (i.e., exhaust air is routed through a HEPA filter system). Redirection of the ventilation system through the HEPA (High Efficiency Particulate Air) filter system was necessary at the peak of the radiation release event to protect aboveground workers at the site and the public in the surrounding areas; however, it has hampered recovery efforts and has exacerbated the inherent safety issues of working underground. As a component of the Carlsbad Environmental Monitoring and Research Center's (CEMRC) WIPP environmental monitoring program, airborne radioactivity samples are collected daily in the exhaust shaft ventilation air as it reaches the surface, and before entering the HEPA filters. Since June 2014, airborne radiation levels entering the HEPA filters have averaged about 1% of a derived air concentration (DAC) for {sup 241}Am. An airborne radioactivity level of one DAC is equivalent to a potential worker internal dose exposure of 50 mSv per work year, if a worker were continuously exposed for a full work year (2000 hours). During Project Reach and throughout decontamination campaigns conducted in Panel 7 for several months in 2015, these unfiltered levels rose to 3-5% of a DAC, with a one week spike reaching almost one DAC in June 2015, during decontamination of the most heavily contaminated exhaust drift in Panel 7. Since that period, the unfiltered levels have never exceeded 2% of a DAC. Going forward, the question is, should WIPP resume unfiltered discharge of its underground ventilation air? Evaluation of more than two years of underground ventilation sampling data indicates that residual radioactivity levels in the underground no longer warrant HEPA filtration in order to meet either worker or environmental protection criteria. In terms of radiological risk at or in the vicinity of the WIPP site, the increased risk from resuming unfiltered ventilation is exceedingly small and would not result in a hypothetical dose to a member of the public in excess of the 100 micro Sv per year limit for periodic confirmatory sampling as required for operating repositories by the Environmental Protection Agency (40 CFR 191). In addition, the requirement for respiratory protection for workers underground exposed to 1-2% of a DAC is questionable, given the added industrial safety risk of heavy equipment operation while wearing respirators. (authors)

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

Integrated Wash-Aid, Treatment, and Emergency Reuse System (IWATERS) for Strontium Contaminations - 17390

The Integrated Wash-Aid, Treatment, and Emergency Reuse System (IWATERS) is being developed to mitigate the release of radioactivity to the urban environment by disseminating, collecting, and processing decontamination wash water. Previously, IWATERS was designed for the treatment of surfaces contaminated with radioactive cesium. Experiments were performed to investigate the expansion of IWATERS for strontium contamination. Bench-scale tests were performed to investigate various counter-ions and chelators to promote ion exchange of strontium from building surfaces. Desorption results for Cs, Sr, and Eu from concrete aggregate and concrete, brick, and asphalt coupons using low- and high-pressure flow with Ca{sup 2+} and Ba{sup 2+} rich wash water are reported along with those from non-porous surfaces such as glass and vinyl siding. Several sorbents were tested for their ability to specifically separate Sr{sup 2+} from these high ionic strength wash waters. Alkali metal and alkaline earth salts were found to be effective counter-ions for decontamination of Sr{sup 2+}, which may facilitate the use of common sorbents, even if their K{sub d} values are <100 mL/g. In practice, low K{sub d} values require the use of large quantities of sorbent. Computer simulations were run using the GoldSim Contaminant Fate Module software package to optimize the efficiency of IWATERS for strontium and account for the technical limitations of the sorbents. This modeling allows greater flexibility to design the wash water recycling system for various end-user specifications including options from rain barrel designs for treating runoff from roofs and facades to larger options like drainage basins for larger-scale wash down activities. (authors)

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

Modular Caustic-Side Solvent Extraction Unit (MCU) Improved Solvent Quality and Processing Results - 17496

The Modular Caustic-Side Solvent Extraction (CSSX) Unit (MCU) is a process for Cs-137 removal from legacy nuclear waste. MCU began hot operations in 2008 with an initial planned operating life of three years and a design life of five years. After eight years of operation, MCU has processed more than 23.8 million liters (6.3 million gallons) of waste. The MCU process uses a solvent developed by Oak Ridge National Laboratory (ORNL) to extract the fission product Cs-137 from liquid salt waste, or salt solution (SS), stored in the Tank Farms at Savannah River Site (SRS) after the material has been treated through the Actinide Removal Process (ARP). Batches of SS are prepared and first fed to the ARP, where it is treated to become clarified SS (CSS), then fed to MCU. Once the Cs-137 has been extracted by the solvent, the CSS becomes decontaminated SS (DSS), which is ultimately processed into a cementitious grout at the Saltstone Production Facility (SPF) and subsequently disposed of at the Saltstone Disposal Facility (SDF). The Cs-137 is then stripped from the solvent into an aqueous strip effluent (SE) stream for incorporation into borosilicate glass at the Defense Waste Processing Facility (DWPF). The solvent used in the MCU process is made of four components: the extractant (extracts Cs-137 from CSS), the modifier (increases solubility of extractant and suppressor in the diluent), the suppressor (suppresses the effects of anions and improves stripping), and the diluent (controls density and viscosity). There have been two solvents utilized during the lifetime of MCU. The first solvent was developed by ORNL and was used to design the plant based on its cesium removal capability. Due to the extended life of MCU, in order to continue to minimize the curies disposed at SDF, a new solvent capable of removing more cesium from the CSS was developed by ORNL and implemented at MCU. The second MCU solvent, Next-Generation Solvent (NGS), uses the same modifier and diluent as the original solvent, but a different extractant and suppressor. Subsequent changes to the overall MCU flowsheet were made for the use of NGS. The success of cesium removal from the CSS is often presented as the cesium decontamination factor (DF) achieved for each batch processed. The DF is the ratio of the concentration of cesium in the CSS to the cesium concentration in the DSS. The BOBCalixC6 based solvent was capable of achieving DFs between 100 and 400 for each DSS batch. The NGS solvent is capable of achieving DFs greater than 40,000 for each DSS batch. The NGS blend solvent is currently meeting the goals of the pure NGS solvent. (authors)

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

Operating experience in the treatment and storage of LLW in Turkey - 15010

Treatment and storage of low-level radioactive wastes (LLW) are being carried out in Radioactive Waste Management Unit (RWMU) at Cekmece Nuclear Research and Training Center (CNRTC) in Istanbul. The wastes are generated from nuclear research and nuclear applications mainly in medicine, biology, ng to these specifications, LLW are classified as exempt waste, very low level waste (VLLW) etc. Decay storage is applied at a designated area for short lived radionuclide. Long lived spent radioactive sources are immobilized in retrievable packages. Generally liquid radioactive wastes are decontaminated by filtration or chemical precipitation. The radioactive sludge or filtration media from liquid waste processing is solidified using cementation. Solid waste is compacted into drums for size reduction. TENORM wastes and contaminated metal wastes are handled for decontamination and size reduction. Then these types of waste are placed in 20 ton bulk containers for long term storage. (authors)

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

Laser size reduction of contaminated Magnox Pond skips - 15166

The Magnox Pond Program Team at Hinkley Point A Power Station have successfully demonstrated the use of an automated Computer Numeric Controlled (CNC) dry milling process to efficiently decontaminate radioactive pond skips that were once used to store and transport Magnox spent fuel assemblies between reactor sites and Sellafield for reprocessing. This decontamination process has ensured ILW skips meet metal melt recycling criteria, and even Out of Scope levels, while at the same time producing dry swarf that is easily disposed of as ILW. However, the dry milling method requires the skips to be size reduced to a very simple flat geometry to simplify the milling operation. Several size reduction techniques (mechanical and thermal) have been evaluated. Based on the advantages of remote handling, high cutting speed, flexibility with complex geometries, reduced secondary waste production and significantly reduced worker exposure, robotically operated fibre delivered laser cutting was chosen. This paper describes the development, installation and operation of an automated laser cutting system commissioned at Hinkley Point A to size reduce contaminated skips in to five flat sections. It was demonstrated that a complete skip can be remotely size reduced in less than 60 minutes. This successful demonstration, coupled with CNC milling, has the potential to save an estimated British Pounds 30 million across the Magnox sites alone, with much higher saving potentials for skips stored at Sellafield, enabling this new size reduction process to meet cost reduction objectives of the UK's Nuclear Decommissioning Authority. (authors)

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

Chooz-A steam generators characterization - 15205

EDF nuclear waste management requires a deep understanding of characterization, classification and waste sorting operations. In fact, French nuclear waste management defines several classes with specific management, treatment and storage facilities. Based on particular criteria, the more the radiological risk of the nuclear waste is important, the more its management will be complex and expensive. During the dismantling of the first French pressurized reactor Chooz-A, decontamination of the primary water circuits (not including the reactor vessel), the steam generators and the pressurizer have been carried out in order to reduce their activity levels. Thanks to these decontamination operations, and a specific characterization methodology, EDF was able to de-classify the 4 steam generators and store them in one piece at the ANDRA Very Low Level Activity disposal facilities instead of the Low and Intermediate Level activity one. This de-classification allowed EDF to avoid important cutting and packaging processes. Therefore, significant dose reductions for the operators and financial gain have been made. To characterize and declare Chooz-A SG activity, EDF-CIDEN uses a methodology defined by the French institute of atomic energy, CEA. The method is based on external gamma spectrometry measurements performed with NaI collimated detectors, associated with MERCURAD simulations providing the transfer functions for the detectors and activity sources. Internal measurements are carried out with a CZT (CdZnTe) probe inside the SG tubes to refine the 3D model. In fact, the primary side represents the main source of activity, and understanding its contamination distribution is important to reduce the model and calculation uncertainties. Measurements eventually provide SG {sup 60}Co global activity, from which the activities of other radionuclides of the spectrum are determined using scaling factors. The final activity declaration takes into account the standard deviation of the measurements in order to cover the uncertainties of the methodology. Thereby, the declaration is 'slightly over-estimative', enough to meet the waste storage facility acceptance criteria. (authors)

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

Next generation solvent performance in the modular caustic side solvent extraction process - 15495

Changes to the Modular Caustic Side Solvent Extraction Unit (MCU) flow-sheet were implemented in the facility. Implementation included changing the scrub and strip chemicals and concentrations, modifying the O/A ratios for the strip, scrub, and extraction contactor banks, and blending the current BoBCalixC6 extractant based solvent in MCU with clean MaxCalix extractant based solvent. During the successful demonstration period, the MCU process was subject to rigorous oversight to ensure hydraulic stability and chemical/radionuclide analysis of the key process tanks (caustic wash tank, solvent hold tank, strip effluent hold tank, and decontaminated salt solution hold tank) to evaluate solvent carryover to downstream facilities and the effectiveness of cesium removal from the liquid salt waste. Results indicated the extraction of cesium was significantly more effective with an average Decontamination Factor (DF) of 1,129 (range was 107 to 1,824) and that stripping was effective. The contactor hydraulic performance was stable and satisfactory, as indicated by contactor vibration, contactor rotational speed, and flow stability; all of which remained at or near target values. Furthermore, the Solvent Hold Tank (SHT) level and specific gravity was as expected, indicating that solvent integrity and organic hydraulic stability were maintained. The coalescer performances were in the range of processing results under the BOBCalixC6 flow sheet, indicating negligible adverse impact of NGS deployment. After the Demonstration period, MCU began processing via routine operations. Results to date reiterate the enhanced cesium extraction and stripping capability of the Next Generation Solvent (NGS) flow sheet. This paper presents process performance results of the NGS Demonstration and continued operations of MCU utilizing the blended BobCalixC6-MaxCalix solvent under the NGS flowsheet. (authors)

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

Boiling water reactor high flow filter system - 15683

Mechanical filtration is an essential element of industrial processes and operations. In the nuclear industry, mechanical filtration is particularly problematic because of the added challenge of minimizing personnel exposure and contamination associated with filter handling. Until recently, there has been no effective filtration system to maintain water clarity and minimize dose for diving operations in Torus/Suppression Pools, or for vacuuming operations to decontaminate Dryer Separation Pits during plant outages. To be effective, the filtration system must provide high flow to speed recovery from low visibility conditions. It must also have a high dirt loading capacity to minimize downtime during outage critical path. This paper looks at a new filtration system that has been introduced to the U.S. commercial nuclear power industry and used in numerous outages since 2012 and compares this Solids Collection Filter (SCF{sup TM}) system to previous filtration systems and practices and its impact on outage diving operations and decontamination efforts. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

EPRI Software for Estimating Decommissioning Waste Volumes - 16046

EPRI has adapted computer software that can estimate the volume and activity content of waste resulting from the dismantlement of nuclear facilities. This software, designated as Automatic Estimation of the Radiological Inventory or AERI, was originally developed by Iberdrola for use by ENRESA in Spain. The software estimates the radiological inventory for all the components and civil structures of the plant or other facility that are included in the physical inventory. The radiological inventory shows the internal and external surface contamination, as well as the volumetric contamination. The processes of the software are able to estimate both the specific activity levels (in Bq/g or pCi/g) for each radionuclide and the surface activity (Bq/cm{sup 2} or dpm/100 cm{sup 2}) for components and civil structures located in the radiological areas. AERI allows the user to extrapolate, through a statistical approach, the contamination levels obtained by sampling data in specific components and civil structures to the entire facility. The main benefits and values obtained with AERI are the following: - The radiological inventory of the plant can be obtained from a limited set of radiological sampling data. Using the mathematical models allows for the optimization of the number of sampling and analysis that needs to be done. The uncertainties of the calculations are reduced with more data. The user must balance the level of uncertainty of the calculations and the number of sample data they will use. - The automatic management of radiological data in short computing times allows for the execution of several scenarios in a reasonable time. Therefore, different radioactive waste and decommissioning hypotheses can be tested. For example, inventory and data can be split into two different set. One of them can be used for obtaining radiological inventory for early dismantling (3 years, as example) and the second one for deferred dismantling (25 years.) - The software supports the planning and implementation of decommissioning strategies for nuclear plants. For instance, several decontamination factors can be evaluated, assuming different level for decontamination of materials, and results for materials from several streams and use of different container types may be included in each evaluation. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗