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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:  Project Prometheus Reactor Module final report Add
 Summary:  This report describes the work of the Naval Reactors program's contractor laboratories - the Knolls Atomic Power Laboratory and the Bettis Atomic Power Laboratory - in NASA's Project Prometheus. Naval Reactors worked as NASA's partner in the design of a civilian space reactor for a 15-20 year mission. The first identified mission was the Jupiter Icy Moons Orbiter (JIMO). Five reactor designs were evaluated by Naval Reactors, with the direct gas Brayton plant deemed the most promising design. NR investigated a number of issues, including plant design, instrumentation and control, and core and plant materials. In performing this research, Naval Reactors determined that the existing state of reactor technology did not support the creation of a plant that would enable mission goals to be met. The Prometheus project was ended in 2005. 
 Source:  http://www.osti.gov/bridge/purl.cover.jsp?purl=/884680-LsvaFN/ 
 Date:   2006 
 Subject(s):  Project Prometheus | Nuclear engineering | Naval Reactors 
 Type:  Text 
 Format:  PDF 
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4Title:  Request for Naval Reactors comment on Project Prometheus Add
 Summary:  A formal request, from the Naval Reactors Prime Contractor Team (NRPCT) to Naval Reactors requesting comment on the reactor safety requirements for NASA's Project Prometheus. Project Prometheus was created in 2003 to design reactors for long-duration space missions. The NRPCT, requesting the review, included engineers from Lockeed Martin, the Knolls Atomic Power Laboratory, and the Bettis Atomic Power Laboratory (the latter two being long-time contractors for Naval Reactors). The letter describes the team's primary goal: "Consistent with Naval Reactors program philosophy...no undue risk to the health and safety of workers, the public, or adverse effects to the environment should result from activities associated with the nuclear reactor of the Prometheus project." 
 Source:  http://www.osti.gov/bridge/product.biblio.jsp?osti_id=883434 
 Date:  28 April 2005 
 Subject(s):  Project Prometheus | Nuclear engineering | Naval Reactors 
 Type:  Text 
 Format:  PDF 
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5Title:  Naval Reactors Prime Contractor Team (NRPCT) Experiences and considerations with irradiation test performance in an international environment Add
 Summary:  This document describes the NRPCT's efforts to identify reactors worldwide for irradiation testing of materials expected to be included in a Prometheus reactor. The Experimental Fast Reactor JOYO in O-arai, Japan was identified as the best facility to support irradiation testing for the project (which was created to support space reactor development for solar exploration; the project ended in 2005). JOYO is a sodium-cooled Liquid Metal Reactor (LMR). Detailed planning information for reactor materials irradiation testing is included in the report. 
 Source:  http://www.osti.gov/bridge/purl.cover.jsp?purl=/883694-B0koQW/ 
 Date:  15 February 2006 
 Subject(s):  Project Prometheus | Nuclear engineering | Naval Reactors 
 Type:  Text 
 Format:  PDF 
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6Title:  Naval reactors physics handbook. Volume 1, Selected basic techniques Add
 Chapter title:  "Reactor physics and its application to nuclear power reactors" 
 Summary:  This chapter, written by physicist Alvin Radkowsky, summarizes the design challenges of submarine reactors in comparison with the natural uranium graphite reactors that had been designed and built during World War II. For example, Radkowsky describes the novelty ("close spacing") and design complexity of the control rod arrangement in PWRs (4). He also describes the parallel track of reactor development overseen by NR, with the intermediate range research supporting the S1G and S2G reactors performed by the Knolls Atomic Power Laboratory; and, research supporting the Submarine Thermal Reactor (STR, or the S1W and S2W reactors) led by the Bettis Atomic Power Laboratory. He notes that while the intermediate range reactor approach had, by 1964, been abandoned in favor of the pressurized water reactor (PWR) design, that "fuel loading densities are often sufficiently high [so] that a substantial fraction of the fissions occurs above thermal neutron energies" (2). As a result, some research relating to the intermediate range reactor could be applied to the design of PWRs. Radkowsky also summarizes some design contrasts between submarine reactors and the reactors for the Shippingport Atomic Power Station, with the latter relying on fuels with high U-238 composition. 
 Source:  http://www.osti.gov/bridge 
 Date:   1964 
 Subject(s):  Reactor physics | Nuclear engineering | Naval Reactors 
 Type:  Text 
 Format:  PDF 
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7Title:  Reactor shielding design manual Add
 Chapter title:  "Introduction and outline of basic shielding theory" 
 Summary:  This manual, edited by Theodore Rockwell of the early Naval Reactors group, is designed to provide an engineering overview of shielding design issues. The introductory chapter provides a pathfinder for the manual as a whole. Rockwell defines "shield engineering" as "the art of [lowering radiation levels] within specified limits of weight, volume, or cost" (4). He notes that neutron and gamma radiation are the primary focus of shield design and describes methods (such as the use of specific materials and shield compositions) used to achieve neutron and gamma-ray attenuation. 
 Source:  http://www.osti.gov/bridge 
 Date:   1956 
 Subject(s):  Reactor shielding | Nuclear engineering | Naval Reactors 
 Type:  Text 
 Format:  PDF 
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8Title:  Documentation of Naval Reactors papers and presentations for the Space Technology and Applications International Forum (STAIF) 2006 Add
 Summary:  This document contains information on the presentations and papers (24 in all) prepared by the Knolls and Bettis Atomic Power laboratories for the Space Technology and Applications International Forum (STAIF) 2006 conference. These presentations describe the work of Naval Reactors and its contractor laboratories, Bettis and Knolls, for NASA's Project Prometheus, which was created to investigate the possible use of nuclear-powered systems for long duration space missions. At the time of the project, Naval Reactors was designated by the Department of Energy as the lead agency for the development of civilian space reactor systems. NR engaged the two contractor laboratories to investigate issues related to deep space reactors. The presentations cover topics such as reactor design, reactor instrumentation, and plant materials. 
 Source:  http://www.osti.gov/bridge 
 Date:   2006 
 Subject(s):  Project Prometheus | Nuclear engineering | Naval Reactors 
 Type:  Text 
 Format:  PDF 
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9Title:  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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10Title:  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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