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DOE OSTI2014

Technical Basis Document for Internal Dosimetry at Sandia National Laboratories Revision 2.

The RPID Project will be implemented at all SNL facilities for activities involving the processing and/or storing of radioactive materials. This project includes activities at the Tech Area (TA) I, TA II, TA III, TA IV, TA V, Coyote Test Field, and environmental restoration sites at SNL, located in Albuquerque, New Mexico, and the Kauai Test Facility(SNL/KTF). Reference to SNL throughout this document includes facilities and activities at the Albuquerque location and at SNL/KTF.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2014

GADRAS isotope ID user s manual for analysis of gamma-ray measurements and API for Linux and Android .

Isotope identification algorithms that are contained in the Gamma Detector Response and Analysis Software (GADRAS) can be used for real-time stationary measurement and search applications on platforms operating under Linux or Android operating sys-tems. Since the background radiation can vary considerably due to variations in natu-rally-occurring radioactive materials (NORM), spectral algorithms can be substantial-ly more sensitive to threat materials than search algorithms based strictly on count rate. Specific isotopes or interest can be designated for the search algorithm, which permits suppression of alarms for non-threatening sources, such as such as medical radionuclides. The same isotope identification algorithms that are used for search ap-plications can also be used to process static measurements. The isotope identification algorithms follow the same protocols as those used by the Windows version of GADRAS, so files that are created under the Windows interface can be copied direct-ly to processors on fielded sensors. The analysis algorithms contain provisions for gain adjustment and energy lineariza-tion, which enables direct processing of spectra as they are recorded by multichannel analyzers. Gain compensation is performed by utilizing photopeaks in background spectra. Incorporation of this energy calibration tasks into the analysis algorithm also eliminates one of the more difficult challenges associated with development of radia-tion detection equipment.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2013

Post-Remediation Radiological Dose Assessment, Painesville Site, Painesville, Ohio

A post-remediation radiological dose assessment was conducted for the Formerly Utilized Sites Remedial Action Program (FUSRAP) Painesville Site by using the measured residual concentrations of the radionuclides of concern following the completion of remedial action. The site’s FUSRAP-related constituents of concern (COCs) are radionuclides associated with past activities conducted by the Manhattan Engineer District (MED) and Atomic Energy Commission (AEC) in support of the Nation’s early atomic energy and weapons program, and include radium-226 (Ra-226), thorium-230 (Th-230), thorium-232 (Th-232), and total uranium (Total U). Remedial actions were conducted at the site in accordance with the Final Record of Decision for the Painesville Site, Painesville, Ohio (ROD) (USACE 2006). In the ROD, the U.S. Army Corps of Engineers (USACE) determined that the U.S. Nuclear Regulatory Commission (NRC) standards for decommissioning of licensed facilities found in Title 10, Part 20.1402 of the Code of Federal Regulations (10 CFR 20.1402) and in the Ohio Administrative Code (OAC) 3701:1-38-22(B) are applicable or relevant and appropriate requirements (ARARs) for the site. On the basis of the ARAR analysis, a total effective dose equivalent (TEDE) goal of 25 millirem per year (mrem/yr) was assumed for the site with a construction worker considered as the average member of the critical group. Derived concentration guideline levels (DCGLs) for residual radioactivity of the COCs in soil were developed for the Painesville Site using the RESRAD (Version 6.22) computer code (Yu et al. 2001), which was designed to estimate radiation doses and risks from RESidual RADioactive materials. The DCGLs represent activity levels above site average background activity corresponding to the 25-mrem/yr criterion for a construction worker scenario. The DCGLs are above the background activity concentrations and represent average concentration guidelines. The DCGLs and average background activity concentrations are provided in Table 1. The Final Feasibility Study Addendum (USACE 2005) provides a complete description of the derivation of the DCGLs for the Painesville Site. A post-remediation radiological dose assessment was performed to determine the potential radiation doses to a construction worker, consistent with the approach used to develop the DCGLs for the Painesville Site given in the Final Feasibility Study Addendum (USACE 2005) and ROD (USACE 2006). A second more conservative dose assessment was performed to determine the potential radiation doses to an urban resident who is assumed to live and garden on the site. To further evaluate the long-term protectiveness of the remedial action, a third dose assessment was conducted to determine the potential radiation doses to a hypothetical subsistence farmer. The primary difference between exposure to an urban resident versus a conservative subsistence farmer would be an increase in the amount of foodstuffs (homegrown produce, milk, and beef) produced/cultivated and consumed on-site.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2014

Recommended values for the distribution coefficient (Kd) to be used in dose assessments for decommissioning the Zion Nuclear Power Plant

ZionSolutions is in the process of decommissioning the Zion Nuclear Power Plant. The site contains two reactor Containment Buildings, a Fuel Building, an Auxiliary Building, and a Turbine Building that may be contaminated. The current decommissioning plan involves removing all above grade structures to a depth of 3 feet below grade. The remaining underground structures will be backfilled. The remaining underground structures will contain low amounts of residual licensed radioactive material. An important component of the decommissioning process is the demonstration that any remaining activity will not cause a hypothetical individual to receive a dose in excess of 25 mrem/y as specified in 10CFR20 SubpartE.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2015

Sustainable Management of Flowback Water during Hydraulic Fracturing of Marcellus Shale for Natural Gas Production

This study evaluated the feasibility of using abandoned mine drainage (AMD) as make- up water for the reuse of produced water for hydraulic fracturing. There is an abundance of AMD sources near permitted gas wells as documented in this study that can not only serve as makeup water and reduce the demand on high quality water resources but can also as a source of chemicals to treat produced water prior to reuse. The assessment of AMD availability for this purpose based on proximity and relevant regulations was accompanied by bench- and pilot-scale studies to determine optimal treatment to achieve desired water quality for use in hydraulic fracturing. Sulfate ions that are often present in AMD at elevated levels will react with Ba²⁺ and Sr²⁺ in produced water to form insoluble sulfate compounds. Both membrane microfiltration and gravity separation were evaluated for the removal of solids formed as a result of mixing these two impaired waters. Laboratory studies revealed that neither AMD nor barite formed in solution had significant impact on membrane filtration but that some produced waters contained submicron particles that can cause severe fouling of microfiltration membrane. Coagulation/flocculation was found to be an effective process for the removal of suspended solids and both bench- and pilot-scale studies revealed that optimal process conditions can consistently achieve the turbidity of the finished water below 5 NTU. Adjusting the blending ratio of AMD and produced water can achieve the desired effluent sulfate concentration that can be accurately predicted by chemical thermodynamics. Co-treatment of produced water and AMD will result in elevated levels of naturally occurring radioactive materials (NORM) in the solid waste generated in this process due to radium co-precipitation with barium sulfate. Laboratory studies revealed that the mobility of barite that may form in the subsurface due to the presence of sulfate in the fracturing fluid can be controlled by the addition of appropriate antiscalants.

03 NATURAL GAS↗
DOE OSTI2015

Fact Sheet: Radiopharmaceuticals.

Radiopharmaceuticals are drugs that use radioactive material to treat or diagnose disease. These drugs can also be referred to as medical radioisotopes. Radiopharmaceuticals help to provide doctors with an image of a patient’s internal organs through a noninvasive procedure. Less frequently, radioisotopes are used for treatment of cancer and other diseases, this is known as radiation therapy or radio immunotherapy.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2015

Tritium glovebox stripper system seismic design evaluation

The use of glovebox confinement at US Department of Energy (DOE) tritium facilities has been discussed in numerous publications. Glovebox confinement protects the workers from radioactive material (especially tritium oxide), provides an inert atmosphere for prevention of flammable gas mixtures and deflagrations, and allows recovery of tritium released from the process into the glovebox when a glovebox stripper system (GBSS) is part of the design. Tritium recovery from the glovebox atmosphere reduces emissions from the facility and the radiological dose to the public. Location of US DOE defense programs facilities away from public boundaries also aids in reducing radiological doses to the public. This is a study based upon design concepts to identify issues and considerations for design of a Seismic GBSS. Safety requirements and analysis should be considered preliminary. Safety requirements for design of GBSS should be developed and finalized as a part of the final design process.

42 ENGINEERING↗
DOE OSTI2015

Advanced Imaging Algorithms for Radiation Imaging Systems

The intent of the proposed work, in collaboration with University of Michigan, is to develop the algorithms that will bring the analysis from qualitative images to quantitative attributes of objects containing SNM. The first step to achieving this is to develop an indepth understanding of the intrinsic errors associated with the deconvolution and MLEM algorithms. A significant new effort will be undertaken to relate the image data to a posited three-dimensional model of geometric primitives that can be adjusted to get the best fit. In this way, parameters of the model such as sizes, shapes, and masses can be extracted for both radioactive and non-radioactive materials. This model-based algorithm will need the integrated response of a hypothesized configuration of material to be calculated many times. As such, both the MLEM and the model-based algorithm require significant increases in calculation speed in order to converge to solutions in practical amounts of time.

Marleau, Peter↗
DOE OSTI2015

GRIZZLY Model of Multi-Reactive Species Diffusion, Moisture/Heat Transfer and Alkali-Silica Reaction for Simulating Concrete Aging and Degradation

Concrete is widely used in the construction of nuclear facilities because of its structural strength and its ability to shield radiation. The use of concrete in nuclear power plants for containment and shielding of radiation and radioactive materials has made its performance crucial for the safe operation of the facility. As such, when life extension is considered for nuclear power plants, it is critical to have accurate and reliable predictive tools to address concerns related to various aging processes of concrete structures and the capacity of structures subjected to age-related degradation. The goal of this report is to document the progress of the development and implementation of a fully coupled thermo-hydro-mechanical-chemical model in GRIZZLY code with the ultimate goal to reliably simulate and predict long-term performance and response of aged NPP concrete structures subjected to a number of aging mechanisms including external chemical attacks and volume-changing chemical reactions within concrete structures induced by alkali-silica reactions and long-term exposure to irradiation. Based on a number of survey reports of concrete aging mechanisms relevant to nuclear power plants and recommendations from researchers in concrete community, we’ve implemented three modules during FY15 in GRIZZLY code, (1) multi-species reactive diffusion model within cement materials; (2) coupled moisture and heat transfer model in concrete; and (3) anisotropic, stress-dependent, alkali-silica reaction induced swelling model. The multi-species reactive diffusion model was implemented with the objective to model aging of concrete structures subjected to aggressive external chemical attacks (e.g., chloride attack, sulfate attack, etc.). It considers multiple processes relevant to external chemical attacks such as diffusion of ions in aqueous phase within pore spaces, equilibrium chemical speciation reactions and kinetic mineral dissolution/precipitation. The moisture/heat transfer module was implemented to simulate long-term spatial and temporal evolutions of the moisture and temperature fields within concrete structures at both room and elevated temperatures. The ASR swelling model implemented in GRIZZLY code can simulate anisotropic expansions of ASR gel under either uniaxial, biaxial and triaxial stress states, and can be run simultaneously with the moisture/heat transfer model and coupled with various elastic/inelastic solid mechanics models that were implemented in GRIZZLY code previously. This report provides detailed descriptions of the governing equations, constitutive equations and numerical algorithms of the three modules implemented in GRIZZLY during FY15, simulation results of example problems and model validation results by comparing simulations with available experimental data reported in the literature. The close match between the experiments and simulations clearly demonstrate the potential of GRIZZLY code for reliable evaluation and prediction of long-term performance and response of aged concrete structures in nuclear power plants.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

K West Basin Sand Filter Backwash Sample Analysis

A sand filter is used to help maintain water clarity at the K West Basin where highly radioactive sludge is stored. Eventually that sand filter will require disposal. The radionuclide content of the solids trapped in the sand filter will affect the selection of the sand filter disposal pathway. The Pacific Northwest National Laboratory (PNNL) was contracted by the K Basin Operations & Plateau Remediation Project (operations contractor CH2M Hill) to analyze the radionuclide content of the solids collected from the backwash of the K West Basin sand filter. The radionuclide composition in the sand filter backwash solids will be used by CH2M Hill to determine if the sand filter media and retained sludge solids will be designated as transuranic waste for disposal purposes or can be processed through less expensive means. On October 19, 2015, K Basin Operations & Plateau Remediation Project staff backwashed the sand filter into the North Load-Out Pit (NLOP) and immediately collected sample slurry from a sampling tube positioned 24 in. above the NLOP floor. The 764 g sand filter backwash slurry sample, KW-105 SFBW-001, was submitted to PNNL for analysis on October 20, 2015. Solids from the slurry sample were consolidated into two samples (i.e., a primary and a duplicate sample) by centrifuging and measured for mass (0.82 g combined – wet centrifuged solids basis) and volume (0.80 mL combined). The solids were a dark brown/orange color, consistent with iron oxide/hydroxide. The solids were dried; the combined dry solids mass was 0.1113 g, corresponding to 0.0146 weight percent (wt%) solids in the original submitted sample slurry. The solids were acid-digested using nitric and hydrochloric acids. Insoluble solids developed upon dilution with 0.5 M HNO 3 , corresponding to an average 6.5 wt% of the initial dry solids content. The acid digestate and insoluble solids were analyzed separately by gamma spectrometry. Nominally, 7.7% of the 60 Co was present in the insoluble solids; less than 1% of other gamma-emitters (i.e., 137 Cs, 154/155 Eu, and 241 Am) were present in the insoluble solids. Aliquots of the acid digestate were analyzed directly using gamma energy analysis (GEA) and after separations for 238 Pu, 239+240 Pu, 237 Np, and 241 Am radioisotopes using alpha energy analysis (AEA). The 90 Sr was measured by liquid scintillation counting (LSC) on the Sr-separated fraction. The plutonium isotopic distribution of the acid digestate was analyzed following Pu separations by thermal ionization mass spectrometry (TIMS). A table summarizes the results for the primary and duplicate samples. The 239+240 Pu concentration (µCi/g dry) relative to 90 Sr and to 137 Cs concentrations (µCi/g dry) was examined. The K West Basin sludge has a 239+240 Pu/ 90 Sr ranging from 0.1 to 1.2 and the 239+240 Pu/ 137 Cs ratio ranging from 0.10 to 0.47. In contrast, the sand filter backwash solids 239+240 Pu/ 90 Sr ratio was 10.6 and the 239+240 Pu/ 137 Cs ratio was 2.0. The ratio differences indicate a relative enhancement of the Pu concentration in the sand filter solids relative to the 137 Cs and 90 Sr sludge concentrations currently in the K West Basin. A dose-to-curie radioisotope evaluation of the sand filter waste form may need to consider this dissimilarity.

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

Production of Positron Emitting Radiometals: Cu-64, Y-86, Zr-89. Final report

This proposal seeks support to increase our production of the radionuclides yttrium-86 and zirconium-89 while continuing to produce copper-64. We have the advantage that we already ship out copper-64 to some 20-25 institutions per week (over 60 different institutions total including 4 Canadian sites) and thus our group already has significant experience with producing, purifying and shipping radioactive materials. A significant portion of the funds requested supported the purchase of a new hotcell for the production of zirconium-89 and yttrium-86.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2016

Land and Water Use Characteristics and Human Health Input Parameters for use in Environmental Dosimetry and Risk Assessments at the Savannah River Site. 2016 Update

Operations at the Savannah River Site (SRS) result in releases of small amounts of radioactive materials to the atmosphere and to the Savannah River. For regulatory compliance purposes, potential offsite radiological doses are estimated annually using computer models that follow U.S. Nuclear Regulatory Commission (NRC) regulatory guides. Within the regulatory guides, default values are provided for many of the dose model parameters, but the use of applicant site-specific values is encouraged. Detailed surveys of land-use and water-use parameters were conducted in 1991 and 2010. They are being updated in this report. These parameters include local characteristics of meat, milk and vegetable production; river recreational activities; and meat, milk and vegetable consumption rates, as well as other human usage parameters required in the SRS dosimetry models. In addition, the preferred elemental bioaccumulation factors and transfer factors (to be used in human health exposure calculations at SRS) are documented. The intent of this report is to establish a standardized source for these parameters that is up to date with existing data, and that is maintained via review of future-issued national references (to evaluate the need for changes as new information is released). These reviews will continue to be added to this document by revision.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2016

Nondestructive Analysis of MET-5 Drum at TA35 Building 2 A-Wing Vault.

Measurements using a mechanically cooled portable high purity germanium HPGe Ortec detective were taken of a type B drum Serial number R-1157 to determine if any radioactive material was inside. The Drum, R-1157, spectrum was analyzed using PeakEasy 4.84 and a visual look at the spectrum showed background radiation.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2016

Regulatory Technology Development Plan - Sodium Fast Reactor: Mechanistic Source Term – Trial Calculation

The potential release of radioactive material during a plant incident, referred to as the source term, is a vital design metric and will be a major focus of advanced reactor licensing. The U.S. Nuclear Regulatory Commission has stated an expectation for advanced reactor vendors to present a mechanistic assessment of the potential source term in their license applications. The mechanistic source term presents an opportunity for vendors to realistically assess the radiological consequences of an incident, and may allow reduced emergency planning zones and smaller plant sites. However, the development of a mechanistic source term for advanced reactors is not without challenges, as there are often numerous phenomena impacting the transportation and retention of radionuclides. This project sought to evaluate U.S. capabilities regarding the mechanistic assessment of radionuclide release from core damage incidents at metal fueled, pool-type sodium fast reactors (SFRs). The purpose of the analysis was to identify, and prioritize, any gaps regarding computational tools or data necessary for the modeling of radionuclide transport and retention phenomena. To accomplish this task, a parallel-path analysis approach was utilized. One path, led by Argonne and Sandia National Laboratories, sought to perform a mechanistic source term assessment using available codes, data, and models, with the goal to identify gaps in the current knowledge base. The second path, performed by an independent contractor, performed sensitivity analyses to determine the importance of particular radionuclides and transport phenomena in regards to offsite consequences. The results of the two pathways were combined to prioritize gaps in current capabilities.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Characterization Report for the Demolition and Remodel of B151 Rooms 1318, 1322 and 1326

In support of a remodel for rooms 1318, 1322, and 1326 in building 151, the Environmental Safety and Health (ES&H) team has worked to develop a detailed characterization of contaminants in each room. This report serves as a summary of the results for all radioactive material and Industrial Hygiene (IH) metals sampling performed of accessible areas.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2017

Test report dot 7A type a liquid packaging

This test report documents the performance of Savannah River National Laboratory’s (SRNL’s) U.S. Department of Transportation (DOT) Specification 7A; General Packaging, Type A shielded liquid shipping packaging and compliance with the regulatory requirements of Title 49 of the Code of Federal Regulations (CFR). The primary use of this packaging design is for the transport of radioactive liquids of up to 1.3 liters in an unshielded configuration and up to 113 mL of radioactive liquids in a shielded configuration, with no more than an A2 quantity in either configuration, over public highways and/or commercial aircraft. The contents are liquid radioactive materials sufficiently shielded and within the activity limits specified in173.435 or 173.433 for A2 (normal form) materials, as well as within the analyzed thermal heat limits. Any contents must be compatibly packaged and must be compatible with the packaging. The basic packaging design is based on the U.S. Department of Energy’s (DOE’s) Model 9979 Type A fissile shipping packaging designed and tested by SRNL. The shielded liquid configuration consists of the outer and inner drums of the 9979 package with additional low density polyethylene (LDPE) dunnage nesting a tungsten shielded cask assembly (WSCA) within the 30-gallon inner drum. The packaging model for the DOT Specification 7A, Type A liquids packaging is HVYTAL.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Land and Water Use Characteristics and Human Health Input Parameters for use in Environmental Dosimetry and Risk Assessments at the Savannah River Site 2017 Update

Operations at the Savannah River Site (SRS) result in releases of relatively small amounts of radioactive materials to the atmosphere and to the Savannah River. For regulatory compliance purposes, potential offsite radiological doses are estimated annually using computer models that follow U.S. Nuclear Regulatory Commission (NRC) regulatory guides. Within the regulatory guides, default values are provided for many of the dose model parameters, but the use of site-specific values is encouraged. Detailed surveys of land-use and water-use parameters were conducted in 1991, 2008, 2010, and 2016 and are being concurred with or updated in this report. These parameters include local characteristics of meat, milk, and vegetable production; river recreational activities; and meat, milk, and vegetable consumption rates, as well as other human usage parameters required in the SRS dosimetry models. In addition, the preferred elemental bioaccumulation factors and transfer factors (to be used in human health exposure calculations at SRS) are documented. The intent of this report is to establish a standardized source for these parameters that is up to date with existing data, and that is maintained via review of future-issued national references (to evaluate the need for changes as new information is released). These reviews will continue to be added to this document by revision.

54 ENVIRONMENTAL SCIENCES↗