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At least 883 records · Page 49
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DOE OSTI2021

Nuclear Criticality Safety [Book Chapter]

Nuclear Criticality Safety is a field of nuclear engineering that involves worker and public safety during the handling, processing, transportation, and storage activities of fissile isotopes ( 233 U, 235 U, and 239 Pu) outside of nuclear reactors. Criticality accidents release energy as a result of accidentally producing a self-sustaining fission chain reaction. If these occur during hands-on operations the results can be lethal to worker within about 4.6 m (15 ft.) of the accident. There have been 22 process criticality accidents in the world between 1953 and 1999 resulting in physical injury, on and off-site dose, or death. Nuclear Criticality Safety involves work by qualified persons to preclude criticality accidents and their consequences by ensuring activities with fissionable materials remain subcritical, i.e., safe, during all normal and credible abnormal conditions. This article presents information about those nuclear fuel cycle steps presenting Nuclear Criticality Safety concerns that must be considered to ensure worker and public safety is assured.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Analysis of Tank 38H (HTF-38-16-26, 27) and Tank 43H (HTF-43-16-28, 29) Samples for Support of the Enrichment Control and Corrosion Control Programs

Savannah River National Laboratory analyzed samples from Tank 38H and Tank 43H to support Enrichment Control Program and Corrosion Control Program. The total uranium in the Tank 38H samples ranged from 20.5 to 34.0 mg/L while the Tank 43H samples ranged from 47.6 to 50.6 mg/L. The U-235 percentage ranged from 0.62% to 0.64% over the four samples. The total uranium and percent U-235 results appear consistent with previous Tank 38H and Tank 43H uranium measurements. The Tank 38H plutonium results show a large difference between the surface and sub-surface sample concentrations and a somewhat higher concentration than previous sub-surface samples. The two Tank 43H samples show similar plutonium concentrations and are within the range of values measured on previous samples. The plutonium results may be biased high due to the presence of plutonium contamination in the blank samples from the cell sample preparations. The four samples analyzed show silicon concentrations ranging from 47.9 to 105 mg/L.

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

Characterization of Tank 11H Samples from Tank Closure Cesium Removal 1A (TCCR 1A) Batch 1

Savannah River Mission Completion (SRMC) is currently operating the Tank Closure Cesium Removal 1A (TCCR 1A) process to remove 137 Cs from tank waste supernate using an ion exchange process. The TCCR 1A unit processes dissolved salt fed from Tank 10H through a series of ion exchange columns containing crystalline silicotitanate (CST, IONSIV TM R9120-B, 30x60) and the effluent is then discharged to Tank 11H. In support of the TCCR 1A program, SRNL analyzed samples taken from Tank 11H (without tank mixing) before, during, and at the completion of TCCR 1A Batch 1 processing. Tank 11H serves as the receipt tank for the filtered and cesium removed product from the TCCR 1A system. Processing of Batch 1 commenced on January 13, 2022 and completed on February 17, 2022, after processing approximately 70,100 gallons. A pre-production sample was collected from the heel remaining in Tank 11H just before Batch 1 processing began. In addition, five interim surface samples were collected from Tank 11H during processing, and both a surface and a variable depth sample (VDS, ~7” from tank bottom) were collected just after processing completed. Analysis of all samples included density and gamma spectroscopy, in addition to a more comprehensive suite of analytes for the pre- and post-production samples. The density of the pre-production sample was the highest of all samples measured and was then observed to decrease for the first two interim samples, followed by becoming fairly consistent for the remainder of the samples (~1.3 g/mL). The density of 1.3 g/mL is similar to the density measured for one of the three Tank 10H qualification samples (HTF-10-21-126). The 137 Cs activity was found to decrease as additional decontaminated effluent from the TCCR columns was added to Tank 11H during processing; however, an increase in activity was observed during periods of no processing which can be attributed to leaching of 137 Cs from the known solids in Tank 11H. The Cs isotope concentrations in the Tank 11H postproduction surface sample were determined to be 99.7-99.8% lower than the concentrations measured in the Tank 10H feed as measured by mass spectrometry.

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

Thermal Neutron Scattering Law Evaluation for Zirconium Carbide and Critical Mass Calculations

Zirconium carbide (ZrC) is a candidate material for use in advanced high temperature reactors, including space nuclear thermal propulsion applications. Thermal neutron scattering laws (TSLs) are generated for carbon bound in ZrC, C(ZrC), and zirconium bound in ZrC, Zr(ZrC), using ab initio lattice dynamics methods. These evaluations are to be submitted for inclusion in ENDF/B-VIII.1 and use the incoherent approximation for inelastic scattering as well as the new mixed elastic scattering treatment. The application of disordered alloy theory is introduced to appropriately capture the isotopic composition of Zr and C in the elastic scattering cross section. Localized higher energy vibrations in the C(ZrC) phonon density of states that are separated from lower energy modes result in quantized oscillations in the inelastic contributions to the TSL with a significant likelihood of large energy down-scattering and up-scattering interactions, where the latter increases in probability with temperature. The quanta of energy transfer during neutron thermalization is substantially greater than classically expected within the thermal neutron energy range. MC21 critical mass calculations of ZrC mixtures with high-enriched uranium demonstrate an impact of the TSLs when compared to free-gas treatment for 235 U concentrations less than 0.2 g/cm 3 . Additional MC21 critical mass calculations with homogenous mixtures of ZrC and reactor-grade graphite also demonstrate sensitivity to the ZrC TSL for thermal spectrum driven fission systems.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

SpectrumSDT: A program for parallel calculation of coupled rotational-vibrational energies and lifetimes of bound states and scattering resonances in triatomic systems

In this work, we present SpectrumSDT – a program for calculations of energies and lifetimes of bound rotational-vibrational states below and scattering resonances above the dissociation threshold on a global potential energy surface of a triatomic system, which may include stable molecules, weekly-bound van-der-Waals complexes, and unbound atom + diatom scattering systems. Large-amplitude vibrational motion is treated explicitly using hyper-spherical coordinates. Three options for the rotational-vibrational interaction are supported: uncoupled (symmetric top rotor), partially coupled (to include interaction between several nearest states only) and full-coupled (vibrating asymmetric-top rotor). In addition to energies and lifetimes, SpectrumSDT is able to integrate ro-vibrational wave functions over the user-defined regions of potential energy surface, which helps to classify these states. In this release of the code, SpectrumSDT is limited to ABA-type molecules with wave functions that do not extend into the regions near Eckart singularities.

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

Resourcing Strategies for the NRU Reactor Shutdown Planning and Execution - 19218

The National Research Universal Reactor (NRU) was commissioned in 1957. The NRU was fundamental for the development of the Candu reactors and, until 2016, quietly responsible for producing 40 percent of the world's supply of medical isotopes for diagnosis and cancer therapy. The Government of Canada decided in 2015 to discontinue operation as of March 31, 2018 due to challenges associated with running a 60-year old nuclear reactor. Anticipating this closing date allowed the Permanent Safe Shutdown Project (PSSP) to transition the NRU from an operating reactor to a state suitable for Storage with Surveillance. One of the biggest challenges to address was human resource management, which was rightly identified as the key to keeping the project within schedule and budget. The dilemma was, how to keep the work force motivated to contribute with the preparation of the shutdown while addressing job security concerns. This paper summarises Canadian Nuclear Laboratories (CNL) challenges and strategies to secure specialized subject matter experts, trades and operators for planning and executing the NRU reactor shutdown Project. The discussion also includes the solutions and lessons learned to date. (author)

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

Automated Clean Chemistry for Bulk Analysis of Environmental Swipe Samples – Final Report

Sample preparation methods for mass spectrometry are being automated using commercial-off-the-shelf (COTS) equipment to shorten lengthy and costly manual chemical purification procedures. This development addresses a serious need in the International Atomic Energy Agency’s (IAEA) Network of Analytical Laboratories (NWAL) to increase efficiency in the Bulk Analysis of Environmental Samples for Safeguards program with a method that allows unattended, overnight operation. In collaboration with Elemental Scientific Inc. (ESI), the prepFAST-MC2 was designed based on COTS equipment. It was modified for uranium/plutonium separations using renewable columns packed with Eichrom TEVA® and UTEVA® resins, with a chemical separation method based on the Oak Ridge National Laboratory (ORNL) NWAL chemical procedure. Initial verification of individual columns yielded small elution volumes with consistent elution profiles and good recovery. Combined column calibration demonstrated ample separation without crosscontamination of the eluent. Automated packing and unpacking of the built-in columns initially showed >15% deviation in resin loading by weight, which can lead to inconsistent separations. Optimization of the packing and unpacking methods led to a reduction in the variability of the packed resin to less than 5% daily. Based on the initial experience at ORNL with the prepFAST-MC2, a second-generation model, the prepFAST-SR, was built by ESI. The newly designed prepFAST-SR has several upgrades compared with the original prepFAST-MC2. The upgrades largely addressed issues with the reliability of unpacking the Eichrom columns and with ventilation to ensure adequate removal of acid vapors. Both systems are currently installed in the Ultra-Trace Forensics Science Center at ORNL. Separations of mixed uranium and plutonium samples containing certified reference materials were analyzed by multi-collector inductively coupled plasma mass spectrometry and yielded good results. Current efforts have demonstrated a wider applicability of the prepFAST system. Near-quantitative removal of metal interferences was achieved with the system in both uranium and plutonium separated fractions. Successful system validation was completed with several archived samples. Isotopic results from archived samples and certified reference materials were well within data quality limits for the IAEA NWAL. Additional COTS equipment has been evaluated for its potential to aid the prepFAST-SR system in reducing the time allotments and cleanroom infrastructure requirements for accurate separations. Overall, these efforts help ensure continued efficient and effective operation of the NWAL. Comparison of the amount of personnel time necessary for successful manual vs. automated chemical separations showed a significant decrease in hands-on time from 9.8 hours to 35 minutes for seven samples, respectively. This documented time savings and reduced labor translates to a significant cost savings per sample. Overall, the system will enable faster sample reporting times at reduced cost by limiting personnel hours dedicated to the chemical separation.

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

SU-F-I-78: Design of Diagnostic Multileaf Collimator Based On the Monte Carlo Simulation for Dose Reduction of Diagnostic Radiography

Purpose: A diagnostics Multileaf Collimator (MLC) was designed for diagnostic radiography dose reduction. Monte Carlo simulation was used to evaluate efficiency of shielding material for producing leaves of Multileaf collimator. Material & Methods: The general radiography unit (Rex-650R, Listem, Korea) was modeling with Monte Carlo simulation (MCNPX, LANL, USA) and we used SRS-78 program to calculate the energy spectrum of tube voltage (80, 100, 120 kVp). The shielding materials was SKD 11 alloy tool steel that is composed of 1.6% carbon(C), 0.4% silicon (Si), 0.6% manganese (Mn), 5% chromium (Cr), 1% molybdenum (Mo), and vanadium (V). The density of it was 7.89 g/m3. We simulated leafs diagnostic MLC using SKD 11 with general radiography unit. We calculated efficiency of diagnostic MLC using tally6 card of MCNPX depending on energy. Results: The diagnostic MLC consisted of 25 individual metal shielding leaves on both sides, with dimensions of 10 × 0.5 × 0.5 cm3. The leaves of MLC were controlled by motors positioned on both sides of the MLC. According to energy (tube voltage), the shielding efficiency of MLC in Monte Carlo simulation was 99% (80 kVp), 96% (100 kVp) and 93% (120 kVp). Conclusion: We certified efficiency of diagnostic MLC fabricated from SKD11 alloy tool steel. Based on the results, the diagnostic MLC was designed. We will make the diagnostic MLC for dose reduction of diagnostic radiography.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2020

Performing k eff Validation of As-Loaded Criticality Safety Calculations Using UNF-ST&DARDS: Applicable Experiment Selection

The general method for performing validation of as loaded criticality safety calculations using UNF ST&DARDS is presented in a paper by Clarity, which includes a description of the UNF-ST&DARDS system. Proof-of-principle analyses were performed in the summer of 2019 for MPC-32 dual purpose canisters (DPCs) containing pressurized water reactor (PWR) fuel assemblies. Summaries of these results are presented in this and a companion paper for this conference. The current paper describes the TSUNAMI-IP calculations performed to select applicable experiments for validation of 11 MPC-32 DPCs. The companion paper discusses the TSUNAMI-3D calculations used to generate sensitivity data to support the experiment selections discussed here. Experiment selection is based on the sensitivity/uncertainty (S/U) methods used to validate criticality safety calculations of as-loaded DPCs containing pressurized water reactor (PWR) spent nuclear fuel (SNF). This process has been demonstrated and is summarized in this paper. The approach is similar to that used in NUREG/CR-7109, which provides an approach for validation of PWR burnup credit (BUC), including major and minor actinides and major fission products. The premise of S/U-based validation is that applicable experiments—those having a similar bias to a given application system—will have similar sensitivities for each isotope and reaction in the two systems. It is assumed that cross sections with larger uncertainties are more likely to contain data errors which contribute to the bias. The integral index c k thus propagates the system sensitivities with the nuclear covariance data to calculate a correlation coefficient representing the similarity of the two systems. In this work, a c k value of 0.8 or higher is interpreted as identifying an experiment with sufficient similarity for use in validation. This paper presents a brief summary of the characteristics of the 11 MPC-32 DPCs used in the proof of-principle analysis for as-loaded criticality safety calculation validation and an overview of the critical experiment suite with which each of these DPC models was compared. A summary and discussion of c k results is also presented, followed by conclusions and a discussion of future work to be performed for validation of UNF ST&DARDS as-loaded criticality safety calculations.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Neptunium mononitride as a target material for Pu-238 production

Deep space exploration requires specialized sources for both thermal and power applications. Radioactive decay heat of plutonium-238 (238Pu) provides these sources in the form of radioisotope thermoelectric generators (RTGs). The 238 Pu is produced via neutron capture reaction involving neptunium-237 ( 237 Np) target material. Continual optimization of 237 Np target materials and evaluation of potential alternative targets for production of 238 Pu RTGs are advantageous for meeting ongoing space power system resource requirements. Current production of 238 Pu for RTGs for the United States space program utilizes neptunium dioxide ( 237 NpO 2 ) targets; however, the use of neptunium mononitride ( 237 NpN) presents an opportunity to increase the mass of 237 Np per target compared to the dioxide form, as well as increase the thermal conductivity of the target. To assess the viability of a 237 NpN target material, the material chemistry must be thoroughly evaluated, including synthesis methods and dissolution and reprocessing schemes. This review presents a summary of synthesis pathways for 237 NpN based on published literature on actinide mononitrides. Specific literature on 237 NpN is limited, necessitating evaluation of other actinide systems to gather parallels. This suggests a need for additional experimental studies on 237 NpN. A particular limitation in the existing literature is a lack of information on the differences in material characteristics, such as morphology, particle size, and trace chemical impurities, as a function of synthesis method. These parameters may affect subsequent reactor performance or dissolution of irradiated targets. The evaluation of existing literature is presented with a focus on the efficacy of 237 NpN targets for 238 Pu production.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2013

Justification to Disregard U-234 Mass in Certain Uranium Systems

The ANSI/ANS-8.1-1998 Standard regarding nuclear criticality safety in operations with fissionable material outside reactors states certain guidelines for dealing with uranium isotopes when setting mass limits. In ANSI/ANS-8.1-1998 235 U limits apply to mixtures “with 234 U, 236 U, or 238 U, provided that 234 U is considered to be 235U” in computing mass. In this evaluation calculations coupled with benchmark data show a superior way of handling the other common uranium isotopes when setting limits for 235 U. Currently 234 U mass must be considered as 235 U mass in order to account for the fact 234 U has a k ∞ = 1.70 meaning criticality can be reached with a finite amount of 234 U mass. 236 U and 238 U have k ∞ ’s equal to 0.71 and 0.31 meaning they remain subcritical in infinite amounts and thus no mass limits are required. In fresh uranium (un-irradiated or reprocessed) enriched to 94% 235 U by weight, bounding ratios of 234 U, 236U, and 238U exist due to the enrichment process. Utilizing these ratios it is shown that 234 U accounts for less than 0.20 of the total mass fraction of 234 U, 236 U, and 238 U and the remaining 0.80 mass fraction is in a ratio of at least 1:1 for 238 U: 236 U. Creating models conservatively based on these two material limits, the resulting k ∞ = 0.91367 ± 0.00208 suggests a mass limit for 234 U is no longer required.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Transuranic Waste Processing Center (TWPC) Legacy Tank RH-TRU Sludge Processing and Compliance Strategy - 13255

The U.S. Department of Energy (DOE) needs to safely and efficiently treat its 'legacy' transuranic (TRU) waste and mixed low-level waste (LLW) from past research and defense activities at the Oak Ridge National Laboratory (ORNL) so that the waste is prepared for safe and secure disposal. The TWPC operates an Environmental Management (EM) waste processing facility on the Oak Ridge Reservation (ORR). The TWPC is classified as a Hazard Category 2, non-reactor nuclear facility. This facility receives, treats, and packages low-level waste and TRU waste stored at various facilities on the ORR for eventual off-site disposal at various DOE sites and commercial facilities. The Remote Handled TRU Waste Sludge held in the Melton Valley Storage Tanks (MVSTs) was produced as a result of the collection, treatment, and storage of liquid radioactive waste originating from the ORNL radiochemical processing and radioisotope production programs. The MVSTs contain most of the associated waste from the Gunite and Associated Tanks (GAAT) in the ORNL's Tank Farms in Bethel Valley and the sludge (SL) and associated waste from the Old Hydro-fracture Facility tanks and other Federal Facility Agreement (FFA) tanks. The SL Processing Facility Build-outs (SL-PFB) Project is integral to the EM cleanup mission at ORNL and is being accelerated by DOE to meet updated regulatory commitments in the Site Treatment Plan. To meet these commitments a Baseline (BL) Change Proposal (BCP) is being submitted to provide continued spending authority as the project re-initiation extends across fiscal year 2012 (FY2012) into fiscal year 2013. Future waste from the ORNL Building 3019 U-233 Disposition project, in the form of U-233 dissolved in nitric acid and water, down-blended with depleted uranyl nitrate solution is also expected to be transferred to the 7856 MVST Annex Facility (formally the Capacity Increase Project (CIP) Tanks) for co-processing with the SL. The SL-PFB project will construct and install the necessary integrated systems to process the accumulated MVST Facilities SL inventory at the TWPC thus enabling safe and effective disposal of the waste. This BCP does not include work to support current MVST Facility Surveillance and Maintenance programs or the ORNL Building 3019 U-233 Disposition project, since they are not currently part of the TWPC prime contract. The purpose of the environmental compliance strategy is to identify the environmental permits and other required regulatory documents necessary for the construction and operation of the SL- PFB at the TWPC, Oak Ridge, TN. The permits and other regulatory documents identified are necessary to comply with the environmental laws and regulations of DOE Orders, and other requirements documented in the SL-PFB, Safety Design Strategy (SDS), SL-A-AD-002, R0 draft, and the Systems, Function and Requirements Document (SFRD), SL-X-AD-002, R1 draft. This compliance strategy is considered a 'living strategy' and it is anticipated that it will be revised as design progresses and more detail is known. The design basis on which this environmental permitting and compliance strategy is based is the Wastren Advantage, Inc., (WAI), TWPC, SL-PFB (WAI-BL-B.01.06) baseline. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Critical and Californium Source-Driven Noise Analysis Subcritical Measurements with an Unreflected Cylindrical Tank of Mixed Uranium-Plutonium Nitrate Solution

The reported experiments were performed by Oak Ridge National Laboratory (ORNL) at the Battelle Northwest Laboratory’s (now Pacific Northwest National Laboratory) critical experiments facility at Hanford, Washington in 1981 and used 16 days of critical facility time, not including 10 days for setup and removal of ORNL equipment. These measurements were to assess the capability of the Cf source-driven noise analysis (CSDNA) method to measure the subcriticality (keff) of mixed U-Pu nitrate solutions. In addition to the CSNDA measurements, measurements were also performed near delayed criticality where CSDNA measurements cannot be performed. This report documents the experiments that were not reported at that time by presenting the ORNL experimental results and any online analysis performed during and shortly after the measurements. The mixed nitrate solution had a U concentration at 188 grams per liter (g/L), a Pu concentration of 280 g/L, free acid normality of 2.80, H ion molarity of 5.9 and a specific gravity of 1.754 g/cm 3 , a 240 Pu isotopic content of 7.981 wt. %, and a 235 U isotopic content of 0.724 wt. %. The stainless-steel tank for the solution had an inside diameter of 35.38 cm, an outside diameter of 35.53 cm, a height of 56.72 cm, and bottom thickness of 0.9525 cm. A Zircaloy pipe with a 3.1496 cm outside diameter, a 2.7788 cm inside diameter, and bottom thickness of 0.635 cm was available for insertion of the Cf source in the center of the fissile solution. The Cf source was also located at the outside surface of the tank (solution height varied from 10 to 53 cm) and in the center of the solution (solution height varied from 10 to 60.7 cm). The CSDNA measurements were not analyzed online to determine the subcritical neutron multiplication factors. At all subcritical states, the break frequency noise analysis data was fitted to obtain the prompt neutron decay constant. The neutron multiplication factors were determined for the two configurations of the measurements near delayed criticality. The subcritical neutron multiplication factors from the CNSDA measurements can be obtained with further analysis. However, the near delayed critical configuration, the prompt neutron decay constants, the count rates, and the measured cross and auto power spectral densities can be calculated directly for benchmarking. Much of data presented in this report are from ORNL notes—not in the ORNL logbooks. For the final benchmark analysis, the data from the Battelle Northwest Laboratory (which operated the critical facility in 1981) critical facility logbook should be consulted and be incorporated where appropriate. The purpose of this report is to document the experimental information for the measurements performed so that at a later date researchers could perform the required uncertainty and calculational analyses and documentation to use these data for an International Criticality Safety Benchmark Program (ICSBEP) or Nuclear Energy Agency benchmark. The data from these measurements are available from the ORNL Records Management Services Department, and the logbook is available from ICSBEP at Idaho National Laboratory. Preparation of the present report is part of a larger cooperative effort between Idaho National Laboratory (INL) and Oak Ridge National Laboratory (ORNL) to document more than 15 undocumented critical and subcritical experiments enumerated in ORNL/TM-2019/18 and performed by ORNL at ORCEF and other USDOE critical experiments facilities using more than 500 operational days of critical facility time.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Data from: "Warming of alpine tundra enhances belowground production and shifts community towards resource acquisition traits"

This archive contains data used to draw conclusions in “Warming of alpine tundra enhances belowground production and shifts community towards resource acquisition traits”, by Yang et al. 2020. Data were collected on Niwot Ridge, in an alpine meadow within the Alpine Treeline Warming Experiment (ATWE) field sites in Colorado, USA. Samples were also processed in the U.S. Geological Survey Forest and Rangeland Ecosystem Science Center, in Boise, Idaho. File formats in this archive include comma-separated values (.csv), portable document format (.pdf), Microsoft Excel (.xlsx), and two types of geospatial files: keyhole markup language (.kml), and ESRI shapefiles (.shp). Leaf scans are .jpg images, and root scans are .tiff/.tif images.The .csv files can be opened using R, Microsoft Excel, or any simple text-editing software such as TextEdit and Notepad. Microsoft Excel files can be opened using Microsoft Excel, and .pdf files can be opened with Adobe Acrobat Reader, Preview, or other compatible programs. Scanned images can be opened using any photo and/or picture viewing software.The .kml file can be opened using Google Earth and Google Maps, and the shapefiles can be opened by any programs compatible with shapefiles, such as the ArcGIS Desktop suite, and QGIS.------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------Measures of belowground net primary productivity (BNPP) are required to understand whether aboveground net primary production (ANPP) changes reflect changes in allocation or are indicative of a whole plant NPP response. Plant functional traits provide a key way to scale from the individual plant to the community level, and provide insight into drivers of NPP responses to environmental change. We used infrared heaters to warm an alpine plant community at Niwot Ridge, Colorado, and applied supplemental water to compensate for soil water loss induced by warming. We measured ANPP, BNPP, and leaf and root functional traits across treatments after 5 years of continuous warming. Community-level ANPP and total NPP (ANPP + BNPP) did not respond to heating or watering, but BNPP increased in response to heating. Heating decreased community-level leaf dry matter content and increased total root length, indicating a shift in strategy from resource conservation to acquisition in response to warming.

13C/12C isotope ratio↗
DOE OSTI2014

SU-E-T-323: The FLUKA Monte Carlo Code in Ion Beam Therapy

Purpose: Monte Carlo (MC) codes are increasingly used in the ion beam therapy community due to their detailed description of radiation transport and interaction with matter. The suitability of a MC code demands accurate and reliable physical models for the transport and the interaction of all components of the mixed radiation field. This contribution will address an overview of the recent developments in the FLUKA code oriented to its application in ion beam therapy. Methods: FLUKA is a general purpose MC code which allows the calculations of particle transport and interactions with matter, covering an extended range of applications. The user can manage the code through a graphic interface (FLAIR) developed using the Python programming language. Results: This contribution will present recent refinements in the description of the ionization processes and comparisons between FLUKA results and experimental data of ion beam therapy facilities. Moreover, several validations of the largely improved FLUKA nuclear models for imaging application to treatment monitoring will be shown. The complex calculation of prompt gamma ray emission compares favorably with experimental data and can be considered adequate for the intended applications. New features in the modeling of proton induced nuclear interactions also provide reliable cross section predictions for the production of radionuclides. Of great interest for the community are the developments introduced in FLAIR. The most recent efforts concern the capability of importing computed-tomography images in order to build automatically patient geometries and the implementation of different types of existing positron-emission-tomography scanner devices for imaging applications. Conclusion: The FLUA code has been already chosen as reference MC code in many ion beam therapy centers, and is being continuously improved in order to match the needs of ion beam therapy applications. Parts of this work have been supported by the European FP7 project ENVISION (grant agreement no. 241851)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2013

Water Management Strategies for Improved Coalbed Methane Production in the Black Warrior Basin

The modern coalbed methane industry was born in the Black Warrior Basin of Alabama and has to date produced more than 2.6 trillion cubic feet of gas and 1.6 billion barrels of water. The coalbed gas industry in this area is dependent on instream disposal of co-produced water, which ranges from nearly potable sodium-bicarbonate water to hypersaline sodium-chloride water. This study employed diverse analytical methods to characterize water chemistry in light of the regional geologic framework and to evaluate the full range of water management options for the Black Warrior coalbed methane industry. Results reveal strong interrelationships among regional geology, water chemistry, and gas chemistry. Coalbed methane is produced from multiple coal seams in Pennsylvanian-age strata of the Pottsville Coal Interval, in which water chemistry is influenced by a structurally controlled meteoric recharge area along the southeastern margin of the basin. The most important constituents of concern in the produced water include chlorides, ammonia compounds, and organic substances. Regional mapping and statistical analysis indicate that the concentrations of most ionic compounds, metallic substances, and nonmetallic substances correlate with total dissolved solids and chlorides. Gas is effectively produced at pipeline quality, and the only significant impurity is N{sub 2}. Geochemical analysis indicates that the gas is of mixed thermogenic-biogenic origin. Stable isotopic analysis of produced gas and calcite vein fills indicates that widespread late-stage microbial methanogenesis occurred primarily along a CO{sub 2} reduction metabolic pathway. Organic compounds in the produced water appear to have helped sustain microbial communities. Ammonia and ammonium levels increase with total dissolved solids content and appear to have played a role in late-stage microbial methanogenesis and the generation of N{sub 2}. Gas production tends to decline exponentially, whereas water production tends to decline hyperbolically. Hyperbolic decline indicates that water volume is of greatest concern early in the life of a coalbed methane project. Regional mapping indicates that gas production is controlled primarily by the ability to depressurize permeable coal seams that are natively within the steep part of the adsorption isotherm. Water production is greatest within the freshwater intrusion and below thick Cretaceous cover strata and is least in areas of underpressure. Water management strategies include instream disposal, which can be applied effectively in most parts of the basin. Deep disposal may be applicable locally, particularly where high salinity limits the ability to dispose into streams. Artificial wetlands show promise for the management of saline water, especially where the reservoir yield is limited. Beneficial use options include municipal water supply, agricultural use, and industrial use. The water may be of use to an inland shrimp farming industry, which is active around the southwestern coalbed methane fields. The best opportunities for beneficial use are reuse of water by the coalbed methane industry for drilling and hydraulic fracturing. This research has further highlighted opportunities for additional research on treatment efficiency, the origin of nitrogen compounds, organic geochemistry, biogenic gas generation, flow modeling, and computer simulation. Results of this study are being disseminated through a vigorous technology transfer program that includes web resources, numerous presentations to stakeholders, and a variety of technical publications.

01 COAL, LIGNITE, AND PEAT↗
DOE OSTI2015

MO-D-213-07: RadShield: Semi- Automated Calculation of Air Kerma Rate and Barrier Thickness

Purpose: To develop the first Java-based semi-automated calculation program intended to aid professional radiation shielding design. Air-kerma rate and barrier thickness calculations are performed by implementing NCRP Report 147 formalism into a Graphical User Interface (GUI). The ultimate aim of this newly created software package is to reduce errors and improve radiographic and fluoroscopic room designs over manual approaches. Methods: Floor plans are first imported as images into the RadShield software program. These plans serve as templates for drawing barriers, occupied regions and x-ray tube locations. We have implemented sub-GUIs that allow the specification in regions and equipment for occupancy factors, design goals, number of patients, primary beam directions, source-to-patient distances and workload distributions. Once the user enters the above parameters, the program automatically calculates air-kerma rate at sampled points beyond all barriers. For each sample point, a corresponding minimum barrier thickness is calculated to meet the design goal. RadShield allows control over preshielding, sample point location and material types. Results: A functional GUI package was developed and tested. Examination of sample walls and source distributions yields a maximum percent difference of less than 0.1% between hand-calculated air-kerma rates and RadShield. Conclusion: The initial results demonstrated that RadShield calculates air-kerma rates and required barrier thicknesses with reliable accuracy and can be used to make radiation shielding design more efficient and accurate. This newly developed approach differs from conventional calculation methods in that it finds air-kerma rates and thickness requirements for many points outside the barriers, stores the information and selects the largest value needed to comply with NCRP Report 147 design goals. Floor plans, parameters, designs and reports can be saved and accessed later for modification and recalculation. We have confirmed that this software accurately calculates air-kerma rates and required barrier thicknesses for diagnostic radiography and fluoroscopic rooms.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2022

Corrosion of U233-Doped Uranium Oxide using Microfluidics Methods

The aim of work this year has been to investigate the role of alpha (α)-radiation and hydrogen (H 2 ) on the corrosion of uranium oxide (UO 2 ) using a microfluidic device. The microfluidic device, termed the Particle-Attached Microfluidic Electrochemical Cell, (PAMEC), enables monitoring of the UO 2 electrochemical corrosion potential (E corr ) under different environments, including de-aerated conditions and in the presence of dissolved H 2 . The Si 3 N 4 window allows us to study morphological and chemical changes under an electron microscope. We have previously demonstrated that the PAMEC matches the results from bulk electrochemical tests with UO 2 [1]. We used the high specific activity uranium (U) isotope, 233 U, (t 1/2 = 160,000 years) incorporated into UO 2 , to generate a localized a-field. The objective of the experiments were to mimic the radiation environment that would be experience at the surface of aged spent nuclear fuel (SNF) during long-term geologic disposal under anoxic conditions. Wittman et al. [2] predicted that in the presence of a pure a-radiation field and under H 2 conditions, the concentration of the radiolytic oxidant H 2 O 2 would be suppressed or even eliminated. In a UO 2 corrosion experiment this would be exhibited through a lowering of the measured UO 2 corrosion potential compared to identical conditions in the absence of dissolved H 2 . We found that the predictions of Wittman and co-workers were supported and that the corrosion potential of the 233 U-doped UO 2 in solution lowered with presence of H 2 gas and increased in the absence of H 2 , when under anoxic conditions. The PAMEC experiments indicate that H 2 O 2 has been eliminated in a solution sparged with H 2 while exposed to an a-radiation field. The corrosion potential of the a-doped 233 U(10%)- 238 UO 2 in a solution sparged with Ar/H 2 matched the corrosion potential of 238 UO 2 in solution sparged with air. This clearly demonstrated that the H 2 O 2 had been eliminated and that the only oxidant present was O 2 in this system in complete agreement with the modeling results of Wittman et al. These results point to the need to improve the Fuel Matrix Degradation (FMD) Process Model training data set that is being used in the FMD surrogate model that is being developed for the repository program. The incorporation of realistic radiation chemistry will improve the scientific basis for the FMD Model.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗