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At least 307 records · Page 17

Qualification of Y-12 Legacy Criticality Accident Alarm System Detectors

The Y-12 National Security Complex has had a criticality accident alarm system (CAAS) capable to detect a criticality accident since 1945. With the exception of Y-12’s new High-enriched Uranium Materials Storage facility (HEUMF), the system installed in Y-12’s existing facilities dates to 1957. This legacy system has received many upgrades since installation and detector stations have been removed and relocated as enriched uranium operations have evolved, but the basic design and electronic configuration of the system is the same as that in 1957.

61 RADIATION PROTECTION AND DOSIMETRY↗

Materials for Advanced Ultra-Supercritical (AUSC) Steam Turbines – AUSC Component Demonstration

The work by the United States Department of Energy (U.S. DOE)/Ohio Coal Development Office (OCDO) Advanced Ultra-Supercritical (AUSC) Steam Boiler and Turbine Materials Consortia from 2001 through September 2015 was primarily focused on lab scale and pilot scale materials testing. The field testing included air-cooled probes and steam-cooled “loops” that were installed into existing utility boilers to gain exposure of these materials to realistic conditions of high temperature and corrosion due to the constituents in the coal ash. Successful research and development resulted in identification of metallic alloy materials and fabrication processes suited for power generation applications with metal temperatures up to approximately 1472°F (800°C). These materials or alloys have shown, in extensive laboratory tests and shop fabrication studies, to have excellent applicability for high-efficiency low CO2 transformational power generation technologies. However, as valuable as these material field tests have been for obtaining information, their scale is significantly below that required to minimize the risk associated with a power company building a multi-billion-dollar AUSC power plant. To decrease the identified risk barriers to full-scale implementation of these advanced materials, the U.S. DOE/OCDO AUSC Steam Boiler and Turbine Materials Consortia identified the key areas of the technology that need to be tested at a larger scale. Based upon the recommendations and outcome of a Consortia-sponsored workshop with the U.S.’s leading utilities, a Component Test (ComTest) Program for AUSC was proposed. The AUSC ComTest program would define materials performance requirements, plan for overall advanced system integration, design critical component tests, fabricate components for testing from advanced materials, and carry out the tests. A Pre-FEED evaluation of the test facility including the evaluation of three potential sites has been completed and previously reported. This report describes the Front-End Engineering Design (FEED), Detailed Engineering and the Environmental Assessment of the ComTest project. A significant amount of work was accomplished during Phase 1 to identity the components, as well as the component size, that would be manufactured from advanced alloys such as INCONEL 740H or HAYNES 282 alloys. Pathways to supply these components for ComTest have been identified as well as any further development that would be required. Every project of this nature has challenges and this project was no exception. The greatest challenge has been the project cost. Largely due to the infrastructure and balance of plant equipment upgrades required for Youngstown Thermal, the estimated project cost for the “base case” project was greater than originally budgeted. The project team spent a significant amount of time exploring opportunities for reducing the project cost. These options included: • Youngstown Thermal was exploring building a Combined Heat and Power (CHP) unit at their facility. The Youngstown Thermal staff, supported by Thermal Engineering Group, worked with the project team to determine whether it was economically desirable to supply ComTest with steam and other utilities, which would be supplied by the CHP project. • The project team also evaluated opportunities to optimize the project scope to reduce cost, such as reducing project scope by eliminating the Superheater Outlet Header test. • Southern Company had offered two sites as possible host locations for ComTest. Both sites had infrastructure in place such as steam from an existing unit, natural gas, cooling water and ample electrical power that would be required to be added at the Youngstown Thermal site. Utilizing this existing infrastructure could reduce the project cost. An additional project challenge was the departure of Babcock and Wilcox from the project in September 2016. This project setback was overcome by GE Power’s willingness to assume the design of the ComTest superheater. Since the project could not proceed without the superheater, this was a very positive project development. The merger of the GE Power and Alstom Power steam turbine divisions created synergies leading to the conclusion that instead of building a small-scale turbine, a more practical and useful endeavor would be to validate the manufacturing technology and supplier readiness for fabricating and machining full scale steam turbine components. Consequently, in March 2017, GE Power decided that a steam turbine test would not be necessary as part of the ComTest demonstration. Removing the operational testing of the steam turbine from ComTest further reduced the total project cost. Additionally, there was an opportunity to move ComTest from Youngstown Thermal to one of the Southern Company sites and operate ComTest at supercritical steam conditions and test the steam turbine main stop/control valve at a temperature and pressure that would be representative of the conditions of a high pressure steam turbine inlet valve. The steam could also possibly be returned to the host boiler so there would be no economic penalty of using this steam for ComTest. Relocating the project could further reduce the project cost due to the infrastructure already in place, including high pressure natural gas and cooling water, and the potential for heat integration, allowing a reduction of operating costs. Due to these potential cost reductions, it was decided to move the project away from Youngstown Thermal in March 2017. As integrating ComTest at the Southern Company site was explored, an issue was raised regarding the integration of the ComTest cycle with the plant steam cycle. Steam would be supplied to ComTest at nominally ~3865 psia and 775°F. The steam would be heated with natural gas to 1400°F for the valve test and then cooled to 1000°F so that the steam would be returned to the host boiler at the same conditions as their high pressure steam turbine inlet. However, there were concerns raised that the addition of the energy (natural gas) to the steam could trigger a New Source Review on the unit. Southern Company did not want ComTest to cause this review, so condensing all of the steam exiting ComTest and returning it to the host steam cycle as condensate was explored to insure there would be no implication that ComTest was adding energy to the host plant’s steam cycle. Ultimately, Southern Company decided that they wanted to “decouple” ComTest from the operating unit, making ComTest self-supporting, similar to the Youngstown Thermal “base case.” Without the steam turbine, the project team reached a consensus that operating the ComTest Superheater and steam turbine valves at the typical pressure of a package boiler (~1000 psig) would not generate enough useful technical information to justify the cost of the test. Therefore, it was decided in the December 2017 project quarterly meeting to cancel the operational phase of ComTest. Throughout the engineering phase of ComTest, manufacturing and supply chain challenges were identified. When GE Power recommended dropping the steam turbine test from ComTest, they identified many important manufacturing issues that should be evaluated and demonstrated. Other challenges had been raised by the GE Power group designing the superheater as well as those engineering components such as pressure relief and steam turbine bypass valves. In the December 2017 meeting, the project team decided that it was more beneficial to the AUSC technology to pursue these manufacturing and supply chain issues as Phase 2 of the ComTest project. This report is organized in the following manner: • Section 2 discusses the ComTest project background and organization. • Section 3 discussed project management. • Section 4 discusses the design of the ComTest AUSC components – the Superheater, steam turbine and Superheater Outlet Header. • Section 5 discusses the implementation of ComTest at Youngstown Thermal and the Southern Company site. • Section 6 discusses the development of the AUSC Supply Chain • Section 7 discusses the recommended plan for Phase 2 of ComTest

01 COAL, LIGNITE, AND PEAT↗

Ion concentration polarization-electrocoagulation hybrid water treatment system

Between two juxtaposed similar ion exchange membranes (AEMs or CEMs), an ion depletion zone and ion enrichment zone are generated under an electric field. As cations are selectively transferred through the CEMs, for example, anions are relocated in order to achieve electro-neutrality, resulting in the concentration drop (increase) in ion depletion (enrichment) zone. The use of a sacrificial metal anode allows electrocoagulation (EC) concurrently with ICP thereby permitting concentration of both ionic and non-ionic impurities to occur at the same time within the same cell or device.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

AUXILIARY FEEDWATER REHEATING TO MITIGATE PRIMARY AND SECONDARY SIDE THERMAL STRESSORS

A consequence of operating steam cycles under time varying loads, is the variability of the feedwater temperature entering the bottom of the steam generators. This variability is a direct result of changes in turbine impulse pressure and subsequently turbine tap pressures in response to the load changes, with a reduction in feed temperature during periods of low turbine demand. These fluctuations over time will increase thermal cycling and threaten the integrity of the tube bundle as well as other components in the feed system. Sensible heat, TES systems have been shown to be an effective load management strategy allowing nuclear reactor systems to operate at effectively 100% full power while storing excess thermal energy for recovery later. A byproduct of the TES system is the availability of large amounts of low grade waste heat that can be used for ancillary applications, including auxiliary feed heating during periods of low turbine demand. The addition of an auxiliary feed heater at the end of the traditional feed train can minimize feed temperature variations during periods of low turbine demand, reducing the effects of thermal cycling and increasing overall thermodynamic efficiency. This paper discusses the design, operation and control strategies for an auxiliary feed water system as an integral component of a coupled Thermal Energy Storage system and Integral Pressurized Water Reactor. The impact on system performance, particularly as it affects reactor dynamic response and reductions in thermal cycling will be addressed. In addition to the improved thermal efficiency, a benefit of this approach is the relocation of thermal stressors from the reactor coolant system to more easily managed Balance of Plant systems

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Re-Evaluating the Current Safety Goals

The Safety Goals adopted by the US NRC consist of two qualitative safety goals (QSGs) backed up by two quantitative health objectives (QHOs). The QHOs establish risk limits for severe accidents in terms of their radiological consequences, in particular, the average individual health risks of early fatality and latent cancers from radiation exposure of members of the public living in the vicinity of a nuclear power plant. This paper is devoted to a re-examination of the coverage of the current safety goals in evaluating the actual risk posed by nuclear power plant operation. Experience of recent Level 3 PRAs suggests that given a high likelihood of evacuation of the close-in population before any release occurs, the current QHOs are satisfied by large margins. However, the experience of an actual severe accident at Fukushima showed that actual human health effects from released radiation were not the dominant consequences; there were no early fatalities and no measurable increases expected in cancer rates above the baseline rates in the Japanese population. Hence, regardless of accident probability, Fukushima-type accidents would satisfy the NRC’s health-related safety goals. However, there were very significant societal costs, in that large numbers of people were relocated for long periods. We argue that, in addition to the risks addressed in current safety goals, societal risk should also be evaluated. The paper discusses specific possibilities for a goal and an associated quantitative objective.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Purification of ultra-high saline and contaminated water by multi-stage ion concentration polarization (ICP) desalination

A water stream is passed between two juxtaposed similar ion exchange membranes (AEMs or CEMs), forming an ion depletion and ion enrichment zones when an electric field is applied. As cations are selectively transferred through the CEMs, for example, anions are relocated in order to achieve electro-neutrality, resulting in the concentration drop (increase) in ion depletion (enrichment) zone. Trifurcation of the output channel allows collection of concentrated, dilute and intermediate streams, with the intermediate stream serving as input to the next stage of a serialized implementation.

Kim, Bumjoo↗

Separate-Effects Validation Experiments for Models of Fracture in Ceramic Nuclear Fuel

A number of advances have recently been made in modeling fracture initiation and propagation in ceramic light water reactor (LWR) fuel. These include the use of techniques such as peridynamics, the extended finite element method (XFEM), the discrete element method (DEM), and phase field fracture. Improved predictions of fracture in LWR fuel can benefit fuel performance simulations in a number of ways, with important implications for radial relocation, pellet-cladding mechanical interactions, and fragmentation during accident conditions. Although these modeling approaches appear promising, to-date most of their applications have been to fresh fuel. A major limitation in applying these approaches with confidence to a wider variety of conditions, such as fragmentation of high-burnup fuel, is that there is very limited data that can be directly used to validate the results of these fracture models. While fracture patterns are available from cross-sections of numerous irradiated fuel rods, the fuel rods have typically been subjected to complex irradiation histories. To validate models of fracture propagation, focused experiments that give snapshots of the initiation and propagation of fracture under well characterized prototypical reactor conditions are needed. Under a multi-institutional U.S. Department of Energy Integrated Research Project (IRP), three separate-effects validation experiments are underway to provide information on the propagation of cracks in LWR fuel under controlled conditions. The primary challenge in developing such an experiment is in providing conditions that result in a radial temperature distribution that is prototypical of LWR in-reactor conditions. Two out-of-reactor experiments are being planned that employ very different approaches to achieve such a temperature distribution. The first of these achieves this by employing a combination of inductive and resistive electrical heating to generate volumetric heating. The second experiment involves slowly heating a pellet in a metal rod to a high temperature, and then quenching that rod in a cold bath. The quenching process generates a temperature profile approximating that in an LWR fuel rod for a moment, but has a very different temperature history from the LWR. Finally, an experiment is being planned for the Idaho National Laboratory TREAT reactor that will use nuclear fission as a heat source, and which will employ a metal heat sink to replicate the thermal conditions provided by coolant in an LWR. These experiments all have their advantages and drawbacks, with varying levels of ability to reproduce actual LWR conditions and varying ability to use instruments to monitor quantities of interest over the course of the experiment that can be used to validate fracture models. For that reason, these different experimental approaches are complimentary to each other, and the combined results are expected to significantly improve confidence in the ability of computational models to simulate fracture in ceramic nuclear fuel.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Separate Effects Testing in TREAT for ATF Fuels

This work performed examinations on high burnup UO2 from commercial power plant with the intention of better understanding fission gas and fragmentation behavior. Fuel fragmentation, relocation, and dispersal (FFRD) is a major factor in fuel lifetime limits imposed on UO2 by regulators, both domestic and foreign, and is a major area of interest by industry groups (e.g., Electric Power Research Institute, [EPRI]). The characterization of the fuel focused on the differences in porosity between the high burnup (rim) zone, the dark (intermediate) zone, and center zones of the fuel as well as the fission gas content of each region. Work involved thermal treatments to analyze fission gas and high resolution microscopy techniques to investigate pore structure.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

An Optimization-Based Strategy for SAV Fleet Repositioning

With the emergence of autonomous technology, shared autonomous vehicles (SAVs) will potentially be the prevalent transportation mode for urban mobility. On one hand, relying on SAV fleets can provide several operational benefits. On the other hand, SAVs can increase travel distance and add congestion due to unoccupied trips such as pickup and repositioning trips. One important aspect for a SAV fleet's success is to serve the incoming requests at reasonably low waiting time. This is achieved by an adequate fleet size that is spatially distributed thoughtfully so that incoming requests can be served by a nearby vehicle. Unfortunately, it is challenging to keep a satisfactory spatial distribution of vehicles due to imbalances in the origin and destination patterns of incoming requests. This paper focuses on the impact of SAV relocation on traveler wait times using a novel optimization-based algorithm for repositioning. POLARIS, an agent-based tool, is used for a case study of Bloomington, Illinois to quantify the benefits of allowing SAV repositioning. On average, the wait times were around 20% lower with repositioning for all adequate fleet sizes. SAVs were available more uniformly across the region's zones, and proportional to trip-making at different times of day. In addition, enabling repositioning led to a higher share of demands being served. These benefits, however, are achieved at the expense of 6% added vehicles miles traveled.

Agent-Based Simulation↗

Experiment Assembly and Testing in Hot Cell in Support of Advanced Fuels Irradiation Testing

Follow-on experimentation of irradiated nuclear fuel is critical to understanding fuel behavior at different stages of the fuel lifecycle. Transient experiments using fuel with high burnup can enlarge understanding of fuel fragmentation, relocation and dispersal under various reactor accident situations. Assembling experiments with such fuel requires remote operations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Methods and systems for evaluating and improving distribution-grid observability

This document describes systems and techniques for evaluating and improving distribution-grid observability. These systems and techniques allow engineers to quantify the observability of a distribution grid, which represents an ability to combine actual measurements and various types of computations (e.g., analytics, estimators, forecasters), from a system model. Distribution engineers can also identify islands of observability where operating parameters, including voltages, currents, and power flows, can be determined from available sensor readings. By exclusion, distribution engineers can similarly identify areas of the distribution grid with observability deficiencies that may require additional instrumentation to maintain proper operation. Distribution engineers, using an iterative or automated process, can determine the observability of the system model with new or relocated sensors to generate a sensor allocation plan. The sensor allocation plan can indicate the number and location of sensors to either maximize observability for a fixed sensor cost or minimize sensor cost for predetermined observability.

Reiman, Anthony P.↗

RELAP5-3D British and SI Unit Conversion Task

The RELAP5-3D code uses conversion factors throughout the code to convert values between British and SI units. Conversion factors are typically in each of the separate routines that perform unit conversion. This results in inconsistent unit conversion and increases the likelihood of errors being introduced in the conversion factors. Unit conversion factors were relocated to a centralized module so that these values are only entered in one location. The unit conversion values were then replaced in the various locations with a unit conversion variable. The code was tested using the verification test set after making these changes and the results were analyzed. As expected, all the cases that use British input were found to have changed. In addition, three other cases changed due to changes in conversion factors.

97 MATHEMATICS AND COMPUTING↗

Combined TREAT-LOC & SATS Integral LOCA Experiment Plan

The Transient Reactor Test Facility (TREAT) loss-of-coolant (LOC) and highburnup (HBu) experiment series, along with the Severe Accident Test Station (SATS) HBu experiment series, are integral LOC accident (LOCA) experiments planned under the Department of Energy (DOE) Advanced Fuels Campaign (AFC) program, which aims to support burnup extension needs by addressing identified R&D priorities in order to achieve an improved understanding of fuel fragmentation, relocation, and dispersal (FFRD) of HBu fuel during LOCA events. Priorities have been identified by the Electric Power Research Institute (EPRI)’s Collaborative Research on Advanced Fuel Technologies (CRAFT) Fuel Performance and Testing Technical Experts Group (FPTTEG). The data produced under this plan will be used to further validate and confirm existing models and inform future R&D and model development. The experimental program was specifically developed to address data gaps and opportunities identified via detailed review of the existing public knowledge base on LOCA FFRD, as well as reviewing specific experimental development activities regarding prototypic LOCA conditions for light-water reactor (LWR) systems. The test program relies on a unique combination of in- and out-of-pile experimental approaches to provide a clear tieback to the existing integral and semi-integral LOCA experiment database, using state-of-the-art facilities. More importantly, the program will systematically investigate the impacts of prototypic HBu fuel/cladding thermomechanical behaviors under postulated LWR LOCA conditions not yet fully investigated. These conditions correspond with prototypic decay-energy heatup (DEH) and stored-energy heatup (SEH) conditions. First, TREAT’s unique capability will enable the first evaluation of the impact of SEH conditions on HBu fuels. The test program will emphasize the development of an improved mechanistic understanding of key phenomena through independent experimental systems, development of a database to support fuel performance modeling tools, world-leading advanced materials characterization, and the most advanced approach to in situ diagnostics ever deployed to evaluate FFRD. The results will represent a significant leap forward in evaluating prototypic conditions and novel data to support modeling development and validation, as well as to inform the technical basis for LOCA-induced FFRD.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

High Power Radiofrequency Operation of the Radiofrequency Quadrupoles in the Spallation Neutron Source

The Spallation Neutron Source (SNS) recently took delivery of a third Radiofrequency Quadrupole (RFQ03) that will ultimately be installed on the front-end (FE) of the SNS Linac. The first RFQ (RFQ01) operated in the SNS FE for more than a decade before being replaced with the second RFQ (RFQ02). RFQ01 was relocated to the Beam Test Facility (BTF) where it operated for five more years. The RFQ02 was initially installed in the BTF for high power testing and used with H- beam for BTF operation. It replaced RFQ01 in the SNS FE in 2017 and has been operating for beam production since then. There are some differences between the three RFQs. RFQ01 has a square cross-section with pi-mode stabilizing loops (PISLs) with the structure being fabricated using two layers of materials, GlidCop outside and OFHC inside. RFQ02 and RFQ03 has an octagonal cross-section with end-wall stabilizer rods and was fabricated using OFHC only. RFQ01 suffered some field flatness distortion incidents that resulted in degradation in beam transmission efficiency and required RF tuning. RFQ02 has performed well but had a melted RF seal in the high energy end wall, that was ultimately mitigated by a redesign of the end flange seals. The SNS decided to order RFQ03 that has a design that followed that of RFQ02 closely, but end-wall contacts were modified to prevent RF seal failure. This report presents the testing, installation, high power RF operation, and design improvements of the RFQ03.

Ren, Haitao↗

Burnt Mountain Alaska RTG Extraction, Transport and Disposal Project - 17359

The Air Force Technical Applications Center (AFTAC), headquartered at Patrick Air Force Base (AFB), Florida, removed excess nuclear material from one of the center's seismic arrays on Burnt Mountain, Alaska, after nearly 15 years of un-operability, and transferred it to a the low-level waste disposal facility at the Nevada National Security Site (NNSS). Ten radioisotope thermoelectric generators (RTGs), were used as a power source by AFTAC at its array 97 kilometers (60 miles) north of the Arctic Circle. Originally placed on Burnt Mountain in 1973, the first RTG was installed to power a single seismometer. Twelve years later, due to the increase in communications equipment and demand for additional power, nine more RTGs were placed at the site. For the next 17 years, the RTGs were operated and maintained by AFTAC's Detachment 460 based at Eielson AFB near Fairbanks, Alaska, and were used because of their high reliability and low maintenance requirements to determine if regional seismic activity was caused by nuclear explosions or naturally-occurring events such as earthquakes, volcanic activity, etc. AFTAC's primary mission is to verify compliance with nuclear test ban treaties. In 1992, a tundra fire encroached on the Burnt Mountain site and damaged some of the data cables. Despite the fact the fire did not impact the generators themselves, it raised public concern among nearby inhabitants about the safety of using radioactive material as a power source at the station. The Air Force, in coordination with multiple agencies and mission partners, made the decision to remove the RTGs and relocate them to NNSS for permanent disposal. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

The Port Hope Area Initiative's Port Granby Project: From Planning to Execution - 17451

The Port Hope Area Initiative's Port Granby Project involves the construction of a new, secure long-term waste management facility (LTWMF) and supporting infrastructure, and the excavation and relocation of some 450,000 m{sup 3} of LLRW and marginally contaminated soil from the former Port Granby Waste Management Facility (PGWMF) to the newly constructed LTWMF. Presented is an overview of the Port Granby Project and a description of the planning needed to execute the project successfully. A comprehensive suite of approximately 40 pre-construction submittals were developed and submitted for review and approval by Canadian Nuclear Laboratories (CNL), the project authority. Importantly, comprehensive training programs were developed and implemented to ensure all project staff fully understood the scope of the planning documents (covering such things as Quality Control, Environmental Monitoring, Radiation Protection, Dust Management, Noise, Radiation Protection etc.) and possess the appropriate competencies for completing the work to which they have been assigned. The effective planning that has been invested in the Port Granby project, from initial design to project planning and implementation, has allowed this complex project to proceed with minimal difficulty. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

The Legacy of Taiwan Lanyu LLW Interim Storage Site - 17491

The Lanyu (Orchid Island) LLW (Low Level Waste) interim storage facility for the Taiwan LLW is located on Lanyu Island off the coast of Taiwan (Republic of China). It is perhaps the only outdoor LLW interim storage facility in the world. The shipment of LLW to the Lanyu LLW interim storage facility began in May 1982 and ceased in June 1995 after strong opposition from the local aboriginal Yamei tribe people. Since the shipment of LLW to the Lanyu LLW storage facility ceased in 1995, the LLW in Taiwan has been stored temporarily at the indoor facilities at the Taipower (Taiwan Electric Power Company) three operating NPPs (nuclear power plants) and the INER (Institute of Nuclear Energy Research). The current Taiwan government has recently begun to lay out a multi-faceted plan to address the Lanyu LLW interim storage facility and evaluate different options of relocating Lanyu LLW to other possible sites. However, many challenges remain, notably the site selection for the LLW final disposal site. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Portable positional scanning lane

A positional lane scanner and method therefor includes a positional lane cargo scanning structure having a first frame portion, a second frame portion, and a top frame portion connected to define a passage lane for a cargo vessel. The first and second frame portions are located on opposite sides and cargo vessels pass therebetween. The first and second frame portions, or both, include a first housing portion and a scanning device mounted within the first housing portion. A lifting member is provided on the top frame portion facing outward, or if the lifting means is on a transport vehicle, such as a reach spreader, a pair of lifting beams may be arranged on the transport vehicle. The lift member removably connects with a boom portion of the transport vehicle for positioning and relocating the scanning structure to be disposed adjacent a cargo vessel lane.

Stihel, Gregory C.↗