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At least 289 records · Page 16
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Transitioning from Construction to Startup for the Waste Treatment and Immobilization Plant's Low-Activity Waste Facility - 17284

The original design of the Waste Treatment and Immobilization Plant (WTP) required all waste be processed through the Pretreatment (PT) Facility in order to separate the low-activity waste and high-level waste streams. The U.S. Department of Energy (DOE), Office of River Protection (ORP) has developed a sequenced approach to completing the WTP Project, beginning with delivering some liquid waste directly to the Low-Activity Waste (LAW) Facility, bypassing the PT Facility. This operating configuration is referred to as direct-feed LAW (DFLAW). The primary driver for the DFLAW approach is to enable startup and commissioning of the LAW Facility for low-activity waste processing prior to the completion and availability of the PT and High-Level Waste facilities. In March 2016, a federal judge amended the Consent Decree between DOE and the State of Washington to require completion of hot commissioning for the LAW Facility no later than December 31, 2023. During the next few years ORP and the WTP contractor will transition the project from completion of construction to the startup and commissioning phase for the LAW Facility and the associated support facilities. The WTP contractor has developed an integrated management process to successfully conduct facility startup testing for the WTP facilities required to support DFLAW. The WTP contractor will also develop a commissioning plan and facility transition plan. These plans will describe the strategy; schedule; and requirements for safe, efficient, and sequential startup, commissioning, and transfer of the WTP facilities to an operations contractor. The WTP contractor has also established a strategy to achieve operational readiness for the facilities required to support DFLAW consistent with DOE O 425.1D, Verification of Readiness to Start Up or Restart Nuclear Facilities and DOE O 413.3B, Program and Project Management for the Acquisition of Capital Assets. ORP has established a strategy for oversight of the WTP contractor's startup, commissioning, and readiness activities. The strategy also encompasses ORP's own activities to achieve readiness. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS2003

Designing Facilities for Collaborative Operations

A methodology for designing operational facilities for collaboration by multiple experts has begun to take shape as an outgrowth of a project to design such facilities for scientific operations of the planned 2003 Mars Exploration Rover (MER) mission. The methodology could also be applicable to the design of military "situation rooms" and other facilities for terrestrial missions. It was recognized in this project that modern mission operations depend heavily upon the collaborative use of computers. It was further recognized that tests have shown that layout of a facility exerts a dramatic effect on the efficiency and endurance of the operations staff. The facility designs (for example, see figure) and the methodology developed during the project reflect this recognition. One element of the methodology is a metric, called effective capacity, that was created for use in evaluating proposed MER operational facilities and may also be useful for evaluating other collaboration spaces, including meeting rooms and military situation rooms. The effective capacity of a facility is defined as the number of people in the facility who can be meaningfully engaged in its operations. A person is considered to be meaningfully engaged if the person can (1) see, hear, and communicate with everyone else present; (2) see the material under discussion (typically data on a piece of paper, computer monitor, or projection screen); and (3) provide input to the product under development by the group. The effective capacity of a facility is less than the number of people that can physically fit in the facility. For example, a typical office that contains a desktop computer has an effective capacity of .4, while a small conference room that contains a projection screen has an effective capacity of around 10. Little or no benefit would be derived from allowing the number of persons in an operational facility to exceed its effective capacity: At best, the operations staff would be underutilized; at worst, operational performance would deteriorate. Elements of this methodology were applied to the design of three operations facilities for a series of rover field tests. These tests were observed by human-factors researchers and their conclusions are being used to refine and extend the methodology to be used in the final design of the MER operations facility. Further work is underway to evaluate the use of personal digital assistant (PDA) units as portable input interfaces and communication devices in future mission operations facilities. A PDA equipped for wireless communication and Ethernet, Bluetooth, or another networking technology would cost less than a complete computer system, and would enable a collaborator to communicate electronically with computers and with other collaborators while moving freely within the virtual environment created by a shared immersive graphical display.

Norris, Jeffrey↗
DOE OSTI2023

Assessment of Baseline Monitoring Data for the Remote-Handled Low-Level Waste Disposal Facility at Idaho National Laboratory

The purpose of this document is to summarize environmental monitoring data collected at the Remote-Handled Low-Level Waste (RHLLW) Disposal Facility during the first four years of facility operations (FY 2019 through FY 2022). This summary provides a “baseline” condition for the facility, which can be used to distinguish contaminant releases from the RHLLW Disposal Facility from pre existing contamination as well as potential future releases from other sources (i.e., upgradient aquifer sources), as specified in the facility monitoring plan. Sufficient data has been collected over the first four years of operation of the RHLLW Disposal Facility to establish baseline conditions for future compliance monitoring of aquifer wells and performance monitoring of vadose zone lysimeters. Except for seven elevated tritium measurements from lysimeter HFEF-South believed to have been impacted by waste disposals, all data was deemed appropriate for establishing baseline concentrations. Monitoring of the aquifer detected indicator analytes gross alpha and gross beta and target analytes tritium and C-14. There were a few C-14 detections prior to an increase in the required detection level (RDL) after the first round of sampling. I-129 and Tc-99 were not detected above RDLs in aquifer samples. Lysimeter samples detected indicator analytes gross alpha and gross beta and target analyte tritium. Target analytes C-14, I-129, and Tc-99 were not detected above RDLs in lysimeter samples. Except for tritium, baseline concentrations of indicator and target analytes measured during the first four years of RHLLW facility operations are expected to represent conditions through the expected 20 year operating period of the facility. Elevated tritium in the aquifer, a result of past discharges of tritium at upgradient facilities, showed a decline over the first four years of operations, and levels are expected to continue to decline with time as a result of dilution and decay. Current tritium levels in lysimeter samples may be elevated due to tritium in the water applied during construction and infiltration testing of the facility. As a result, tritium concentrations in lysimeters may decline with time as a result of dilution and decay. Statistical measures (i.e., mean, standard deviation, and 99% upper confidence level [UCL]) were calculated for all detected analytes using the four years of concentration data. These measures will be used to evaluate future monitoring results and to demonstrate the facility is performing, or not performing, as established in the facility performance assessment (PA). Future monitoring results will also be combined with the baseline data in this report to further establish temporal trends and natural variability in the measurements. Based on the data collected during the baseline period, it is recommended the gross alpha action level for performance monitoring increase from 10 to 20 pCi/L due to several baseline measurements exceeding the preliminary action level of 10. An action level of 20 pCi/L is slightly greater than the 99% UCL, is protective of the aquifer, and would reduce unnecessary sampling. It is also recommended that tritium be added to the lysimeter analyte list for post-baseline period monitoring. Tritium, while not a dose concern, is a good tracer that can provide valuable information on water flow in and around the RHLLW Disposal Facility. Recommendations for a tritium action level are provided if considered necessary.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Accelerating Nuclear Facility Mission Completion Through an Improved Chemical Hazard Control Strategy - 20460

Code of Federal Regulation (CFR) 10 CFR 830, Nuclear Safety Management [1], explicitly identifies hazardous materials in addition to radioactive materials when establishing the scope of materials that must be evaluated in Hazard Category (HC) 1, 2, and 3 nuclear facilities. This inclusion has led to a near decade-long progression of more and more rigorous application of chemical hazard controls in these facilities. The progressive addition of more rigorous chemical hazard controls has occurred across the Department of Energy (DOE) Complex, having been driven by contractors and regulators alike. However, chemical hazards in other DOE facilities have been, and continue to be, controlled using traditional industrial safety centered strategies. Thus, identical chemical hazards are controlled using dramatically different approaches based solely on the facility type. Not only are the controls different, the analysis and derivation rigor and requirements are also significantly different. In particular, chemical control derivation and application in non-nuclear facilities are much less onerous than the comprehensive hazard analysis process used in nuclear facilities, even in circumstances in which chemicals are not initiators to, or do not exacerbate a nuclear event. While the derivation and control approaches used in these two types of facilities are disparate, the safety results are similarly effective based on a review of reported chemical events from 2014 through 2018. While both approaches achieve comparable safety outcomes, the approach used in nuclear facilities results in significantly greater preparation costs and implementation costs. Furthermore, the problem with the differing approach is not relegated to the control preparation and implementation burden. The operational impact of potentially unnecessary controls slows mission completion without a corresponding increase in safety, not to mention the dilution of the importance of Technical Safety Requirements. The proposed solution must first include a critical and thorough examination of DOE Complex and top-tier chemical industry chemical hazard control practices, coupled with a similar examination of the driving facility regulatory requirements. Results from the examination can then be used to fuel a bold departure from current chemical control practices in Hazard Category 1, 2, and 3 nuclear facilities. Development of an improved and consistent chemical hazard control process would follow, one that continues to ensure safety while accelerating mission completion and significantly reducing the burden associated with chemical hazard control development and implementation. The improved process must establish a hierarchy where radiological hazards are addressed using safety class and safety significant controls and chemical hazards are addressed using overarching safety management programs (with limited unique exceptions). The improved Hazard Category 1,2, and 3 nuclear facility chemical hazard control strategy, while compliant with 10 CFR 830 [1], would be founded on the requirements in 10 CFR 851, Worker Safety and Health Program [2], and the complementary DOE O 151.1D, Comprehensive Emergency Management System [3]. In this way, the Hazard Category 1, 2, and 3 nuclear facility chemical hazard control strategy would, wherever appropriate, be the same process used at other DOE Complex facilities. Ultimately, the improved process will maintain safety and significantly reduce costs and life cycle risk through faster mission completion. (authors)

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

Facility Informatics with GeoVisipedia (Final Report)

Characterization and understanding of nuclear fuel cycle facilities is of considerable interest to DNN R&D and the nonproliferation community due to their importance in nuclear weapons manufacture. Satellite imagery is a key source of information about these facilities, particularly in denied areas. Such imagery can be useful in its own right but annotations applied to the imagery vastly expand the value of the imagery and its usefulness by providing object descriptions, context and potentially linkages to other relevant information. GeoVisipedia (Geospatial Visual Wikipedia) is a new and novel approach to collecting, organizing, sharing and preserving knowledge about nuclear facilities to support nonproliferation efforts. Our approach is unique in that it emphasizes geospatial visual information about nuclear facilities rather than focusing exclusively on textual information. GeoVisipedia users, such as all-source analysts and decision makers, have access to richly annotated nuclear facility imagery with hyperlinks to visual and textual information that provides users with details and context about the facility. A major accomplishment of this project was to develop technology that allows annotations from one satellite image of a facility to be propagated to other satellite imagery of the facility. New satellite imagery of the facility is automatically populated with all previous annotations of the facility. This technology breakthrough provides users with the annotations, and hence knowledge and insight, from all the Subject Matter Experts (SMEs) who have previously annotated imagery of the facility. This capability was demonstrated in a GeoVisipedia prototype, awarded a patent and is described in several publications. In the latter part of the project an informal collaboration was established with the DNN R&D Multi Informatics for Nuclear Operations Scenarios (MINOS) venture. A second prototype of GeoVisipedia was developed that directly supports the MINOS venture. This prototype is called the Nuclear Facility Information Explorer (NFIX) and is tailored to the High Flux Isotope Reactor (HFIR) facility, a testbed for the MINOS venture. NFIX downloads sensor metadata and operator logs, parses and displays this information. NFIX has been licensed and deployed to several laboratories participating in the MINOS venture.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2023

Survey of U.S. Research Reactor Auxiliary Facilities Used for Material Testing and Basic Neutron Science

A catalogue of and attributes of the Materials Irradiation and Testing facilities (MIF) and the Basic Neutron Science facilities (BSF) of auxiliary (AUX) facilities of research reactors are compiled. The survey of these facilities is drawn from the set of U.S. university and DOE research reactor facilities and should be considered as a reference point when comparing commonalities in international AUX facilities. The size and shielding capabilities of the facilities, the typical characterization equipment used, and the neutron flux and irradiation capabilities of the facilities are listed. Short descriptions of experimental activities and current practices are detailed for MIFs and BSFs. The attributes are important in consideration of the reconfiguration of the facilities for purposes other than stated mission, i.e., for proliferation of weapons-usable nuclear material (WUNM). An evaluation of the capacity for production rate of WUNM from neutron beams-on-targets or a sample that has been placed in an irradiation position for a period is provided. Further evaluation of reconfiguration of these MIF and BSF are recommended to refine the proliferation risks. Specifically, MIF hot cells and potential configurations, additional modeling of facility throughput and development of methods for determining levels of concern for MIF and BSF have all been identified as steps for refinement of determination of the risks associated with these facilities. Future work will focus on: 1. A model of projected throughput for different MIF and BSF configurations that will be used to provide a window of potential operational misuse for a facility and better understand the capability of production and the rate of processing for WUNM in these facilities. 2. A misuse study for various configurations of MIF hot cell layouts detailing capacity, specifications, shielding constraints for MIF operations, identification of a maximum shielding thickness, and reasonable need associated with each activity performed in a MIF. 3. Development of a rule of thumb rating system utilizing data gathered from previous reports to show levels of concern at a given power level, flux, and experiment set.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1993

Research and test facilities

A description is given of each of the following Langley research and test facilities: 0.3-Meter Transonic Cryogenic Tunnel, 7-by 10-Foot High Speed Tunnel, 8-Foot Transonic Pressure Tunnel, 13-Inch Magnetic Suspension & Balance System, 14-by 22-Foot Subsonic Tunnel, 16-Foot Transonic Tunnel, 16-by 24-Inch Water Tunnel, 20-Foot Vertical Spin Tunnel, 30-by 60-Foot Wind Tunnel, Advanced Civil Transport Simulator (ACTS), Advanced Technology Research Laboratory, Aerospace Controls Research Laboratory (ACRL), Aerothermal Loads Complex, Aircraft Landing Dynamics Facility (ALDF), Avionics Integration Research Laboratory, Basic Aerodynamics Research Tunnel (BART), Compact Range Test Facility, Differential Maneuvering Simulator (DMS), Enhanced/Synthetic Vision & Spatial Displays Laboratory, Experimental Test Range (ETR) Flight Research Facility, General Aviation Simulator (GAS), High Intensity Radiated Fields Facility, Human Engineering Methods Laboratory, Hypersonic Facilities Complex, Impact Dynamics Research Facility, Jet Noise Laboratory & Anechoic Jet Facility, Light Alloy Laboratory, Low Frequency Antenna Test Facility, Low Turbulence Pressure Tunnel, Mechanics of Metals Laboratory, National Transonic Facility (NTF), NDE Research Laboratory, Polymers & Composites Laboratory, Pyrotechnic Test Facility, Quiet Flow Facility, Robotics Facilities, Scientific Visualization System, Scramjet Test Complex, Space Materials Research Laboratory, Space Simulation & Environmental Test Complex, Structural Dynamics Research Laboratory, Structural Dynamics Test Beds, Structures & Materials Research Laboratory, Supersonic Low Disturbance Pilot Tunnel, Thermal Acoustic Fatigue Apparatus (TAFA), Transonic Dynamics Tunnel (TDT), Transport Systems Research Vehicle, Unitary Plan Wind Tunnel, and the Visual Motion Simulator (VMS).

Source record↗
NASA NTRSDate not supplied

Mars Sample Return (MSR) Sample Receiving Facility (SRF) Assessment Study (MSAS)

NASA, in partnership with the European Space Agency (ESA), is seeking to return Martian geological and atmospheric samples to Earth for scientific study in the early 2030s. Due to the possibility that the samples could contain extraterrestrial life, Mars Sample Return (MSR) is classified as a Category V: Restricted Earth Return mission by the NASA Planetary Protection Office. As a result of this classification, a MSR Sample Receiving Facility (SRF) must not only provide a pristine environment to ensure samples are protected from terrestrial contamination for scientific investigations, it must also provide high-containment (biosafety level 4 [BSL-4]-equivalence) to isolate the samples from Earth’s biosphere until the samples are deemed safe for release and/or sterilized. The nominal utilization period for a SRF is anticipated to be 2-5 years and is intended to enable curation activities, biohazard assessment, select early science activities, and the rapid release of samples to the scientific community. However, to account for possible delays in schedule or the identification of extant life, this anticipated period of time must be flexible to accommodate schedule extensions and contingency plans. Due to requirements for high-level biological containment and cleanliness, a traditional receiving/curation facility cannot be utilized for MSR. Therefore, beginning in 2022, NASA Johnson Space Center is performing a MSR SRF Assessment Study (MSAS) to investigate the most optimal facility modality for a MSR SRF, as well as start to define programmatic early estimate of costs and schedules before the initial design phase begins. NASA is partnering with industry contractors (architectural and engineering firms with BSL-4 and cleanroom technology experience, as well as other contracted infrastructure and construction specialists) along with selected experts from NASA, ESA, existing U.S. BSL-4 facilities, and other U.S. government agencies, to carry out the assessment study. The MSAS should also aid in the future refinement of the science requirements (e.g., contamination control, equipment accommodations) before site-specific design would commence. As part of the MSAS, NASA is planning to assess an array of possibilities for a MSR SRF. One of the main considerations is the facility modality and whether an existing BSL-4 facility can be utilized (for some or all functions); or, if new constructure would be required, would a traditional fixed facility or a modular facility the best choice. MSAS will also investigate the ability of the modalities to accommodate two different facility capability endmembers: 1) a minimal facility focusing on biohazard assessment and curation tasks with a small footprint, and 2) an enhanced facility with additional capabilities to enable expedited processing and the completion of time-sensitive and (some) sterilization-sensitive science. The assessment is intended to generate information that will inform the selection of facility modalities for high-level conceptual design development. While the assessment study will focus on SRF requirements for accommodating curation, science, and sample safety assessment infrastructure, it will also consider an array of other factors, such as ease of access for international users, decommissioning, repurposing, future sale or lease following MSR’s use of the facility, and uncontained preparatory laboratory spaces. Upon completion of the study, the preferred modality and refined requirements would be utilized for site-specific design but will not be finalized until NASA’s completion of the National Environmental Policy Act (NEPA) process.

A.D. Harrington↗
DOE OSTI2017

Optically-based Sensor System for Critical Nuclear Facilities Post-Event Seismic Structural Assessment

The U.S. Department of Energy (DOE) has ownership and operational responsibility for a large enterprise of nuclear facilities that provide essential functions to DOE missions ranging from national security to discovery science and energy research. These facilities support a number of DOE programs and offices including the National Nuclear Security Administration, Office of Science, and Office of Environmental Management. With many unique and “one of a kind” functions, these facilities represent a tremendous national investment, and assuring their safety and integrity is fundamental to the success of a breadth of DOE programs. Many DOE critical facilities are located in regions with significant natural phenomenon hazards including major earthquakes and DOE has been a leader in developing standards for the seismic analysis of nuclear facilities. Attaining and sustaining excellence in nuclear facility design and management must be a core competency of the DOE. An important part of nuclear facility management is the ability to monitor facilities and rapidly assess the response and integrity of the facilities after any major upset event. Experience in the western U.S. has shown that understanding facility integrity after a major earthquake is a significant challenge which, lacking key data, can require extensive effort and significant time. In the work described in the attached report, a transformational approach to earthquake monitoring of facilities is described and demonstrated. An entirely new type of optically-based sensor that can directly and accurately measure the earthquake-induced deformations of a critical facility has been developed and tested. This report summarizes large-scale shake table testing of the sensor concept on a representative steel frame building structure, and provides quantitative data on the accuracy of the sensor measurements.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2019

Natural and Engineered Features Supporting Environmental Performance of Idaho National Laboratory’s Remote-Handled Low-Level Waste Disposal Facility

The Idaho National Laboratory’s (INL’s) Remote-Handled Low-Level Waste (RH LLW) Disposal Facility has been designed and constructed to receive waste generated at the INL site in support of the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) and Naval Reactors (NR) missions. The facility has been designed to receive legacy RH LLW currently in storage at INL facilities and new waste generated by nuclear research programs. RH LLW to be received at the facility is solid waste consisting of surface contaminated debris; ion exchange resins used to purify cooling and canal water; activated metals generated during reactor core change outs and Naval spent fuel management; and nuclear Research & Development. Disposal of INL's RH LLW poses unique disposal challenges due to high radiation levels (contact rates up to 600 Sv/hr [60,000 Rem/hr]) that require remote handling and adequate shielding for transportation and disposal. The facility has been designed to meet the requirements of DOE Order 435.1 [1], including the all-pathways dose, air pathway dose, inadvertent intruder doses, and radon emissions limits. This paper presents an overview of natural and engineered design features that support the safety case and enhance the environmental performance of the facility. These include location selection, use of reinforced pre-cast concrete vaults; a robust hydraulic drainage system; use of steel waste canisters for waste disposal; the facility monitoring system; facility layout; and crediting waste forms. The facility performance assessment (PA) [2] developed to meet the requirements of DOE Order 435.1 “Radioactive Waste Management” credits each of these protective features. Releases from waste forms disposed of at this facility are limited by corrosion of the steel waste canisters which will ultimately allow contact with infiltrating water followed by transport of radionuclides from the vault system. The steel waste canisters were credited for their ability to limit water contact with the waste based on the demonstrated ability of the reinforced pre-caste concrete vaults to provide long-term structural protection of the steel waste canisters. Verification of hydraulic and concrete performance was conducted through laboratory data, field infiltration tests supported by a robust monitoring system, and numerical modeling. Releases into the vault system from activated metals are based on site-specific corrosion data. Radionuclide transport from the vault system from all waste forms is based on site-specific climatological, hydrologic, and stratigraphic data. At INL, the groundwater pathway historically and appropriately has received the most attention. By crediting aspects of the natural system, the engineered vault system, and the waste, estimated radiologic doses via the groundwater pathway are minimal. Air pathway doses were reduced by differentially crediting corrosion of aluminum and stainless steel activated metals resulting in the pathway being screened out of the PA. Groundwater and inadvertent intruder doses were reduced through selective facility layout. This paper provides an overview of the system-wide protective features incorporated into the facility design along with an extensive reference list to supporting technical papers and analyses that provide data and interpretation.

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

Natural and Engineered Features Supporting Environmental Performance of Idaho National Laboratory's Remote-Handled Low-Level Waste Disposal Facility - 19436

The Idaho National Laboratory's (INL's) Remote-Handled Low-Level Waste (RH LLW) Disposal Facility has been designed and constructed to receive waste generated at the INL site in support of the US DOE Office of Nuclear Energy (NE) and Naval Reactors (NR) missions. The facility has been designed to receive legacy RH LLW currently in storage at INL facilities and new waste generated by nuclear research programs. RH LLW to be received at the facility is solid waste consisting of surface contaminated materials; ion exchange resins used to purify cooling and canal water; and activated metal components generated during reactor core change outs and Naval spent fuel management; and nuclear R and D. Disposal of INL's RH LLW poses unique disposal challenges due to high radiation levels (contact rates up to 600 Sv/hr [60,000 Rem/hr]) that require remote handling and adequate shielding for transportation and disposal. The facility has been designed to meet the requirements of DOE Order 435.1, Radioactive Waste Management, including the all-pathways dose, air pathway dose, inadvertent intruder doses, and radon emissions limits. This paper presents an overview of natural and engineered design features that support the safety case and enhance the environmental performance of the facility. These features include location selection, use of reinforced pre-cast concrete vaults; a robust hydraulic drainage system; use of steel waste canisters for waste disposal; the facility monitoring system; facility layout; and crediting waste forms. The facility performance assessment (PA) developed to meet the requirements of DOE Order 435.1, credits each of these protective features. Releases from waste forms disposed of at this facility are limited by corrosion of the steel waste canisters which will ultimately allow contact with infiltrating water followed by transport of radionuclides from the vault system. The steel waste canisters were credited for their ability to limit water contact with the waste based on the demonstrated ability of the reinforced pre-caste concrete vaults to provide long-term structural protection of the steel waste canisters. Verification of hydraulic and concrete performance was conducted through laboratory data, field infiltration tests supported by a robust monitoring system, and numerical modeling. Releases into the vault system from activated metals are based on site-specific corrosion data. Radionuclide transport from the vault system from all waste forms is based on site-specific climatological, hydrologic, and stratigraphic data. At INL, the groundwater pathway historically and appropriately has received the most attention. By crediting aspects of the natural system, the engineered vault system, and the waste, estimated radiologic doses via the groundwater pathway are minimal. Air pathway doses were reduced by differentially crediting corrosion of aluminum and stainless steel activated metals resulting in the pathway being screened out of the PA. Groundwater and inadvertent intruder doses were reduced through selective facility layout. This paper provides an overview of the system-wide protective features incorporated into the facility design along with an extensive reference list to supporting technical papers and analyses that provide data and interpretation. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Contaminant Migration Modeling to Support the In Situ Decommissioning of Hardened Nuclear Facilities - 20215

The decommissioning of hardened nuclear facilities provides a unique challenge in balancing current worker dose with exposure to future receptors and media. In situ decommissioning (ISD) is a cost-effective and safe option for the closure of such facilities but requires a detailed understanding of the threat that radioactive and hazardous constituents remaining in the facilities pose to groundwater and surrounding surface water. The Savannah River Site (SRS) extensively employs contaminant migration models within the decommissioning process to develop site-specific removal, grouting, and monitoring strategies in support of safe and effective final end states of nuclear facilities. This work discusses the use of contaminant migration modeling in the ISD process at SRS, including specific examples from the closure of P-Reactor and R-Reactor, Building 235-F, F-Canyon and F/H Laboratory Complex. Migration models developed during the closure of P-Reactor and R-Reactor were divided into four source areas: the reactor vessel, the process area, the disassembly basin, and the purification wing. Each source area was assigned a specific inventory and migration pathway and was then subject to varying hypothetical removal, capping, and grouting schemes. Modeling indicated that groundwater Maximum Concentration Limits (MCLs) might be exceeded if no action was taken for eleven and ten constituents of concern (COCs) at P-Reactor and R-Reactor, respectively. These exceedances could occur in as few as 200 to 500 years. Alternatively, the migration modeling demonstrated that the selected ISD actions reduced contaminant mobility which allowed for significant radioactive decay and resulted in fewer predicted exceedances of groundwater MCLs (five COCs for P-Reactor; eight COCs for R-Reactor). In response to the modeling results for P-Reactor and R-Reactor, effectiveness monitoring programs were developed to target contaminants, specific to each reactor, that may migrate to groundwater. Contaminant migration modeling also revealed that roof collapse was a large factor in the release of COCs to the environment, giving rise to roof improvements, and an inspection and vegetation control program to ensure roof stability over time. Contaminant migration modeling is also aiding in the closure planning for hardened facilities in F Area. Building 235-F housed the Actinide Billet Line, which produced Np-237 billets for irradiation in SRS reactors, and the Plutonium Fuel Form (PuFF) facility that produced Pu-238 heat sources for the space program. As a result of these missions, areas within Building 235-F contain considerable residual amounts of both Pu-238 and Np-237. Contaminant migration modeling of Building 235-F was originally performed in 2012 to identify the feasibility of ISD and the amount of radioactive material removal required to prevent the exceedance of groundwater MCLs. The original model indicated that a 60% reduction in the PuFF facility Pu-238 inventory could keep groundwater concentrations below standards, while Np-237 did not pose a threat to groundwater. However, updates to the model with an emphasis on source impact pathways revealed that, due to the orientation of the source areas relative to groundwater flow, no amount of reasonable removal of Pu-238 would keep groundwater concentrations below MCLs and that Np-237 could be a large contributor to localized MCL exceedances. With this insight, the refined 2019 model is being used to assist in the development of grouting plans specific to each facility source area, where bentonite may be utilized to slow the migration of Pu-238 and its daughter products from the PuFF facility and a reducing grout may decrease Np-237 transport by ensuring the nuclide remains in the less mobile +IV oxidation state. The beginning phases of contaminant migration modeling are underway for F-Canyon and associated facilities using lessons learned from the Reactors and Building 235-F. A contaminant migration pathway, similar to the pathway used for the reactor vessels, is being developed for the hot and warm canyons. The F-Canyon inventories are being spatially refined to identify specific and localized source zones, comparable to Building 235-F, that may impact groundwater. Contaminant migration modeling provides vital input within the ISD process, from facility investigation to post closure effectiveness monitoring, making it a valuable tool in the development of safe and effective end states for hardened nuclear facilities at SRS. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS1998

Refurbishment and Automation of the Thermal/Vacuum Facilities at the Goddard Space Flight Center

The thermal/vacuum facilities located at the Goddard Space Flight Center (GSFC) have supported both manned and unmanned space flight since the 1960s. Of the 11 facilities, currently 10 of the systems are scheduled for refurbishment and/or replacement as part of a 5-year implementation. Expected return on investment includes the reduction in test schedules, improvements in the safety of facility operations, reduction in the complexity of a test and the reduction in personnel support required for a test. Additionally, GSFC will become a global resource renowned for expertise in thermal engineering, mechanical engineering and for the automation of thermal/vacuum facilities and thermal/vacuum tests. Automation of the thermal/vacuum facilities includes the utilization of Programmable Logic Controllers (PLCs) and the use of Supervisory Control and Data Acquisition (SCADA) systems. These components allow the computer control and automation of mechanical components such as valves and pumps. In some cases, the chamber and chamber shroud require complete replacement while others require only mechanical component retrofit or replacement. The project of refurbishment and automation began in 1996 and has resulted in the computer control of one Facility (Facility #225) and the integration of electronically controlled devices and PLCs within several other facilities. Facility 225 has been successfully controlled by PLC and SCADA for over one year. Insignificant anomalies have occurred and were resolved with minimal impact to testing and operations. The amount of work remaining to be performed will occur over the next four to five years. Fiscal year 1998 includes the complete refurbishment of one facility, computer control of the thermal systems in two facilities, implementation of SCADA and PLC systems to support multiple facilities and the implementation of a Database server to allow efficient test management and data analysis.

Donohue, John T.↗
NASA NTRS2005

D-Side: A Facility and Workforce Planning Group Multi-criteria Decision Support System for Johnson Space Center

"To understand and protect our home planet, to explore the universe and search for life, and to inspire the next generation of explorers" is NASA's mission. The Systems Management Office at Johnson Space Center (JSC) is searching for methods to effectively manage the Center's resources to meet NASA's mission. D-Side is a group multi-criteria decision support system (GMDSS) developed to support facility decisions at JSC. D-Side uses a series of sequential and structured processes to plot facilities in a three-dimensional (3-D) graph on the basis of each facility alignment with NASA's mission and goals, the extent to which other facilities are dependent on the facility, and the dollar value of capital investments that have been postponed at the facility relative to the facility replacement value. A similarity factor rank orders facilities based on their Euclidean distance from Ideal and Nadir points. These similarity factors are then used to allocate capital improvement resources across facilities. We also present a parallel model that can be used to support decisions concerning allocation of human resources investments across workforce units. Finally, we present results from a pilot study where 12 experienced facility managers from NASA used D-Side and the organization's current approach to rank order and allocate funds for capital improvement across 20 facilities. Users evaluated D-Side favorably in terms of ease of use, the quality of the decision-making process, decision quality, and overall value-added. Their evaluations of D-Side were significantly more favorable than their evaluations of the current approach. Keywords: NASA, Multi-Criteria Decision Making, Decision Support System, AHP, Euclidean Distance, 3-D Modeling, Facility Planning, Workforce Planning.

Tavana, Madjid↗
DOE OSTI2015

Andra's strategy and approach for operational safety of French underground facility Cigeo - 15127

According to the French Act on radioactive waste of 28 June 2006, Andra shall design a reversible repository. In response to this demand, Andra is developing the industrial project namely 'Cigeo', a reversible geological disposal facility for HLW and ILW located in Meuse/Haute-Marne. The previous safety cases of 2005, 2009 and 2012 give a sound basis to assess Cigeo's safety both for the operational and post-closure periods. Regarding operational safety, in this new stage of the project, the challenging issues for the preparation of the safety case are the following: - To adapt the various regulatory frameworks (nuclear and non-nuclear) to the underground facility - To design the underground facility and handling equipment's to reduce at first risks in the underground facility as such radioactive particles release, gas explosion, fire and co-activity and limit their consequence. - To establish waste acceptance criteria to be fulfilled by waste producers in line with operational safety and design options. The objective of the paper is to present the experience of Andra in operation safety of the underground facility of Cigeo. Operational safety for Cigeo is assessed as much as possible according to the existing French regulations and guidance for nuclear facilities. Indeed, compared to existing nuclear facilities, a geological repository is a specific case due to the existence of underground facilities with particularities such as: - ventilation management in a large scale underground facility, - fire risk management. As an input for the industrial design of Cigeo for operation, Andra defined a fire guidance and a specific reference safety guidance, based on the adaptation of the French nuclear regulatory texts for and when feasible to mines and tunnels regulatory texts and on its own experience gained through safety operational reports since 2005 and their reviews by the Safety Authority. - underground co-activity of nuclear operation with construction works. According to the common 'defense-in depth' principle for nuclear facilities, these guidelines for the industrial design of Cigeo for operation consider the successive levels of defense of: - Preventing the risk and its spreading by design and construction; - Detecting incidents and early fighting by design and operation dispositions (detection and control systems, survey procedure); - Controlling incidents and limiting of their consequences by safeguarding systems, specific procedures. At that stage, design options for Cigeo constitute an adequate response to the operational safety functions and requirements: - Two levels of containment are set up: the first one is ensured by the waste package, the second one is ensured from the surface facility to the disposal vaults by the transfer cask during the transfer of waste packages in the disposal vaults by the disposal vault itself and a nuclear ventilation with HEPA filter; - Favorable options for fire risk prevention and mitigation: transport system minimizing heat load (cable car for the transfer from the surface, electric transfer engines), twin access tunnels to ease fire-fighting and evacuation...; - Permanent complete physical separation of nuclear zone and construction zone (including ventilation systems). In practice, operation safety assessment of Cigeo, designed following the aforesaid reference safety guidelines, follows a common risk analysis (internal aggressions such as auxiliary losses, chemical and toxic risks, fall or collision, fire, human factors and external aggressions such as earthquake, air traffic, flooding) with the following steps: - Identification of the major risks and the design basis scenarios that have to be considered for dimensioning of the installations and the safety devices; - On the basis of these different situations, those who will have to be quantified are identified; one or more representative and envelop situations are chosen in order to evaluate robustness of the installation. Quantitative evaluation of these situations permits to: - identify the consequences on man, environment and installation, and in particular to gain confidence in the maintaining of the safety functions; - identify or define the technic and/or organisational dispositions classified as 'important elements for protection' (IEP); - Situations beyond dimensioning permit to verify the robustness of the installation. They permit, on one hand to identify possible complementary material or organisational dispositions; on the other hand, to define the 'hard core' i.e. structures, systems and components that permit to maintain the confinement function of radioactive substance. (authors)

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

The status of LILW disposal facility construction in Korea - 15158

In this paper, we discuss the experiences during the construction of the first LILW disposal facility in South Korea. In December 2005, the South Korean Government designated Gyeongju-city as a host city of Low- and Intermediate-Level Radioactive Waste (LILW) disposal site through local referendums held in regions whose local governments had applied to host disposal facility in accordance with the site selection procedures. The LILW disposal facility is being constructed in Bongil-ri, Yangbuk-myeon, Gyeongju. The official name of the disposal facility is called 'Wolsong Low and Intermediate Level Radioactive Waste Disposal Center (LILW Disposal Center)'. It can dispose of 800,000 drums of radioactive wastes in a site of 2,100,000 square meters. At the first stage, LILW repository of underground silo type with disposal capacity of 100,000 drums was completed by the end of June 2014. The Wolsong Low and Intermediate Level Radioactive Waste Disposal Center consists of surface facilities and underground facilities. The surface facilities include a receipt and inspection facility, an interim storage facility, a radioactive waste treatment building, and supporting facilities such as main control center, equipment and maintenance shop. The underground facilities consist of a construction tunnel for transport of construction equipment and materials, an operation tunnel for transport of radioactive waste, an access shaft for workers, and six silos for final disposal of radioactive waste. (authors)

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

In-Situ Gamma-Ray Assay of the 235-F Plutonium Fuel Form Facility at the Savannah River Site - 16293

The Plutonium Fuel Form (PuFF) Facility is located in Building 235-F near the geographic center of the Savannah River Site. Between 1978 and 1984, the facility was used to produce approximately 165 kilograms of iridium-encapsulated Pu-238 spheres and pellets for use as radioisotope thermal generators, primarily for the space program. The facility is divided between two cell lines, the east cell line used to process powdered Pu-238 oxide raw material into fuel forms and the west cell line used to encapsulate the fuel forms in iridium. Scoping in-situ gamma-ray assays were performed in the PuFF Facility in 2006 and estimated that hundreds of grams of Pu-238 holdup were within the facility. Using this holdup estimate as a source term, SRS has performed a risk analysis that indicated a seismic event that induces a full-facility fire in 235-F could lead to a 28,800 rem dose to a co-located worker. Based on this risk assessment, SRS is taking steps to decontaminate the facility. One of the first steps taken has been to improve upon the quality of the in-situ gamma-ray assay data. Carts and collimators were specially designed to survey the equipment in the PuFF Facility. The cell interiors were assayed along with the furnaces and storage coolers that protrude beneath the cells. While the previous scoping work consisted of 32 measurements, the current series of measurements included nearly 400 measurements using both high-purity germanium and lanthanum bromide detectors. Data analysis for the current set of measurements was conducted with greater rigor as well. MCNP5 was used to evaluate a variety of possible physical distributions for the Pu-238 source term and estimate cross-talk between neighboring cells. Data analysis was performed using three gamma-rays emitted by Pu-238 (99.85 keV, 152.7 keV, and 766.4 keV) providing three independent estimates of the mass of Pu-238 holdup in each of the cells. This also allowed the application of matrix correction factors to adjust gamma-ray attenuation. The weighted mean of these three results was used as the best estimate of Pu-238 holdup in the PuFF facility. This work lowered the estimate of holdup in the PuFF facility by approximately 100 grams compared to the previous estimate, a 0.55's difference. The uncertainty in the Pu-238 holdup was also reduced substantially relative to the 2006 scoping assay. Further radiological measurements are planned for the near future including in-situ gamma-ray assays of wing cabinets in the east cell line maintenance area, the transfer tube between the two cell lines, and some of the furnaces and storage coolers. A germanium gamma imager, the GeGI, will also be used to map out the distribution of the Pu-238 source term in the PuFF Facility. (authors)

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

Implementation of Increased Salt and Decreased Sludge Volumes in the Defense Waste Processing Facility - 19597

The Savannah River Site (SRS) Liquid Waste System (LWS) safely stores and treats high-level radioactive waste. The LWS consists of 51 waste storage tanks (eight of which are operationally closed and filled with grout), waste evaporators, treatment facilities, and solidification facilities such as the Defense Waste Processing Facility (DWPF) and Saltstone Production Facility (SPF). The disposition of high-level waste (HLW) into a glass waste form poured into canisters at the DWPF began in 1996. DWPF processed a sludge only flowsheet until 2008 when the Actinide Removal Process (ARP) and Modular Caustic Side Solvent Extraction Unit (MCU) began treating the salt waste from the underground liquid storage tanks by striking with monosodium titanate (MST) to remove actinides, when necessary to meet downstream acceptance criteria, and removing the MST and entrained sludge solids (SS) by filtration before using an organic solvent to strip the cesium from the salt, creating two new concentrated HLW streams. This change caused the DWPF to begin processing under both a sludge-only and a coupled operations flowsheet where salt streams would only be added when available. This strategy allows for continuous glass canister production at DWPF. Recently, DWPF completed a 10-month melter replacement in December 2017 and the Concentrate, Storage, and Transfer Facility (CSTF) has completed repairs to the 3H Evaporator system. These facility outages have caused delays in preparing the next batch of sludge feed for DWPF. Concurrent with these facility outages, the Salt Waste Processing Facility (SWPF) has been undergoing system testing and commissioning, working toward startup. SWPF is designed to process salt waste in the same manner as ARP/MCU, but with much higher throughput volumes produced per unit time. In order to process the SWPF salt streams efficiently, DWPF will implement a coupled operation only flowsheet. Based on the facility outages mentioned above, efforts have been made to identify strategies to minimize sludge volumes and maximize salt volumes in the DWPF until the next sludge batch is ready for processing in DWPF. Savannah River Remediation (SRR) recently performed an evaluation that examines the effects of reducing the volume of sludge and maximizing the addition of the cesium-rich concentrated salt solution in each Sludge Receipt and Adjustment Tank (SRAT) batch in order to extend the existing sludge batch as long as possible until the next sludge batch is ready. These effects include a reduction of weight percent solids in the DWPF processing vessels, an increase in cycle time, a new canister heat generation value that remains within regulatory limits, and higher gamma source that remains within bulk facility gamma shielding. Balancing the above parameters resulted in a small reduction of sludge volume per SRAT batch under SWPF operation and maximizing the salt volumes for both ARP/MCU and SWPF operation. This slight reduction in sludge volume and increase in salt volumes allows SRR to mitigate some of the schedule impact incurred by the facility outages. This strategy was evaluated for both ARP/MCU and SWPF operation. This strategy has been successfully implemented under ARP/MCU operation using small step changes in the current process. This paper provides an overview of the evaluation performed for DWPF with ARP/MCU and DWPF with SWPF operations. (authors)

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