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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.

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At least 163 records · Page 9
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Fit for Purpose Remote Technology Decommissioning Solutions - 18630

The Magnox Decommissioning Project is actively progressing clean-up and decommissioning of 12 reactor sites across England, Scotland and Wales. Funded and directed by the United Kingdom's Nuclear Decommissioning Authority (NDA) and managed by the Cavendish Fluor Partnership (CFP), the Magnox Decommissioning Project introduces a wide range of radiological and industrial hazards warranting deployment of remote technologies to safely de-inventory, decontaminate and decommission a vast array of nuclear facilities. The Magnox Decommissioning Project encompasses 22 former Magnox nuclear power production reactors across 10 reactor sites and five research/test reactors at two former research sites. Decommissioning activities have been underway at some sites for nearly 20 years with early focus on retrieving and stabilising wastes and draining and stabilising spent fuel ponds. The Magnox decommissioning strategy invokes a programme perspective to drive consistency in tackling similar decommissioning and waste challenges across multiple sites. Programme central teams have worked with technology suppliers to foster learning from industry experience to develop and deploy tooling and robotics. The programme approach has also facilitated learning from one site to the next. In many cases, this central programme approach has promoted an evolution of technologies resulting in optimisation of simple remote tooling. In other instances, collaboration with suppliers is producing state-of-the-art technologies to tackle more unique challenges including, most recently, early demonstration of laser cutting for reactor vessel dismantling. This paper will outline the programme management approach and the scalable optioneering process Magnox employs to select the engineered controls and technologies used in decommissioning and waste management activities. It will provide an overview of the broad range of remote tooling and robotics deployed across the 12 Magnox sites by exploring some specific examples. Lastly, the paper will summarise key lessons learned in the now extensive history of selecting and using remote technologies to decommission Magnox nuclear facilities. (authors)

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

Ultrasonic imaging in liquid sodium

The fourth generation of nuclear reactor can use liquid sodium as the core coolant. When the reactor is operating, sodium temperatures can reach up to 600 deg. C. During maintenance periods, when the reactor is shut down, the coolant temperature is reduced to 200 deg. C. Because molten sodium is optically opaque, ultrasonic imaging techniques are developed for maintenance activities. Under-sodium imaging aims at i) checking the health of immersed structures. It should also allow ii) to assess component degradation or damage as cracks and shape defects as well as iii) the detection of lost objects. The under-sodium imaging system has to sustain high temperature (up to 300 deg. C) and hostility of the sodium environment. Furthermore, specific constraints such as transducers characteristics or the limited sensor mobility in the reactor vessel have to be considered. This work focuses on developing a methodology for detecting damages such as crack defects with ultrasound devices. Surface-breaking cracks or deep cracks are sought in the weld area, as welds are more subject to defects. Traditional methods enabled us to detect emerging cracks of submillimeter size with sodium-compatible high-temperature transducer. The presented approach relies on making use of prior knowledge about the environment through the implementation of differential imaging and time-reversal techniques. Indeed, this approach allows to detect a change by comparison with a reference measurement and by focusing back to any change in the environment. It is a means of analysis and understanding of the physical phenomena making it possible to design more effective inspection strategies. Difference between the measured signals reveals the acoustic field scattered by a perturbation (a crack for instance), which may occur between periodical measurements. The imaging method relies on the adequate combination of two computed ultrasonic fields, one forward and one adjoint. The adjoint field, which carries the information about the defects, is analogous to a time-reversal operation. One of the interests of the presented method is that the time-reversal operation is not done experimentally but numerically. Numerical simulations have been carried out to validate the practical relevance of this approach. The preliminary numerical results show a nice agreement between the guessed and the actual positions of the defect. After water-tests, in sodium-tests must be done in order to validate the water/sodium transposition. For this purpose, an under-sodium device is under development, which can move the transducers with four degrees of freedom in a 1.5 m{sup 3} sodium pot. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Development of Advanced Remote Cutting Technology for Reactor Internals Segmentation - 18274

Since the Three Mile Island decommissioning in the 1980's, Westinghouse has been involved in dismantling and segmentation of commercial reactors. With dozens of components segmented and packaged across the globe, Westinghouse has continued to evaluate the state of the art cutting technologies for their applications in the nuclear industry. While there have been well documented issues of certain cutting technologies in field applications, there are advancements of these technologies which prove them applicable for specific areas of the decommissioning process. During 2017, Westinghouse embarked on a revision of its plasma arc cutting (PAC) technology as part of an extensive innovation program for reactor internals segmentation. While these systems were used in Yankee Rowe and Three Mile Island, new plasma arc cutting systems improve the versatility and have much better precision of cutting. While bringing this technology into the future, Westinghouse is adapting the mounting and application so that it is compatible with the existing mechanical cutting portfolio. This means that in the event that a mechanical cutting technique is unable to perform as intended, a plasma cutting alternative will be available with minimal time to switch cutting methods. A test program for underwater PAC was performed, and the test gave a good reference of cutting speed and handling in different plate thicknesses which can be used when evaluating the use of the new cutting technology for future projects. PAC technology has proven that it could have a future in reactor vessel internals segmentation projects as a complement to the existing mechanical cutting methods under water. Thicknesses up to at least 25 mm can be cut with a much higher speed than the existing mechanical cutting methods in use. This and other cutting technology developments as part of the innovation program provide exciting new possibilities for dismantling and segmentation projects. While each of the techniques presents definite benefits for operators and, more importantly, customers, there are risks that need to be mitigated through engineered processes and controls. (authors)

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DOE OSTI2016

Code Validation of a SBLOCA test in SMART Integral Test Loop

Integrated small modular reactors (iSMRs) are favored because of their unique characteristics, such as enhanced safety without any large break loss of coolant accidents (LBLOCAs), low initial construction costs, and multi-use of desalination and electricity generation. Many iSMRs have been developed thus far. In Korea, an iSMR called SMART was developed after a thorough validation of the components and systems, and finally received its design certificate in 2012 for the first time in the world. To ensure the performance of its safety systems, an integral test loop, SMART-ITL, is constructed. It contains the primary system, secondary system, passive safety systems, and an auxiliary system. The passive safety systems include a passive residual heat removal system (PRHRS), passive injection system with a core makeup tank (CMT) and a safety injection tank (SIT), and an automatic depressurization system (ADS). The height is preserved while the diameter is scaled down to 1:7. A series of SBLOCA tests have been performed with 4 trains of PRHRS, 1 train of CMT and SIT, and ADS. In parallel with the experiments, simulation of a representative test has been conducted to track the underlying thermal hydraulics. SMART-ITL is an integral test loop for SMART and a series of SBLOCA tests for confirming the performance of passive safety systems have been conducted. In this paper, a system code simulation on a SBLOCA test using SMART-ITL is presented. The reactor vessel water level, break flow rate, and CMT and SIT water levels are predicted well. (authors)

42 ENGINEERING↗
DOE OSTI2020

Computational Fluid Dynamics of Bioreactors with Micro-Aeration

Recent work has probed the potential for using Zymomonas mobilis as a flexible platform for the production of advanced fuels and intermediates. One set of recombinant organisms requires a micro-aerobic environment to promote the production of the desired product (2,3-BDO), but the economics and process-control implications of this approach are uncertain at industrial scale. Computational multi-physics models have the potential to describe the fermentation in detail at multiple scales, reducing the risk of scale-up. A two-phase Euler-Euler CFD model is used alongside a simple model of oxygen uptake to describe this bioreaction. A few reactor vessels, including airlift and CSTR bioreactors, are analyzed at multiple scales, regions of oxygen-depletion and over-oxygenation are characterized, and the bioreactors are compared. We show that it is difficult to achieve homogenous oxygen concentrations in industrial-scale CSTR reactors and that advanced bioreactor designs are needed for economical operation.

09 BIOMASS FUELS↗
DOE OSTI2017

Analysis of COBRA-TF Critical Heat Flux Models for Triangular Fuel Assembly Pitch

The Coolant Boiling in Rod Arrays-Two Fluid (COBRA-TF) code is a best estimate code which is used to perform thermal-hydraulic analyses of a light water reactor vessel. The COBRA-TF heat transfer package has models calculating the critical heat flux (CHF). In order to verify code's predictive capability for CHF for triangular nuclear fuel assembly pitch, experiment results and system parameters of KS-1 test facility that is single loop model of the VVER primary system located in Kurchatov Institute were simulated using COBRA-TF. In order to calculate CHF for triangular nuclear fuel assembly pitch there is a look-up table generated. In addition, there are some correction factors for critical heat flux values listed the table. In this study, once KS-1 experiment results were simulated using COBRA-TF, correction factors were used to correct CHF results from COBRA-TF calculations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

The High Capacity Type B Waste Transport Cask to Serve Dismantling Project - 18344

The United States have already announced around 20 reactor shutdown and many more in the next 10 year. Financial constraints on plant operators and the low cost of fossil energy contribute to accelerate the shutdown of additional reactors. The industry has been gaining experience with reactor dismantling and is continuously trying to improve the cost efficiency of the upcoming projects. To do so, larger and more capacitive transport casks are needed to transport increasing volume of Low Level Waste (LLW) at a more attractive cost. Waste disposal is also a major cost element in a dismantling project. Consequently, the capacity to segment, package and transport the waste in the most optimized configuration will result in significant cost savings. While for LLW, the constraints for package requirements are limited, activated waste resulting mainly from reactor vessel and vessel internals segmentation require additional shielding to meet disposal and transportation regulations. The key element to succeed is the availability on the market of a licensed type B transport package with a larger payload that allows at the same time an optimization of the waste segmentation. AREVA TN has designed and licensed versatile high capacity type B casks for the transport of spent fuel canisters as part of the NUHOMS{sup R} system that can also be used to transport radioactive waste. While it will take a few more years to transport spent fuel to a consolidated interim storage facility and even more to a repository, the type B casks can be used to transport activated LLW from dismantled reactors to disposal sites. A first of a kind project will be performed with AREVA's partner NorthStar at Vermont Yankee Nuclear Power Plant. AREVA's team will perform the reactor segmentation and package the material into an AREVA's large capacity type B cask. This unmatched cask capacity is expected to significantly decrease the LLW disposal costs, as well as the onsite operations costs, while optimizing the schedule. The package will be transported by rail to waste disposal facilities. This paper will present the main characteristics of the package, provide an update of the overall schedule, and explain the innovations that contribute to simpler and more cost effective transport and operation. Finally, it will emphasize the benefits of using larger type B casks for immediate and future transportation needs. (authors)

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DOE OSTI2017

Upfront Update on Barsebaeck 1-2 Reactor Internals Segmentation Project - 17382

Westinghouse is currently executing a contract, awarded in June 2015 from Barsebaeck Kraft AB, to segment all the reactor vessel (RV) internal parts in the two units 1 and 2. The scope of work covers the dismantling and segmentation of the RV internals under water, including the upfront engineering studies and equipment manufacturing and qualification. It also includes the loading of the segmented parts into waste cassettes, handling the cassettes, reactor internals sampling (73 samples), complementary water filtration system to catch the segmentation chips and contribute to maintaining water clarity during the segmentation work. Before leaving the site, the pool environment must be restored to the initial condition and all equipment will be decontaminated and shipped off site. The site work started in August 2016 with the mobilization on the Unit 2 and the actual cutting of the first component (steam dryer) started in September. Cutting will be performed with different types of Westinghouse designed cutting equipment. This paper presents the conceptual design of the reactor internals segmentation and packaging process that will be implemented at Barsebaeck, including the planning, methodology, equipment, waste management and packaging strategy. A first feedback on the site work will be provided for the segmentation of the Unit 2 reactor internals. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Progress in Fast Modular Reactor Conceptual Design

The Fast Modular Reactor (FMR) is a 100-MW(thermal) gas-cooled fast reactor being developed by General Atomics Electromagnetic System with the goal of developing a FMR for flexible and dispatchable power to the U.S. electricity market in the mid-2030s. The conceptual design aims to develop and verify simplified design features. These include an inert helium gas coolant, pellet-loaded fuel rods, installations with air cooling as ultimate heat sink, and small and passive heat removal systems. The goal is to ensure the development of a safe, maintainable, cost-effective, and distributed nuclear energy-generating station. The baseline technologies selected to achieve this goal are a helium coolant that is an inert gas with no chemical reaction with structural components, not activated, single phase, enabling high-temperature operation and a high thermal efficiency Brayton cycle; conventional uranium dioxide (UO 2 ) fuel, which is the most widely used and well-known fuel material, capable of high burnup (100 MWd/kg) and a long fuel life; and silicon carbide composite (SiGA®) cladding and internal structures that are chemically inert in the helium environment, exceptionally radiation tolerant, and being derisked by accident tolerant fuel technology development. Further, the reactor was specifically designed with passive safety features, including high-temperature in-core materials and a reactor vessel cooling system consisting of cooling panels of naturally circulating water. The passive safety of the core was confirmed for the depressurized loss-of–forced cooling accident, which showed the peak cladding temperature at ~1600°C during the transient, which is below the current design limit of 1800°C. The conceptual design of the FMR has been conducted for the reactor system, vessel system, generator and turbomachine, instrumentation and control, residual heat removal system, plant service system, and containment, as well as pre-application licensing documents.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Irradiated stainless steel 347 and irradiated, welded, and hydrogen charged Zircaloy-4 property data for the target solution vessel and support lines of SHINE

In support of the reactor-vessel development effort of SHINE medical technologies, irradiation of its structural materials and post-irradiation testing have been performed in the High Flux Isotope Reactor at Oak Ridge National Laboratory. This irradiation was performed on welded and unwelded AISI 347 steel and Zircaloy-4 samples at low temperature (~60°C) in contact with the reactor coolant. Hardness, Charpy, and tensile data have been taken from all cases of unirradiated specimens, while post-irradiation tensile and hardness data have been taken from AISI 347 base metal and Zircaloy-4 base metal and weld specimens so far. Additionally, 3-point bend data have been taken from explosion-welded specimens. The expected changes to hardness, yield strength, ultimate tensile strength, uniform elongation, and total elongation as a result of the low temperature and low dose neutron irradiation are reported here.

36 MATERIALS SCIENCE↗
DOE OSTI2015

A Scoping Analysis Of The Impact Of SiC Cladding On Late-Phase Accident Progression Involving Core–Concrete Interaction

The overall objective of the current work is to carry out a scoping analysis to determine the impact of ATF on late phase accident progression; in particular, the molten core-concrete interaction portion of the sequence that occurs after the core debris fails the reactor vessel and relocates into containment. This additional study augments previous work by including kinetic effects that govern chemical reaction rates during core-concrete interaction. The specific ATF considered as part of this study is SiC-clad UO 2 .

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Emergency response and solutions for damaged fuel assembly handling events - Lessons learned from a 2013 fuel positioning issue

In 2013, during the refueling of a French 1300 MWe reactor, a fuel assembly (FX3599) was damaged by a fuel assembly positioning tool in the reactor vessel. The upper part of a corner rod of the fuel assembly was twisted and the top end grid was torn off. Because of the protruding fuel rod, the assembly could not be easily transferred for repair to the fuel storage pool via the transfer tube. This unexpected and rarely encountered situation was assessed jointly by EDF and AREVA NP. The chosen option was to directly perform the field service inside the reactor building in cells generally used for intermediate storage during refueling. This operation, performed by EDF and AREVA, involved removing the twisted rod and positioning it in a special-purpose canister by means of a gripper-type tool for removal (KIPCI gripper) and a gripper for holding the rod in place. The special-purpose canister was designed, ordered and supplied by a sub-supplier within one week. The operation went through 3 phases. First a preparation step in the fuel building: The stripper was disassembled and technical checks carried out to ensure correct operation. The gripper tightening force was checked in real-life conditions with the gripper suspended from auxiliary crane. The fuel rod removal was simulated several times by trained operators. Second, the equipment was installed in the reactor building. And third, the operation took place in the reactor cavity: the special-purpose canister was opened, the radioactive waste canister placed in the special-purpose canister, the twisted fuel rod was then grasped by the KIPCI gripper and removed from the fuel assembly. The twisted fuel rod was positioned above the canister and dropped into it. Once the fuel rod had been removed, the canister and fuel assembly FX3599 were transferred to the fuel building and refueling resumed. The canister containing the fuel rod is currently in interim storage in the reactor pool until appropriate action is taken for it to be shipped and eventually reprocessed. EDF has performed improvement in its handling procedures and in its handling tools to avoid a similar new event.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Experimental Studies of Coolant Temperature and Velocity on an Integrated Water Model of a Fast Reactor in Different Operating Regimes

The results of experimental studies of a next-generation fast reactor performed using an integrated water model (scale 1:10) are presented: local velocity over height and radius of the top chamber in a plane in a direction from core center to the intermediate heat exchangers and coolant temperature in the top chamber and other elements of the circulation loop with forced circulation and emergency cool-down by natural convection. The data were obtained using a specially designed system for performing measurements which is incorporated in the bench and provides high accuracy and rate of acquisition. Stable temperature stratification of the coolant is observed in the top, cold, and pressure chambers of the reactor, elevator enclosure, and reactor vessel cooling system as well as at the exit from the intermediate and autonomous heat exchangers in different operating regimes.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

SIT Makeup Effect on the Loss of Residual Heat Removal during Mid-loop Operation

For the loss of cooling during the mid-loop operation, the plant needs to have a sufficient time for operator action without active makeup measures such as a loss of alternating current (AC) power. This concern is raised after the Fukushima accident. Especially, in addition to the loss of residual heat removal (LORHR), if the AC power or active makeup measures are unavailable, the passive makeup using gravity feed or pressurized feed will provide sufficient time before core uncovery. This paper investigates means of the passive makeup measures using safety injection tank (SIT) after the LORHR event to prevent or delay the core uncovery and fuel failure for the typical advanced power reactor 1400 (APR1400). Unfortunately, the refueling water storage tank of APR1400 cannot be used for gravity makeup since it is located on the bottom of the containment with a lower elevation than the reactor vessel. So, the SIT makeup and the effect on the core uncovery are investigated. The results show that gravity makeup of a SIT provides insufficient means but the slightly pressurized SIT is required for sufficient operator action time for the LORHR with the station black out (SBO) during the mid-loop operation for the APR1400. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Benchmarking of Probabilistic Fracture Mechanics Models in Grizzly

In 2020, as part of contract/task order 31310019N0006/31310020F0060, the Nuclear Regulatory Commission tasked Idaho National Laboratory (INL) with benchmarking the MOOSE/GRIZZLY code against the NRC’s Fracture Analysis of Vessels - Oak Ridge (FAVOR), version 16.1 code. Both codes can predict large light-water reactor vessel integrity deterministically and probabilistically. In fact, such predictions are the primary purpose of the FAVOR code, and are a main focus of the GRIZZLY code, although GRIZZLY is capable of modeling a wider range of geometries and problems. The vessel integrity modeling capabilities of the GRIZZLY code are still under development and being gradually expanded, and the capability to model vessel fracture by crack growth has not yet been integrated into GRIZZLY (although it does exist in pre-release versions). Thus, the present benchmark study focused on comparing predictions for the conditional probability of crack growth initiation (CPI) between GRIZZLY and FAVOR v16.1. The present benchmarking study expands on a previous study completed in 2020, which focused on single plate regions, various flaw types, and full vessel studies. The previous study was limited in scope and revealed some relatively large discrepancies, particularly in the case of weld regions. In the present study, detailed investigations into the sources of discrepancies between GRIZZLY and FAVOR were performed, with a focus on weld regions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Successes and Challenges of the Integrated Waste Treatment Unit's 50-Day Demonstration Run - 20151

The Integrated Waste Treatment Unit (IWTU) has been tasked with using steam reforming technology to treat 3.4 million liters (900,000 gallons) of radioactive sodium-bearing waste (SBW) at the Idaho National Laboratory (INL). The project has been hampered by a number of processing issues associated with process equipment during simulant testing, with the two main issues involving the fluidizing gas distribution system in the main reactor vessel, the denitration mineralization reformer (DMR), and high differential pressures in the main filter vessel, the process gas filter (PGF). Following the replacement of the fluidizing gas distribution system in the DMR, a successful 30-day demonstration run was performed that revealed unacceptable bridging in the filter bundles of the PGF. The bundles were replaced with new ones, and a 50-day demonstration run was performed. The main purpose of this run was to operate the IWTU over a range of operating conditions. Overall, the run was a success; however, there were multiple challenges that arose from the run. (authors)

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

Modelling of transport processes in concrete exposed to elevated temperatures – An alternative formulation for sorption isotherms

There is a significant need to understand, analyse and assess moisture transport in cementitious materials exposed to elevated temperatures in order to confidently predict the behaviour and ultimately the development of damage in safety critical applications such as nuclear reactor vessels, structures exposed to fire and well bore grouts. In view of this need a rigorous and robust formulation to describe water retention curves (sorption isotherms) as a function of temperature based on the evolution of physical parameters is presented. The model formulation is successfully validated against independent sets of experimental data up to temperatures of 80 °C. It is then further validated under isothermal drying conditions and then high temperature conditions through the numerical reproduction of laboratory experiments following implementation in a fully coupled hygro-thermo-mechanical finite element model. The new formulation is found to work well under a variety of conditions in a variety of cementitious material types.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Preliminary Evaluation of Removing Spent Nuclear Fuel from Shutdown Nuclear Power Plant Sites - 2018 - 19101

This is a technical paper that does not take into account the contractual limitations under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (Standard Contract) (10 CFR Part 961). Under the provisions of the Standard Contract, DOE does not consider spent nuclear fuel in multi-assembly canisters to be an acceptable waste form, absent a mutually agreed to contract amendment. To the extent discussions or recommendations in this paper conflict with the provisions of the Standard Contract, the Standard Contract provisions prevail. The U.S. Department of Energy Office of Integrated Waste Management (DOE-IWM) continues to conduct evaluations of removing spent nuclear fuel (SNF) from shutdown nuclear power plant sites. The 14 shutdown sites included in the evaluation to date have nuclear power reactors that have been permanently shut down and the sites have been decommissioned or are undergoing decommissioning. The 14 shutdown sites are Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, Zion, Crystal River, Kewaunee, San Onofre, Vermont Yankee, and Fort Calhoun. The Oyster Creek Nuclear Generating Station ceased operations in September 2018 and will be included in future evaluations. The evaluations of the shutdown sites contain four components: - A characterization of the SNF and greater-than-Class C low-level radioactive waste (GTCC waste) inventory; - A description of the on-site infrastructure at the shutdown sites; - An evaluation of the near-site transportation infrastructure and experience at the shutdown sites - An evaluation of the actions necessary to prepare for and remove SNF and GTCC waste. The primary sources for the inventory of SNF were the DOE GC-859 and RW-859 spent nuclear fuel inventory databases. This inventory has been augmented with data obtained from the shutdown sites for SNF discharges that occurred after June 2013. The primary sources for information on the conditions of site and near-site transportation infrastructure and experience included field observations and information collected during visits to the 14 shutdown sites, information provided by managers and staff at the shutdown sites, and Google Earth imagery. Extensive photographs taken during the shutdown site visits have proven to be particularly useful in documenting the current conditions at and near the shutdown sites. Tribal and state representatives, State and Regional Group representatives, and Federal Railroad Administration representatives have participated in 11 of the shutdown site visits. In addition to visiting the shutdown site, site visits often include activities such as meeting with local railroads to discuss the near-site railroad infrastructure and meeting with community engagement panels. Oyster Creek permanently ceased power operations on September 17, 2018 and fuel was removed from the reactor vessel on September 25, 2018. Planning for the Oyster Creek site visit has begun and is conducted in three stages. In the first stage, data and documents that are applicable to Oyster Creek are collected. In the second stage, questions are developed based on the data and documents collected during the first stage. These questions are submitted to the Oyster Creek site approximately one month before the site visit. The third stage involves establishing specific schedules and participants. Participants will likely include tribal and state representatives, the Council of State Governments - Eastern Regional Conference, and the Federal Railroad Administration. State participation will likely include a representative of the State Rail Safety Participation Program, the New Jersey Department of Transportation, and the New Jersey Bureau of Nuclear Engineering. The site visit is tentatively expected to occur in April or May of 2019. (authors)

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