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DOE OSTI2019

Revisiting an airgap split-optics mitigation for beam filamentation in high power lasers

We extend the split-optic approach for mitigating filamentation in a thick optical component previously proposed for small beams to conditions relevant to high-power lasers. The split-optic approach divides a thick optic into two thinner optics separated by an airgap to reduce filamentation through diffraction management. Our numerical study focuses on filamentation of a flat-top beam with intensity modulation noise sources passing through a split-optic system. The improvement in the distance to collapse in glass is shown to be potentially substantial (>30%), yet has limited increase with the airgap size, unlike the common understanding when considering a collapse of a whole beam or a sole perturbation on a beam. The improvement in the collapse distance in glass asymptotes to an upper bound value that depends mainly on the beam mean intensity and its contrast for any airgap size above some value that depends mainly on the shortest spatial periods comprising the excitation noise source. Examining the difference in the simulation results for a periodic versus a randomly generated perturbation source-term suggests that the observed effect is governed by the statistical interference dynamics of the beam while propagating through the airgap

47 OTHER INSTRUMENTATION↗

DOE OSTI2026

Model Data Archive Associated with Manuscript "Fire-altered Carbon Pools Create Disturbance Memory in Stream Dissolved Organic Carbon"

This data package supports the publication “Fire-altered Carbon Pools Create Disturbance Memory in Stream Dissolved Organic Carbon” by Li et al. (2026). The package contains processed model inputs, configuration files, restart files, simulation outputs, scripts, and visualization products used to evaluate post-fire dissolved organic carbon (DOC) dynamics in the Naches River Watershed, Washington, USA, following the 2021 Schneider Springs Fire. The modeling workflow couples ELM-BGC, the biogeochemistry-enabled Energy Exascale Earth System Model Land Model; ATS, the Advanced Terrestrial Simulator for integrated surface-subsurface hydrology; and PFLOTRAN, a reactive transport model for multicomponent aqueous geochemistry. Together, these models simulate how wildfire-induced changes in vegetation, litter, coarse woody debris, and soil organic matter influence DOC production, transport, and reaction from burned hillslopes to stream networks. The archive includes preprocessed meteorological, geospatial, hydrologic, and biogeochemical forcing data; ELM-BGC-derived DOC source terms; ATS mesh files; PFLOTRAN reactive-transport inputs; model configuration files; spin-up and transient restart files; watershed-scale diagnostic outputs; stream concentration time series; and figures or visualization files used to inspect and reproduce key results. File types include Hierarchical Data Format 5 (HDF5) files for gridded forcing and model-coupling data, model input and configuration files for ELM-BGC, ATS, and PFLOTRAN, restart and simulation-output files generated by the modeling workflow, tabular or time-series diagnostic outputs, scripts for post-processing and figure generation, and image or visualization products associated with the manuscript. Use of the package depends on the intended task. Re-running the simulations requires the relevant modeling software, including ELM-BGC, ATS, and PFLOTRAN as ATS's geochemical engine. Inspecting outputs and reproducing figures requires Python with scientific plotting libraries such as Matplotlib, and three-dimensional model outputs may be viewed with ParaView. Geographic information system files or maps may be inspected with ArcGIS Pro or comparable GIS software. The data package is intended to enable traceability, reuse, and partial reproduction of the coupled land-to-watershed hydro-biogeochemical modeling workflow used to test how wildfire disturbance affects terrestrial carbon pools and downstream DOC dynamics.

ATS↗

DOE OSTI2022

Integrated Disposal Facility FY2011 Glass Testing Summary Report [Erratum]

Pacific Northwest National Laboratory was contracted by Washington River Protection Solutions, LLC to provide the technical basis for estimating radionuclide release from the engineered portion of the disposal facility (e.g., source term). Vitrifying the low-activity waste at Hanford is expected to generate over 1.6 x 10 5 m 3 of glass (Certa and Wells 2010). The volume of immobilized low-activity waste (ILAW) at Hanford is the largest in the DOE complex and is one of the largest inventories (approximately 8.9 x 10 14 Bq total activity) of long-lived radionuclides, principally 99 Tc (t 1/2 = 2.1 x 10 5 ), planned for disposal in a low-level waste (LLW) facility. Before the ILAW can be disposed, DOE must conduct a performance assessment (PA) for the Integrated Disposal Facility (IDF) that describes the long-term impacts of the disposal facility on public health and environmental resources. As part of the ILAW glass testing program PNNL is implementing a strategy, consisting of experimentation and modeling, in order to provide the technical basis for estimating radionuclide release from the glass waste form in support of future IDF PAs. The purpose of this report is to summarize the progress made in fiscal year (FY) 2011 toward implementing the strategy with the goal of developing an understanding of the long-term corrosion behavior of low-activity waste glasses.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2018

Field-Scale Lysimeter Studies of Low-Activity Waste Form Degradation

Before vitrified and cementitious nuclear waste can be disposed of at the U.S. Department of Energy’s Hanford Site Integrated Disposal Facility (IDF), a performance assessment (PA) must be conducted. A field lysimeter is one method capable of confirming that laboratory experiments provide reasonable model parameter estimates for PA release models. Field lysimeters are also useful in demonstrating that release models can accurately predict material degradation in a field environment. This report describes the implementation plan for a study of LAW waste form degradation using the Field Lysimeter Test Facility (FLTF) on the Hanford Site. This facility provides an opportunity to study glass and cementitious waste forms that will be produced during Hanford cleanup activities in field-relevant conditions. Results from this study will: 1) provide field experimental data on degradation of various waste forms; 2) be used to refine process model descriptions of contaminant (source term) release from the waste forms; 3) reduce uncertainties about the representativeness of laboratory testing results for determining long-term waste form performance under field conditions; and 4) improve confidence in the IDF PA by providing data that verify parameters and assumptions used in the PA modeling. It is expected that this study will lead to improved stakeholder confidence in the safe disposal of LAW at the IDF and aid in the adoption of informed facility designs with the potential to reduce operational costs.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2018

Probabilistic Economic Valuation of Safety Margin Management

The goal of this project was to explore a new methodology for evaluation of potential safety margin upgrades to add a new instrument to the Risk Informed Safety Margin Characterization (RISMC) toolkit. To this end, the RAVEN code developed by the Idaho National Laboratory was used to perform a cost-benefit analysis of several potential power plant safety upgrades to compare the material cost of the upgrades with the probabilistic benefit of avoiding accident scenario consequences to find the most cost-effective safety upgrades. By coupling RAVEN to MELCOR to drive Monte-Carlo method stochastic parameter analysis and using data from existing nuclear codes, as well as MELCOR’s radionuclide relocation tracking capabilities, the radionuclide release source term activity in a severe accident scenario can be estimated. Severe Accident Mitigation upgrades were benchmarked against the Hardened Containment Ventilation upgrade mandated by the USNRC for Mk1 BWRs using dollars spent per Curie of release prevented as a metric by which to evaluate the economic efficacy of the upgrades. Additionally, Fault Tree/Event Tree-based analyses were performed to compare older methods of evaluation to the methodology pioneered in this work, and Monthly Operating Report (MORP) outage data was analyzed for insights into power plant outage optimization and reduction.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2018

Advanced Integration of High Dose Neutron, Microcalorimeter and Voltammetry Sensor Technologies in an Electro-Refining Process

In this report we examine measurements from two different sensors, for detecting a diversion of plutonium in an electro-refining (ER) process. One method is based on counts from a high dose neutron detector (HDND). The other is based on spectra from a micro-calorimeter sensor. The spectra comes from the ER-salt, from fission product waste and from a source term. The goal is to differentiate between normal and off normal operating conditions using neutron counts or changes in counts in the spectra.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2018

A safety and licensing roadmap to identify the research and development gaps of commercial molten salt reactors

Molten salt reactors (MSRs) both inherently and through good design possess highly desirable safety attributes. Successful commercial deployment of MSRs is heavily dependent on establishing their safety characteristics in a well-coordinated licensing requirements program. The approach outlined is based on adapting the MSR licensing process to the current LWR–based NRC process that exists today, which has been the traditional approach used by new technologies such as the Pebble Bed Reactor licensing application, the Clinch River Breeder Reactor licensing application, and more recently the NuScale licensing application. NRC has recognized some of the unique passively safe aspects of advanced non-LWR reactors and has introduced an advanced reactor strategic plan which will address policy issues that need to be addressed to accommodate advanced reactor concepts within the existing regulatory framework. Examples of some of these policy issues are the use of functional containment, the use of mechanistic source terms, and more reliance on performance-based and risk-informed regulatory processes, such as the Licensing Modernization Project, which emphasizes the use of risk-informed processes to select licensing basis events, safety categorization of SSC, and well-defined defense-in-depth systems. Applying performance-based, risk-informed regulatory processes is, consequently, vital to substantially decreasing the cost, time, and uncertainty of commercial MSR reactor licensing while providing for the health and safety of the public and protecting the environment. This document is intended to provide a roadmap for the MSR licensing and safety activities necessary to provide a success path to commercialization.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2018

Thermal Test Suite for Code Maintenance and Validation of Diablo

This document describes a series of test problems that ensure that the Thermal capabilities of Diablo are working correctly. Diablo utilizes the Finite Element Method to discretize the problem in space and a trapezoidal rule finite difference scheme to discretize the problem in time. This suite will test the codes capabilities to handle various boundary conditions, source terms, and initial conditions. It will also test the coupling between thermal and solid effects, and the use of various linear and nonlinear solvers.

42 ENGINEERING↗

DOE OSTI2019

Rad-NESHAP Review of TA-21-257 Remediation Activities

Environmental Protection and Compliance (EPC) Division personnel reviewed the remediation work at Technical Area 21, Building 257 (TA-21-257) for compliance with Radionuclide NESHAP1 requirements. Remediation activities include removal of Building 257, the associated waste lines, and waste storage tanks. We used project data and the most recent CAP88 plume dispersion model to estimate the emissions source term and the resulting off-site dose to members of the public. Emissions monitoring requirements from this work are being met by measuring ambient air concentrations at public receptor locations using LANL’s Airnet system. All requirements under 40 CFR 61 Subpart H are being met for the currently planned scope of operations.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Inventory Data Package for the Hanford Site Composite Analysis

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Advanced Reactor Siting Policy Considerations

This report describes considerations for and challenges to revising the US Nuclear Regulatory Commission (NRC) guidance regarding population for siting an advanced reactor. Current regulations in 10 Code of Federal Regulations (CFR) Part 100, Reactor Site Criteria, and guidance support in Regulatory Guide (RG) 4.7, Revision 3, General Site Suitability Criteria for Nuclear Power Stations, are focused on large light water reactors (LWRs), addressing the distance that a nuclear reactor should be sited from a densely populated center. This report analyzes the alternatives to using a formula distance as the primary figure of merit for siting small modular reactors (SMRs) or non-light water reactors (non-LWRs). Some reactor vendors use a business plan that sites reactors close to an industrial partner to supply heat or to back-fit a reactor at a fossil plant site to take advantage of existing infrastructure. Factors such as smaller source terms (fission product release to the environment), passive safety systems, advances in barrier technology, advances in simulation and modeling, and improved understanding of societal risk may allow for siting these types of reactors closer to densely populated centers than has historically been accepted for large LWRs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2019

Completion of Level 2 Milestone – M2AT 19IN030401044, Complete construction of the gas and moisture supply systems and receive the custom-designed tube furnace from the vendor

The Advanced Gas Reactor (AGR) program has designed and is now constructing the Air/Moisture Ingress Experiment (AMIX) facility to test irradiated tristructural isotropic (TRISO) fuels under air or moisture-ingress accidents postulated for high-temperature gas-cooled reactors. AMIX will ultimately be installed in the Fuel Conditioning Facility at the Materials and Fuels Complex (MFC) at Idaho National Laboratory. AMIX consists of a gas supply (capable of supplying mixtures of air/helium and steam/helium), a tube furnace capable of accepting samples up to 6 cm in diameter and heating them to 1650°C, gamma detectors for measuring the release of condensable, gamma-emitting fission products, and cold traps with gamma detectors for measuring the release of fission gases. Data obtained from AMIX experiments will be used to qualify this new generation of TRISO fuels with the U.S. Nuclear Regulatory Commission, and inform models and simulations of TRISO fuel accident performance and source term.

42 ENGINEERING↗

DOE OSTI2019

Monitoring an Electro-Refining Process (Rev. 1)

In this report we examine measurements from two different sensors for detecting a diversion of plutonium in an electro-refining (ER) process. One method is based on counts from a high dose neutron detector (HDND). The other is based on spectra from a micro-calorimeter sensor. The spectra come from ER-salt, fission product waste, and from a source term. The goal is to differentiate between normal and off normal operating conditions using neutron counts or changes in counts in the spectra. Unfortunately, there is no electro-refining process that we can actually monitor to obtain real data from various operating conditions so we will use simulated data to illustrate our approach.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2019

Modeling of Pyroprocessing Hot Cell for Process Monitoring Evaluation

In this report, we examine studies undertaken to determine the applicability of applying measurements with a neutron counter within a pyroprocessing hot cell. The studies focus on the modeling of the hot cell and neutron counter in Monte Carlo N- Particle Code (MCNP) to determine expected radiation effects and applicability of neutron counters. The neutron counter of focus was the High Dose Neutron Detector (HDND) developed at Los Alamos National Laboratory (LANL). We studied the HDND to determine what would be the detector response under normal background conditions with operating process equipment as well as the background’s potential effect on counting applications involving a U/TRU ingot produced by the pyroprocessing electrorefiner. To perform this task, we were required to develop models of the pyroprocessing process cell and unit operations and generate source terms for various different times in the process. The data for the pyroprocessing materials came from SNL’s Safeguards and Security Process Model (SSPM) as part of MPACT’s advanced integration mission. This report will discuss the development of the pyroprocessing process cell and unit operations models in MCNP as well as various simulations to determine the applicability of the HDND for process monitoring efforts in the hot cell.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2019

Failures and Implications of Heat Pipe Systems

Under Department of Energy (DOE), Office of Nuclear Energy (NE), Gateway for Accelerated Innovation in Nuclear (GAIN), Sandia National Laboratories (SNL) was awarded DOE-NE GAIN voucher GA-19SN020107, "Risk-informed mechanistic source term calculations for a sodium fast reactor." Under this GAIN voucher, SNL supported the industry partners development in preparation for licensing and commercialization by providing subject matter expertise on heat pipe technologies, providing computer code training and support, and perform first-of-a-kind experiments demonstrating the safety/risk impacts of heat pipe breach failures. The experiments that were performed had two primary goals: measure the peak heat fluxes that lead to heat pipe dry out and subsequent wall breach; and observe the consequences that result from catastrophic failure of a heat pipe wall. Intentional breaching of the heat pipe walls took advantage of heat pipe physics and operating limits. Large and nearly instantaneous heat fluxes were applied to the heat pipe to first cause localized dry out at the evaporator section which then leads to melting of the heat pipe wall. The hour glass heat pipe (Test 1) experienced dry out at 112 W/cm 2 and after 45 seconds, wall temperatures measure about 1,280°C and intentional failure of the heat pipe wall was achieved. The cylindrical heat pipe (Test 2) experienced dry out at 125 W/cm 2 and after 65 seconds, wall temperatures exceeded 1,400°C and intentional failure of the heat pipe wall was achieved. Both experiments characterize the parameters needed to lead to heat pipe wall failure. Furthermore, the failure of the heat pipes characterizes the safety/risk impacts from sodium-oxygen reactions that occur following the intentional failure. There were two major conclusions of these intentional failure tests: the heat pipes were able to continue operating beyond expected performance limits, and the failure behavior validated decades of operational experience.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2019

Inventory Data Package for the Hanford Site Composite Analysis

The disposed inventory estimate is perhaps the most important component of the composite analysis (CA), as it directly affects the future radiological impacts following site closure. The goal is to estimate radionuclide inventory from site inception to closure. Because of the inclusive nature of a CA, all relevant contaminants are identified and initially considered. Then, subsets of contaminants appropriate for quantitative analysis are selected. Reducing the number of radionuclides for inclusion in the quantitative analysis helps focus budget and resources on simulating only those radionuclides that are likely to contribute to the total dose to the receptor above a threshold value. Inventory is reported within the geographical boundary of the Hanford Site 200 Areas and includes contributing sources on the Central Plateau. The inventory covers the radiological inventories of historical discharges and residuals that have the potential to release after closure. The reported inventory is focused on the source terms needed for transport modeling and the evaluation of risk associated with the ground water pathway. Risk from the air pathway was considered. Inventories of sites with approved performance assessments (PAs) will be integrated using their release curves directly with vadose zone transport modeling.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Final Report on the NA-22 sponsored project from FY17-19: Next-Generation Correlated Fission Event Simulations

Through this multi-organization correlated fission physics research and development project, the CGMF and FREYA fission event generators have been implemented into the MCNP6.2 code and released to the public while a variety of dedicated differential experiments on correlated fission signatures have been performed to help validate the complex fission theory and physics implemented in these codes. In addition to enabling these new capabilities, a multitude of efforts in modeling and transport simulation has been directed toward a variety of applications including general nuclear nonproliferation, subcritical neutron multiplication, and nuclear criticality safety applications. While there is still much more to be done in the area of correlated fission physics, including the study of b-delayed fission observables, more robust validation and an extension into more application spaces, one of the key successes of this project can be summarized as the delivery of a newly demonstrated and consistently integrated simulation capability spanning several complex nuclear physics codes. From the special nuclear materials source term to the radiation transport to the time-coincident detector response modeling, new and existing users around the world, especially throughout the nuclear nonproliferation communities, can now begin to model these kinds of highly complex special nuclear material applications. Another successful aspect of this project included fruitful collaborations, mentoring opportunities and community outreach. One extended collaboration workshop was held in Santa Fe, NM in March 2017. Two other limited collaboration meetings took place including a brief gathering during the mid-project Schubert review at LANL and a single all-day special session on “Correlated Fission Physics, Transport and Applications” held during the 2019 Annual Meeting of the Institute of Nuclear Materials Management (INMM) in Palm Desert, CA. Finally, as part of this project, we have mentored 14 undergraduate students, PhD candidates, and postdoctoral researchers listed in Appendix A. The remainder of this report contains detailed information followed by highlights of the goals and accomplishments in Appendix B broken down by task within this project.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2020

Oak Ridge National Laboratory Pool Critical Assembly MCNP6 Criticality Calculation Input File

This report provides an MCNP ® input file for a k-eigenvalue (kcode) calculation of the Oak Ridge National Laboratory (ORNL) Pool Critical Assembly (PCA). That is, this report provides the input file used for the first of the calculations described to produce a fission source term for the ORNL PCA.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗