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At least 919 records · Page 51

Predictive Contaminant Transport Simulation with the P2R Model for the Composite Analysis Recharge Sensitivity Case

The Plateau-to-River (P2R) model is a groundwater flow and contaminant fate and transport (F&T) simulation model used to support remedial activities conducted by the Central Plateau Cleanup Company (CPCCo) at the Hanford Site in south-central Washington State. The overall objective of the saturated zone modeling effort is to provide a basis for making informed remedial action decisions based on descriptions of current and expected future contaminant concentrations in groundwater at decision points within and downgradient of the Central Plateau of the Hanford Site. Specifically, the purpose of this environmental calculation file (ECF) is to describe a recharge sensitivity case of the CA base case. The recharge sensitivity case implements a change in the activity contribution from the vadose zone in the A Trenches Area model (for tritium [H-3] and iodine-129 [I-129]), BC Cribs and Trenches model (for I-129 and technetium-99 [Tc-99]), or the Plutonium Uranium Extraction (PUREX) Area model (for H-3 and I-129). All other simulated inventories are identical to the CA base case. The simulation of fate and transport of contaminants reported in this case will support dose predictions as part of the updated Hanford Site CA.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Hanford Site Composite Analysis: Dose Sensitivity Analyses

This environmental calculation file (ECF) documents the methodologies, assumptions, and results of five analyses that evaluate parameter changes to the revised Hanford Site Composite Analysis (CA) (DOE-RL-2019-52, Composite Analysis for Low-Level Waste Disposal in the Hanford Site Central Plateau). Specifically, this ECF describes the impact on total dose and analyte-specific doses for five sensitivity cases where changes to the following exposure-related parameters were implemented – (1) inventory, (2) recharge rate, (3) external dose conversion factors (DCFs) based on Federal Guidance Report (FGR) No. 15 in (EPA-402-R-19-002, Federal Guidance Report No. 15, External Exposure to Radionuclides in Air, Water, and Soil) instead of FGR No. 12 (EPA 402-R-93-081, Federal Guidance Report No. 12, External Exposure to Radionuclides in Air, Water, and Soil) for groundwater pathway scenario; (4) use of mean instead of 95th percentile intake assumptions for groundwater pathway scenario; and (5) consideration of limited sources based on U.S. Department of Energy (DOE) owned waste sites regulated under DOE O 435.1, Radioactive Waste Management. It should be noted that dose assessments for both base and sensitivity cases are based on the all-pathways representative person groundwater exposure scenario presented in DOE/RL-2022-52, Composite Analysis for Low-Level Waste Disposal in the Hanford Site Central Plateau.

54 ENVIRONMENTAL SCIENCES↗

Tritium Fires: Simulation and Safety Assessment

This is the Sandia report from a joint NSRD project between Sandia National Labs and Savannah River National Labs. The project involved development of simulation tools and data intended to be useful for tritium operations safety assessment. Tritium is a synthetic isotope of hydrogen that has a limited lifetime, and it is found at many tritium facilities in the form of elemental gas (T 2 ). The most serious risk of reasonable probability in an accident scenario is when the tritium is released and reacts with oxygen to form a water molecule, which is subsequently absorbed into the human body. This tritium oxide is more readily absorbed by the body and therefore represents a limiting factor for safety analysis. The abnormal condition of a fire may result in conversion of the safer T 2 inventory to the more hazardous oxidized form. It is this risk that tends to govern the safety protocols. Tritium fire datasets do not exist, so prescriptive safety guidance is largely conservative and reliant on means other than testing to formulate guidelines. This can have a consequence in terms of expensive and/or unnecessary mitigation design, handling protocols, and operational activities. This issue can be addressed through added studies on the behavior of tritium under representative conditions. Due to the hazards associated with the tests, this is being approached mainly from a modeling and simulation standpoint and surrogate testing. This study largely establishes the capability to generate simulation predictions with sufficiently credible characteristics to be accepted for safety guidelines as a surrogate for actual data through a variety of testing and modeling activities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

User Guide to the Advanced Dimensional Depletion for Engineering of Reactors (ADDER) Software (V.1.01)

The Advanced Dimensional Depletion for Engineering of Reactors (ADDER) software is being developed in the Research and Test Reactor (RTR) Program at Argonne National Laboratory to meet the reactor design and analysis needs of the Conversion Program. ADDER is a flexible tool that (1) provides a depletion capability through coupling external neutronics codes with a built-in CRAM solver or external depletion code and (2) provides a user-friendly interface to perform fuel management and criticality search operations. The ADDER software is a Python 3 application written using modern software development practices subject to a compliant implementation of NQA-1 and applicable Department of Energy software quality assurance standards. This report is the user guide for the software release referred to as ADDER v1.0.1. The motivation for a software to have flexible capabilities that ADDER possesses is the need to support a wide variety of geometries that are commonly required in analysis of research and test reactors. These reactors can have complex fuel, experiment, or control material shuffling patterns that persist over several years with many fuel management and partial refueling intervals. The scale of fuel management analysis can require tracking of an inventory that is multiple times the core loading. Many reactors, both power and non-power reactors of various types, will find the features of ADDER useful to facilitate key tasks that a fuel or core design engineer must perform with the convenience of concise input and validated functionality.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

FY2020 Rotating Bed Reactor for Iodine Removal from WTP Melter Effluent to Support DFLAW Start-Up

In Performance Assessment (PA) of the Hanford Integrated Disposal Facility (IDF), 129 I is a significant contributor to the risk. The 2017 IDF PA accounted for 0.22% of the total 129 I inventory in Hanford tank waste as residing in the solidified liquid waste from the Effluent Treatment Facility (ETF). However, recent estimates suggest that higher percentages of the iodine processed at the Hanford Waste Treatment and Immobilization Plant (WTP) may be present in the liquid waste routed to the Liquid Effluent Retention Facility (LERF) for eventual treatment at the ETF. In addition, a recent integrated melter test performed at the Vitreous State Laboratory (VSL) at The Catholic University of America (CUA) showed the possibility that a significant fraction (>50%) of the total iodine in the melter feed could be present in the aqueous waste stream destined for the ETF. This stream represents a combination of the Effluent Management Facility (EMF) evaporator overhead condensate and the liquid waste from the WTP off-gas caustic scrubber. Higher levels of 129 I in this stream could affect the LERF and ETF Hazard Categorizations as well as the acceptability of the solidified ETF liquid secondary waste brine for disposal at the IDF. Accordingly, there is a need to investigate and develop potential methods and processes to mitigate this risk. Washington River Protection Solutions, LLC (WRPS) tasked Atkins and VSL to perform development and testing work to evaluate potential methods for removal of iodine species from the waste stream routed to the LERF. In particular, the testing investigated the use of an Atkins-SpinIonic™ Rotating Bed Reactor (RBR) system as a potential alternative to conventional column-based ion exchange or absorption processes. The present report provides results obtained from testing with an RBR system for the removal of iodine from simulants of the projected WTP waste stream that discharges to the LERF.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

MIS Shelf-Life Final Report for Plutonium Oxide Item 1000089 (SSR136 and SSR136A) from Rocky Flats Peroxide Precipitation and Calcination Process

A plutonium dioxide material from the Material Identification and Surveillance (MIS) Program inventory has been studied to determine the gas generation and corrosion behavior in a storage environment. Sample 1000089 represents product-quality plutonium oxides produced in the Rocky Flats peroxide precipitation process currently stored in 3013 containers. This study followed over time the gas pressure and composition of a sample with nominally 0.5 wt% water in a sealed container with an internal volume scaled to 1/500th of the volume of a 3013 container. Gas compositions had been measured periodically over almost 13 years. The maximum observed gas pressure was 204 kPa and was related to a temperature excursion. The maximum observed gas pressure from gas generation was 126 kPa. The increase over the initial pressure of 86.7 kPa was due to the generation of hydrogen, oxygen, and nitrogen. Carbon dioxide, methane and carbon monoxide were minor components of the headspace gas. The material exhibited unusual behavior in that the atmosphere reached flammable levels of hydrogen and oxygen within the first 30 days and remained flammable for the duration of the experiment. It has been determined that the sample inside the reactor was a mixture of the AR material, material calcined at 800 °C, and material calcined at 950 °C. The experiment was terminated in 2012 and unloaded in 2016. A new reactor, SSR136A, was loaded with the original sample freshly calcined at 950 °C. The total pressure inside SSR136A remained close to the initial value of 93 kPa. The total pressure is composed of He (steady levels after the reloading) and hydrogen which is 10% of the reactor’s total pressure. Nitrogen and oxygen compose trace amounts of the reactor headspace. The internal components of SSR136 were inspected in 2021, five years after it was removed from the array. The inside surfaces of the inner bucket exposed to the original sample for 4,576 days had an etched appearance, and pit-like features were observed in the microscopic analysis. No corrosion was observed on the SSR136A inner bucket that held the freshly calcined sample that was exposed for 1,303 days.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

SCALE Modeling of the Fast Spectrum Heat Pipe Reactor

As part of the severe accident analysis collaboration with Sandia National Laboratories (SNL) and the Nuclear Regulatory Commission (NRC), SCALE models were developed for a fast-spectrum heat pipe reactor. These models were based on the Idaho National Laboratory (INL) Design A concept, which is an alternative design to the Los Alamos National Laboratory (LANL) Special Purpose Reactor (SPR), also known as the Megapower reactor. The model contains 1,134 heat pipes, surrounded by hexagonal fuel elements, with a potassium working fluid; the fuel is UO 2 with 19.75 wt% 235 U enrichment. The model contains axial beryllium oxide (BeO) reflectors above and below the active fuel region along with a radial alumina reflector containing 12 B 4 C control drums. The center of the core is left unfueled to make room for two shutdown control rods, one annular and one solid. The active region of the core was discretized into twenty axial and five radial zones to analyze spatial variations in power and burnup. Infinite lattice unit cell sensitivity studies were used to perform verification between the SCALE and INL models. The eigenvalue results agreed well with the reported results to within roughly 50 percent mille (pcm). Full-core model verification was performed by analyzing system eigenvalues with differing configurations of control drum and shutdown rod positions. These full core results all had eigenvalue differences less than 310 pcm. Control drum and shutdown rod worths were also compared, with differences of 3.2% or less. Using the verified model, the isotopic inventory and decay heat, as well as temperature feedback coefficients, were calculated and provided to SNL as input to the MELCOR severe accident code to analyze potential releases from this class of reactor. The results of the MELCOR analysis are provided in a different report.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

GREET-Based Interactive Life-Cycle Assessment of Biofuel Pathways: User Manual

An interactive, web-based tool was developed to streamline the process design for biofuel pathways that the Department of Energy’s (DOE) Bioenergy Technologies Office (BETO) is developing. The tool utilizes the latest life-cycle analysis (LCA) data in the Greenhouse gases, Regulated Emissions, and Energy use in Technologies (GREET) model developed by Argonne National Laboratory (Argonne National Laboratory, 2021). It provides techno-economic analysis (TEA) modelers with a interative interface that generates real-time LCA results based on life-cycle inventory (LCI) data and generates useful insights into key emissions drivers. Economic implications of the LCA results are also included in the tool. Overall, the tool aims to assist TEA researchers with the development of economically viable and environmentally beneficial biofuel technologies. This manual introduces the interface and analysis capabilities of the tool.

09 BIOMASS FUELS↗

Measurement of radiolytic hydrogen generation from surrogate zr-based materials in a "mini-canister"

This report describes a test campaign to investigate radiolytic hydrogen generation from surrogate materials resembling the (Zr-based) cladding of commercial spent nuclear fuel (SNF) with an inventory of residual water post-dryout. The radiolytic hydrogen generation rates from residual waters in the SNF-in-canister system are important for predicting the in-canister environment for sealed SNF dry storage canisters over time and the impact to continued safe dry storage.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Status of the Slide Rule Update - Phase 4 Plutonium Configurations – Delayed Gamma [Slides]

The fourth phase of Slide-Rule update gives a calculation scheme for DFG doses and its application to five plutonium material, as well as comparisons between radiation transport and depletion codes such as MCNP, SCALE, COG and FISPACT. Overall, DFG dose rates calculated by each laboratory led to consistent results. Extra effort is under way to identify the cause of the remaining differences, especially for case 2, by comparing precisely the gamma source, and particularly nuclides inventories.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

MIS Shelf-Life Final Report for Plutonium Oxide Item CAN92 (SSR147) from Rocky Flats Analytical Laboratory Operations

A plutonium/uranium oxide material from the Material Identification and Surveillance (MIS) Program inventory has been studied to determine the gas generation and corrosion behavior in a storage environment. Sample CAN92 represents plutonium/uranium oxides stored in 3013 containers. The material originated in the analytical laboratory at Rocky Flats. This study followed over time the gas pressure and composition of a sample with nominally 0.5 wt% water in a sealed container with an internal volume scaled to 1/400th of the volume of a 3013 container. Gas compositions had been measured periodically over 5 ½ years. The maximum observed gas pressure from gas generation was 117 kPa. The increase over the initial pressure of 77.5 kPa was largely due to the generation of nitrogen and carbon dioxide with minor amounts of hydrogen and nitrous oxide. The internal components of SSR147 were inspected 6.5 years after it was removed from the array. The inside surfaces of the inner bucket exposed to the original sample for 1,937 days had an etched appearance and general corrosion. No pitting was observed on the inner bucket that held the sample.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Safety Testing of AGR-5/6/7 Compacts 2-2-2 and 2-2-4

Compacts 2-2-2 and 2-2-4 from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program’s final irradiation campaign (AGR-5/6/7) were each separately subjected to a thermal soak at 1600°C for 300 h under flowing helium to simulated conditions experienced during a postulated high-temperature gas-cooled reactor (HTGR) depressurization conduction cooldown event. The safety testing of the fuel compacts’ performance under accident conditions was conducted in the Oak Ridge National Laboratory (ORNL) Core Conduction Cooldown Test Facility (CCCTF), which utilizes a stand-alone hot cell at the Irradiated Fuels Examination Laboratory (IFEL). The CCCTF heats fuel to temperatures up to 1900°C in a non-oxidizing environment while continuously monitoring the sweep gas for radioactive 85 Kr to detect release levels associated with hermetic failure of the tristructural-isotropic (TRISO) coating surrounding each fuel kernel. In addition, certain metallic radionuclides that escape the fuel compact are collected on deposition cups that are periodically exchanged with a new cup to obtain information on the overall retention behavior of the TRISO coating layers. Because cesium can diffuse through intact pyrocarbon layers, abnormal degradation of the silicon carbide (SiC) layer in the absence of holistic TRISO coating failure is indicated by release of 134 Cs at levels equivalent to an individual particle inventory in the absence of significant 85 Kr release (Hunn et al. 2014). Additional description of the CCCTF system is provided in Appendix A.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

PELICAN Design, Test Planning, and Commissioning Results

To support design efforts for the Versatile Test Reactor (VTR) core assemblies, an experimental facility has been designed and constructed at Argonne National Laboratory to match the hydraulic flow conditions within the VTR’s primary heat transport system (PHTS). This facility, the Pressure drop Experimental Loop for Investigations of Core Assemblies in advanced Nuclear reactors, PELICAN, provides the ability to measure pressure drop across a full-scale fuel assembly containing prototypic axial reflectors, fuel, and plena components. The PELICAN facility was designed and built to offer maximum flexibility, allowing testing from short sub-sections all the way to the full-length core fuel assemblies. The report presents the high-level program objectives, a summary of the facility design, instrumentation, and control systems, and the testing procedure. The outcomes from facility characterization efforts and first test matrix results are then presented and then, finally, the conclusion contains a summary of the future work to be performed. This test facility was designed to match the hydraulic conditions of the flowing sodium in the VTR using water as a surrogate fluid. To do so, the water is elevated to a temperature of 110°C where its viscosity matches that of sodium, and with a 50-HP centrifugal pump, is capable of generating full scale flow rates to achieve prototypic Reynolds and Euler number flow conditions. To prevent boiling, the system is maintained at elevated pressure of at least 2.7-3.0 bar (40-44 psig). A set of 15 tests have been used to perform checkout activities in order to commission the device, survey the capabilities of PELICAN, and generate experimental data at isothermal conditions and over a range of flow rates up to 44 kg/s (700 GPM) to produce datasets used in the verification and validation of VTR design and modeling efforts. Initial testing focused on facility characterization, which assessed the operational capabilities of various control systems. The thermal control system was qualified, including the thermal response of the loop to the heat added by self- and auxiliary heaters, the chiller and heat exchanger systems for removing excess heat, and their coupled ability to maintain steady-state isothermal conditions as desired. Additionally, the pressure control system was verified to ensure necessary operating environments that promote pump health and prevent boiling of loop inventory at high temperatures could be met. With these systems in place, the first set of matrix tests have been carried out using orifice plates with inner diameters of 2.25-inches and 2.5-inches. Orifice plate flow behavior is well covered in scientific literature, and the results find good agreement with the measured pressure drop vs flow rate and those of theoretical predictions. This initial testing not only helps to validate the experiment, but it also produces a high quality data set for a geometry that is easily reproducible for simulations. Finally, the report is concluded with a discussion of the work to come and provides a snapshot of the test articles in the experimental pipeline whose designs are being inspired by the current designs from the VTR.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Engineering evaluations of nuclear material storage containers against the packaging requirements at TA-55 (FY22 Rules of Use Report)

A set of obligations in the yearly storage container surveillance program, ensures that a retrospective approach is applied with loaded containers in inventory to confirm compliance with all “users” to ensure nuclear material storage container requirements are being followed. Validation that packaged containers are properly used, within a specific set of container types prescribed “bounding conditions of use,” is based on meeting requirements as identified in operating procedures associated with TA55-DOP-091, TA-55 Nuclear Material Packaging, and PA-RD-01022, Nuclear Material Packaging Requirements. The surveillance plan obligation requires the application of local area nuclear material accountability software or (LANMAS) to produce queries. The data queries are used to assess attributes of containers in storage these investigations are conducted on a bi-annual basis. This rules of use (ROU) compliance process ensures proper usage of storage containers as containment systems and therefore provides effective worker protection. It is structured as a supporting effort by implementing aspects of PA-AP-01207, Nuclear Material Container Safety Management at TA-55.

36 MATERIALS SCIENCE↗

Examining United States Nuclear Weapon Educational Initiatives_A Quantitative Research Study on the Satisfaction of Participants Within the Nuclear Weapons Community

This study focused on a few United States nuclear weapons educational initiatives exploring six classroom environment factors that may or may not influence Satisfaction in a virtual or live classroom environment within the nuclear weapons community. The research explored Personalization, Involvement, Student Cohesiveness, Task Orientation, Innovation, and Individualization. The United States nuclear weapons community needed to attract new and talented individuals to the nuclear weapons profession. This study added to the nuclear weapons community by researching factors that may aid academic instructors, and curriculum developers create new and inventive ways of incorporating these factors into existing or new nuclear weapons educational courses. Convenience sampling was the data collection method used to gather quantitative research data. One hundred and seven voluntary participants utilized the College and University Classroom Environment Inventory with the Los Alamos National Laboratory's consent. There is not much concurrent research on Satisfaction within the live and virtual classroom environment. This research was vital for future research into how to increase participant satisfaction. The study provided the nuclear weapons community with grounded analysis to help create and guide new classes and course creation, possibly increasing Satisfaction in a live and virtual classroom environment. According to binary logistic regression, there was a statistically significant relationship among the dependent variable, Satisfaction, and two independent variables, Involvement, and Task Orientation in the classroom environment within the nuclear weapons community. Future nuclear educational programs may increase Satisfaction in the classroom environment by increasing participant Involvement and Task Orientation within a class using the CUCEI as a reference.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗

Design, Deployment, and Characterization of the World’s First Flexible Large Power Transformer

GE Research and its partner Prolec GE have designed, built and deployed in the field the world’s first flexible power transformer. The flexible power transformer is a transmission class 3-phase autotransformer configurable in impedance and in voltage which allows it to serve as a universal spare for multiple units in a given fleet. However, the key innovation in this new concept is the online adjustable leakage impedance which allows the transformer to change its impedance without interrupting the transmission line operation. The flexible power transformer can be designed with up to three low voltage transmission class ratings and up to 12 impedance values changeable both online and offline. This report provides an overview of the design, manufacturing, testing, and commissioning of the 165kV, 60MVA prototype built including the results of the field performance validation tests. The prototype was specified in collaboration with Cooperative Energy, the utility host. It was designed and tested in the factory according to IEEE standard C57.12.00 and followed all protocols for transportation, installation, and commissioning of a power transformer. In addition to the prototype, a flexible protection system capable of automatically adjusting its settings upon the transformer impedance was also developed and deployed in the for testing and validation. On September 3, 2021 the prototype was energized in Cooperative Energy’s substation in Columbia, Mississippi to become the world’s first flexible power transformer in operation. Its performances and impact on the grid operation were demonstrated through different field tests. Results obtained confirm that the impedance of the flexible transformer can be varied under load through its full range, from 4.3% to 9.3%, without adverse impacts on the line operation, the protection system, the transformer stability and health condition. Results also proved that the flexible transformer is very effective in controlling power transfer through the line or load sharing between units operating in parallel. Indeed, it was proven that higher impedances decrease the thruput power of the transformer while lower impedances increase it. Up to 26MW was controllable on a line loading of 45MVA. It was also possible to demonstrate that the variation of the transformer impedance has no effect on the circulating current between units in parallel, except a minor transient during the impedance change. It was also proven that the flexible protection relay can update its protection settings automatically when the impedance change was detected. The prototype has operated continuously for more than 12 months now with a peak load exceeding 50MVA corresponding to >80% of its ONAN power rating. No alarm, trip or sign of failure has been reported by the utility. In addition to the development and deployment of the flexible transformer prototype, investigations were carried out on new nanodielectric fluids to replace the mineral oil used in power transformers with the goal of reducing their footprint and weight. The key parameters that were targeted for improvement included the breakdown voltage to reduce clearances between windings and tank hence the footprint; viscosity and thermal conductivity to increase the cooling efficiency and therefore to reduce the winding material. Several nanodielectric mixtures with mineral oil including with alumina (Al2O3), titania (TiO2) and Borum Nitrate (BN) with different surfactants have been analyzed and tested. Unfortunately, despite encouraging results no nanofluid candidate has been found viable to replace mineral oil. With the formulations tested, breakdown voltages are generally similar to mineral oil at lower particle contents and worse at higher particle contents. Viscosity appreciably increased at particles concertation over 2 wt% and thermal conductivity increased slightly at 5wt% and appears to be 10-15% higher at 10 wt% particle content. It is recommended to continue investigations to find solutions that can help increase the power density of future flexible power transformers. Flexible power transformers can significantly help the future power grid by providing more flexibility and resiliency. Indeed, by providing voltage and impedance flexibility, flexible power transformers reduce the need for multiple spares, hence inventory costs for utilities. With their online controllable impedance, they can provide support to the grid and help manage short-circuit currents, power flow, line congestion, and grid stability which will become more important with higher penetrations of intermittent renewable resources. During the field validation tests, it was demonstrated that up to 26MW was controllable on a line loading of 45MVA when the transformer impedance was varied from its minimum to its maximum range. Also, with the impedance range, the short-circuit currents could be reduced by up to 38% at the load side of the transformer. With its controllable impedance, flexible transformers can be used in future strategies of grid resilience to help better prepare the grid to face forecasted severe events including storms, heat waves and contingencies. The flexible power transformers can also find role in other applications including high voltage transmission cables such as offshore wind farms where solutions for energizing the cables and managing the reactive power are of critical importance. The designed flexible power transformer is now fully validated and ready for commercialization. Further analysis on the benefits of flexible power transformers for grid stability and short-circuit management including current limiting capability, reclosure and line restoration, control of inrush current, sizing of flexible AC components (FACTS) would help its rapid adoption by the industry.

24 POWER TRANSMISSION AND DISTRIBUTION↗

Tier 2 Safety Basis Document: Weapons and Complex Integration - Building 225

This Safety Basis Document (SBD) covers operations in the Weapons and Complex Integration (WCI) Building 225 (B225) facility. B225 is located in the southwestern quadrant of Lawrence Livermore National Laboratory (LLNL) and is operated and owned by WCI. This SBD satisfies the requirements of LLNL ES&H Manual Document 3.1, Nonnuclear Safety Basis Program (Ref. 2). B225 is classified as a Low Hazard Facility based on the inventory of radiological materials. Other hazards in B225: chemicals, radiation generating devices, and industrial hazards are managed at the Light Science and Industry (LSI) level. Biological and explosive hazards are not authorized in B225. There are four general Administrative Controls (ACs) identified as the B225 Operational Safety Requirements (OSRs). These administrative controls are deemed sufficient to ensure safe operations with respect to co-located workers and the public. Other safety features and administrative requirements are required by the LLNL ES&H Manual, including those to protect the facility workers and the environment, but these have not been selected as credited controls to protect the co-located workers or the public, which are the target populations of concern for the SBD.

61 RADIATION PROTECTION AND DOSIMETRY↗

Knowledge Transfer and Training in Safeguards for Decommissioned Nuclear Facilities (Project Final Report for FY2022)

The International Atomic Energy Agency (IAEA) applies safeguards to nuclear facilities that are not operating, including those undergoing decommissioning, and the IAEA’s effort in this area is both considerable and increasing. Specifically, the IAEA Department of Safeguards’ Division of Concepts and Planning (SGCP-003: Safeguards Approaches) identified an R&D need to “Develop safeguards implementation guidelines for facilities under decommissioning and safeguards concepts for post-accident facilities under decommissioning”. Nuclear facilities undergoing decommissioning are not exempt from safeguards agreements between the IAEA and Host State, and, accordingly, the requirement for verification of no diversion of nuclear material and detection of undeclared activities at decommissioned facilities remain even after facility shutdown. However, the effort required to meet safeguards objectives diminishes as nuclear material and essential equipment are removed during the decommissioning process which shifts the emphasis from verification of ever-diminishing fissile or source material inventories to verification of changes in facility design and equipment operability.

96 KNOWLEDGE MANAGEMENT AND PRESERVATION↗