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All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

Site Environmental Report for Calendar Year 2013: DOE Operations at The Boeing Company, Santa Susana Field Laboratory, Area IV

This Annual Site Environmental Report (ASER) for 2013 describes the environmental conditions related to work performed for the Department of Energy (DOE) at Area IV of the Santa Susana Field Laboratory (SSFL). The Energy Technology Engineering Center (ETEC), a government-owned, company-operated test facility, was located in Area IV. The operations in Area IV included development, fabrication, operation and disassembly of nuclear reactors, reactor fuel, and other radioactive materials. Other activities in the area involved the operation of large-scale liquid metal facilities that were used for testing non-nuclear liquid metal fast breeder reactor components. All nuclear work was terminated in 1988, and all subsequent radiological work has been directed toward environmental restoration and decontamination and decommissioning (D&D) of the former nuclear facilities and their associated sites. Liquid metal research and development ended in 2002. Since May 2007, the D&D operations in Area IV have been suspended by the DOE, but the environmental monitoring and characterization programs have continued. Results of the radiological monitoring program for the calendar year 2013 continue to indicate that there are no significant releases of radioactive material from Area IV of SSFL. All potential exposure pathways are sampled and/or monitored, including air, soil, surface water, groundwater, direct radiation, transfer of property (land, structures, waste), and recycling. Due to the suspension of D&D activities in Area IV, no effluents were released into the atmosphere during 2013. Therefore, the potential radiation dose to the general public through airborne release was zero. Similarly, the radiation dose to an offsite member of the public (maximally exposed individual) due to direct radiation from SSFL is indistinguishable from background. All radioactive wastes are processed for disposal at DOE disposal sites and/or other licensed sites approved by DOE for radioactive waste disposal. No liquid radioactive wastes were released into the environment in 2013.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2014

Phase 2 Methyl Iodide Deep-Bed Adsorption Tests

Nuclear fission produces fission products (FPs) and activation products, including iodine-129, which could evolve into used fuel reprocessing facility off-gas systems, and could require off-gas control to limit air emissions to levels within acceptable emission limits. Research, demonstrations, and some reprocessing plant experience have indicated that diatomic iodine can be captured with efficiencies high enough to meet regulatory requirements. Research on the capture of organic iodides has also been performed, but to a lesser extent. Several questions remain open regarding the capture of iodine bound in organic compounds. Deep-bed methyl iodide adsorption testing has progressed according to a multi-laboratory methyl iodide adsorption test plan. This report summarizes the second phase of methyl iodide adsorption work performed according to this test plan using the deep-bed iodine adsorption test system at the Idaho National Laboratory (INL), performed during the second half of Fiscal Year (FY) 2014. Test results continue to show that methyl iodide adsorption using AgZ can achieve total iodine decontamination factors (DFs, ratios of uncontrolled and controlled total iodine levels) above 1,000, until breakthrough occurred. However, mass transfer zone depths are deeper for methyl iodide adsorption compared to diatomic iodine (I2) adsorption. Methyl iodide DFs for the Ag Aerogel test adsorption efficiencies were less than 1,000, and the methyl iodide mass transfer zone depth exceeded 8 inches. Additional deep-bed testing and analyses are recommended to (a) expand the data base for methyl iodide adsorption under various conditions specified in the methyl iodide test plan, and (b) provide more data for evaluating organic iodide reactions and reaction byproducts for different potential adsorption conditions.

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

Annual Site Environmental Report, Department of Energy Operations at the Energy Technology Engineering Center – Area IV, Santa Susana Field Laboratory

This Annual Site Environmental Report (ASER) for 2014 describes the environmental conditions related to work performed for the DOE at Area IV of the Santa Susana Field Laboratory (SSFL). The ETEC, a government-owned, company-operated test facility, was located in Area IV. The operations in Area IV included development, fabrication, operation and disassembly of nuclear reactors, reactor fuel, and other radioactive materials. Other activities in the area involved the operation of large-scale liquid metal facilities that were used for testing non-nuclear liquid metal fast breeder reactor components. All nuclear work was terminated in 1988, and all subsequent radiological work has been directed toward environmental restoration and decontamination and decommissioning (D&D) of the former nuclear facilities and their associated sites. Liquid metal research and development ended in 2002. Since May 2007, the D&D operations in Area IV have been suspended by the DOE, but the environmental monitoring and characterization programs have continued. Results of the radiological monitoring program continue to indicate that there are no significant releases of radioactive material from Area IV of SSFL. All potential exposure pathways are sampled and/or monitored, including air, soil, surface water, groundwater, direct radiation, transfer of property (land, structures, waste), and recycling.

Frazee, Brad↗
DOE OSTI2013

Site Environmental Report For Calendar Year 2012. DOE Operations at The Boeing Company Santa Susana Field Laboratory, Area IV

This Annual Site Environmental Report (ASER) for 2012 describes the environmental conditions related to work performed for the Department of Energy (DOE) at Area IV of Boeing’s Santa Susana Field Laboratory (SSFL). The Energy Technology Engineering Center (ETEC), a government-owned, company-operated test facility, was located in Area IV. The operations in Area IV included development, fabrication, operation and disassembly of nuclear reactors, reactor fuel, and other radioactive materials. Other activities in the area involved the operation of large-scale liquid metal facilities that were used for testing non-nuclear liquid metal fast breeder reactor components. All nuclear work was terminated in 1988, and all subsequent radiological work has been directed toward environmental restoration and decontamination and decommissioning (D&D) of the former nuclear facilities and their associated sites. Liquid metal research and development ended in 2002. Since May 2007, the D&D operations in Area IV have been suspended by the DOE, but the environmental monitoring and characterization programs have continued. Results of the radiological monitoring program for the calendar year 2012 continue to indicate that there are no significant releases of radioactive material from Area IV of SSFL. All potential exposure pathways are sampled and/or monitored, including air, soil, surface water, groundwater, direct radiation, transfer of property (land, structures, waste), and recycling.

Liu, Ning↗
DOE OSTI2015

Deep Borehole Disposal Remediation Costs for Off-Normal Outcomes

This memo describes rough-order-of-magnitude (ROM) cost estimates for a set of off-normal (accident) scenarios, as defined for two waste package emplacement method options for deep borehole disposal: drill-string and wireline. It summarizes the different scenarios and the assumptions made for each, with respect to fishing, decontamination, remediation, etc.

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

Strategy for Fuel Rod Receipt, Characterization, Sample Allocation for the Demonstration Sister Rods

The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), Office of Fuel Cycle Technology, has established the Used Fuel Disposition Campaign (UFDC) to conduct the research and development activities related to storage, transportation, and disposal of used nuclear fuel and high-level radioactive waste. The mission of the UFDC is to identify alternatives and conduct scientific research and technology development to enable storage, transportation and disposal of used nuclear fuel (UNF) and wastes generated by existing and future nuclear fuel cycles. The UFDC Storage and Transportation staffs are responsible for addressing issues regarding the extended or long-term storage of UNF and its subsequent transportation. The near-term objectives of the Storage and Transportation task are to use a science-based approach to develop the technical bases to support the continued safe and secure storage of UNF for extended periods, subsequent retrieval, and transportation. While low burnup fuel [that characterized as having a burnup of less than 45 gigawatt days per metric tonne uranium (GWD/MTU)] has been stored for nearly three decades, the storage of high burnup used fuels is more recent. The DOE has funded a demonstration project to confirm the behavior of used high burnup fuel under prototypic conditions. The Electric Power Research Institute (EPRI) is leading a project team to develop and implement the Test Plan to collect this data from a UNF dry storage system containing high burnup fuel. The Draft Test Plan for the demonstration outlines the data to be collected; the high burnup fuel to be included; the technical data gaps the data will address; and the storage system design, procedures, and licensing necessary to implement the Test Plan. To provide data that is most relevant to high burnup fuel in dry storage, the design of the test storage system must closely mimic real conditions high burnup SNF experiences during all stages of dry storage: loading, cask drying, inert gas backfilling, and transfer to an Independent Spent Fuel Storage Installation (ISFSI) for multi-year storage. To document the initial condition of the used fuel prior to emplacement in a storage system, “sister ” fuel rods will be harvested and sent to a national laboratory for characterization and archival purposes. This report supports the demonstration by describing how sister rods will be shipped and received at a national laboratory, and recommending basic nondestructive and destructive analyses to assure the fuel rods are adequately characterized for UFDC work. For this report, a hub-and-spoke model is proposed, with one location serving as the hub for fuel rod receipt and characterization. In this model, fuel and/or clad would be sent to other locations when capabilities at the hub were inadequate or nonexistent. This model has been proposed to reduce DOE-NE’s obligation for waste cleanup and decontamination of equipment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Radiation Hardened Electronics Destined For Severe Nuclear Reactor Environments

Post nuclear accident conditions represent a harsh environment for electronics. The full station blackout experience at Fukushima shows the necessity for emergency sensing capabilities in a radiation-enhanced environment. This NEET (Nuclear Energy Enabling Technologies) research project developed radiation hardened by design (RHBD) electronics using commercially available technology that employs commercial off-the-shelf (COTS) devices and present generation circuit fabrication techniques to improve the total ionizing dose (TID) hardness of electronics. Such technology not only has applicability to severe accident conditions but also to facilities throughout the nuclear fuel cycle in which radiation tolerance is required. For example, with TID tolerance to megarads of dose, electronics could be deployed for long-term monitoring, inspection and decontamination missions. The present work has taken a two-pronged approach, specifically, development of both board and application-specific integrated circuit (ASIC) level RHBD techniques. The former path has focused on TID testing of representative microcontroller ICs with embedded flash (eFlash) memory, as well as standalone flash devices that utilize the same fabrication technologies. The standalone flash devices are less complicated, allowing better understanding of the TID response of the crucial circuits. Our TID experiments utilize biased components that are in-situ tested, and in full operation during irradiation. A potential pitfall in the qualification of memory circuits is the lack of rigorous testing of the possible memory states. For this reason, we employ test patterns that include all ones, all zeros, a checkerboard of zeros and ones, an inverse checkerboard, and random data. With experimental evidence of improved radiation response for unbiased versus biased conditions, a demonstration-level board using the COTS devices was constructed. Through a combination of redundancy and power gating, the demonstration board exhibits radiation resilience to over 200 krad. Furthermore, our ASIC microprocessor using RHBD techniques was shown to be fully functional after an exposure of 2.5 Mrad whereas the COTS microcontroller units failed catastrophically at <100 krad. The methods developed in this work can facilitate the long-term viability of radiation-hard robotic systems, thereby avoiding obsolescence issues. As a case in point, the nuclear industry with its low purchasing power does not drive the semiconductor industry strategic plans, and the rapid advancements in electronics technology can leave legacy systems stranded.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

Project 57 Air Monitoring Report: October 1, 2013, through December 31, 2014

On April 24, 1957, the Atomic Energy Commission (AEC, now the Department of Energy [DOE]) conducted the Project 57 safety experiment in western Emigrant Valley north east of the Nevada National Security Site (NNSS, formerly the Nevada Test Site) on lands withdrawn by the Department of Defense (DoD) for the Nevada Test and Training Range (NTTR). The test was undertaken to develop (1) a means of estimating plutonium distribution resulting from a nonnuclear detonation; (2) biomedical evaluation techniques for use in plutonium-laden environments; (3) methods of surface decontamination; and (4) instruments and field procedures for prompt estimation of alpha contamination (Shreve, 1958). Although the test did not result in the fission of nuclear materials, it did disseminate plutonium across the land surface. Following the experiment, the AEC fenced the contaminated area and returned control of the surrounding land to the DoD. Various radiological surveys have been performed in the area and in 2007, the DOE expanded the demarked contamination area by posting signs 200 to 400 feet (60 to 120 meters) outside of the original fence. Plutonium in soil is thought to attach preferentially to smaller particles. Therefore, redistribution of soil particulates by wind (dust) is the mechanism most likely to transport plutonium beyond the boundary of the Project 57 contamination area. In 2011, DRI installed two instrumentation towers to measure radiological, meteorological, and dust conditions. The monitoring activity was implemented to determine if radionuclide contamination was detectable in samples of airborne dust and characterize meteorological and environmental parameters that influence dust transport. Collected data also permits comparison of radiological conditions at the Project 57 monitoring stations to conditions observed at Community Environmental Monitoring Program (CEMP) stations around the NTTR. Biweekly samples of airborne particulates are submitted for laboratory assessment of gross alpha and gross beta radioactivity and for determination of gamma-emitting radionuclides. Annual average gross alpha values at the Project 57 monitoring stations are in the same range as the highest two values reported for the CEMP stations surrounding the NTTR. Annual average gross beta values at the Project 57 monitoring stations are slightly higher than the lowest value reported for the CEMP stations surrounding the NTTR. Gamma spectroscopy analyses on samples collected from the Project 57 stations identified only naturally occurring radionuclides. No manmade radionuclides were detected. Thermoluminescent dosimeters (TLDs) indicated that the average annual radioactivity dose at the monitoring stations is higher than the dose determined at surrounding CEMP stations but approximately half of the estimated national average dose received by the general public as a result of exposure to natural sources. The TLDs at the Project 57 monitoring stations are exposed to both natural sources (terrestrial and cosmic) and radioactive releases from the Project 57 contamination area. These comparisons show that the gross alpha, gross beta, and gamma spectroscopy levels at the Project 57 monitoring stations are similar to levels observed at the CEMP stations but that the average annual dose rate is higher than at the CEMP stations. Winds in excess of approximately 15 mph begin to generate dust movement by saltation (migration of sand at the ground surface) or direct suspension in the air. Saltated sand, PM10 (inhalable) dust, and PM2.5 (fine particulate dust) exhibit an approximately exponential increase with increasing wind speed. The greatest concentrations of dust occur for winds exceeding 20 mph. During the reporting period, winds in excess of 20 mph occurred approximately 1.6 percent of the time. Preliminary assessment of individual wind events suggests that dust generation is highly variable likely because of the influence of other meteorological and environmental parameters. Although winds sufficient to generate significant amounts of dust occur at the Project 57 site, they are infrequent and of short duration. Additionally, the potential for wind transport of dust is dependent on other parameters whose influence have not yet been assessed.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2016

Silicon Carbide Temperature Monitor Processing Improvements. Status Report

Silicon carbide (SiC) temperature monitors are used as temperature sensors in Advanced Test Reactor (ATR) irradiations at the Idaho National Laboratory (INL). Although thermocouples are typically used to provide real-time temperature indication in instrumented lead tests, other indicators, such as melt wires, are also often included in such tests as an independent technique of detecting peak temperatures incurred during irradiation. In addition, less expensive static capsule tests, which have no leads attached for real-time data transmission, often rely on melt wires as a post-irradiation technique for peak temperature indication. Melt wires are limited in that they can only detect whether a single temperature is or is not exceeded. SiC monitors are advantageous because a single monitor can be used to detect for a range of temperatures that occurred during irradiation. As part of the process initiated to make SiC temperature monitors available at the ATR, post-irradiation evaluations of these monitors have been previously completed at the High Temperature Test Laboratory (HTTL). INL selected the resistance measurement approach for determining irradiation temperature from SiC temperature monitors because it is considered to be the most accurate measurement. The current process involves the repeated annealing of the SiC monitors at incrementally increasing temperature, with resistivity measurements made between annealing steps. The process is time consuming and requires the nearly constant attention of a trained staff member. In addition to the expensive and lengthy post analysis required, the current process adds many potential sources of error in the measurement, as the sensor must be repeatedly moved from furnace to test fixture. This time-consuming post irradiation analysis is a significant portion of the total cost of using these otherwise inexpensive sensors. An additional consideration of this research is that, if the SiC post processing can be automated, it could be performed in an MFC hot cell, further reducing the time and expense of lengthy decontaminations prior to processing. Sections of this report provide a general description of resistivity techniques currently used to infer peak irradiation temperature from silicon carbide temperature monitors along with some representative data, the proposed concepts to improve the process of analyzing irradiated SiC temperature monitors, the completed efforts to prove the proposed concepts, and future activities. The efforts detailed here succeeded in designing and developing a real-time automated SiC resistivity measurement system, and performed two initial test runs. Activities carried out include the assembly and integration of the system hardware; the design and development of a preliminary monitor fixture; the design of a technique to automate the data analysis and processing; the development of the communication, coordination, and user software; and the execution and troubleshooting of test run experiments using the box furnace. Although the automation system performed as required, the designed fixture did not succeed in establishing the needed electrical contacts with the SiC monitor.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

Uranium-Bearing Evaporite Mineralization Influencing Plume Persistence. Literature Review and DOE-LM Site Surveys

This report on evaporite mineralization was completed as an Ancillary Work Plan for the Applied Studies and Technology program under the U.S. Department of Energy (DOE) Office of Legacy Management (LM). This study reviews all LM sites under Title I and Title II of the Uranium Mill Tailings Radiation Control Act (UMTRCA) and one Decontamination and Decommissioning site to provide (1) a summary of which sites have evaporite deposits, (2) any available quantitative geochemical and mineralogical analyses, and (3) references to relevant reports. In this study, “evaporite” refers to any secondary mineral precipitate that occurs due to a loss of water through evaporative processes. This includes efflorescent salt crusts, where this term refers to a migration of dissolved constituents to the surface with a resulting salt crust, where “salt” can refer to any secondary precipitate, regardless of constituents. The potential for the formation of evaporites at LM sites has been identified, and may have relevance to plume persistence issues. Evaporite deposits have the potential to concentrate and store contaminants at LM sites that could later be re-released. These deposits can also provide a temporary storage mechanism for carbonate, chloride, and sulfate salts along with uranium and other contaminants of concern (COCs). Identification of sites with evaporites will be used in a new technical task plan (TTP), Persistent Secondary Contaminant Sources (PeSCS), for any proposed additional sampling and analyses. This additional study is currently under development and will focus on determining if the dissolution of evaporites has the potential to hinder natural flushing strategies and impact plume persistence. This report provides an initial literature review on evaporites followed by details for each site with identified evaporites. The final summary includes a table listing of all relevant LM sites regardless of evaporite identification.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2016

A Brief User's Guide to the Excel ® -Based DF Calculator

To understand the importance of capturing penetrating forms of iodine as well as the other volatile radionuclides, a calculation tool was developed in the form of an Excel ® spreadsheet to estimate the overall plant decontamination factor (DF). The tool requires the user to estimate splits of the volatile radionuclides within the major portions of the reprocessing plant, speciation of iodine and individual DFs for each off-gas stream within the Used Nuclear Fuel reprocessing plant. The Impact to the overall plant DF for each volatile radionuclide is then calculated by the tool based on the specific user choices. The Excel ® spreadsheet tracks both elemental and penetrating forms of iodine separately and allows changes in the speciation of iodine at each processing step. It also tracks 3 H, 14 C and 85 Kr. This document provides a basic user's guide to the manipulation of this tool.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Actual Waste Demonstration of the Nitric-Glycolic Flowsheet for Sludge Batch 9 Qualification

For each sludge batch that is processed in the Defense Waste Processing Facility (DWPF), the Savannah River National Laboratory (SRNL) performs qualification testing to demonstrate that the sludge batch is processable. Testing performed by the Savannah River National Laboratory has shown glycolic acid to be effective in replacing the function of formic acid in the DWPF chemical process. The nitric-glycolic flowsheet reduces mercury, significantly lowers the catalytic generation of hydrogen and ammonia which could allow purge reduction in the Sludge Receipt and Adjustment Tank (SRAT), stabilizes the pH and chemistry in the SRAT and the Slurry Mix Evaporator (SME), allows for effective rheology adjustment, and is favorable with respect to melter flammability. In order to implement the new flowsheet, SRAT and SME cycles, designated SC-18, were performed using a Sludge Batch (SB) 9 slurry blended from SB8 Tank 40H and Tank 51H samples. The SRAT cycle involved adding nitric and glycolic acids to the sludge, refluxing to steam strip mercury, and dewatering to a targeted solids concentration. Data collected during the SRAT cycle included offgas analyses, process temperatures, heat transfer, and pH measurements. The SME cycle demonstrated the addition of glass frit and the replication of six canister decontamination additions. The demonstration concluded with dewatering to a targeted solids concentration. Data collected during the SME cycle included offgas analyses, process temperatures, heat transfer, and pH measurements. Slurry and condensate samples were collected for subsequent analysis

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

Report on Status of Shipment of High Fluence Austenitic Steel Samples for Characterization and Stress Corrosion Crack Testing

The goal of the Mechanisms of Irradiation Assisted Stress Corrosion Cracking (IASCC) task in the LWRS Program is to conduct experimental research into understanding how multiple variables influence the crack initiation and crack growth in materials subjected to stress under corrosive conditions. This includes understanding the influences of alloy composition, radiation condition, water chemistry and metallurgical starting condition (i.e., previous cold work or heat treatments and the resulting microstructure) has on the behavior of materials. Testing involves crack initiation and growth testing on irradiated specimens of single-variable alloys in simulated Light Water Reactor (LWR) environments, tensile testing, hardness testing, microstructural and microchemical analysis, and detailed efforts to characterize localized deformation. Combined, these single-variable experiments will provide mechanistic understanding that can be used to identify key operational variables to mitigate or control IASCC, optimize inspection and maintenance schedules to the most susceptible materials/locations, and, in the long-term, design IASCC-resistant materials. In support of this research, efforts are currently underway to arrange shipment of “free” high fluence austenitic alloys available through Électricité de France (EDF) for post irradiation testing at the Oak Ridge National Laboratory (ORNL) and IASCC testing at the University of Michigan. These high fluence materials range in damage values from 45 to 125 displacements per atom (dpa). The samples identified for transport to the United States, which include nine, no-cost, 304, 308 and 316 tensile bars, were relocated from the Research Institute of Atomic Reactors (RIAR) in Dimitrovgrad, Ulyanovsk Oblast, Russia, and received at the Halden Reactor in Halden, Norway, on August 23, 2016. ORNL has been notified that a significant amount of work is required to prepare the samples for further shipment to Oak Ridge, Tennessee. The preliminary work for sample shipment between Halden and Oak Ridge includes fabrication of an inner cask sample container, decontamination and preparation of a Type A container, preparation of new activity calculations, all necessary paperwork, and handling. ORNL will continue to work to track progress of sample preparation and shipment status, and to work toward an agreement that covers material shipping costs between the Halden Reactor and the Oak Ridge National Laboratory.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Gary Laughlin IHNS Intro to 2015 HPC Annual Report

The International, Homeland, and Nuclear Security (IHNS) Program Management Unit (PMU) oversees a broad portfolio of Sandia’s programs in areas ranging from global nuclear security to critical asset protection. We use science and technology, innovative research, and global engagement to counter threats, reduce dangers, and respond to disasters. The PMU draws on the skills of scientists and engineers from across Sandia. Our programs focus on protecting US government installations, safeguarding nuclear weapons and materials, facilitating nonproliferation activities, securing infrastructures, countering chemical and biological dangers, and reducing the risk of terrorist threats. We conduct research in risk and threat analysis, monitoring and detection, decontamination and recovery, and situational awareness. We develop technologies for verifying arms control agreements, neutralizing dangerous materials, detecting intruders, and strengthening resiliency. Our programs use Sandia’s High-Performance Computing resources for predictive modeling and simulation of interdependent systems, for modeling dynamic threats and forecasting adaptive behavior, and for enabling decision support and processing large cyber data streams. In this report, we highlight four advanced computation projects that illustrate the breadth of the IHNS mission space.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2016

US Efforts in Support of Examinations at Fukushima Daiichi – 2016 Evaluations

Although it is clear that the accident signatures from each unit at the Fukushima Daiichi Nuclear Power Station (NPS) [Daiichi] differ, much is not known about the end-state of core materials within these units. Some of this uncertainty can be attributed to a lack of information related to cooling system operation and cooling water injection. There is also uncertainty in our understanding of phenomena affecting: a) in-vessel core damage progression during severe accidents in boiling water reactors (BWRs), and b) accident progression after vessel failure (ex-vessel progression) for BWRs and Pressurized Water Reactors (PWRs). These uncertainties arise due to limited full scale prototypic data. Similar to what occurred after the accident at Three Mile Island Unit 2, these Daiichi units offer the international community a means to reduce such uncertainties by obtaining prototypic data from multiple full-scale BWR severe accidents. Information obtained from Daiichi is required to inform Decontamination and Decommissioning activities, improving the ability of the Tokyo Electric Power Company Holdings (TEPCO) to characterize potential hazards and to ensure the safety of workers involved with cleanup activities. This document reports recent results from the US Forensics Effort to use information obtained by TEPCO to enhance the safety of existing and future nuclear power plant designs. This Forensics Effort, which is sponsored by the Reactor Safety Technologies Pathway of the Department of Energy Office of Nuclear Energy Light Water Reactor (LWR) Sustainability Program, consists of a group of US experts in LWR safety and plant operations that have identified examination needs and are evaluating TEPCO information from Daiichi that address these needs. Examples presented in this report demonstrate that significant safety insights are being obtained in the areas of component performance, fission product release and transport, debris end-state location, and combustible gas generation and transport. In addition to reducing uncertainties related to severe accident modeling progression, these insights are being used to update guidance for severe accident prevention, mitigation, and emergency planning. Furthermore, reduced uncertainties in modeling the events at Daiichi will improve the realism of reactor safety evaluations and inform future D&D activities by improving the capability for characterizing potential hazards to workers involved with cleanup activities.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

High Throughput Method of Extracting and Counting Strontium-90 in Urine

A method has been developed for the rapid extraction of Sr-90 from the urine of individuals exposed to radiation in a terrorist attack. The method employs two chromatographic ion-exchange materials: Diphonix resin and Sr resin, both of which are commercially available. The Diphonix resin reduces the alkali ion concentrations below 10 mM, and the Sr resin concentrates and decontaminates strontium-90. Experimental and calculational data are given for a variety of test conditions. On the basis of these results, a flowsheet has been developed for the rapid concentration and extraction of Sr-90 from human urine samples for subsequent beta-counting.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2017

CoDCon Dynamic Modeling.

The Co-Decontamination (CoDCon) Demonstration project is designed to test the separation of a mixed U and Pu product from dissolved spent nuclear fuel. The primary purpose of the project is to quantify the accuracy and precision to which a U/Pu mass ratio can be achieved without removing a pure Pu product. The system includes an on-line monitoring system using spectroscopy to monitor the ratios throughout the process. A dynamic model of the CoDCon flowsheet and on-line monitoring system was developed in order to expand the range of scenarios that can be examined for process control and determine overall measurement uncertainty. The model development and initial results are presented here.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Data Validation Package June 2016 Groundwater and Surface Water Sampling at the Old and New Rifle, Colorado, Processing Sites September 2016

Sampling Period: June 14–17 and July 7, 2016. Water samples were collected from 36 locations at New Rifle and Old Rifle, Colorado, Disposal/Processing Sites. Planned monitoring locations are shown in Attachment 1, Sampling and Analysis Work Order. Duplicate samples were collected from New Rifle locations 0216 and 0855, and Old Rifle location 0655. One equipment blank was collected after decontamination of non-dedicated equipment used to collect one surface water sample. See Attachment 2, Trip Report for additional details. Sampling and analyses were conducted as specified in the Sampling and Analysis Plan for U.S. Department of Energy Office of Legacy Management Sites (LMS/PRO/S04351, continually updated, http://energy.gov/lm/downloads/sampling-and- analysis-plan-us-department-energy-office-legacy-management-sites). New Rifle Site Samples were collected at the New Rifle site from 16 monitoring wells and 7 surface locations in compliance with the December 2008 Groundwater Compliance Action Plan [GCAP] for the New Rifle, Colorado, Processing Site (LMS/RFN/S01920). Monitoring well 0216 could not be sampled in June because it was surrounded by standing water due to the high river stage from spring runoff, it was later sampled in July. Monitoring well 0635 and surface location 0322 could not be sampled because access through the elk fence along Interstate 70 has not been completed at this time. Old Rifle Site Samples were collected at the Old Rifle site from eight monitoring wells and five surface locations in compliance with the December 2001 Ground Water Compliance Action Plan for the Old Rifle, Colorado, UMTRA Project Site (GJ0-2000-177-TAR).

54 ENVIRONMENTAL SCIENCES↗