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All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Structural, Criticality, and Radiation Dose Calculations to support SNF Loading into a DOE Standard Canister

The DOE Standard Canister Demonstration Project includes the development of an internal support structure (ISS) for the 4.6-m long and 45.7-cm diameter canister. This study presents structural evaluations, criticality safety assessments, and dose rate calculations conducted within the scope of the ISS design process to support smooth canister loading operations and to ensure safe storage, transportation, and disposal of Peach Bottom 1 Core II (PB2) and Fort St. Vrain (FSV) SNF, currently stored at the CPP-603 facility of the INL. The ISS includes a 316L stainless steel basket that holds twelve PB2 SNF rods. Further, six individual, 78.7-cm long, 316L stainless steel columns are equally spaced and welded to the inner canister wall at the lower end of the shell. These columns represent the FSV basket and can hold one FSV SNF element. An A92014 T6 aluminum spacer disc is bolted to the bottom plate of the PB2 basket to vertically restrain the FSV SNF element after the PB2 basket is placed inside the canister on top of the FSV basket. The structural evaluations of the ISS followed applicable ASME BPVC.III.3 guidelines and included finite element (FE) analyses of the PB2 basket structure; analyses of welds and bolds; buckling analyses of selected components, and acceptability assessments of the expected basket deformations under loading operations. The criticality safety assessments used the Monte Carlo N-Particle (MCNP) software architecture version 6.2 including ENDF/B-V continuous energy cross-section libraries, considering intact SNF in a single storage overpack or two multi-storage overpack configurations, and intact or failed SNF configured for disposal. The dose rate computations are based on source terms taken from the DOE Spent Fuel Database. The isotopic composition was decay corrected for the year 2022 using the ORIGEN module in the SCALE suite. A 19-group photon spectrum and 27-group neutron-source spectra were generated and used in MCNP to calculate estimated dose-equivalent rates, both on DOE Standard Canister contact and at a radial distance of from the canister surface. The results of this study indicate a structurally sound system that can uphold its criticality safety functions throughout its intended operational phases. Further, they increase confidence that sufficient radiological protection is technically achievable.

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

Structure and transport of aqueous electrolytes: From simple halides to radionuclide ions

Molecular simulations are used to compare the structure and dynamics of conventional and radioactive aqueous electrolytes: chloride solutions with sodium, potassium, cesium, calcium, and strontium. The study of Cs{sup +} and Sr{sup 2+} is important because these radioactive ions can be extremely harmful and are often confused by living organisms for K{sup +} and Ca{sup 2+}, respectively. Na{sup +}, Ca{sup 2+}, and Sr{sup 2+} are strongly bonded to their hydration shell because of their large charge density. We find that the water molecules in the first hydration shell around Na{sup +} form hydrogen bonds between each other, whereas molecules in the first hydration shell around Ca{sup 2+} and Sr{sup 2+} predominantly form hydrogen bonds with water molecules in the second shell. In contrast to these three ions, K{sup +} and Cs{sup +} have low charge densities so that they are weakly bonded to their hydration shell. Overall, the structural differences between Ca{sup 2+} and Sr{sup 2+} are small, but the difference between their coordination numbers relative to their surface areas could potentially be used to separate these ions. Moreover, the different decays of the velocity-autocorrelation functions corresponding to these ions indicates that the difference in mass could be used to separate these cations. In this work, we also propose a new definition of the pairing time that is easy to calculate and of physical significance regardless of the problem at hand.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2025

Ultra-sensitive radon assay using an electrostatic chamber in a recirculating system

Rare event searches such as neutrinoless double beta decay and Weakly Interacting Massive Particle detection require ultra-low background detectors. Radon contamination is a significant challenge for these experiments, which employ highly sensitive radon assay techniques to identify and select low-emission materials. This work presents the development of ultra-sensitive electrostatic chamber (ESC) instruments designed to measure radon emanation in a recirculating gas loop, for future lower background experiments. Unlike traditional methods that separate emanation and detection steps, this system allows continuous radon transport and detection. This is made possible with a custom-built recirculation pump. A Python-based analysis framework, PyDAn, was developed to process and fit time-dependent radon decay data. Radon emanation rates are given for various materials measured with this instrument. A radon source of known activity provides an absolute calibration, enabling statistically-limited minimal detectable activities of 20 uBq. These devices are powerful tools for screening materials in the development of low-background particle physics experiments.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Measurement of Mass-Independent Perturbations of Cadmium in Nuclear Debris Samples

The isotopic composition of cadmium in nuclear debris was measured by multi-collector inductively coupled plasma-mass spectrometry (MC-ICP-MS). Mass-independent perturbations due to fission product decay on the isotopes of Cd were observed. Relative fission yields for masses 111, 112, 114 and 116 were determined in each sample of nuclear debris and compared with reported values found in irradiated plutonium and uranium fuels. Mass spectrometry measurements of the valley fission product endpoints are extremely challenging due to the low cumulative fission yields and small amounts of each isotope that are produced. To overcome these challenges, a new purification method for Cd was developed and validated by examination of a number of geological reference materials.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

A New Era of Discovery: The 2023 Long-Range Plan for Nuclear Science (V.1.2)

Nuclear science is the investigation of how protons and neutrons are formed from elementary particles and how the forces between those particles produce both nuclei and the vast variety of nuclear phenomena that occur in the universe. It has evolved into a broad field that addresses profound scientific questions: Where does the mass of visible matter come from? How do stars ignite, live, and die? How do nuclei illuminate the search for new laws of nature? This science points the way to using nuclei to build new technologies that benefit society. The 2015 Nobel Prize in physics was shared by nuclear physicists Art McDonald and Takaaki Kajita for the discovery of neutrino oscillations, which confirmed that neutrinos have mass. Our progress on big questions like this one since 2015 has been remarkable owing to new experimental tools, theoretical breakthroughs, powerful computational techniques, and the talented people who make these innovations possible. Focusing on these new tools, the Facility for Rare Isotope Beams (FRIB) at Michigan State University is already producing exciting results on decays of never-before-produced isotopes a year after it was completed on time and on budget. The energy upgrade of the Continuous Electron Beam Accelerator Facility (CEBAF) at the Thomas Jefferson National Accelerator Facility (Jefferson Lab) was also completed on schedule and on budget—new data from this facility are revealing the spectrum, structure, and dynamics of protons, neutrons, nuclei, and mesons. On the theory front, we can now calculate the distribution of quarks inside the proton from first principles. The implementation of artificial intelligence (AI) and machine learning (ML) techniques has led to improved data analysis and increased efficiency in running experiments and theoretical calculations. The impact of nuclear science goes beyond expanding the frontiers of knowledge about matter in the universe. We simultaneously develop a STEM work force that advances the security, technology, health, and wealth of our nation. Some connections are obvious. Expert scientists trained to work with radioactive nuclei are in demand in nuclear security arenas and are highly sought after by various government agencies and private industries. Graduate students and postdoctoral fellows (postdocs) obtain extensive computational, modeling, and data science skills that are similarly in high demand. Less obvious but equally important is the connection between these trained scientists and success in other professions, including medicine, energy, and entrepreneurial pursuits. The workforce that enables discovery in nuclear science also makes breakthroughs in technologies with tremendous impact on the nation’s economic advancement.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Structural, Criticality, and Radiation Dose Calculations to Support SNF Loading into a DOE Standard Canister

The DOE Standard Canister Demonstration Project includes the development of an internal support structure (ISS) for the 4.6-m-long and 45.7-cm-diameter canister. This study presents structural evaluations, criticality safety assessments, and dose rate calculations conducted within the scope of the ISS design process to support smooth canister loading operations and to ensure safe storage, transportation, and disposal of Peach Bottom 1 Core II (PB2) and Fort St. Vrain (FSV) SNF, currently stored at INL’s CPP-603 facility. The ISS includes a 316L stainless-steel basket that holds 12 PB2 SNF rods. The PB2 basket rests on top of six individual, 78.7-cm-long, 316L stainless-steel columns that are equally spaced and welded to the inner canister wall at the lower end of the shell. These columns represent the FSV basket and can hold one FSV SNF element. An A92014 T6 aluminum spacer disc is bolted to the bottom plate of the PB2 basket to vertically restrain the FSV SNF element after the PB2 basket is placed inside the canister above the FSV basket. The structural evaluations of the ISS followed applicable ASME BPVC.III.3 guidelines and included finite element (FE) analyses of the PB2 basket structure; analyses of welds and bolds; buckling analyses of selected components, and acceptability assessments of the expected basket deformations under loading operations. The criticality safety assessments used the Monte Carlo N-Particle (MCNP) software architecture Version 6.2, including ENDF/B-V continuous-energy cross-section libraries, considering intact SNF in a single storage overpack or two multistorage overpack configurations and intact or failed SNF configured for disposal. The dose rate computations are based on source terms taken from the DOE Spent Fuel Database. The isotopic composition was decay corrected for the year 2022 using the ORIGEN module in the SCALE suite. A 19-group photon spectrum and a 27-group neutron-source spectrum were generated and used in MCNP to calculate estimated dose-equivalent rates, both on DOE Standard Canister contact and at a radial distance of 1 m from the canister surface. The results of this study indicate a structurally sound system that can uphold its criticality safety functions throughout its intended operational phases. Furthermore, this study provides confidence that sufficient radiological protection is technically achievable.

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

ARIES Blend Lot Book Data Analysis

The Advanced Recovery Integrated Extraction System (ARIES) Program being executed at Los Alamos TA-55 produces unpolished plutonium oxide powder from pit disassembly in batches of material referred to as a blend lot. Each blend lot is sampled and analyzed by Los Alamos group C-AAC (Chemistry Division, Actinide Analytical Chemistry Group) for radioactive and nonradioactive constituents contained in the material. A blend lot, which nominally consists of 12 kg(Pu), is then packaged into three DOE-STD-3013 containers that are analyzed by nondestructive analysis (NDA) to determine the fraction and total amount of plutonium mass relative to impure plutonium oxide. The analytical and NDA data are presented in a Blend Lot Book (BLB). The data are used to confirm that a blend lot grouping of three cans meets the requirements for off-site shipment specified in the formal interface control document (ICD). The analytical and NDA data from 49 blend lots were reviewed to provide summary-level information on the constituents present in the blend lots. The results provide an overall perspective in historical amounts for constituent averages over all blend lots and similar information in the variation between constituent levels. Note that to properly compare the can contents, it would be most appropriate to decay all of cans’ constituents to a particular date. Previously, this analysis only examined statistical data as determined at the time of blend lot chemical analysis. The chemical analysis dates span the period from June 2011 to May 2015 (4 years), which is 28% of one Pu-241 half-life. Therefore, the approximate nature of the aggregate information must be properly understood in this light. However, to address this issue, a new table is included in this revision that decays the 48 blend lots to a single date.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Overview of Salt Repository Research and Development for Spent Nuclear Fuel and High-level Nuclear Waste in the United States - 20307

This paper summarizes the current US Department of Energy Office of Nuclear Energy's (DOE-NE) work towards developing and executing a research and development program that addresses both scientific and technical issues related to long-term disposal of spent nuclear fuel (SNF) and high-level waste (HLW) in a hypothetical bedded salt based geological repository. A primary goal of the program is to create a generic Geologic Disposal Safety Assessment (GDSA) Framework that can be used to help guide decisions on siting a possible future bedded salt repository. The generic GDSA work includes analysis of the impacts of heat generation caused by decay of short-lived radionuclides. We report progress in four primary areas. First, we discuss the development and recent modifications of a research and development road-map. Second, we briefly describe an experimental approach to better understand thermal processes in salt. Third, we highlight collaborations with the international research community that leverage salt-based repository science around the world. Finally, we discuss how our findings are being used to aid in the development of a generic safety assessment for a bedded salt repository containing SNF and HLW. (authors)

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

MO-FG-BRA-03: A Monte-Carlo Study of Cellular Dosimetry of Radioactive Gold-Palladium Nanoparticles

Purpose: Radioactive gold-palladium nanoparticles ({sup 103}Pd:Pd@Au NPs) are being developed for prostate cancer brachytherapy. Photons emitted by the radioisotope palladium (photon energy: 20.1 and 23.0 keV), interacting with gold-coating of NPs, lead to enhanced energy distribution in nucleus. Here, a simple cellular model was studied using detailed track-structure method. Methods: Geant4-DNA was used with auger electrons enabled. Biological cell was modeled as a sphere of radius r=5 µm that were immersed in a fluid containing large number of NPs at different concentrations (S=1, 2.15, 5.1, 17.2 mg-Au/g-H2O). Nucleus was modeled as a concentric sphere (r=3µm). Thickness of gold-coating on {sup 103}Pd core was 15nm, 20nm and 25nm, respectively. A scenario of NP diffusion was investigated, where S=5.1 mg-Au/g-H2O outside cell and S=1 mg-Au/g-H2O in cytoplasm. 10{sup 10} {sup 103}Pd decays were simulated for each combination of NP concentration and gold-coating. Results: A uniform increase in energy deposition (Edep) is observed in cell nucleus and the energy enhancement ratio (EER) is 1.16, 1.22 and 1.3 for 15nm, 20nm and 25nm of gold -coatings, respectively. Edep at the center of nucleus is increased by a factor of 1.47, 2.51 and 5.54 when the NP concentration in the cytoplasm increases from 1 mg-Au/g-H2O to 2.15, 5.10 and 17.2 mg-Au/g-H2O, respectively. When NPs diffuse into cytoplasm, the mean value of Edep in nucleus increases from 0.42 to 1.13 MeV per 10{sup 9} decays (GBq-Second) of {sup 103}Pd and the maximum value increases from 0.54 to 2.5 MeV per GBq-Second. Conclusion: These results suggest that {sup 103}Pd:Pd@Au NPs constitute a promising nanotherapeutic agent. Ongoing studies use transmission electron microscopy (TEM) images of prostate cancer.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2012

Long-Term Assessment of Critical Radionuclides and Associated Environmental Media at the Savannah River Site

During the operational history of the Savannah River Site (SRS), many different radionuclides have been released from site facilities. However, only a relatively small number of the released radionuclides have been significant contributors to doses and risks to the public. At SRS dose and risk assessments indicate tritium oxide in air and surface water, and Cs-137 in fish and deer have been, and continue to be, the critical radionuclides and pathways. In this assessment, indepth statistical analyses of the long-term trends of tritium oxide in atmospheric and surface water releases and Cs-137 concentrations in fish and deer are provided. Correlations also are provided with 1) operational changes and improvements, 2) geopolitical events (Cold War cessation), and 3) recent environmental remediation projects and decommissioning of excess facilities. For example, environmental remediation of the F- and H-Area Seepage Basins and the Solid Waste Disposal Facility have resulted in a measurable impact on the tritium oxide flux to the onsite Fourmile Branch stream. Airborne releases of tritium oxide have been greatly affected by operational improvements and the end of the Cold War in 1991. However, the effects of SRS environmental remediation activities and ongoing tritium operations on tritium concentrations in the environment are measurable and documented in this assessment. Controlled hunts of deer and feral hogs are conducted at SRS for approximately six weeks each year. Before any harvested animal is released to a hunter, SRS personnel perform a field analysis for Cs-137 concentrations to ensure the hunter's dose does not exceed the SRS administrative game limit of 0.22 millisievert (22 mrem). However, most of the Cs-137 found in SRS onsite deer is not from site operations but is from nuclear weapons testing fallout from the 1950's and early 1960's. This legacy source term is trended in the SRS deer, and an assessment of the ''effective'' half-life of Cs-137 in deer (including the physical decay half-life and the environmental dispersion half-life) is provided. The ''creek mouth'' fisherman is the next most critical pathway at SRS. On an annual basis, three species of fish (panfish, catfish, and bass) are sampled from the mouths of the five SRS streams. Three composites of up to five fish of each species are analyzed from each sampling location. Long-term trending of the Cs-137 concentrations in fish and the subsequent doses from consumption of SRS fish is provided.

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

Vacuum drying of spent fuel storage canisters/casks - 'just how much water could be left in there, anyway?' - 15233

The process of moving spent fuel assemblies from wet storage in the spent fuel pool to dry storage in an inert gas environment within a canister or cask typically involves two main stages. In the first stage, the bulk of the water is removed by a draining or siphoning operation, conducted with pressure inside the canister maintained at 1.3 to 2 atm. This process removes most of the water in the cask cavity, but residual water can be left behind, in three main regions. The largest region, typically, is the bottom of the cavity below the end of the siphon tube. A second general class of locations, also located at the bottom of the cask cavity, consists of the 'dashpots' at the ends of the various non-fuel tubular structures within the fuel assemblies. For PWR assemblies, these are the control rod guide sleeve tubes and instrumentation thimbles. For BWR assemblies, these are the water rods of various designs and control blade ends in some configurations. A third class of locations consists of porous material within the cavity that can absorb water, such as thin ceramic neutron poison plates, and ostensibly intact fuel rods that nevertheless might have pinholes or fine cracks in the cladding that could allow pool water to creep inside the fuel rod. To remove residual water from these locations, the second stage of drying operations utilizes vacuum drying, by reducing the pressure sufficiently to convert the water to vapor that can then be pumped out of the canister cavity. In this operation, the cask cavity pressure is slowly reduced from the nominal condition of (1.3-2.0 atm (1000-1520 torr), to a final target pressure below 3 torr (0.004 atm). This is typically accomplished in three to five stages, with prescribed 'hold' points defined in procedures or in the Technical Specifications, to preclude sudden pressure spikes due to flashing of water to steam, or development of large local temperature gradients in the fuel rods or other cask internal components. This paper discusses the various locations in the system where residual water may 'hang up' in the cask, and presents a discussion of the physical basis for the effectiveness of vacuum drying for removing it. Results and conclusions based on realistic qualitative and quantitative evaluations of the transient heat transfer behavior of the system are presented, based on detailed modeling of a range of spent fuel canister designs and decay heat loads, most notably the 24P Dry Shielded Canisters (DSCs) stored at the Rancho Seco Independent Spent Fuel Storage Installation (ISFSI), the NAC-MPC modules at Maine Yankee, the 24P DSCs stored at Calvert Cliffs, and the Holtec HI-STORM modules at Hope Creek. The evaluations presented here show that ice formation within a spent fuel canister is highly unlikely, due to the availability of a large thermal source term in the form of the decay heat load of the spent fuel. The formation of ice is an operational concern, since it mainly affects operation of equipment external to the canister (i.e., the hoses, valves, and fittings of the vacuum drying equipment.) Evaluations are presented for quantifying the amount of residual water that could remain in the cask cavity, even after successful drying operations, since vacuum drying is a limit process, and by definition, cannot remove all of the water from the cavity, no matter how long the operation goes on, even if all of it is in vapor form. (authors)

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

Long-Term Assessment of Critical Radionuclides and Associated Environmental Media at the Savannah River Site - 13038

During the operational history of the Savannah River Site (SRS), many different radionuclides have been released from site facilities. However, only a relatively small number of the released radionuclides have been significant contributors to doses and risks to the public. At SRS dose and risk assessments indicate tritium oxide in air and surface water, and Cs-137 in fish and deer have been, and continue to be, the critical radionuclides and pathways. In this assessment, statistical analyses of the long-term trends of tritium oxide in atmospheric and surface water releases and Cs-137 concentrations in fish and deer are provided. Correlations also are provided with 1) operational changes and improvements, 2) geopolitical events (Cold War cessation), and 3) recent environmental remediation projects and decommissioning of excess facilities. For example, environmental remediation of the F- and H-Area Seepage Basins and the Solid Waste Disposal Facility have resulted in a measurable impact on the tritium oxide flux to the onsite Fourmile Branch stream. Airborne releases of tritium oxide have been greatly affected by operational improvements and the end of the Cold War in 1991. However, the effects of SRS environmental remediation activities and ongoing tritium operations on tritium concentrations in the environment are measurable and documented in this assessment. Controlled hunts of deer and feral hogs are conducted at SRS for approximately six weeks each year. Before any harvested animal is released to a hunter, SRS personnel perform a field analysis for Cs-137 concentrations to ensure the Hunter's dose does not exceed the SRS administrative game limit of 0.22 milli-sievert (22 mrem). However, most of the Cs-137 found in SRS onsite deer is not from site operations but is from nuclear weapons testing fallout from the 1950's and early 1960's. This legacy source term is trended in the SRS deer, and an assessment of the 'effective' half-life of Cs-137 in deer (including the physical decay half-life and the environmental dispersion half-life) is provided. The 'creek mouth' fisherman is the next most critical pathway at SRS. On an annual basis, three species of fish (panfish, catfish, and bass) are sampled from the mouths of the five SRS streams. Three composites of up to five fish of each species are analyzed from each sampling location. Long-term trending of the Cs-137 concentrations in fish and the subsequent doses from consumption of SRS fish is provided. (authors)

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

New approach to gamma scans for processed natural uranium at a FUSRAP site - 15470

The Superior Steel Site is part of the FUSRAP because the former facility processed natural uranium metal in support of the Atomic Energy Commission (AEC) fuel-element development program in the 1950's. Previous investigations of the site have identified uranium contamination both inside and outside of the building associated with the uranium work and some limited remediation was performed in the past. As part of the remedial investigation (RI) process for the site, gamma scan surveys were completed in order to define the nature and extent of elevated residual uranium associated with AEC activities. One of the challenges with investigating for processed natural uranium versus unprocessed natural uranium is that the processed uranium is practically devoid of its decay progeny, which produce gamma photon radiation emissions commonly relied upon to detect the presence of uranium. To address this challenge, a sodium iodide (NaI) detector was designed and fabricated to optimize the signal to noise ratio for U-238 low energy gammas (59-127 keV) and to reduce interferences from other radionuclides that may be present from non-FUSRAP sources. Gamma scan surveys were performed with the optimized NaI detector in conjunction with typical, commercially available 0.35 liter NaI detectors. The results of this survey and a comparison between the detectors is presented. (authors)

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

Californium Electrodepositions at Oak Ridge National Laboratory

Electrodepositions of californium isotopes were successfully performed at Oak Ridge National Laboratory (ORNL) during the past year involving two different types of deposition solutions, ammonium acetate (NH 4 C 2 H 3 O 2 ) and isobutanol ((CH 3 ) 2 CHCH 2 OH). A californium product that was decay enriched in 251 Cf was recovered for use in super-heavy element (SHE) research. This neutron-rich isotope, 251 Cf, provides target material for SHE research for the potential discovery of heavier isotopes of Z=118. The californium material was recovered from aged 252 Cf neutron sources in storage at ORNL. These sources have decayed for over 30 years, thus providing material with a very high 251 Cf-to- 252 Cf ratio. After the source capsules were opened, the californium was purified and then electrodeposited using the isobutanol method onto thin titanium foils for use in an accelerator at the Joint Institute for Nuclear Research in Dubna, Russia. Another deposition method, ammonium acetate, was used to produce a deposition containing 1.7 0.1 Ci of 252 Cf onto a stainless steel substrate. This was the largest single electrodeposition of 252 Cf ever prepared. The 252 Cf material was initially purified using traditional ion exchange media, such as AG50-AHIB and AG50-HCl, and further purified using a TEVA-NH 4 SCN system to remove any lanthanides, resulting in the recovery of 3.6 0.1 mg of purified 252 Cf. The ammonium acetate deposition was run with a current of 1.0 amp, resulting in a 91.5% deposition yield. Purification and handling of the highly radioactive californium material created additional challenges in the production of these sources.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2020

Best Practices for Shielding Analyses of Activated Metals and Spent Resins from Reactor Operation

This report investigated best practices for performing shielding evaluations of Type B waste packages, as defined in 10 CFR Part 71 regulations on packaging and transportation of radioactive material, or packages for which the contents are not defined before loading and may include a broad range of nuclides, geometries, and non-fuel materials. The following non-fuel waste streams were analyzed: Activated metals from decommissioned commercial power reactors, including Type 304 stainless steel, reactor vessel steel, and Inconel, Control blades from boiling water reactors, Neutron-activated corrosion products on surfaces of activated metals, and Spent resins from power plant operations. Measured elemental compositions, including major constituents and impurities, for steel and Inconel samples from commercial power reactors were used in activation calculations to determine radionuclide inventories in activated metals. For a simplified cask model, 60 Co contribution to the total external package dose rate at 30 days after shutdown varied from approximately 60% to 95%, depending on the activated metal, initial cobalt impurity concentration in the metal, and the thickness of the overpack gamma shield. Its maximum contribution to the total external package dose rate of approximately 100% was reached within the time interval of 2 to 5 years after shutdown and was maintained for up to 45 to 60 years after shutdown, depending on material, initial cobalt impurity concentration, and shield thickness. Thereafter, the 60 Co contribution to external package dose rate decreased with increasing decay time. Cobalt-60 is primarily produced by neutron reactions with the cobalt impurity in steel and Inconel. Other important radionuclides in activated metals contributing to package external dose rate are radionuclides with relatively short decay times, including 51 Cr, 59 Fe, 58 Co, and 54 Mn. These radionuclides may be represented as an equivalent 60 Co activity/source because 60 Co gamma ray emissions are bounding in terms of source strength and energy to other important radionuclides identified in the analyzed activated metals. Approaches for modeling the neutron-activated corrosion products that may be attached to activated reactor components were analyzed in this report. It was demonstrated that a surface source is more conservative than a uniform volumetric source for the treatment of neutron-activated corrosion products with respect to external package dose rates. An analysis of the maximum radionuclide loadings reported on spent resins identified the radionuclides 137 Cs, 60 Co, 134 Cs, 65 Zn, and 58 Co as the primary contributors to external package dose rate. For a resin cooled for 3.08 years, the external package dose rate was entirely produced by the reported 137 Cs and 60 Co inventory. The neutron sources from actinides found on spent resins or activated metals produced negligible dose rates and may be ignored in dose rate analyses. Effects of idealized waste material, source geometry, and spatial material/source distributions on external package dose rates were determined based on dose rate results for a simplified cask model under normal conditions of transport. Type 304 stainless steel, zirconium, and aluminum with adjusted mass densities based on a maximum content weight were analyzed for material modeling effects on external package dose rate. These materials produced identical external package dose rates within the statistical uncertainties of the dose rate estimates. Among four different source geometry configurations with homogeneous material of different mass densities, uniform volumetric source distribution, and the same source strength, the geometry configuration with lowest mass density (i.e., minimum self-shielding effect among the four cases) was most conservative. Spatial source distributions that better represent localized peak 60 Co activity values were more conservative than a uniform volumetric source distribution, assuming the same weight and total source strength per package. The increase in external dose rate caused by localized source peaks can be as much as the ratio between source peak activity density to average activity density, depending on the location of the activated metal with peak activity density. Therefore, the shielding analysis may be simplified if localized peak activities and the average activity per package can be measured/determined and documented at the time of cask loading. For simplicity, external package dose rates may be determined based on average source activity and a uniform volumetric source distribution. The dose rate results from that calculation model multiplied by the ratio of peak activity density to average activity density will produce maximum dose rate values for conservative estimates.

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

Regulation of the Shutdown Defueling and Decommissioning Reactors in the United Kingdom - 16576

Nuclear policy, including for decommissioning, in the United Kingdom is set by the UK Government. The Nuclear Decommissioning Authority, NDA, is responsible for managing the clean-up of the UK's nuclear legacy which includes the first generation of Magnox power stations. Magnox Ltd (MXL), owned by Cavendish Fluor Partnership, is the licensee and the management and operations contractor responsible for ten Magnox reactor sites and two restoration sites in the UK, working for the sites' owner, NDA. The work of the company is diverse covering electricity generation at Wylfa (until December 2015), defueling, and preparation for an interim end state, known as Care and Maintenance (C and M) which will precede the final site clearance. The C and M phase is expected to last over 75 years during which the sites are maintained in a passively safe and secure state while radiation levels are left to decay naturally. Preparations for this phase include removing fuel from the reactors and storage ponds; retrieving intermediate level waste from current stores and treating it or converting it to forms suitable for long-term storage or disposal and making sure of the integrity of critical safety systems of the de-fuelled reactors. The Decommissioning, Fuel and Waste programme within the Office for Nuclear Regulation, ONR, regulates nuclear safety in the 12 Magnox nuclear licensed sites. ONR works with the environmental regulators who are responsible for environmental protection on nuclear sites. The focus of this paper is the regulation of nuclear safety by ONR across the Magnox sites and the interactions with nuclear security and the environmental aspects required to achieve effective, proportionate and efficient regulation of these sites in the UK. ONR's strategy for regulating the Magnox defueling and decommissioning sites is focused on enabling the licensee to complete the initial stages of decommissioning and reduce the nuclear and radioactive hazards on its sites. To this effect, ONR's revised inspection guidance is centred on inspecting the (remaining) important systems for safety identified in the safety cases. Other interventions focus on engaging with the licensee on safety significant decommissioning projects to gain evidence of the licensee's readiness prior to and during these activities rather than formal assessment of the associated safety cases. In light of reduced hazards and risks across the de-fueled and decommissioning Magnox reactors, ONR's revised regulatory expectation regarding evolution of emergency arrangements during decommissioning is based on key transition points that follow electricity generation. Implementation of these revised strategies as well as organisational restructuring within the team responsible for regulation of the Magnox fleet has resulted in improved efficiency and consistency in regulatory approach whilst securing the required safety standards in managing decommissioning activities. (authors)

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

Oxidation of Spent Fuel During Fuel Handling Accidents Test Plan

The transfer of commercial spent nuclear fuel (CSNF) from transport or dry storage cask into final repository cask represents a scenario where fuel elements are potentially at risk to exposure to air and elevated temperatures. In the event of an accident during handling operations, the facility cooling may cease to function, and the temperature of the fuel would begin to rise due to decay heat generation. CSNF is generally stored in an inert environment, but handling operations could be performed in an air environment. If any damaged fuel elements are present during this postulated accident the fuel could begin to oxidize from UO 2 to U 3 O 8 . If the oxidation event is severe, volume expansion from oxidation could result in large-scale cladding failure and large releases of CSNF debris into the transfer facility. Depending on the design of the fuel handling facility, any SNF release could have an impact on off-site dose.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Remediation Strategy for Former PChP Uranium Production Site in Ukraine - 18217

The paper describes current status and results of the investigation carried out at the one of the largest Uranium production facilities in the former Soviet Union (SU) located in Ukraine, 'Pridneprovsky Chemical Plant' (PChP). A large volume of new site specific contamination data and parameters needed for safety assessment, prioritization and environment remediation planning have been received during different phases of the site characterization and hazards investigation, carried out in the frame of State remediation programs and also with financial support of the EC projects during 2009-2017. Specific features of this site are significant amount of sludge materials, containing high activity concentration of {sup 230}Th and also number of the buildings, which were used for Uranium production in past and now containing large amount of spilled and dispersed U-residues with very high activity concentration of {sup 226}Ra, {sup 230}Th - as well as other radionuclides of {sup 238}U decay series, which also remain in many locations at this site. This site is still used for industrial purposes, which are not related to the Uranium production, and therefore many people are visiting this site daily, posing potential high risks for them to be exposed. The paper presents the understanding about the identified hazards and the priority between number of many other objects, which are required for the remediation, and also describes the problems faced during remediation planning because lack of safety management, experience and technical capabilities as well as funds available in Ukraine, which are needed to bring PChP site in safe conditions. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗