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1Title:  Materials performance in operating PWR steam generators Add
 Summary:  This paper describes a challenge to the operation of pressurized water reactors on naval vessels: Steam generator U-tube leakage, primarily due to secondary chemistry problems. As described in the abstract, chemistry problems are centered in "those areas of the steam generators where limited coolant circulation and high heat flux have caused impurities to concentrate." Circulation problems (leading to cracking and corrosion) in Inconel U-tubes can be produced by "sludge deposits accumulated on the tube sheet or on tubing supports." In terms of prevention, the paper notes that "at the present time, all U.S. manufacturers of PWR's are recommending that their customers use an all-volatile treatment of the secondary coolant." It continues by providing water chemistry case studies on the three methods then used to maintain secondary chemistry: "A phosphate treatment, an all-volatile treatment, and a zero-solids treatment" (and the importance of moving from the first treatment method and attempting to reverse sludge problems in plants with extensive past use of phosphate treatments). Note: Some portions of the reproduced text are not legible. 
 Source:  http://www.osti.gov/bridge 
 Date:   1975 
 Subject(s):  Nuclear engineering | Naval Reactors 
 Type:  Text 
 Format:  PDF 
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2Title:  Corrosion and wear handbook Add
 Chapter title:  Introduction 
 Summary:  The introduction describes the handbook's purpose: "to accumulate and correlate the pertinent corrosion and wear information" that was the product of the first eight years in developing pressurized water reactor (PWR) technology for naval nuclear propulsion (3). The primary focus of the handbook is corrosion data related to the primary coolant system and steam generators in PWRs. The chapter provides a basic overview of PWR technology and emphasizes the importance of managing corrosion, noting that "only by closely controlling the amount of corrosion products in the primary system can this portion of the nuclear plant be made available for maintenance and repair within a reasonable period of time" after reactor shutdown (5). It includes summary information on stainless steel ("the major material of construction for water-cooled nuclear reactors") and carbon steel (5). 
 Source:  http://www.osti.gov/bridge 
 Date:   1957 
 Subject(s):  Nuclear engineering | Naval Reactors 
 Type:  Text 
 Format:  PDF 
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3Title:  Reactor compartment package characteristics for several submarine and surface plants Add
 Summary:  This graphic shows reactor compartment package characteristics for some submarine and surface ship reactor plants. After decommmissioning, the reactor plant(s) in a submarine or ship are removed and packaged for storage at the Puget Sound Naval Shipyard. The compartments are then shipped to and stored at the Hanford Site in Washington state. The primary system components housed inside the reactor compartment include: the reactor pressure vessel, reactor shielding, main coolant pumps, pressurizer system, and steam generators. 
 Source:  http://www.fas.org/man/dod-101/sys/ship/eng/reactor.html 
 Reference:  United States Department of the Navy. Draft environmental assessment on the disposal of decommissioned, defueled naval reactor plants from USS Enterprise (CVN-65). U.S. Department of the Navy, 2011, pages 2-2 - 2-5. 
 Date:   unknown  
 Subject(s):  Nuclear engineering | Naval Reactors 
 Type:  Image 
 Format:  GIF 
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4Title:  Simplified view of S8G naval nuclear propulsion plant Add
 Summary:  A simplified view of the S8G reactor used to power the Ohio-class Trident ballistic missile submarines. The S8G plant's two turbines provide 60,000 shp (thermal power, shaft horsepower), approaching twice the power produced by the S6G plant used to drive the Los Angeles-class attack submarines. Admiral Hyman Rickover, head of Naval Reactors when the Trident submarine was designed in the early 1970s, supported the 60,000 shp plant, which contributed to the submarine's large size (560 feet long, with a submerged displacement of 18,700 tons). 
 Source:  http://www.robse.dk/pages/SSBN/OhioFami.asp 
 Reference:  Polmar, Norman, and Thomas B. Allen. Rickover: Controversy and Genius, a Biography. New York: Simon and Schuster, 1984, pages 564-578. 
 Date:   unknown  
 Subject(s):  S8G | Nuclear engineering | Naval Reactors 
 Type:  Image 
 Format:  GIF 
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5Title:  Pressurized-water naval nuclear propulsion system Add
 Summary:  A simplified view of the major primary and secondary components in a naval nuclear propulsion plant. The fuel elements, containing Uranium-235 pellets, are enclosed in the reactor vessel. Pressurized water is used to moderate neutrons in the reactor core and serves as the heat transfer medium. Heated water moves to the steam generator, where the heat transfer takes place between the primary and secondary loops. The main coolant pump then returns the relatively cool water to the reactor core. The pressurizer enables primary loop pressure control through heaters (to increase pressure) and spray (to reduce pressure). The steam produced in the steam generator is used to drive turbines for propulsion and electrical power. 
 Source:  http://www.fas.org/man/dod-101/sys/ship/eng/reactor.html 
 Reference:  Hewlett, Richard G., and Francis Duncan. Nuclear Navy, 1946-1962. Chicago: University of Chicago Press, 1974, pages 131-135. 
 Date:   unknown  
 Subject(s):  Nuclear engineering | Naval Reactors 
 Type:  Image 
 Format:  JPEG 
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6Title:  Defueling the S2G reactor Add
 Summary:  This report describes the defueling of Seawolf's S2G reactor plant at Electric Boat in January 1959. This defueling was accomplished as part of the Seawolf's conversion from the sodium-cooled, intermediate range S2G reactor to a pressurized water reactor (PWR), owing to problems with the sodium-cooled design. These serious problems, which plagued the S1G (or Mark A) prototype and S2G shipboard plants, demonstrated the clear superiority of the PWR design in submarine propulsion. The report describes the importance of training (for Knolls Atomic Power Laboratory, Electric Boat, and Navy personnel who worked on the defueling) consisting of lectures and dry-runs that took place in the fall of 1958. The dry-runs enabled workers to check the condition of refueling equipment and time estimates for the completion of maintenance steps. (The summary on page 18 describes the importance of dry-runs and recommends some best practices for accomplishing them.) The dry-runs also contributed to the success in minimizing radiation exposure when the refueling was performed: "No individuals were exposed to more than the maximum permissible daily dose, 50 [millirem]" (3). The report provides an overview of the steps performed in defueling the sodium-cooled reactor. It also provides a summary of lessons learned, including: failure of a brazed joint in a cup designed to catch sodium drippage from fuel elements, which was identified during the dry run operation and fixed by using cups with welded joints; and, gas leakage from a transfer cask. Also, there was a report of difficulty in grappling an S2G fuel rod that was being removed, due to wear in the grappling equipment. After completion of the refueling, the S2G's fuel rods were shipped via train to the Idaho National Laboratory's Expended Core Facility. 
 Source:  http://www.osti.gov/bridge 
 Date:   1959 
 Subject(s):  S2G | Nuclear engineering | Naval Reactors 
 Type:  Text 
 Format:  PDF 
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7Title:  TMI-2 Lessons Learned Task Force: Status report and short-term recommendations Add
 Summary:  This document, known as NUREG-0578, was created by the Lessons Learned Task Force, an interdisciplinary group formed by the Nuclear Regulatory Commission in the aftermath of the Three Mile Island (TMI-2) accident, which occurred on 28 March 1979. Of particular interest is the section on short-term recommendations, in which the task force proposes changes to operating procedures given the circumstances of the TMI-2 accident (a loss of feed in the secondary system, followed by a loss of coolant accident [LOCA] in the primary system of the pressurized water reactor, with resulting core damage). Several recommendations stand out. First, providing emergency power for critical services, such as pressurizer level indicator, pressurizer heaters, and power-operated control values. Second, performing periodic checking of primary system safety and relief valves. Third, and critically, ensuring that operators are trained to better diagnose "low reactor coolant level and inadequate core cooling using existing reactor instrumentation (flow, temperature, power, etc.)" (8). While the recommendations as a whole are focused on commercial power reactor plants, many of these operational recommendations are applicable to the pressurized water reactors operated in the Navy's submarine and surface force. 
 Source:  http://www.osti.gov/bridge 
 Date:   1979 
 Subject(s):  Reactor safety | Nuclear engineering | Naval Reactors 
 Type:  Text 
 Format:  PDF 
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8Title:  Pressure vessel and piping codes applicable to the PWR reactor plant Add
 Summary:  This document provides information on standards compliance for the pressurized water reactor (PWR) installed in the Shippingport Atomic Power Station at the time of publication. The ASME standard, Boiler and Pressure Vessel Code, sections I and VIII, are referenced in the compliance summary, which maps code compliance to specific areas and components of the reactor plant. 
 Source:  http://www.osti.gov/bridge 
 Date:   1957 
 Subject(s):  Shippingport Atomic Power Station | Nuclear engineering | Naval Reactors 
 Type:  Text 
 Format:  PDF 
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9Title:  An evaluation of data on zirconium-uranium alloys Add
 Summary:  This document, compiled by Frank Rough of the Battelle Memorial Institute, contains a review of information on zirconium-uranium alloys. As noted in the introduction, "because of the similar properties and fabricational characteristics of these materials, the cladding of zirconium-uranium alloys with Zircaloy has proven to be very successful, with good metallurgical bonds being obtained" (7). This review addresses issues such as the corrosion of zirconium-uranium alloys in high temperature/high pressure systems and the impact of neutron irradiation upon these alloys. These and other issues are addressed and mapped to an extensive bibliography. As described by historians Thomas Hewlett and Francis Duncan in their book Nuclear Navy, Naval Reactors was deeply involved in the development of zirconium production in the United States, with the need to produce tonnage lots of zirconium to support early prototype and submarine reactor core construction. Beyond this, improvements in the technology were needed, such as the development of Zircaloy-2, a material superior to the zirconium-uranium alloy used in the first Mark I/S1W core. 
 Source:  http://www.osti.gov/bridge 
 Date:   1955 
 Subject(s):  Zirconium/Zircaloy | Nuclear engineering | Naval Reactors 
 Type:  Text 
 Format:  PDF 
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10Title:  Mechanical properties of Zircaloy-2 Add
 Summary:  To summarize: "Zircaloy-2 is a zirconium-tin alloy developed for use in water cooled nuclear reactors. It possesses good corrosion resistance to high-temperature water, excellent nuclear characteristics, and sufficiently good mechanical properties for use as a structural material in reactor cores and as a fuel element material" (1). The report analyzes changes in Zircaloy-2 properties caused by changes in operating conditions, including temperature, hydrogen concentration, and the presence of small notches in the material. As noted in the Hewlett/Duncan book, Nuclear Navy, "the study of zirconium alloys [in the first half of the 1950s] resulted in the development of a new material called Zircaloy-2, which was far superior to the material used in the [Mark I/S1W] core." 
 Source:  http://www.osti.gov/bridge 
 Date:   1961 
 Subject(s):  Zirconium/Zircaloy | Nuclear engineering | Naval Reactors 
 Type:  Text 
 Format:  PDF 
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