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All sourcesNASA NTRSDOE OSTI
NASA NTRSDate not supplied

New Hayabusa2 and OSIRIS REx Curation Facilities at Nasa Johnson Space Center

Asteroids are made up of rocky material that is left over from the formation of our solar system [1]. This material holds the key to unlocking information about the history of the planets and our sun. Hayabusa2 and OSIRIS-REx are two sample return missions that have collected rocky material from carbonaceous asteroids which are rich in water, carbon, and organic compounds [1]. Hayabusa2 is a spacecraft that collected samples from asteroid Ryugu launched by the Japan Aerospace Exploration Agency (JAXA). Samples were returned to Earth in December 2020 and 10 percent of the collected material is currently housed in a state-of-the-art laboratory at NASA Johnson Space Center (JSC) dedicated to Ryugu material [1]. OSIRIS-REx is a spacecraft launched by NASA that collected samples from asteroid Bennu. The samples are on their journey to Earth and will arrive in September 2023. They will be stored in a separate cleanroom dedicated to Bennu material. From these two missions, scientists will examine dust particles and rocky material to research clues on how the early solar system formed and how life began [2]. Because these samples are so precious and rare, it is important to keep them free from Earth contamination. In September of 2020, a massive construction project began to provide curation facilities to meet the needs of two incoming asteroid collections, Hayabusa2 and OSIRIS-REx. Despite numerous setbacks caused by the coronavirus pandemic, these facilities have been completed in the fall of 2021. Some of these setbacks included strict PPE requirements, keeping safe working distances as well as access to the facilities as NASA was operating under a Stage 3 condition in response to the pandemic. These facilities will provide curation laboratories to safely house the precious asteroid samples as well as facilities to manipulate samples, do sample preparation and process samples into and out of the labs for future studies by the scientific community. The lab suite consists of several different rooms including an initial anteroom (ISO 7), a staging area (ISO 7), two gowning areas (one for OSIRIS-REx and one for Hayabusa2; both ISO 6)), the Haybusa2 Curation Laboratory (ISO 5), the OSIRIS-REx Curation Laboratory (ISO 5), an ultramicrotomy laboratory (ISO 7) and a thin section laboratory (not a cleanroom). To maintain cleanliness levels necessary for these lab spaces quarterly to monthly Balazs organic and inorganic sampling has been conducted on a regular basis before, during and after the construction project to monitor the cleanliness levels in the new labs that will house the asteroid samples. These new facilities feature a Ruuvi (Grafana) monitoring system that measures temperature, humidity, and pressure of the lab suite. This is a small puck shaped sensor that sends signals to the Grafana cloud where the data can be plotted and set up on dashboards within the lab suite. These sensors are easy to set up and install and they run off a cell battery that is good for 3-5 years before needing replacement. This system can deliver text messages and emails to the appropriate personnel when values fall outside of set limits. In addition to the monitoring system, this new lab suite features viewing windows from several different locations and sliding motion censored doors. We have obtained some state-of-the-art equipment to outfit the lab spaces (desiccators, microscopes, ultramicrotome) and are in the process of outfitting the suite with additional items such as: nitrogen gloveboxes, additional desiccators, micromanipulators, etc. The Hayabusa2 samples provided to NASA by JAXA included 23 millimeter sized grains and 4 containers with finer material. OSIRIS-REx is expected to return as much as 400 grams of asteroid Bennu. Samples will be stored in a nitrogen glove box and separated into nitrogen sealed transfer containers. Small Bennu and Ryugu particles will be handled and prepared (e.g., ultra-thin sections for scanning electron microscopy (SEM) or transmission electron microscopy (TEM)) using an Axis Pro micromanipulator and Leica EM UC7 ultramicrotome. Analyzing samples at the atomic scale will provide greater insight to the origins of the solar system and life formation on Earth. All the curation preparation for these sample collections is essential to ongoing research and efforts to understand our solar system, both nowand for future generations.

Curation↗
NASA NTRSDate not supplied

Artemis Crewed Surface Operations Training Dust Mitigation Test/Training Facility (DMTF)

Based on several Apollo mission(s) reports and lessons learned1 on lunar dust impacts, it is imperative for successful Artemis missions, to develop dust mitigation methods & train crew in dust cleaning techniques for efficient use of lunar surface time and for crew health. Longer durations and increased number of Extra Vehicular Activities (EVAs) for Artemis need better training. Apollo missions lacked training in a dust facility. Current facilities do not provide adequate fidelity to train crew in dust cleaning operations in suited, pressurized environments. Mimicking dust cleaning actions in existing facilities (eg- Neutral Buoyancy lab) does not provide in depth understanding of actions and time needed to clean suit or positions crew need to take to remove dust from suit. Available dust bins do not support suited/pressurized ops, personnel not trained in suited ops, more overhead/travel. A local facility at Johnson Space Center (JSC) w/ adequate fidelity supporting suited/pressurized ops, personnel in suited ops is a crucial need for rapid development/testing and training. This facility will provide NASA capabilities to: 1. Provide a dust containment bin to perform suited pressurized crewed training/testing activities for dust interaction/mitigation leveraging existing JSC facility infrastructure and 2. Perform end-to-end simulated lunar dust testing of fluid, mechanical, and electrical systems to provide “test like you fly” confidence for industry developed solutions.

EVA training↗
DOE OSTI2017

EFFECT OF INCLINED WAVES ON DEEPLY EMBEDDED NUCLEAR FACILITIES

Several conceptual designs for advanced reactors show that most if not all of the structure is embedded in the soil. Consequently, analysis techniques that have been verified and validated for the analysis of light water reactors on surface foundations may not be directly applicable to deeply embedded reactors. Specifically two effects that may change the in-structure response (ISR) of embedded facilities are: (i) the effect of the kinematic interaction becomes more prominent as the embedment depth increases; (ii) non-vertically propagating waves, which are usually not considered for surface-founded facilities, may be more important for deeply embedded facilities. A preliminary study is presented in this article to understand the response of embedded nuclear facilities to non-vertically propagating seismic waves. The inclined wave field is generated from source-to-site simulations using representative earthquake faults with different dip angles. The embedment depth of the nuclear containment facility is also varied to obtain a better understanding of the effect of embedment. This analysis of the soil-structure ensemble is conducted using the application called Multi-hazard Analysis for STOchastic time-DOmaiN phenomena (MASTODON), developed on top of Idaho National Lab’s open source MOOSE finite element framework. The Domain Reduction Method implemented in MASTODON to incorporate the seismic excitation within the region of interest is verified and then used to reduce the computational effort in analyzing this problem. Preliminary analysis using linear soil and structure material properties and tied contact between the soil and structure indicate that embedding the nuclear facility reduces significantly the response of the structure in all the earthquake scenarios considered.

42 ENGINEERING↗
DOE OSTI2018

The Fermilab Test Beam Facility Data Acquisition System Based on $otsdaq$

The Real-Time Systems Engineering Department of the Scientific Computing Division at Fermilab has deployed set of customizations to our Off-The-Shelf Data Acquisition solution (otsdaq) at the Fermilab Test Beam Facility (FTBF) to read out the beamline instrumentation in support of FTBF users. In addition to reading out several detectors which use various detection technologies and readout hardware, the FTBF Data Acquisition system (DAQ) can perform basic track reconstruction through the facility in real time and provide data to facility users. An advanced prototype smart coincidence module, known as the NIM+, performs trigger distribution and throttling, allowing the beamline instrumentation to be read out at different rates. Spill data are saved to disk for studies of the facility performance, and hit data are also made available on the FTBF network for experiments' use. A web-based run control and configuration GUI are provided, and the online monitoring snapshots created after each beam spill are viewable from any computer connected to the Fermilab network. The integrated DAQ system for the facility provides users with high-precision tracking data along the beamline and a single location for obtaining data from the facility detectors, which set the baseline for testing their own detectors.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

FEMP Facility Evaluation (Audit) Decision Tree

This document is a resource to be used in concert with FEMP Audit Definitions, FEMP Consolidated Facility Management Guidance, and agency best practices and expert judgment—for agencies to use to determine the best approaches to evaluate their facilities. It features a decision tree that provides a framework for agencies to interpret gathered facility data to understand the existing conditions; benchmark the facility against others in the portfolio to determine whether an on-site or remote audit is appropriate for a facility; determine the best remote or on-site audit level of detail to meet agency facility auditing goals, such as meeting Energy Independence and Security Act of 2007 requirements and developing projects; and improve their energy and water management programs.

auditing↗
DOE OSTI2016

Comparison of Calculated and Experimental Results for a Boiling/Condensing Experimental Facility

A new experimental facility is being developed for materials irradiation and testing at the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR). Details of this facility have been presented before. A prototype of this facility, the Thermosyphon Test Loop (TSTL) has been built, and experimental data have been obtained and analyzed. Pretest calculations for this facility with the RELAP5-3D code have been presented previously as well as other calculations with the TRACE code. The results of both codes were very different. RELAP5-3D predicted much higher pressures and temperatures than TRACE. This paper compares calculated results with the TSTL experimental data. Comparison of calculations with the codes RELAP5-3D and TRACE with experimental data of the new TSTL facility has shown that TRACE results agree well with the data and that RELAP5-3D calculates very high pressures and temperatures. The TRACE code is well suited to model this facility and is being used for future calculations. (authors)

42 ENGINEERING↗
DOE OSTI2017

Computer Security for I and C Systems at Nuclear Facilities - 116

Cyber-attacks on Instrumentation and Control (I and C) systems may jeopardize the safety and security of nuclear facilities by contributing to sabotage or aiding in the unauthorized removal of nuclear material. The effects of cyber-attacks on I and C systems important to safety to safety may result in a wide range of consequences, such as a temporary loss of process control or unacceptable radiological consequences. Public awareness of cyber-attacks that affect I and C systems may also undermine confidence in the safety and security of nuclear facilities. The International Atomic Energy Agency (IAEA) recognizes the need for the guidance on the protection of I and C systems that provide safety, security, or auxiliary functions at nuclear facilities from cyber-attacks. To meet this need, The IAEA has developed a new Nuclear Security Series publication with the provisional title Computer Security of Instrumentation and Control Systems at Nuclear Facilities which builds upon and expands the guidance published in Nuclear Security Series No. 17 Computer Security at Nuclear Facilities (2011) . This publication will be formally released in 2017 as Technical Guidance and will be the second international consensus publication dedicated to Computer Security at Nuclear Facilities. It was developed in parallel with the new IAEA Nuclear Safety publications providing recommendations on I and C systems at research reactors (RR) and nuclear power plants (NPP) . The purpose of this paper is to provide an overview of this publication and its application. The paper will also discuss its relation to other IAEA Nuclear Security Series (NSS) publications on computer security. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Progress Towards the Validation of a new RELAP5-3D model of the High Temperature Test Facility

Validation is a key step in the development of any type of systems model. As the next generation of reactors approaches, the need for codes that have been validated for these new types of systems continues to grow. An example of a prominent option is the Reactor Excursion Leak Analysis Program (RELAP5-3D), developed by Idaho National Laboratory. This code was developed for the purpose of systems level thermal-hydraulic modeling of light water reactors (LWRS) and postulated transients that can occur in LWRS.RELAP5-3D has been substantially validated against LWR data. Due to its long history as a reactor safety analysis tool, there has been an effort to adapt RELAP5-3D for the purposes of advanced reactor concepts such as prismatic high-temperature gas-cooled reactors (HTGRs). However, RELAP5-3D has not nearly been validated and verified for HTGRs to the degree of LWRs, warranting verification and validation opportunities with computational benchmarks and existing experimental facilities. Examples of such facilities include the modular high-temperature gas-cooled reactor (MHTGR) 350 and the high temperature engineering test reactor (HTTR) from Japan. The MHTGR 350 is a benchmark design concept for code-to-code verification purposes; therefore, it does not provide any experimental data for validation opportunities The HTTR provides useful multiphysics validation data but does not have the in-core instruments to generate thermal-hydraulic experimental data to help with RELAP5-3D validation. Consequently, a facility that could provide key in-core temperatures for thermal-hydraulic validation was still needed. The High Temperature Test Facility (HTTF) is an integral effects facility for HTGR thermal hydraulics developed and operated by Oregon State University. HTTF represents ¼ length scale of the General Atomics MHTGR and is rated for a total power of 2.2 MW. Axially, the core consists of an upper and lower reflector and 10 blocks, numbered from bottom to top (Block 1 is right above lower reflector). The core is heated via graphite resistive heater rods, with respective channels distributed throughout the core. The primary coolant is helium and heat can radiate out of the core to the reactor cavity cooling system (RCCS), which is cooled by water. The primary purpose of the facility is to investigate pressurized conduction cooldown (PCC) and depressurized conduction cooldown (DCC) transients, which are also referred to as the pressurized and depressurized loss of forced cooling respectively. Two experiments were chosen to perform the validation study with a RELAP5-3D model of HTTF. These experiments are PG-27 (PCC) and PG-29 (DCC). These were chosen based off of the quality of available experimental data before and during the experiment which led to their inclusion in the HTGR Thermal Hydraulics Benchmark.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Gas migration in low- and intermediate-level waste (LILW) disposal facility in Korea

The first low- and intermediate-level waste (LILW) disposal facility is under construction in saturated granite in Korea. The safety assessment report (SAR) identified that different gases, such as hydrogen, carbon dioxide, and methane are generated at the disposal facility due to the corrosion of metal wastes and steel drum, and microbial degradation of organic matters. Reinforced concrete plays a role as an engineered barrier at the disposal facility, so its properties with regard to gas migration were evaluated in laboratory-scale experiments. Then modeling of gas migration was carried out to evaluate gas pressure build-up in the disposal facility. The gas entry pressure and relative gas permeability of the concrete was determined to be 0.97±0.15 bar, and the relative gas permeability decreased exponentially with increasing water content. The results of the modeling showed that most of hydrogen gas was dissolved in groundwater and did not significantly influence pressure build-up inside the disposal facility based on the reference case of gas generation. (authors)

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

Evaporation Of Hanford Waste Treatment Plant Direct Feed Low Activity Waste Effluent Management Facility Core Simulant

The Hanford Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) vitrification facility will generate an aqueous condensate recycle stream (LAW Melter Off-Gas Condensate, LMOGC) from the off-gas system. The baseline plan for disposition of this stream during full WTP operations is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation, and recycled to the LAW vitrification facility. However, during the Direct Feed LAW (DFLAW) scenario, planned disposition of this stream is to evaporate it in a new evaporator, in the Effluent Management Facility (EMF), and then return it to the LAW melter. It is important to understand the composition of the effluents from the melter and new evaporator, so that the disposition of these streams can be accurately planned and accommodated. Furthermore, alternate disposition of the LMOGC stream would eliminate recycling of problematic components, and would reduce the need for closely integrated operation of the LAW melter and the Pretreatment Facilities. Long-term implementation of this option after WTP start-up would decrease the LAW vitrification mission duration and quantity of glass waste, amongst the other operational complexities such a recycle stream presents. In order to accurately plan for the disposition path, it is key to experimentally determine the fate of contaminants. To do this, testing is needed to accurately account for the buffering chemistry of the components, determine the achievable evaporation end point, identify insoluble solids that form, and determine the distribution of key regulatory-impacting constituents. The LAW Melter Off-Gas Condensate stream will contain components that are volatile at melter temperatures, have limited solubility in the glass waste form, and represent a materials corrosion concern, such as halides and sulfate. Because this stream will recycle within WTP, these components will accumulate in the Melter Condensate stream, exacerbating their impact on the number of LAW glass containers that must be produced. Diverting the stream reduces the halides and sulfates in the recycled Condensate and is a key outcome of this work. This overall program examines the potential treatment and immobilization of this stream to enable alternative disposal. The objective of this task was to demonstrate evaporation of a simulant of the LAW Melter Off-gas Condensate expected during DFLAW operations, in order to predict the composition of the effluents from the EMF evaporator to aid in planning for their disposition. This document describes the results of that test using the core simulant. This simulant formulation is designated as the “core simulant”; other additives will be included for specific testing, such as volatiles for evaporation or hazardous metals for measuring leaching properties of waste forms. The results indicate that the simulant can easily be concentrated via evaporation. During that the pH adjustment step in simulant preparation, ammonium is quickly converted to ammonia, and most of the ammonia was stripped from the simulated waste and partitioned to the condensate. Additionally, it was found that after concentrating (>12x) and cooling that a small amount of LiF and Na 3 (SO 4 )F precipitate out of solution. With the exception of ammonia, analysis of the condensate indicated very low to below detectable levels of many of the constituents in the simulant, yielding very high decontamination factors (DF).

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NASA NTRS1967

Technical facilities catalog, volume II

This Technical Facilities Catalog has been developed in response to an increasing demand for general and specific information on NASA technical facilities. The Catalog is intended to: a)provide NASA management, facilities managers, program and project managers, individual research engineers and scientists, and other agencies of the Government with knowledge of existing NASA research, development, test and evaluation facilities and b)provide the reader with a basic concept of the capabilities and functions of the wide variety of NASA technical facilities and the technological areas they support.

Wind tunnel↗
DOE OSTI2013

Development and Implementation of a Scaled Saltstone Facility at Savannah River National Laboratory - 13346

The Savannah River National Laboratory (SRNL) has supported the Saltstone Production Facility (SPF) since its conception. However, bench scaled tests have not always provided process or performance data related to the mixing, transfer, and other operations utilized in the SPF. A need was identified to better understand the SPF processes and to have the capabilities at SRNL to simulate the SPF unit operations to support an active low-level radioactive waste (LLW) processing facility. At the SPF, the dry premix is weighed, mixed and transferred to the Readco '10-inch' continuous mixer where it is mixed with the LLW salt solution from the Salt Feed Tank (SFT) to produce fresh Saltstone slurry. The slurry is discharged from the mixer into a hopper. The hopper feeds the grout pump that transfers the slurry through at least 457.2 meters of piping and discharges it into the Saltstone Disposal Units (SDU) for permanent disposal. In conjunction with testing individual SPF processes over several years, SRNL has designed and fabricated a scaled Saltstone Facility. Scaling of the system is primarily based on the volume capacity of the mixer and maintaining the same shear rate and total shear at the wall of the transfer line. At present, SRNL is utilizing the modular capabilities of the scaled Saltstone Facility to investigate the erosion issues related to the augers and paddles inside the SPF mixer. Full implementation of the scaled Saltstone Facility is still ongoing, but it is proving to be a valuable resource for testing alternate Saltstone formulations, cleaning sequences, the effect of pumping Saltstone to farther SDU's, optimization of the SPF mixer, and other operational variables before they are implemented in the SPF. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

A French Industrial Management Plan for Very Low Level Radioactive Wastes (VLLW) at the CIRES Facility - 17012

Since 2003, very-low-level waste has been disposed of at the Andra CIRES disposal facility, the first disposal facility in the world for this type. It is now the main route in terms of volumes for the management of French radioactive waste. The 2015 national inventory has identified 1.1 million m{sup 3} of VLLW to be produced until 2030. Several waste management scenarios were then analyzed, including issues associated with the flow reduction of the waste to be stored. The key avenues to reduce the volumes of waste, technical and economic viability of which still needing to be challenged, are at this stage: - Favor the reuse of large amounts of metallic VLLW to be produced within the next decade. - To service the recycling of metals, different opportunities for reuse should be considered -including within the non-nuclear sector. - 30% to 50% of VLLW produced in the so-called 'nuclear waste' areas probably have zero or non-detectable radioactivity levels and questions about the introduction of a clearance level. - A disposal route based on a concept derived from waste storage facility for non-nuclear and non-hazardous wastes could be an alternative to the clearance level for the least active inert waste. Therefore volumetric optimizations appear possible, even if the waste management costs will be higher than the current route. Furthermore, efficiency gains and progresses in operating modes would increase the technical capacity of the CIRES of about 250 000 m{sup 3}. In all cases studied, it appears essential to create around 2030 a new VLLW disposal facility, particularly for the dismantling of the current fleet of reactors. According to the proposed optimizations, the necessary capacity for a second VLLW disposal facility would range between 600 000 m{sup 3} and 1.2 million m{sup 3}. (authors)

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DOE OSTI2015

Remote monitoring of Alpha Gamma hot cell facility by ARG-US RFID system - 15177

Argonne National Laboratory (Argonne) has developed a monitoring and tracking system for nuclear materials based on radio frequency identification (RFID) technology. The system, called ARG-US ('Watchful Guardian'), has been in fielding testing and applications at selected U.S. Department of Energy (DOE) sites since 2010. ARG-US utilizes battery-powered tags to remotely and continuously monitor the environmental parameters, e.g., radiation levels, of the tagged items such as containers of radioactive materials. To demonstrate that the system can monitor nuclear and radiological facilities as well, a compact system was installed in the Alpha Gamma Hot Cell Facility (AGHCF) at Argonne in 2013. AGHCF is a Hazard Category 2 non-reactor nuclear facility that is being decommissioned. It is also being used to support de-inventory of other facilities on the Argonne site. As such, large quantities of radioactive materials and wastes are being discharged from the hot cell, loaded into transport containers, and shipped away. The deployed ARG-US system monitors worker areas, material transfer areas, cell exhaust HEPA filters, and?on a rover/robot?in-cell locations of AGHCF. Thus far, the system is performing as designed and yielding results that can be corroborated by existing surveillance means. The in-cell unit is particularly valuable, as it yields information that would otherwise be hard to obtain. In time, the benefits of automated monitoring 24/7 by ARG-US should become apparent and may lead to a fuller system with more distributed tags in the hot cell areas until the facility is fully decommissioned. (authors)

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

Integration of performance assessments and 3116 basis documents within the Savannah River Site liquid waste facilities - 15375

The end of the Cold War has left a legacy of over 136 million liters (36 million gallons) of radioactive waste in the aging waste tanks at the Department of Energy's Savannah River Site (SRS). A robust program is in place to remove waste from these tanks, treat the waste to separate it into a relatively small volume of high-level waste and a large volume of low-level waste, and to actively dispose of the low-level waste on-site. This process allows the US Department of Energy (DOE) to move forward with closure of the cleaned waste tanks and associated ancillary structures. To support performance-based, risk decision making for closure and disposal activities, three individual performance assessments (PAs) have been developed - salt-stone Disposal Facility, F-Tank Farm and H-Tank Farm. In addition, three Waste Determinations (WDs) have been made by the Secretary of Energy, in consultation with the Nuclear Regulatory Commission (NRC), under the authority of Section 3116 of the Ronald W. Reagan National Defense Authorization Act for Fiscal Year 2005 (NDAA Section 3116). To support these waste determination decisions, individual '3116 Basis Documents' were developed that describe the assumptions and conditions under which the criteria outlined in NDAA Section 3116 are met. The development of each of the PAs represented a long and arduous effort that involved the development of both deterministic and probabilistic conceptual models of the facilities, the development of assumptions and associated ranges, uncertainty and sensitivity analyses, model data collection and analyses, writing of the actual performance assessment, internal and external peer reviews, and ultimately approval by DOE. Similarly, the development of the 3116 Basis Documents represents similar time and financial commitments. At the end of these efforts, after years of preparation and millions of dollars in investment, we have thousands of pages of highly technical text, tables of data and hundreds of graphs and charts all contained in a series of thick binders that fit neatly on the shelf of a bookcase. We were done - right? Actually, we were just getting started. Performance Assessments and 3116 Basis Documents are living documents that shape day-to-day and long-term decision making in the operation of a low-level waste disposal facility (e.g., salt-stone Disposal Facility) or in area closure activities (e.g., F-Tank Farm or H-Tank Farm). Although these documents do not contain specifically described requirements per se, they do utilize specific assumptions that, if changed significantly, could result in no longer being able to demonstrate that there is reasonable assurance that the required performance objectives or the criteria from NDAA Section 3116 can be met. Savannah River Remediation LLC (SRR), the contractor responsible for Liquid Waste operations at SRS, has not only developed the PAs and 3116 Basis Documents to support this work and shepherded these documents through the DOE approval process, but has also developed the processes and procedures to ensure the successful implementation and integration of these documents into day-to-day decision making within the associated facilities. This paper describes these processes, including the Un-reviewed Waste Management Question process and the PA Maintenance process, and provides specific examples of their successful implementation. (authors)

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

Aging Infrastructure: Application of System Health Monitoring at the Defense Waste Processing Facility - 16069

The disposition of high level waste (HLW) into borosilicate glass canistered waste forms began at SRS in 1996. Since that time, the Defense Waste Processing Facility (DWPF) has produced nearly four thousand HLW canisters. The DWPF is now approaching 20 years of radioactive service and 25 years of operation since the commencement of nonradioactive testing in 1990. Many facility systems are challenged by equipment or component obsolescence, which places a premium on proper work planning, engineering, and maintenance resource allocation to ensure facility reliability. To this end, Savannah River Remediation (SRR) implements programs and procedures for performance monitoring of structures, systems, and components (SSCs) within all of its processing facilities. As these programs have matured, they have helped to identify vulnerabilities within facility systems and provided a framework for ranking system assets to ensure correct priority is established and maintained. This paper provides an overview of how SRR structures and implements system performance monitoring with specific focus on the recent replacement of the DWPF Safety Grade Purge Flowmeters. This case study will discuss: 1) How the performance monitoring process identified an availability risk due to the minimal vendor support for the in-use flow transmitters, 2) The design changes developed as part of the project to transition to a different measurement technology, which includes operational lessons learned from the original installation, 3) The execution of the replacement effort, which was complicated by the fact that the purge system is a Vital Safety System (VSS) and not easily taken out of service for extended maintenance, and 4) The need to consider future maintainability, particularly given that continued acceleration in technical innovation will likely increase the frequency of equipment obsolescence. (authors)

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

Developing a New Model to Estimate the Costs to Decommission DOE's Active Facilities - 16368

The Active Facilities Data Collection System (AFDCS) is a major means the U.S. Department of Energy (DOE or Department) uses to estimate its out-year environmental liability for the active and surplus facilities across the Department. The previous AFDCS cost model was based on cost assumptions that date back to the mid-1990s. The model is being replaced with a model that reflects more recent facility disposition experience and costs. DOE's Office of Environmental Management's (DOE-EM) historical cost collection database provides the basis for the new AFDCS cost model. This paper describes the process that was used to update the AFDCS facility data inputs, analyze this data, and formulate historical costs and parameters. This process created a robust and reliable cost model that will provide DOE with a valid estimate of the facilities' disposition costs. (authors)

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

Preliminary Long-term Safety Assessment for the Planned LLW Disposal Facility at Al-Tuwaitha, Iraq - 18142

Before 1991, Iraq had a significant nuclear program which encompassed ten nuclear sites spread out across the country. The largest nuclear site was located at the Al-Tuwaitha site near Baghdad where 18 nuclear facilities were concentrated. During the first Gulf War in 1991, all sites suffered significant damage from bombing and subsequent looting. Currently, the Ministry of Science and Technology is responsible for the ten heavily damaged nuclear sites and for the management of radioactive waste in Iraq in general. The task of decommissioning the damaged nuclear facilities and managing the resulting waste is being supported by the International Atomic Energy Agency (IAEA), the European Commission (EC), and the United States (US). In the framework of its Instrument for Nuclear Safety Cooperation (INSC), the EC has launched a cooperation project to support the planned construction of an engineered repository for short-lived low- and intermediate level waste for the safe disposal of radioactive waste resulting from the decommissioning activities, as well as for waste from the continuing use of radioisotopes in medicine and the oil and gas industry in the country. The main objectives of this project, which was awarded to a Consortium led by NUKEM Technologies Engineering Services GmbH and joined by DBE Technology GmbH of Germany and JAVYS a.s. of Slovakia, are the development of an initial conceptual level basic design and preliminary safety assessment report (PSAR) and the subsequent detailed engineering design and supporting pre-construction safety assessment report (PCSAR). The current paper focuses on the development of the PSAR. Special focus is given to the long-term calculations and difficulties associated with the limited data base in regard to site characterization as well as the expected inventory for the future repository. Because of the ongoing security concerns in Iraq, the Ministry of Science and Technology had already selected the Al-Tuwaitha Site as the best suited location for the planned surface repository at the start of the project. Although some site characterization work had already been carried out prior to the start of the project, certain data for modelling the geology and hydrology of the site are still lacking. Missing data have been compensated by conservative assumptions and the assessment of alternative scenarios. Consequently, for the development of the simplified GoldSim{sup R} computer model used to carry out the long-term calculations supporting the PSAR, existing site information was used and, as appropriate, amended or replaced by conservative assumptions. Due to the nature of the waste, data on the expected inventory are associated with large uncertainties. Although certain estimates related to the future volume of the waste exist, little to no information is available on the expected radionuclide vector and respective activities. Therefore, the long-term calculations were carried out for a generic inventory including a broad range of radionuclides that are to be expected as being important considering the kind of waste to be disposed of in the facility. Based on the results of these calculations, the maximum total capacity of the facility was estimated and converted into activity limits for individual radionuclides that will be integrated into the existing general WAC for the facility. Compliance with these WAC will then guarantee the safety of the repository in regard to potential radionuclide releases via the groundwater pathway. In parallel to the determination of activity limits for the WAC, long-term calculations were carried using different estimates for the expected total activity to determine the general suitability of the proposed technical solution of the repository for receiving the total amount of Iraqi radioactive waste principally acceptable for surface disposal. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗