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DOE OSTI2014

RADTRAN 6 Technical Manual

This Technical Manual contains descriptions of the calculation models and mathematical and numerical methods used in the RADTRAN 6 computer code for transportation risk and consequence assessment. The RADTRAN 6 code combines user-supplied input data with values from an internal library of physical and radiological data to calculate the expected radiological consequences and risks associated with the transportation of radioactive material. Radiological consequences and risks are estimated with numerical models of exposure pathways, receptor populations, package behavior in accidents, and accident severity and probability.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Radiological Worker II Training, Course 20301 (Live), Course 12909 (Test)

Radiological worker training is the basic building block for any additional radiological training you may receive. Upon completing radiological worker training, you will have the basic knowledge needed to work safely, using proper radiological practices, in areas where radiological hazards exist. You will also have a better understanding of the hazards and responsibilities associated with radiological work to help prevent the carelessness that can occur when working continually with or around radioactive material. This course does not qualify you for any specific radiological work. You may be required to take additional training at individual facilities to address facility- and job-specific hazards and procedures.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2012

Treatment of Uranium Slugs at the CEA Marcoule site - 12026

The decladding units on the Marcoule nuclear site in southeast France were commissioned in the early 1960's to receive spent fuel cartridge for interim storage in pools and to prepare them for dissolution. After decladding, the uranium slugs were sent to the reprocessing plant for dissolution. The units were shut down in 1997. All the operational and process equipment was dismantled and transferred to the ANDRA waste repository between 1998 and 2001. During cleanup of the G2-G3 interim storage pool all the internal components (baskets and machinery) mounted on the civil engineering structures were cut up. In November 2007, a UNGG fuel slug in its aluminum cartridge was found at the bottom of the pool. Similarly, while iron slugs (used to control plutonium-producing reactors) were being sorted, five de-cladded uranium slugs were found between 2006 and 2008. As the processing facilities had already been dismantled, it was decided to send these fuel cartridges to a laboratory for dissolution. The project was to recover and condition the cartridge and slugs and transport them safely to the laboratory. This required the design of dedicated equipment due in particular to the high dose rate and pyro-phoricity hazard. The procedure also required safety clearance by the regulatory authorities for the operations within the facility and for the transport of radioactive materials. The cartridge was dissolved in May 2009, 18 months after it was discovered. Based on the experience acquired, the retrieval of the uranium slugs was completed in March 2011 after a year of preparation. (authors)

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

Standards and Guidelines for the Design of Trunnions of RAM Transport Packages - 17365

Trunnion systems of packages for the transport of radioactive materials have to guarantee the safe handling of package during crane operations (lifting, tilting) and to secure package tie down to the transport vehicle, if the trunnions are used as attachment points during transport. The design of trunnions is based on IAEA Regulations SSR-6, the supporting Advisory Material SSG-26 and further appropriate technical standards and/or guidelines. For package approval procedures in Germany the guideline BAM-GGR 012 has to be applied. If a package is handled inside a german nuclear power plant, the standard KTA 3905 has to be fulfilled additionally. In this paper the requirements of KTA 3905 [4] concerning the trunnion systems as load attaching points (LAP) are discussed in connection with the recommendations in the guideline BAM-GGR 012 [3]. This guideline is prepared at BAM Federal Institute for Materials Research and Testing for analysis and assessment of bolted lid and trunnion systems of Type B(U) transport packages. The quality assurance questions concerning trunnion systems are discussed as well. (authors)

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

Total Integrated Dose to Spent Fuel Pool Level Instrument Probes and Refueling Floor Personnel Dose Rates Meeting NRC Order EA-12-051 for a Typical Boiling Water Reactor - Paper 22

In reviewing the events after the March 11, 2011 Fukushima accident, the Nuclear Regulatory Commission (NRC) issued Order EA-12-051, Reference 1, requiring new reliable spent fuel pool (SFP) instrumentation for pool level determination under beyond design basis event conditions. In 2013, URS prepared a calculation for a typical Boiling Water Reactor to determine the Total Integrated Dose (TID) potentially to be imparted to radiation sensitive underwater probes (one primary and one backup) planned for meeting this Order. Additionally, URS calculated the radiological dose rates to an individual standing on the refueling floor for various SFP fill levels. NEI 12-02, Reference 2, defines the levels at which SFP condition identification is necessary as the following: Level 1 - level that is adequate to support operation of the normal fuel pool cooling system. Level 1 represents the higher of the following two points: - The level at which reliable suction loss occurs due to uncovering of the coolant inlet pipe, weir or vacuum breaker (depending on the design), or - The level at which the water height, assuming saturated conditions, above the centerline of the cooling pump suction provides the required net positive suction head specified by the pump manufacturer or engineering analysis. Level 2 - level that is adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck. Level 2 is based on either of the following: - 10 feet (+/- 1 foot) above the highest point of any fuel rack seated in the spent fuel pools, or - A designated level that provides adequate radiation shielding to maintain personnel radiological dose levels within acceptable limits while performing local operations in the vicinity of the pool. This level shall be based on either plant-specific or appropriate generic shielding calculations, considering the emergency conditions that may apply at the time and the scope of necessary local operations, including installation of portable SFP instrument channel components. Level 3 - level where fuel remains covered and actions to implement make-up water addition should no longer be deferred. Level 3 corresponds nominally (i.e. +/- 1 foot) to the highest point of any fuel rack seated in the spent fuel pool. The URS calculation did not define these levels or determine instrumentation setpoints. However, it determined refueling floor dose rates that could be used as a tool to determine the appropriate water depths for these levels. In determining the TID for the SFP probes and the radiological dose rates to the individual, the following radioactive sources were considered: 1. Spent Nuclear Fuel (SNF) stored in the spent fuel storage racks, 2. Local Power Range Monitors (LPRMs) stored above the top of the racks on hangers on the SFP walls 3. Radioactive material laden filters stored on platforms on top of the racks, and 4. Control Rod Blades (CRBs) stored above the top of the racks on hangers on the SFP walls. The TID for the SFP probes includes both the dose from normal operation (pool full of water) for the remaining plant life (an assumed 40 years, conservatively including an allowance for lifetime extension SFP loading) and, consistent with the NEI 12-02 guidelines, Reference 2, the assumed drop in elevation of the pool water level to the top of the spent fuel racks for a 6 hour period. The acceptable TID for the radiation-sensitive probe as defined by the manufacturer was 2x10{sup 9} rad. The individual on the pool deck was assumed to be standing on the edge of the pool with an unobstructed view of its contents, as a worst case scenario. Dose rates to this individual determined from direct shine from the same 4 sources were determined for SFP water levels varying from 1 ft above the SNF racks to 21 ft above the racks in one foot increments. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Irreversible Wash Aid Additive for Cesium Mitigation. Small-Scale Demonstration and Lessons Learned

The Irreversible Wash Aid Additive process has been under development by the U.S. Environmental Protection Agency (EPA) and Argonne National Laboratory (Argonne). This process for radioactive cesium mitigation consists of a solution to wash down contaminated structures, roadways, and vehicles and a sequestering agent to bind the radionuclides from the wash water and render them environmentally immobile. The purpose of this process is to restore functionality to basic services and immediately reduce the consequences of a radiologically-contaminated urban environment. Research and development have resulted in a down-selection of technologies for integration and demonstration at the pilot-scale level as part of the Wide Area Recovery and Resiliency Program (WARRP) under the Department of Homeland Security and the Denver Urban Area Security Initiative. As part of developing the methods for performing a pilot-scale demonstration at the WARRP conference in Denver in 2012, Argonne conducted small-scale field experiments at Separmatic Systems. The main purpose of these experiments was to refine the wash water collection and separations systems and demonstrate key unit operations to help in planning for the large scale demonstration in Denver. Since the purpose of these tests was to demonstrate the operations of the system, we used no radioactive materials. After a brief set of experiments with the LAKOS unit to familiarize ourselves with its operation, two experiments were completed on two separate dates with the Separmatic systems.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2015

Mobile platform for radiological characterization of sites under or after decommissioning - 15139

Nowadays, nuclear industry is facing a crucial need in establishing radiological characterization for the assessment and the monitoring of any remediation work. Regarding its experience in this domain, the French Commission for Atomic Energy and Alternative Energies (CEA) of Fontenay-aux-Roses, established an important feedback and developed a sound methodology for radiological characterization. A clean-up project is managed from the initial surface characterization to the final one, once volumes of contaminated soil have been removed. Samples are collected via recognized methods and data is gathered by real-time acquisition with different conveyance such as all-in-vehicles, cart, walking, helicopter or drone. Accurate radiological analyses are performed on site thanks to the mobile laboratories cluster, called SMaRT. It enables to realize long-time gamma spectroscopy, chemical preparation for alpha spectroscopy or beta liquid scintillation counting. Data is processed with geo-statistics via the software Kartotrak, 2D or 3D mapping are created with the associated uncertainty and probability map. Drilling profiles with geological data (layers composition, particles size, etc.) are studied in order to understand the behavior of the contamination in depth. Source term, contaminated volumes and sanitary impact are calculated in order to determine the optimum excavation depth. Finally, data is stored in an online site that insures the traceability of the whole process. This just-in-time production model allows a fast response and control of costs and avoids the disadvantages of radioactive material transport. This whole innovative process, developed inside the mobile platform, requires all these tools that have been used in France and continuously evolving for more than 10 years thanks to a learning organization. (authors)

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

Fukushima Media Involvement: Lessons Learned and Challenges - 13261

Only days after the Fukushima nuclear reactor disaster on March 11, 2011, the DOE's Pacific Northwest National Laboratory, or PNNL, found itself in a maelstrom of media attention following its announcement of the detection of minute levels of radioactive material originating from the damaged reactors 4,500 miles away. Because PNNL develops state-of-the-art ultra-sensitive radionuclide detection and monitoring systems for national security applications, and has some of the equipment operating on its Richland campus, there was little surprise when one of these sophisticated systems led PNNL to be the first to detect measurable radionuclides in the United States. On Wednesday, March 16, 2011, that system detected minuscule levels of short-lived radioactive xenon, a telltale element derived from either weapons testing or a major reactor disruption. Immediately after the detection was announced, a flurry of inquiries nearly overwhelmed staff as governments, scientific organizations, the general public, and reporters struggled to understand and estimate what impacts this disaster might have on health and environment. Over the course of about three weeks, PNNL's News and Media Relations staff and its scientists and engineers responded to more than 100 requests for information, and engaged in dozens of personal interviews with international, national, regional, and local media. While many of the interviews and resulting stories were accurate and well done, not all communication went flawlessly. In the midst of chaos and confusion, which are part of any significant crisis, hiccoughs are sure to occur. Addressed here is 'the rest of the story'. (authors)

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

From the Lost Radium Files: Misadventures in the Absence of Training, Regulation, and Accountability

Purpose: Radium was the foundation of brachytherapy in the early decades of the 20th century. Despite being a most precious and perilous substance, it was mislaid with surprising frequency. This essay explores how it was lost, the efforts taken to recover it, and measures instituted to prevent mishandling. Methods and Materials: Review of contemporary literature, government publications, archives, and lay press. Results: Radium is a particularly dangerous substance because of its long half-life, its gaseous daughter (radon), and the high-energy emissions of its decay products. Despite the hazard, it was unregulated for most of the century. Any physician could obtain and administer it, and protocols for safe handling were generally lacking. Change came with appreciation of the danger, regulation, mandated training, and the institution of a culture of accountability. Unfortunately, careless management of medical radionuclides remains a global hazard. Conclusion: Responsible stewardship of radioactive material was not a high priority, for practitioners or the federal government, for much of the 20{sup th} century. As a result, large quantities of radium had gone astray, possibly subjecting the general public to continued radiation exposure. Lessons from the radium era remain relevant, as medical radionuclides are still mishandled.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2019

Current Status of Decommissioning Plans and Chemical Decontamination of Metal Scraps in Chinshan NPP - 19304

The decommissioning of nuclear facilities has become a significantly attractive and worldwide topic due to the end of lifetime of many nuclear power plants within next coming years. In Taiwan, Chinshan nuclear power plant owned by the Taiwan Power Company will be the first candidate to be decommissioned according to the Nuclear Reactor Facilities Regulation Act. The Chinshan nuclear power plant located in New Taipei City has two units of 636 MWe boiling water reactors (BWRs) for the generation of 9 billion kWh of electricity per year. The date of cessation of operation of Unit No. 1 and Unit No. 2 will be December in 2018 and July in 2019, respectively. Based on the Regulations for the Review and Approval of Applications for a Decommissioning Permit of Nuclear Reactor Facilities, Taiwan Power Company have submitted a decommissioning plan of Chinshan nuclear power plant to the authority, Atomic Energy Council (AEC). After the review process, the decommissioning plan of Chinshan nuclear power plant has been permitted by AEC in 2017. Decontamination is generally utilized during the period of decommissioning of nuclear facility in order to minimize the radiation exposure of personnel and the cost of management of radioactive wastes. The successful decontamination also probably provides the benefits of lowering the level of radioactive wastes, facilitating potential releases, permitting the reuse of metals and facilitating the radiation principle of As Low As Reasonably Achievable (ALARA) management. Decontamination aims at the removal of radioactive materials from the contaminated surfaces by several technologies such as washing, heating, chemical reaction and mechanical process. The selection of proper decontamination technologies are depended on the requirement of the target materials and the purpose of the decontamination. Chemical decontamination has been proven to be an effective in reducing the radioactivity for metallic materials by using chemical reagents in contact with the contaminated surfaces. During the operation of Chinshan nuclear power plant, chemical decontamination was performed for Unit No.1 and No.2 in 2000 to decontaminate the Reactor Recirculation System (RRS) and Reactor Water Clean-up Systems (RWCU) by using LOMI-NP-LOMI process. Based on the previous experience of chemical decontamination, the dismantled metallic piping is provided as sample by Chinshan nuclear power plant for the evaluation of the feasibility of chemical decontamination. The provided metallic sample is composed of carbon steel with dose rate around 0.2 mSv/hr. In this report, a multistep decontamination process using strongly oxidizing solution and organic acid solution is performed to remove the contaminated oxide layer from the metal surface, and the resulting contact dose rate and decontamination factor (DF) are also shown to discuss the performance in relation to the process control factors. In this paper, the recent progress in decommissioning programs and preliminary results of the decontamination experiment are described to provide usable information regarding the future decommissioning for Chinshan nuclear power plant. (authors)

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

Solubility of {sup 238}U radionuclide from various types of soil in synthetic gastrointestinal fluids using “US in vitro” digestion method

238U radionuclide is a naturally occuring radioactive material that can be found in soil. In this study, the solubility of 238U radionuclide obtained from various types of soil in synthetic gastrointestinal fluids was analysed by “US P in vitro” digestion method. The synthetic gastrointestinal fluids were added to the samples with well-ordered, mixed throughly and incubated according to the human physiology digestive system. The concentration of 238U radionuclide in the solutions extracted from the soil was measured using Induced Coupling Plasma Mass Spectrometer (ICP-MS). The concentration of 238U radionuclide from the soil samples in synthetic gastrointestinal fluids showed different values due to different homogenity of soil types and chemical reaction of 238U radionuclide. In general, the solubility of 238U radionuclide in gastric fluid was higher (0.050 – 0.209 ppm) than gastrointestinal fluids (0.024 – 0.050 ppm). It could be concluded that the US P in vitro digestion method is practicle for estimating the solubility of 238U radionuclide from soil materials and could be useful for monitoring and risk assessment purposes applying to environmental, health and contaminated soil samples.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2019

Initial Assessment of Cover Degradation Features at the Mexican Hat, Utah, UMTRCA Disposal Site - 19556

The Mexican Hat, Utah, Uranium Mill Tailings Radiation Control Act (UMTRCA) Title I Disposal Site is located on the Navajo Reservation in southeast Utah. The US Nuclear Regulatory Commission licensed the site to the US Department of Energy Office of Legacy Management (LM) for custody and long-term care. LM manages long-term surveillance and maintenance activities at the site with the principal goal of protecting human health and the environment. The disposal cell was designed to encapsulate low-level radioactive mill tailings and other residual radioactive materials (RRM) in a way that minimizes the need for active maintenance and limits radon gas emanation and precipitation infiltration. The disposal cell has a multicomponent cover, consisting of a 60 cm thick (24 inch), low-permeability, bentonite-amended radon barrier, overlain by a 15 cm thick (6 inch) sandy gravel bedding/filter layer and a 30 cm thick (12 inch) riprap erosion-protection layer. Under UMTRCA, designs intended to control and contain RRM shall be effective for up to 1000 years, to the extent reasonably achievable and, in any case, for at least 200 years. In 2016, multiple subtle depressions were identified on the rock cover along the toe and lower portions of the northeast side slope of the disposal cell. Due to concerns regarding potential impacts to disposal cell performance and erosion resistance, evaluations of the depressions and cell cover design and construction were initiated. The evaluations included reviews of the disposal cell as-built drawings and supporting design calculations for the rock cover components. Also, in 2017, a weather monitoring station was installed at the site to collect continuous site-specific meteorological data, and repeatable ground-based light imaging, detection, and ranging (lidar) topographical and aerial photogrammetry surveys were implemented to monitor for potential changes in the surface depression features. The evaluations also included visual inspections of the depressions, and in 2018, small-area manual excavations of the rock cover components (test pits) were performed to observe the conditions of the cover components, scan for radioactivity, and observe evidence of erosion on the surface of the radon barrier material. Voids and incisions were discovered within the radon barrier on the north and northeast side slopes of the disposal cell. Visual observations at test pits indicate that the bedding/filter material is highly segregated and may not be in conformance with the disposal cell engineering and construction design specifications. Post-construction development of a weak to strongly cemented material has also been observed immediately below the base of the bedding layer in test pits with observed radon barrier degradation. A series of radiological surveys have been performed at the site, and radiological readings have not exceeded natural background conditions in the vicinity of the site. Thus, the site remains protective of human health and the environment. Implementation of continuous radiological monitoring at the site will be initiated in fall 2018 through the deployment of radon cups and thermoluminescent dosimeters to obtain data that provides supporting evidence that the disposal cell remains protective of human health and the environment. Evaluations to determine the cause(s) of the depression features and associated radon barrier erosion are ongoing. Possible mechanisms of cover degradation identified thus far include segregated bedding material that would allow for higher precipitation runoff velocities, resulting in erosion of the radon barrier; post-construction cementation of radon barrier or aeolian material that could negatively affect the performance of the radon barrier; cation exchange leading to the development of cementation; and aeolian deposition affecting precipitation runoff. Moreover, geotechnical sampling and testing is planned to verify spatial cover conditions relative to design specifications. The resulting information will be used to identify possible causes for the surface depressions and associated radon barrier degradation, gain insight on potential long-term performance implications along the side slopes of the disposal cell, and identify potential corrective actions. (authors)

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

VLLW/Exempt Waste Assay at AWE Using Cronos - 12001

Previous studies have indicated that waste bag photon monitors, such as the CANBERRA Cronos-11, have the potential to assay photon emitting uranium (U) and plutonium (Pu) contaminated wastes at the new UK VLLW/exempt waste thresholds of 1 Bq/g U and 0.15 Bq/g Pu. However, the technique has limitations associated with the operation of its plastic scintillation photon detectors. These include the absence of isotopic information and a variable detector background that is modified by the chemical and physical composition of the waste as well as the presence of any naturally occurring radioactive material (NORM). Hence, it is vital to operate Cronos in a low/non-fluctuating background environment and segregate wastes according to isotopic fingerprint, NORM concentration and waste composition. This paper presents results from subsequent Cronos measurements on a range of uncontaminated waste materials, contaminated wastes from Pu and U facilities and a National Physical Laboratory (NPL) waste package standard. The results were consistent with other assay techniques and show that Cronos can be used for exempt waste assay provided that it is operated in a low/non-fluctuating background area and the waste is well segregated before monitoring. (author)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Oak Ridge National Laboratory EPA Approval Letters and Historical Documentation for a Modification in Applying 40 CFR Part 61 Appendix D

Appendix D of Title 40 Part 61, “Methods for Estimating Radionuclide Emissions,” of the Code of Federal Regulations (CFR) provides a procedure that US Department of Energy (DOE) facility owners and operators can use to estimate radionuclide emissions to the atmosphere for dose calculations instead of measuring emissions for minor sources under 40 CFR Part 61, Subpart H, “National Emission Standards for Emissions of Radionuclides Other than Radon from Department of Energy Facilities.” The procedure assumes that any radioactive material heated above 100°C is completely vaporized and emitted to the atmosphere. In 1991, the DOE Oak Ridge Reservation (DOE-ORR) requested approval to use different release fractions (RFs) for uranium because of its high melting and boiling points (1,132°C and 3,818°C, respectively). In response to the request, Environmental Protection Agency (EPA) Region IV approved the use of modified RFs for elemental uranium provided no reaction had taken place to alter its chemical form. In 2015, DOE-ORR requested approval to use different RFs for radioactive tungsten, also because of its high melting and boiling points (3,410°C and 5,660°C, respectively). EPA Region IV approved the use of modified RFs for heated radioactive tungsten metal. In accordance with the two precedents set for heating uranium and radioactive tungsten metals, in 2016, DOE-ORR requested approval to use modified RFs in similar fashion for other radioactive solid metals and compounds with melting and boiling points above 500°C that might be heated above 100°C in future research projects and experiments, and again, the EPA Region IV granted approval to use modified RFs for the list of compounds. This document contains the EPA approval letters and historical documentation used in the process to obtain approval for the use of alternative Appendix D emission factors. The approval to DOE-ORR allows modifying the existing regulatory RFs to 1 when radioactive solid metals and compounds are heated to temperatures greater than or equal to the boiling point of the solid, to 10 -3 when radioactive solid metals and compounds are heated to temperatures greater than or equal to 90% of the melting point and less than the boiling point of the solid, and to 10 -6 when radioactive solid materials are heated to temperatures above ambient air temperature but below 90% of the melting point of the solid.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2012

MAR flow mapping of Analytical Chemistry Operations (Preliminary Report)

The recently released Supplemental Directive, NA-1 SD 1027, updates the radionuclide threshold values in DOE-STD-1027-92 CN1 to reflect the use of modern parameters for dose conversion factors and breathing rates. The directive also corrects several arithmetic errors within the original standard. The result is a roughly four-fold increase in the amount of weapons-grade nuclear material allowed within a designated radiological facility. Radiological laboratory space within the recently constructed Radiological Laboratory Office and Utility Building (RLUOB) is slated to house selected analytical chemistry support activities in addition to small-scale actinide R&D activities. RLUOB is within the same facility operations envelope as TA-55. Consolidation of analytical chemistry activities to RLUOB and PF-4 offers operational efficiency improvements relative to the current pre-CMRR plans of dividing these activities between RLUOB, PF-4, and CMR. RLUOB is considered a Radiological Facility under STD-1027 - 'Facilities that do not meet or exceed Category 3 threshold criteria but still possess some amount of radioactive material may be considered Radiological Facilities.' The supplemental directive essentially increases the allowable material-at-risk (MAR) within radiological facilities from 8.4 g to 38.6 g for 239 Pu. This increase in allowable MAR provides a unique opportunity to establish additional analytical chemistry support functions in RLUOB without negatively impacting either R&D activities or facility operations. Individual radiological facilities are tasked to determine MAR limits (up to the Category 3 thresholds) appropriate to their operational conditions. This study presents parameters that impact establishing MAR limits for RLUOB and an assessment of how various analytical chemistry support functions could operate within the established MAR limits.

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

EMP Attachment 3: DOE-SC PNNL Site Dose Assessment Guidance

This Dose Assessment Guidance (DAG) describes methods to use to determine the maximally exposed individual (MEI) location and estimate dose impact to that individual under the U.S. Department of Energy (DOE) Office of Science (DOE-SC) Pacific Northwest National Laboratory (PNNL) Site Environmental Monitoring Plan (EMP). This guidance applies to public dose from radioactive material releases to the air from DOC-SC PNNL Site operations. This guidance is an attachment to PNNL’s EMP (PNNL-20919) and addresses a discrete, vital subject area of the EMP that is subject to revision independent of the main text of the EMP document.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2017

Mathematical Calibration of Modular Plastic Scintillators for Various Assay Applications - 17322

Plastic scintillators provide an inexpensive option to assay and characterize radioactive material. Due to the low cost it is generally possible to an array of large numbers of plastic scintillators to achieve high measurement efficiencies. The beta only thin plastic scintillators used in the Argos contamination monitors provide a convenient and reliable scintillator package that can be grouped together to form an assay system. Mathematical calibrations, using MCNP, have been produced for this detector that allow for rapid construction of various custom measurement geometries. The calibrations were constructed by comparing a series of measurements made with a number of beta and beta/gamma emitting nuclides. Characterization of the beta scintillators utilized beta emitting nuclides that included spectra from low average energy, e.g. C-14, to relatively high average energy, e.g. Sr/Y-90. In order to perform the calibration, custom versions of the standard Argos scintillators were produced that provide an MCA output for spectroscopic investigation. The spectra were then compared to detailed MCNP models in an iterative process to energy calibrate and efficiency calibrate each detector type. The validated models and energy calibrations were then tested against a number of volumetric geometries to confirm that the model would translate well to any other geometry. These validated models can then be used to rapidly construct custom measurement geometries for material measurement and screening. Two examples of measurement applications include a beta screening system for dirt on a conveyor belt and a flow through beta liquid assay system. The mathematical calibration process, validation results, and a number of application examples are discussed. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

An evaluation of nearest neighbor averaging's improvements to the measurement sensitivity of wide-area radiological scanning instruments - 15560

Through the Sandia National Laboratories (SNL) New Mexico Small Business Assistance (NMSBA) program, Sandia worked with the Environmental Restoration Group (ERG) Inc. to verify and validate a novel algorithm used to determine the scanning Critical Level (Lc) and Minimum Detectable Concentration (MDC) (or Minimum Detectable Areal Activity) for the 102F scanning system. Through the use of Monte Carlo statistical simulations the algorithm mathematically demonstrates accuracy in determining the Lc and MDC when a nearest-neighbor averaging (NNA) technique was used. To empirically validate this approach, SNL prepared several spiked sources and ran a test with the ERG 102F instrument on a bare concrete floor known to have no radiological contamination other than background naturally occurring radioactive material (NORM). The tests conclude that the NNA technique increases the sensitivity (decreases the Lc and MDC) for high-density data maps that are obtained by scanning radiological survey instruments. It is further described in Sandia Laboratories document number SAND2014-16522. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗