Search NASA⌕ Search

SEARCH · Search NASA

Results for “Nuclear Plant Management”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 307 records · Page 17
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Emulsified Oil Bio-Barrier to Remediate CEE in the Distal Portion of a Groundwater Plume - 20210

A non-time critical (NTC) removal action (RA) was designed at the Savannah River Site (SRS) to inject an emulsified oil mixture with a bioaugmentation culture to prevent trichloroethylene (TCE) discharging to surface water above maximum contaminant levels (MCLs). The emulsified oil mixture is expected to sequester the TCE and then be broken down into harmless compounds through microbial biodegradation. At 15 locations 8,290 L (2,190 gal) of oil mixture, buffer and chase water were injected into the subsurface creating two bio-barriers. At each location half of the oil mixture was injected, then 2 L (0.528 gal) of an enriched bioaugmentation culture was injected, and then the second half of the oil mixture was injected into the subsurface. Finally, at each location 151 L (40 gal) of buffer mixed with 575 L (200 gal) of dilution water was then injected followed by 1,135 L (300 gal) of chase water. SRS estimates this created two emulsified oil barriers totaling 73 m (240 ft) long, approximately 3 m (10 ft) high and 4.9 m (16 ft) wide in the subsurface. Similar emulsified oil barriers have been effective for 3 to 5 years, and SRS anticipates a similar longevity for these barriers. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS2004

Research in Support of the Use of Rankine Cycle Energy Conversion Systems for Space Power and Propulsion

This is the report of a Scientific Working Group (SWG) formed by NASA to determine the feasibility of using a liquid metal cooled nuclear reactor and Rankine energy conversion cycle for dual purpose power and propulsion in space. This is a high level technical report which is intended for use by NASA management in program planning. The SWG was composed of a team of specialists in nuclear energy and multiphase flow and heat transfer technology from academia, national laboratories, NASA and industry. The SWG has identified the key technology issues that need to be addressed and have recommended an integrated short term (approx. 2 years) and a long term (approx. 10 year) research and development (R&D) program to qualify a Rankine cycle power plant for use in space. This research is ultimately intended to give NASA and its contractors the ability to reliably predict both steady and transient multiphase flow and heat transfer phenomena at reduced gravity, so they can analyze and optimize designs and scale-up experimental data on Rankine cycle components and systems. In addition, some of these results should also be useful for the analysis and design of various multiphase life support and thermal management systems being considered by NASA.

Lahey, Richard T., Jr.↗
DOE OSTI2021

Nuclear Power Fault Diagnostics and Preventative Maintenance Optimization

Operation and maintenance costs for nuclear power plants are very large. Reactors are starting to shut down even after their operating licenses have been extended, because they are not price competitive compared to other energy sources. The nuclear industry is witnessing early closure of nuclear power plants due to economic reasons despite excellent safety records. Therefore, it is imperative to reduce costs to prevent these early closures. This paper showcases recent research into advanced fault diagnostics techniques and preventative maintenance optimization to reduce these maintenance costs. This report focuses on the condensate and feedwater system for both pressurized and boiling water reactor systems. The computerized maintenance management system, which contains the plant’s digital record of all the corrective- and preventative-maintenance work orders, was used as a ground truth to locate potential faults and label the process data as healthy or faulty. Various feature extraction techniques were utilized to further differentiate the faults from the healthy data. Support vectors machines were used to categorize other test sets of process data as healthy or faulty through a cross validation procedure. Similar faults were not found within this system leading to preventative maintenance optimization. Unnecessary amounts of preventative maintenance lead to inflated maintenance costs. This paper summarizes the steps for preventative maintenance optimization from component health determination to recommendation for action. This optimization was completed for condensate pumps, condensate booster pumps, and the respective motors that drive them.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

A reinforcement learning approach to long-horizon operations, health, and maintenance supervisory control of advanced energy systems

In this work, we develop a Reinforcement Learning (RL) approach to the supervisory control problem for advanced energy systems, such as novel nuclear reactors and other demand-driven, mission-critical, and component-health-sensitive energy plants. The inclusive problem landscape considered captures the stochastic confluence of plant performance, component health evolution, power demand from the grid, diverse maintenance actions, and operator-defined goals and constraints, all considered over meaningfully long-enough reasoning horizons. Key aspects of the proposed approach are a receding horizon control-inspired technique dictating time- or event-triggered supervisory policy (re-)constructions, as well as additional capability-enabling contributions such as timescale compression, to handle long reasoning horizons and uncertainty in parts of the problem, and practical yet demonstrably-effective handling of hybrid action spaces with continuous and discrete decision variables. The resulting algorithm consists of a simulation-based RL agent constructing stochastic supervisory control policies over nontrivial action spaces and for long horizons, applying the learned policy to the system for a much shorter interval, and perpetually repeating, to construct the next long-horizon policy. That next policy will only be applied, again, for a short interval, yet originally far-in-time events move progressively closer, their associated uncertainty decreases, and new events and aspects enter the reasoning horizon. The proposed methodology bridges fundamental receding horizon concepts with the unequivocally stronger and more scalable reasoning of contemporary RL. Numerical examples using Soft Actor–Critic Deep RL illustrate the operation and efficacy of the proposed technique for a power plant tasked with health-aware load following missions in a dynamic electricity market landscape.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2020

Investigation of Process Emissions during Thermal Cutting of Clad Metals from the Nuclear Sector - 20183

Typically, reactor pressure vessels as well as steam generators comprising low alloyed steel with a cladding of stainless steel. Recently the permissible values for Cr (VI) in the air of the working areas have been reduced in Germany from 100 μg/m{sup 3} to 1 μg/m{sup 3} air. This results in the danger that even with low-alloy steel with claddings these limit values would be exceeded when using flame cutting technologies. In order to ensure compliance with the permissible limit values, exhaust gas systems with gas cleaning measures are provided. The process emission test shall replace the assumptions required for the design of these systems with measured values. NUKEM Technologies Engineering Services GmbH (NUKEM) has chosen autogenous flame cutting as an example of a thermal cutting system for process emission testing, as there are no proven values in the literature for the Cr (VI) and NiOx emissions produced. The autogenous flame cutting can be used for a variety of cutting tasks within decommissioning projects. In order to simulate the material to be cut, mock ups were generated and cut from information about the material specification for the reactor pressure vessel from existing nuclear power plants including possible claddings. The cutting length needed for analysis in the test will be maximum 100 mm. The airborne emissions from the thermal cutting process will be captured and analyzed as follows: - Dust emission: The generated particles will be collected by a hood equipped with strong suction device and suitable filter elements. The resulting filter is gravimetrically measured before and after the tests. - Particle Size Distribution: The airborne particle size distribution measurement will result in figures for the size and number of particles arising from the thermal cutting process. - Gas Analysis: The gas analysis enables the determination of e.g. CO{sub 2}, CO, O{sub 3}, NO, NO{sub 2}. - Cr (VI) and NiOx Analysis: For the Cr (VI) and NiOx analysis, samples of the filter residue will be taken. As a result Cr (VI) in relation to total Cr as well as the Ni content will be determined by ICP-optically emission spectroscopy and put in relation to the total dust amount. The process emissions test itself as well as the necessary gas analyses and spectroscopic analyses were carried out at the Institute of Materials Science (IW) of Leibniz Universitaet Hannover (LUH). The results of the process emission tests are used to confirm the design of the gas cleaning systems. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2026

Supplement Analysis for the Proposed Shipment of High Burnup Research Cask to Idaho National Laboratory for Research and Development Purposes

DOE needs to continue with the HBUSFDP and transfer the HBURC to a facility capable of examination and testing of HBU SNF. These types of facilities do not exist at NAPS but are available at DOE national laboratories. The removal and eventual opening of the TN-32B research cask will provide DOE with the opportunity to gather data that will support continued safe storage of SNF at existing nuclear power plants, or a future federal consolidated storage facility. This effort will also help build public trust and confidence by demonstrating DOE’s ability to safely and securely transport commercial SNF using the certified Atlas rail consist.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Product Consistency Test Results for the LAW Phase 5 Glasses

This report provides the results of the Vapor Hydration Test solutions and the Product Consistency Test leachates from the Low-Activity Waste Phase 5 glasses, a series of simulated nuclear waste glasses fabricated at the Pacific Northwest National Laboratory. The series included quenched and canister-centerline cooled versions of the glasses. These data will be used in the development of enhanced property/composition models for waste glass vitrification at Hanford. Several of the blanks had detectable amounts of sodium and/or silicon. The measured concentrations (mg/L) of boron, sodium, and silicon for the low-activity test reference material glass included with the Product Consistency Tests were much lower than the expected low-activity test reference material test results. A review of the Product Consistency Test data indicated that there was generally little difference between the normalized values based on targeted or measured glass composition. Heat treatment had various impacts on the normalized concentration values. Most of the glasses had NC B , NC Na , and NC Si values that were greater than the Hanford Tank Waste Treatment and Immobilization Plant low-activity waste constraint of 4 g/L. The release rates for boron, sodium, and silicon were highly correlated for the study glasses.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

A Novel Approach to Simulate Realistic Concrete Microstructures under Irradiation

The concrete biological shield of light water reactors is exposed to high neutron and gamma irradiation doses in the long term. Irradiation deteriorates the physical and mechanical properties of concrete. Such effects need to be investigated to predict the concrete’s performance in the event of a lifetime extension of a nuclear power plant. This work combines high-resolution characterization techniques with fast-Fourier transform (FFT)-based 2-D simulations to evaluate the radiation-induced volumetric expansion (RIVE) and damage in concrete microstructures under neutron irradiation. Two concrete microstructures from samples provided by the Japan Concrete Aging Management Program (JCAMP) were characterized using micro x-ray fluorescence (mXRF) to obtain elemental intensity maps, and energy-dispersive x-ray spectroscopy to complement mXRF with local elemental information for Na. Minerals and cement paste are then identified based on the elemental composition to produce high-resolution phase maps, resulting in a more accurate representation of the microstructures compared to previous work. Simulations of radiation-induced volumetric expansion (RIVE), creep, and damage in JCAMP concrete use the fast Fourier transform (FFT)-based code Microstructure Oriented Scientific Analysis of Irradiated Concrete (MOSAIC) combined with the irradiated minerals, aggregates, and concrete (IMAC) database, which contains mineral-specific RIVE models. Overall, the simulation results are in fair agreement with experimental data.

Cheniour, Amani↗
DOE OSTI2021

Sandia Academic Alliance Program Collaboration Report: 2020-2021 Accomplishments

University partnerships play an essential role in sustaining Sandia’s vitality as a national laboratory. The SAA is an element of Sandia’s broader University Partnerships program, which facilitates recruiting and research collaborations with dozens of universities annually. The SAA program has two three-year goals. SAA aims to realize a step increase in hiring results, by growing the total annual inexperienced hires from each out-of-state SAA university. SAA also strives to establish and sustain strategic research partnerships by establishing several federally sponsored collaborations and multi-institutional consortiums in science & technology (S&T) priorities such as autonomy, advanced computing, hypersonics, quantum information science, and data science. The SAA program facilitates access to talent, ideas, and Research & Development facilities through strong university partnerships. Earlier this year, the SAA program and campus executives hosted John Myers, Sandia’s former Senior Director of Human Resources (HR) and Communications, and senior-level staff at Georgia Tech, U of Illinois, Purdue, UNM, and UT Austin. These campus visits provided an opportunity to share the history of the partnerships from the university leadership, tours of research facilities, and discussions of ongoing technical work and potential recruiting opportunities. These visits also provided valuable feedback to HR management that will help Sandia realize a step increase in hiring from SAA schools. The 2020-2021 Collaboration Report is a compilation of accomplishments in 2020 and 2021 from SAA and Sandia’s valued SAA university partners.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Dose Rate Effects on Degradation of Nuclear Power Plant Electrical Cable Insulation at a Common Dose

The intent of this report is to address an identified knowledge gap in relating accelerated aging of nuclear electrical cables to service aging: dose rate effects (DRE). Here, DRE refer to gamma radiation-induced polymer degradation being a function of dose rate in addition to total absorbed dose. The concern raised is that historical qualification conducted at higher dose rates to simulate service lifetime may underestimate insulation degradation that occurs at lower dose rates in service. In the work described herein, common nuclear cable insulation materials—cross-linked polyethylene (XLPE) and ethylene propylene diene elastomer (EPDM)—were subjected to accelerated aging at ambient temperature (26°C) at different gamma dose rates of 100, 200 and 1800 Gy/h for select exposure durations to achieve constant total doses of 170, 210 and 300 kGy to evaluate DRE. First, the cable insulation material types investigated are described. Then, the accelerated aging experimental process involving gamma irradiation applied to the insulation specimens at room temperature and different dose rates is discussed. Then, the experimental characterization techniques used to perform this work are elucidated. These include elongation at break (EAB), mass change, yellowness index (YI), carbonyl index (CI), density, indenter modulus (IM), and relaxation constant (t). Theory of polymer degradation is discussed, and characterization results and discussion are provided. Finally, concluding remarks are made. The findings from this work and cited prior work reveal that DRE are material dependent, even between similar material categories (e.g., XLPE). In the case of the EPDM studied, degradation of ductility was observed to be greater at higher dose rate for the same total dose, indicating accelerated gamma aging to be more conservative than extended aging. Thus, conclusions regarding the conservatism of historical qualification likely require additional consideration for specific materials and conditions in question. The results of this study support the contention that, due to inherent limitations and uncertainties associated with prediction of cable remaining useful life from accelerated aging experiments, trending of installed cable insulation health status will be more practical and useful for safe and efficient cable aging management repair and replace decisions than lifetime prediction from historical qualification. The combination of material robustness demonstrated by the qualification process and ongoing monitoring of cable health status combine to provide confidence in continued safe use of existing nuclear cables. Additional research into effective and efficient condition monitoring methods for non-destructive evaluation of installed cables is needed to support aging cable management, including material studies to inform interpretation of measured results.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS2001

Power Management for Space Advanced Life Support

Space power systems include the power source, storage, and management subsystems. In current crewed spacecraft, solar cells are the power source, batteries provide storage, and the crew performs any required load scheduling. For future crewed planetary surface systems using Advanced Life Support, we assume that plants will be grown to produce much of the crew's food and that nuclear power will be employed. Battery storage is much more costly than nuclear power capacity and so is not likely to be used. We investigate the scheduling of power demands by the crew or automatic control, to reduce the peak power load and the required generating capacity. The peak to average power ratio is a good measure of power use efficiency. We can easily schedule power demands to reduce the peak power from its maximum, but simple scheduling approaches may not find the lowest possible peak to average power ratio. An initial power scheduling example was simple enough for a human to solve, but a more complex example with many intermittent load demands required automatic scheduling. Excess power is a free resource and can be used even for minor benefits.

Jones, Harry↗
DOE OSTI2020

Quantifying the impact of climate change on erosion - 20397

The New York State Energy Research and Development Authority (NYSERDA) is the owner of the Western New York Nuclear Service Center (WNYNSC), a 1,351-ha (3,338-ac) site located approximately 48 km (30 mi) south of Buffalo, New York. In 1962, Nuclear Fuel Services, Inc. (NFS) entered into Agreements with the Atomic Energy Commission and New York State to construct the first commercial reprocessing plant of nuclear fuel in the United States. NFS, a private company, built and operated the spent fuel reprocessing plant and waste disposal facilities, processing 640 Mg (metric tons, or 705 short tons) of spent nuclear fuel from 1966 to 1972 under an Atomic Energy Commission license. Nuclear fuel reprocessing operations ended in 1972 and never reopened, leaving behind radioactive and chemical wastes. Erosion can play an important role in the fate and transport of waste at sites where disposal of long-lived waste is anticipated. The statistical characterization of key processes related to erosion is essential to the understanding of site stability through time. One of the key processes governing erosion is extreme precipitation and it is critical that trends in the distribution of extreme precipitation events through time be represented. The evidence of climate change is increasingly well documented and projected impacts on extreme precipitation events should be incorporated in performance assessment studies when relevant. Not evaluating future climate states in a performance assessment is contradictory to good modeling practice. Specifically, excluding climate change limits development of modeling information that could aid in effective decision making. The current climate literature provides both observational evidence and climate model projections of climate trends and/or climate change in the late 20. and early 21. centuries for North America and the northeast United States. In this work, this information was used to assess the impacts of potential changes in climate on erosion processes. The goal was to understand how projections of future climate relate to the performance of the WNYNSC through time. In the first stage of this work multi-temporal historical aerial images were analyzed in conjunction with orthophotography and Lidar data to develop probability distributions for variables representing important erosion processes. The information from these analyses was then used in conjunction with simulated data from the West Valley Erosion Working Group (EWG). Analysis of EWG simulations provides estimates for the change in erosion rates through time that is driven by changes in climate. The time-varying rates of erosion change were applied to the historical aerial imagery data in order to inform time-varying rates of erosion that are driven by changes in climate. Ultimately, this process resulted in the identification of locations for features like gully heads using both the Lidar dataset and projection of the estimated location from the historical aerial photo under consideration back to the Lidar dataset. The distance between the estimate of the location from the historical image and that of the Lidar dataset was the estimated distance the feature has moved. This distance was then divided by the number of years between the Lidar dataset and the year of the historical aerial image of interest to get a rate of movement through time. This was done for several dozen points on each historical aerial image. The analyses from the EWG were used to characterize the relative impact of climate change on the erosion rates. This relative impact was quantified by comparing the LEM based estimates of erosion that were derived using historical climate data with LEM based estimates of erosion that were derived using projections of future climate. The relative increase in the erosion rates was then applied to the historical estimates of erosion derived from the analysis of the historical aerial imagery. This approach was used since the LEM-based estimated of the historical erosion rates from the past century have a bias towards underestimating erosion. This underestimation is hypothesized to be a consequence of an inability of the LEMs to account for the impacts of land-use change (i.e. deforestation) which had been shown to significantly increase erosion in similar Northern hardwood forest ecosystems. In summary, gully head retreat rate and gully widening rate were characterized using statistical probability distributions from the historical aerial image analysis. Simulated data from the EWG was used to estimate the climatically-driven changes in erosion rates through time. The changes through time from the EWG were applied to the gully head retreat rate and gully widening rate characterized using the historical aerial image analysis. This information can be used to inform PPA models to project future risks from a given site. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

ECAR-1930 BASELINE CHARACTERIZATION DATABASE VERIFICATION REPORT - NBG-18 BILLET 635-14

The purpose of this Engineering Calculations and Analysis Report is to present the data being collected in the Baseline Graphite Characterization program, which is directly tasked with supporting the Idaho National Laboratory’s (INL’s) research and development efforts on the Next Generation Nuclear Plant (NGNP)/Very High Temperature Reactor (VHTR). This program is populating a comprehensive database that will reflect the baseline properties of nuclear-grade graphite with regard to individual grade, billet, and position within individual billets. The physical and mechanical property information being collected will be transferred to the NGNP Data Management and Analysis System (NDMAS), and from that database will help populate handbook of property data available to member nations of the Generation IV International Forum (GIF). The transfer of this data from the applicable technical lead to the dissemination databases available to other end users requires a full review of the test procedures and data collection efforts through an analysis of the multiple summary spreadsheets and values being collected. This report represents that analysis for a single billet of nuclear grade graphite (NBG-18 billet 635-14) and facilitates the release of the associated data to the NDMAS custodians.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Verification and Validation of START: A Spent Nuclear Fuel Routing and Decision Support Tool

The Stakeholder Tool for Assessing Radioactive Transportation (START) is a web-based geospatial decision-support tool being developed by the US Department of Energy’s Office of Integrated Waste Management (IWM) to support federal interim storage for spent nuclear fuel (SNF) and associated transportation. START provides many functions for the IWM program including: serving as a communications tool for conveying geospatial data and information, an options analysis tool for exploring potential transport modes and routes for transporting SNF from nuclear power plants to future federal interim storage facilities, an emergency response planning tool for Tribes and States to identify training needs along potential SNF transport corridors, an environmental analysis tool for estimating potential radiation dose exposure from incident-free and incident-case SNF transport conditions, and a systems analysis support tool providing route-related inputs for system throughput analysis. As part of the START development process, a verification and validation (V&V) effort is being undertaken. In the initial V&V phase, several outputs of the START tool were checked such as the total distance, population and population densities within the buffer zone, and incident free dose. The V&V process is fluid as it will be utilized after each version change to ensure that the core functionalities of the tool are maintained and the results are consistent with the previous versions. Efforts have also been put into developing scripts to aid in the process of automating certain sections of the V&V work. As some of the V&V efforts use Environmental Systems Research Institute’s (ESRI) tools upon which the START framework is built, the START data were compared against outputs from tools like Quantum Geographic Information System (QGIS) for buffer zone populations and route lengths to ensure independence of the V&V process. Good agreement was observed between the START results and the independent V&V studies with the majority of the differences falling between 1% and 5% for populations within the buffer zone and route distance. This presentation describes the development and design of the START tool, V&V methods employed for various metrics of interest and their respective results, and future plans.

START, transportation, GIS, V&V↗
DOE OSTI2025

Brine Availability Test in Salt (BATS) (FY25 Final Closeout)

This fiscal year 2025 (FY25) closeout report summarizes the final experiments conducted and data collected in phase 2 of the Brine Availability Test in Salt (BATS 2). The design and interpretation of the test was previously funded by the US Department of Energy’s Office of Nuclear Energy (DOE-NE) Spent Fuel and Waste Disposition Program, under the Disposal Research and Development (R&D) program of the Office of Spent Fuel & Waste Science and Technology (SFWST) but they are not funding it anymore (since May 2025). The experiment was located underground at the Waste Isolation Pilot Plant (WIPP), southeastern New Mexico, which is a DOE Office of Environmental Management (DOE-EM) site managed by the Carlsbad Field Office (CBFO). DOE-EM funds the WIPP Test Coordination Office (TCO), which provided critical implementation support for the continued execution of BATS. A high-level summary of processes expected in a salt repository are given in Kuhlman & Mills (2025).

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

Damaged Fuel in the United States - 20325

Throughout the history of commercial nuclear power plant operations in the U.S., many fuel assemblies have lost the capability to perform all of their desired functions. Since they can no longer be handled, stored, or transported in accordance with established regulations, they are classified as damaged fuel. The causes of these failed assemblies are diverse and plant-specific. The majority of these failed assemblies are contained in damaged fuel cans to be used in conjunction with storage and/or transportation systems. These systems have a limited number of slots that can be filled with damaged fuel cans. A damaged fuel can is generally a stainless-steel container that confines damaged spent nuclear fuel (SNF) and is closed at one end by mesh endpoints that allow gaseous and liquid media to escape but minimize the dispersal of gross particulate material. Out of extreme caution, a few reactors have loaded high burnup fuel into damaged fuel cans. Damaged fuel is not licensed for storage or transport in the U.S. because relevant regulations do not specify exactly how to classify damaged fuel. Instead, these regulations license/certify packages that specify approved contents. Damaged fuel must be included among the approved contents to be considered acceptable. In many cases, damaged SNF is encapsulated in damaged fuel cans to ensure it can confine gross fuel particles, debris, and/or damaged assemblies to known volumes within loaded casks. A damaged fuel can may then be utilized in the same way as an assembly in a storage and transportation system. Some storage cask systems utilize top and bottom plugs to confine debris in damaged fuel. The most recent domestic documentation on damaged SNF was published by the U.S. Energy Information Administration (EIA), which used data from U.S. reactors compiled from 1968 to June 30, 2013, to produce Form GC-859, 'Nuclear Fuel Data Survey.' According to this form, there were 136,821 boiling water reactor (BWR) SNF assemblies and 104,647 pressurized water reactor (PWR) SNF assemblies, for a combined total of 241,468. Of these, 4,521 were classified as failed. Some were also disassembled and the fuel rods or pieces of fuel rods combined to make consolidated assemblies. These consolidated assemblies may include damaged fuel or were perhaps consolidated as part of a demonstration project. The GC-859 data includes 2,550 consolidated assemblies containing 0 - 264 entire fuel rods. These consolidated assemblies could be placed in single assembly canisters and stored in the spent fuel pool. For dry storage and transportation, a single assembly canister is generally placed in each damaged fuel can. In addition to the consolidated assemblies, 2,391 un-canistered fuel rod pieces exist, which were removed from 494 assemblies. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗