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All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Analysis of measurements from an array of radioxenon samplers near to Hartlepool Nuclear Power Station

As part of a scientific research and development project, the radionuclide fingerprint of an operating advanced gas-cooled reactor 25 (AGR) has been studied across several facets (Goodwin et al., 2024). One part of this project was to deploy an array of 26 radioxenon samplers to the region for a period of around 1 year, to measure any radioxenon emissions from the reactors of the 27 Hartlepool nuclear power station at a range of tens of kilometres away. The array of 3 sensors was operational for around 12 28 months from March 2022 and detected many occurrences of isotopes of radioxenon. Here we provide a detailed analysis and 29 interpretation of the data and where possible, attribution of detections to a source or region. Whilst a key part of this work is to 30 measure any emissions of radioxenon from Hartlepool, this work presents one of the most comprehensive efforts to determine the 31 source of a great number of (mostly) 133Xe detections. A combination of different types of atmospheric dispersion modelling 32 techniques, including the use of stack monitoring data from nearby civil radioxenon-emitting nuclear facilities has enabled the 33 majority of detections to be attributed to one or more possible sources. Whilst emissions from Hartlepool have been detected on 34 the systems, the majority of detections are associated with a medical isotope production facility in Fleurus, Belgium (IRE).

Advanced gas-cooled reactor↗
DOE OSTI2020

Design Optimization of A generic Fissile Solution for Mo99 production using Electron Beam-based Neutron Generator using MCNP+CFD

Molybdenum-99 ( 99 Mo) is a medical isotope used in 80% of all medical imaging procedures today. However, 99 Mo is not directly administered to patients for imaging; its decay product, technetium-99m ( 99 mTc), is a metastable isotope with a half-life of ~6 hours. Such a short half-life does not allow for production, separation, and application to a patient before decaying further. Therefore, the solution is to produce the precursor to 99 mTc, 99 Mo. 99 Mo has a half-life of 66 hours, long enough to produce the material and send it to the desired location before it fully decays. Under National Nuclear Security Administration’s (NNSA) highly enriched uranium (HEU) minimization mission, the Material Management and Minimization (M3) office leads the Molybdenum-99 (Mo-99) program. A major goal of the Mo-99 program is to develop methods of producing 99 Mo using low enriched uranium (LEU) opposed to the traditionally used HEU domestically in the U.S. Internationally, LEU is the new standard for use in research reactors and isotope production facility. In this study, we introduce a generic fissile solution system for 99 Mo production facility meeting the design requirement of 1) LEU usage as fissile material and 2) subcritical fissile system coupled with a horizontal electron beam (E-beam) accelerator. The focus of this research is to optimize a fissile solution configuration for maximum 99 Mo production yield using MCNP model, and ultimately implement the optimized design with computational fluid dynamics (CFD) model to understand thermal hydraulic behavior and solution convection characteristic. The report consists of two parts: 1) isotope yield calculation and criticality analysis with a wide range of parametric test matrix, and 2) thermal hydraulic analysis for solution mixing and cooling assessment.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2023

Production of zirconium-88 via proton irradiation of metallic yttrium and preparation of target for neutron transmission measurements at DICER

Abstract A process for the production of tens to hundreds of GBq amounts of zirconium-88 ( 88 Zr) using proton beams on yttrium was developed. For this purpose, yttrium metal targets (≈20 g) were irradiated in a ~16 to 34 MeV proton beam at a beam current of 100–200 µA at the Los Alamos Isotope Production Facility (IPF). The 88 Zr radionuclide was produced and separated from the yttrium targets using hydroxamate resin with an elution yield of 94(5)% (1σ). Liquid DCl solution in D 2 O was selected as a suitable 88 Zr sample matrix due to the high neutron transmission of deuterium compared to hydrogen and an even distribution of 88 Zr in the sample matrix. The separated 88 Zr was dissolved in DCl and 8 µL of the obtained solution was transferred to a tungsten sample can with a 1.2 mm diameter hole using a syringe and automated filling station inside a hot cell. Neutron transmission of the obtained 88 Zr sample was measured at the Device for Indirect Capture Experiments on Radionuclides (DICER).

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Proposed guidance for preparing and reviewing a molten salt non-power production or utilization facility application

Development of non-power molten salt reactors (MSRs) are under consideration to further establish an MSR experience base, support the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Section 50.43(e), and provide any additional analyses needed for development of a full-scale MSR. Guidance provided in this report is based on MSRs operating with liquid fuel (i.e., fuel dissolved within a molten salt). These reactors, unless owned by the DOE or DOD, will require licensing by the US Nuclear Regulatory Commission (NRC) staff. Standard review plan (SRP) guidance for large light water reactors (LWRs) is available in NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants; Light Water Reactor (LWR) Edition. However, NRC staff observed that NUREG-0800 is very cumbersome to apply to non-power reactors “because of the great differences in complexity and hazards between non-power reactors and nuclear power plants.” Therefore, a program to develop performance-based guidance applicable to non-power reactors was initiated. In 1996, NUREG-1537, Parts 1 and 2, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors, was published. Part 1, the format and content guide, suggests a uniform format for presenting information in non-power reactor applications that is acceptable to the NRC staff, but conformance with the format and content is not required. Part 2, the SRP, ensures the quality and uniformity of the staff review of an application. Unfortunately, the application guidelines and SRP do not provide adequate guidance for all advanced non-LWR technologies and applications. This discrepancy eventually led to the 2012 development of interim staff guidance (ISG) for NUREG-1537, which includes criteria for describing and reviewing aqueous homogeneous reactors (AHRs). Specifically, NUREG-1537 ISG, 2012 expanded the original document to address three areas: 1. updated criteria for heterogeneous non-power reactors, 2. criteria for licensing AHRs, and 3. criteria for licensing a Part 50-licensed isotope production facility. In 2015, the US Department of Energy (DOE) opted to build on the AHR NUREG-1537 ISG experience by performing a gap analysis of the guidance that would be used to license a non-power MSR. MSRs represent one of the advanced non-LWR technologies selected by DOE for development through a multiyear cost share award with Southern Company Services. Under this Advanced Reactor Concepts 2015 (DOE Advanced Reactor Concepts [ARC] 15) award program, the DOE tasked Oak Ridge National Laboratory (ORNL) to evaluate the guidance changes that the NRC may need to consider when licensing an MSR non-power reactor. ORNL staff, with support from Boston Government Services, LLC, focused on five system-related chapters in NUREG-1537 that were considered most relevant to inform the effort that would be required for a non-power MSR applicant. ORNL documented this review in a technical report, ORNL/TM-2018/834, Proposed Guidance for Preparing and Reviewing Molten Salt Non-Power Reactor License Applications (NUREG-1537). The report was subsequently shared with industry and the NRC. The 2018 review was limited in scope, focusing on key system chapters based on the expected significance of each chapter relative to expected differences in addressing advanced non-LWR technologies, specifically non-power MSRs, compared with heterogeneously fueled non-power reactors. In the ORNL report, proposed generic adaptations were suggested for the following NUREG-1537 chapters: Chapter 4, “Reactor Description”; Chapter 5, “Reactor Coolant Systems”; Chapter 6, “Engineered Safety Features”; Chapter 9, “Auxiliary Systems”; Chapter 11, “Radiation Protection Program and Waste Management” The inclusion of Chapter 11 in the previous review effort was intended to provide guidance for categorizing the waste-handling process for an MSR operating with homogenous fuel. The introductions from Parts 1 and 2 of the 2012 NUREG-1537 ISG provide guidance for the application and review of production facilities. After a period of operation, non-power MSRs with homogenous fuel will include gaseous and soluble fission products. The gaseous fission products will be collected and held for decay in an off-gas system. There might also be an initiative to polish or filter the soluble fission products in the fuel salt by some mechanical or chemical means. The treatment and handling of fission products in the non-power MSR fuel salt and the description of this process in the safety analysis report (SAR) must be very precise to avoid the waste treatment facility being construed as a co-located special nuclear material (SNM) fuel cycle facility (see Section 2.3 of this report). Subsequent to the release of ORNL/TM-2018/834, NRC staff expressed a desire to continue the regulatory gap analysis that was begun in that report. This would provide additional clarity and information addressed in certain sections of the original report, while also providing new guidance on certain topics not addressed in the original report. This revision would benefit the NRC staff reviewing applications involving non-power MSR designs and would help developers understand how the NRC staff might approach the review of such applications. The focus of this report is to provide infrastructure support to the NRC staff for the regulatory review of non-power MSRs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

First STAX detector installation at the National Institute for Radioelements (IRE)

The Source Term Analysis of Xenon (STAX) project has been installing stack detectors at medical isotope production facilities to measure radioxenon emissions to investigate the effect of radioxenon releases on nuclear explosion monitoring. This paper outlines the installation of the first STAX detection system at the National Institute for Radioelements (IRE) in Fleurus, Belgium which has been operating for over three years and transferring collected data to the STAX repository. Information about the equipment installed, the data flow established, and calculations for determination of radioxenon releases from the facility are presented. Further. data quality was investigated to confirm values reported by STAX automated data processing and in a comparison of collected STAX data with data collected by IRE for regulatory reporting.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Source Term Analysis of Xenon (STAX): An effort focused on differentiating man-made isotope production from nuclear explosions via stack monitoring

An overview of the hardware and software developed for the Source Term Analysis of Xenon (STAX) project is presented which includes the data collection from two stack monitoring systems installed at medical isotope production facilities, infrastructure to transfer data to a central repository, and methods for sharing data from the repository with users. STAX is an experiment to collect radioxenon emission data from industrial nuclear facilities with the goal of developing a better understanding of the global radioxenon background and the effect industrial radioxenon releases have on nuclear explosion monitoring. The final goal of this work is to utilize collected data along with atmospheric transport modeling to calculate the contribution of a peak or set of peaks detected by the International Monitoring System (IMS) to provide desired discriminating information to the International Data Centre (IDC) and National Data Centers (NDCs). Types of data received from the STAX equipment are shown and collected data was used for a case study to predict radioxenon concentrations at two IMS stations closest to the Institute for RadioElements (IRE) in Belgium. The initial evaluation of results indicate that the data is very valuable to the nuclear explosion monitoring community.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2023

Modeling and Analysis Support for Acceleratro-based Production of Mo-99

NorthStar Medical Radioisotopes, LLC, is in the process of commissioning a medical isotope production facility, which will use high-power electron accelerators to produce molybdenum-99 (Mo-99), the parent of technetium-99m (Tc-99m), through photonuclear reactions in molybdenum-100 (Mo-100). In this approach, a target comprising multiple thin disks of enriched molybdenum metal is bombarded with a 40-MeV electron beam. Electrons that impinge on the molybdenum target produce bremsstrahlung x-rays that cause the nuclear reaction. Because enriched Mo-100 is expensive, there is a desire to use as much beam power as possible to achieve maximum production yield and minimize the size of the target. This requirement leads to very high beam power density (heat deposition in the target), which creates challenging requirements for target cooling. In the latest concept developed by NorthStar, a stack of target (sintered molybdenum) disks is irradiated from two sides by a 40-MeV electron beam, with a total power of 250 kW (125 kW from each side). The target disks are cooled by pressurized helium gas flowing through channels between the disks. Inconel windows in the target housing form a pressure boundary between the helium within the housing and the evacuated beam tube.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2020

Chemical Modeling of the Capture of Radioiodine Species

Iodine is a common fission product resulting from the transmutation of uranium fuel in nuclear power reactors. The short-lived radioactive isotope, 131 I, with a half-life of around eight days, would be a particular health concern if released into the environment. Capture materials, such as activated carbon, are used routinely in nuclear power plants and medical isotope production facilities to remove this hazard during normal operations to reduce emissions to acceptable levels. Iodine, unlike other common fission products such as noble gases, is challenging because of the wide range of molecular forms and oxidation states (from +VII in IF 7 to -I in iodide) that it displays, which can interconvert both in the gas phase and in the solution phase for example. The work in this project is focused on three computational tasks. In the first, we consider the binding of iodine-containing molecules of varying oxidation states to the major component of activated carbon – pristine graphene-like sheets. The second task focused on molecular binding and possible chemical transformations resulting from the reaction of molecular iodine with chemical functional groups on the edge of the graphene sheets or surface defects in the sheet itself. The final task was associated with radiolytic breakdown of 131 I to 131 Xe in iodine containing molecules and investigated the resulting molecular fragmentation process both in the gas phase and on the graphene-like surface. For each of the tasks, the intent was to learn about the thermodynamics and kinetics of the chemical transformations and use the calculations to provide parameters for input into kinetic models describing these processes for interpretation of experimental data in future work.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2022

Using STAX data to predict IMS radioxenon concentrations

The noble gas collection and measurement stations in the International Monitoring System (IMS) are heavily influenced by releases from medical isotope production facilities. The ability to reliably model the movement of radioxenon from the points of release to these IMS samplers has improved enough that a routine aspect of the analysis of IMS radioxenon data should be the prediction of the effect of releases from civilian nuclear facilities on the sample concentrations. In this work, predicted concentrations at IMS noble gas systems in Germany and Sweden based on measured releases from Institute for Radioelements (IRE) in Belgium and atmospheric transport modeling for a four-month period are presented and discussed.

63 RADIATION, THERMAL, AND OTHER ENVIRON. POLLUTAN↗
DOE OSTI2025

Incorporating civilian radioxenon background estimates in anomaly detection

A nuclear explosion screening exercise in 2023 (Maurer et al., 2023) found challenges with discerning anomalous radioxenon activity concentrations relative to elevated background concentrations. Research has continued into methods to detect anomalous radioxenon concentrations by comparing samples to estimates of atmospheric radioxenon background concentrations caused by releases at nuclear reactors or medical isotope production facilities. A new approach estimates the sample concentrations using time-varying radioxenon release rates obtained using optimization techniques that constrain the facility release rates to plausible amounts based on historical data or facility knowledge. The purpose of the optimization is to determine whether any combination of plausible release rates from emitting facilities can explain a series of radioxenon measurements at one or more sampling stations. A case study uses radioxenon data collected at three locations in western Europe for a month in 2021 and considers releases from 77 locations. Fewer samples are identified as being anomalous using a simplistic flagging rule than from an application of the current International Monitoring System (IMS) activity concentration-level rule.

Environmental sciences↗
DOE OSTI2022

New capability for indirect neutron capture measurements: The DICER instrument at LANSCE

The Device for Indirect Capture on Radionuclides (DICER) implements a new indirect technique for (n,γ) studies in which the neutron capture rate is determined from analysis of resonance neutron transmission data. The DICER instrument and associated radionuclide production at the Isotope Production Facility (IPF), both at the Los Alamos Neutron Science Center (LANSCE), as well radioactive sample fabrication, have been under development in the last few years. First measurements on a radioactive sample ( 88 Zr, t ½ =83.4 days), which was recently reported to have an extremely large thermal neutron capture cross section and resonance integral, are planned for the winter of 2021. A performance overview, brief details on the 88Zr fabrication and proof of good operation results will be presented.

47 OTHER INSTRUMENTATION↗
DOE OSTI2023

The new Device for Indirect Capture Experiments on Radionuclides at LANSCE: Efforts on measuring the resonance(s) responsible for the extremely large 88 Zr (n,γ) cross section

The thermal neutron capture cross section of 88 Zr was recently reported to be the second largest in nature with the largest resonance integral measured. Presumably, these very large values are caused by a resonance or resonances very near thermal energy. Determining their energies and widths, and hence the shape of the cross section away from thermal energies,is useful for applications. The short half-life (83.4 days) and associated large background, renders direct measurements of the neutron capture cross section impossible using current techniques. However, it is possible to measure the total neutron cross section, and hence the resonance properties, using the newly commissioned Device for Indirect Capture Experiments on Radionuclides (DICER) at the Los Alamos Neutron Science Center (LANSCE). Transmission measurements are utilized as a surrogate method to perform capture measurements. The 88 Zr needed for a DICER measurement was produced at the Isotope Production Facility (IPF) and cleanly separated from the production target material. A description of the new instrument, efforts and preliminary results on 88 Zr will be presented.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Second group of irradiation capsules: property data of irradiated welded stainless steel 347 and irradiated, welded, and hydrogen charged Zircaloy-4 for SHINE

Zircaloy-4 and AISI 347 stainless steel were irradiated to low doses (10 20 and 10 21 n/cm 2 ) at low temperature (~60°C) in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). These materials were under consideration for some of the structural components of the medical isotope production facility developed by SHINE Technologies. Zircaloy-4 base metal, as welded, welded and post-weld heat treated (PWHT), and hydrogen charged samples were irradiated. AISI 347 base metal from Sandmeyer Steel, Penn Stainless, and Rolled Alloys suppliers were irradiated. AISI 347 gas tungsten arc welding (GTAW) and flux-cored arc welding (FCAW) samples of Penn Stainless and Rolled Alloys were also irradiated. The irradiated materials were tested with microhardness and tensile tests at room temperature in air. The AISI 347 materials had more elongation after irradiation than the Zircaloy-4 materials. In fact, the Zircaloy-4 welded without PWHT and with H charging had the lowest elongation of any of the materials. This report is the final in a series of reports on this project.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Neutron moderation at IPF

We discuss here the possibility of using the Isotope Production Facility (IPF) at LANSCE to test the feasibility and performance of a graphite cube to create a neutron target. The idea of a neutron target enables the measurement of neutron induced reactions in inverse kinematics. This idea is part of the LANSCE strategy to stay a worldwide leader for neutron-induced research. The proof of-principle is the core of the LDRD project 20240004DR. The first steps are a series of activation experiments with different neutron energy distributions. Figs. 1 and 2 show a graphite cube built and simulated from pieces available at LANSCE. First activation experiments have already been performed at lower neutron energies utilizing (p,n) reactions on Li and Be.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Vision 0.3.2 Users Guide: Revision 0

Vision is a research software application developed by Pacific Northwest National Laboratory (PNNL) that incorporates the scientific and operational expertise for reviewing data from treaty monitoring radionuclide stations. These stations are part of a worldwide network to monitor for nuclear explosions, and the data they produce are critical to make the determination of whether a sample is from a nuclear explosion or some other source (i.e., nuclear reactor or medical isotope production facility). Stations deliver their measurements and system status to the International Monitoring System (IMS), which forwards it via email to all subscribers. Vision provides tools for viewing and analyzing these stations’ state of health data.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2020

Beowulf v2.5.3 User Guide: Revision 5

This document is a user guide for the Beowulf (rebranding of Watchmen) tool. It describes how operators can access and utilize the application's functionality. Beowulf is a research software application developed by Pacific Northwest National Laboratory (PNNL) that incorporates the scientific and operational expertise for reviewing data from treaty monitoring radionuclide stations. These stations are part of a worldwide network to monitor for nuclear explosions, and the data they produce are critical to make the determination of whether a sample is from a nuclear explosion or some other source (i.e., nuclear reactor or medical isotope production facility). Stations deliver their measurements and system status to the International Monitoring System (IMS), which forwards it via email to all subscribers. Beowulf is capable of processing data from several radioxenon station types and development is in progress on a solution for particulate stations. Screening of data in Beowulf may be done by a number of different users such as radionuclide analysts, evaluators, and data quality experts. This guide is provided to assist those users in navigating the application. The term Beowulf is used generically throughout this document to refer to any of the various components in the software application. The user interface that is viewed with a web browser is the primary focus of this user guide. Other components include a database to store measurements and state of health (SOH) data; and the data loader that monitors incoming emails, parses the data, populates the database, does the initial analysis, and routes data for review.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2024

Production, Purification and Characterization of Radioisotopes (Final Report)

This collaborative proposal between the University of Alabama at Birmingham (UAB, PI Suzanne Lapi), and Los Alamos National Laboratory (LANL, collaborator Michael Fassbender) aims to develop novel techniques for isotope production. This proposal takes advantage of complementary techniques available at the UAB cyclotron facility and LANL isotope production facility (IPF) and involved UAB faculty, postdocs, graduate students and undergraduates as well as LANL scientists and staff. This project involved the development of separation techniques for radioisotope production at UAB, using radiotracer analogues produced using the TR24 cyclotron, which would directly translate to separations of the desired isotopes produced at LANL via neutron spallation. Extended visits to LANL by UAB students involved in this initiative were originally envision but were hampered due to COVID restrictions.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2025

Baseline Hypothetical Facility for the Production of 131 I and 99 Mo using Activation Targets

This report describes a hypothetical facility for production of medical radioisotopes via activation under the Proliferation Resistance and Optimization (PRO-X) program. The facility uses neutron activation of non-special nuclear material (SNM) to produce the medical isotopes 131 I and 99 Mo at a throughput of 60 Ci/week of 131 I and 5 Ci/week of 99 Mo. The hypothetical design was carried out using a 10 MWt research reactor. The precursors used for the activation process were TeO2 for 131 I and MoO 3 for 99 Mo. The processes are performed in 3 hot cells used for target receipt, extraction, purification low specific activity (LSA) generator introduction, and packaging. A fourth hotcell is used for waste processing. The hot cell processing area takes up a footprint of 15.4 m 2 with the total footprint of the facility, including space for administrative offices, non-rad labs, quality assurance, and radiation buffer areas set at 763 m 2 . Waste is produced at a weekly rate of 257.8 g low activity solid waste and 8032.7 mL of low activity liquid waste, 8032 mL of which is water. This baseline hypothetical facility for production of medical isotopes via activation was then compared and contrasted to the hypothetical facility for production of medical isotopes via fission products to show the differences in approach for the two production modes. The two production modes had several highlighted differences including the overall facility and hot cell layout, the type and amount of waste produced by the respective facilities, and economic factors impacting production mode. Finally, a decision tree for which production mode might be more beneficial for an entrant into medical isotope production was developed based on the differences examined and the desired output of medical isotopes desired by the entrant.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗