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    <copyright>Copyright © 2026. National Academy of Sciences. All rights reserved.</copyright>
    <docs>http://blogs.law.harvard.edu/tech/rss</docs>
    <managingEditor>tris-trb@nas.edu (Bill McLeod)</managingEditor>
    <webMaster>tris-trb@nas.edu (Bill McLeod)</webMaster>
    <image>
      <title>Transport Research International Documentation (TRID)</title>
      <url>https://trid.trb.org/Images/PageHeader-wTitle.jpg</url>
      <link>https://trid.trb.org/</link>
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    <item>
      <title>NUCLEAR SHIP RESEARCH</title>
      <link>https://trid.trb.org/View/399957</link>
      <description><![CDATA[This literature review contains some 600 single-page summaries of investigations on various aspects of nuclear-powered ships. In addition to the summary of the research, bibliographical data, supplementary notes, and subject terms are provided for each entry.]]></description>
      <pubDate>Thu, 21 Jul 1994 00:00:00 GMT</pubDate>
      <guid>https://trid.trb.org/View/399957</guid>
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    <item>
      <title>MARITIME FUEL ASSEMBLY MECHANICAL AND HYDRAULIC EVALUATION</title>
      <link>https://trid.trb.org/View/400142</link>
      <description><![CDATA[This report describes the results of mechanical and hydraulic evaluations of the fuel assembly of a nuclear reactor core designed for maritime applications. Tests were carried out to assess the following  aspects of fuel assembly operation: natural frequency and damping; torsional and cantilever bending; dynamic and static deflection; pressure drop and lift-off; flow induced vibration; relative motion between fuel rod and spacer grid; and spacer grid cyclic strength.]]></description>
      <pubDate>Thu, 21 Jul 1994 00:00:00 GMT</pubDate>
      <guid>https://trid.trb.org/View/400142</guid>
    </item>
    <item>
      <title>DYNAMICS OF THE INTEGRATED SELF-PRESSURIZED WATER REACTOR AFTER FAILURE IN THE CONTROL SYSTEM</title>
      <link>https://trid.trb.org/View/146360</link>
      <description><![CDATA[The entire primary system with core, steam-generator, primary coolant pumps and an integrated-type pressurizer is located in the reactor pressure vessel.  Simulation results show the transient behaviour of relevant system-variables during demanding applications, such as ice-ramming manoeuvres.  Order from NSFI as No. 17886.]]></description>
      <pubDate>Wed, 19 Dec 1979 00:00:00 GMT</pubDate>
      <guid>https://trid.trb.org/View/146360</guid>
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    <item>
      <title>STEADY STATE HEAT TRANSFER ANALYSIS OF MOLTEN CORE-REACTOR VESSEL SYSTEM</title>
      <link>https://trid.trb.org/View/72381</link>
      <description><![CDATA[This paper extends the steady state analysis of molten core and reactor vessel systems by L.S. Tong.  Three other cases of interest were considered and found to be significant for the Maritime CNSG.  Due to the design differences between the CNSG and land-based nuclear plants, there is a greater possibility that a postulated molten core could be contained within the reactor vessel with water only outside of the vessel.]]></description>
      <pubDate>Wed, 26 Apr 1978 00:00:00 GMT</pubDate>
      <guid>https://trid.trb.org/View/72381</guid>
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    <item>
      <title>INVESTIGATIONS INTO THE TECHNICAL AND ECONOMIC ADVANTAGES AND DISADVANTAGES OF THE "POISON CORE" FUEL ROD CONCEPT USING A ROUGH DESIGN OF A REFERENCE CORE BASED ON THE EFDR-50</title>
      <link>https://trid.trb.org/View/57891</link>
      <description><![CDATA[Fuel rods with a central poison core contain a core of boran graphite besides the fuel in form of UO sub 2 ring tablets. The boron serves here as solid burnable neutron poison to compensate for the burn-up excess reactivity. A reactor core constructed of such fuel rods has a particularly uniform power and burn-up distribution at relatively high specific fuel power. Besides this, it exhibits safety technical advantages, in particular the boric acid use and thus the tritium production in the primary circuit can be reduced with this concept in pressurized-water reactors. The present report contains the results of the lattice studies, a rough design of a reference core, as well as fuel circuit cost estimates. (Atomindex citation 08:307163)]]></description>
      <pubDate>Tue, 14 Mar 1978 00:00:00 GMT</pubDate>
      <guid>https://trid.trb.org/View/57891</guid>
    </item>
    <item>
      <title>EXPERIENCES DURING THE EXCHANGE OF FUEL ELEMENTS OF THE NS OTTO HAHN</title>
      <link>https://trid.trb.org/View/54701</link>
      <description><![CDATA[The step by step procedure to remove the fuel elements after 241,588 nautical miles and 25,230 hrs of operation is described.  Significant changes and characteristics of the second core are listed, and the refuelling procedure is discussed.]]></description>
      <pubDate>Sat, 29 Oct 1977 00:00:00 GMT</pubDate>
      <guid>https://trid.trb.org/View/54701</guid>
    </item>
    <item>
      <title>ANTICIPATED CORE LIFE - A MERCHANT SHIP REACTOR</title>
      <link>https://trid.trb.org/View/54740</link>
      <description><![CDATA[The optimum core life for minimum operating cost is discussed in terms of the following four factors: 1) Intervals between each ship inspection and survey including reactor inspection, 2) Availability of refuelling and related equipment, 3) Capacity factor of reactor determined by ship service plan and, 4) Technical restriction influences by core design parameters such as core power density and reactivity control.  Results from four studies are summarized.]]></description>
      <pubDate>Thu, 13 Oct 1977 00:00:00 GMT</pubDate>
      <guid>https://trid.trb.org/View/54740</guid>
    </item>
    <item>
      <title>NS SAVANNAH CORE 2, PRODUCT QUALITY REPORT</title>
      <link>https://trid.trb.org/View/3094</link>
      <description><![CDATA[Principal product characteristics of the N.S. Savannah Core 2 fuel are presented graphically.  Data are based on measurements and analyses made throughout the manufacturing process.  Data sheets show a frequency distribution and average value compared to design limits.]]></description>
      <pubDate>Fri, 17 Jun 1977 00:00:00 GMT</pubDate>
      <guid>https://trid.trb.org/View/3094</guid>
    </item>
    <item>
      <title>NS SAVANNAH FUEL DESIGN AND DEVELOPMENT PROGRAM. NS SAVANNAH PROTOTYPE FUEL ASSEMBLY TENSILE AND COMPRESSIVE LOADING TESTS</title>
      <link>https://trid.trb.org/View/2948</link>
      <description><![CDATA[The results of tensile and compressive tests performed at room temperature on a dummy N.S. Savannah prototype Core 2 fuel assembly in a pre-irradiated condition showed conclusively that the assembly will withstand the two-thousand pound force associated with the refueling process of the assembly.  The test equipment and layout and the test procedure and data are provided.]]></description>
      <pubDate>Fri, 17 Jun 1977 00:00:00 GMT</pubDate>
      <guid>https://trid.trb.org/View/2948</guid>
    </item>
    <item>
      <title>THE ECONOMICS OF NUCLEAR FUEL IN MARITIME APPLICATIONS</title>
      <link>https://trid.trb.org/View/20</link>
      <description><![CDATA[The report presents and updates the results of a detailed investigation ( see PB-169935 ) of the economics of nuclear fuel as a source of energy for merchant ships.  The basic uncertainties, dependences, magnitudes and trends of fuel cost elements and trends with a number of central station water reactor applications are established. ( Author )]]></description>
      <pubDate>Tue, 30 Oct 1973 00:00:00 GMT</pubDate>
      <guid>https://trid.trb.org/View/20</guid>
    </item>
    <item>
      <title>PHOENIX FUEL EVALUATION IN A MARITIME REACTOR DESIGN</title>
      <link>https://trid.trb.org/View/30</link>
      <description><![CDATA[The Consolidated Nuclear Steam Generator - II reactor concept was used as a reference design to compare fuel cycle costs for Pu fueled cores with a high 240Pu content ( Phoenix fuel ) to costs for an enriched U base case core.  The Phoenix fuel was found to be competitive with the slightly enriched U reference fuel.  Results are presented in graphical and tabular form, and the results and calculational methods are discussed. ( H.D.R. )]]></description>
      <pubDate>Tue, 30 Oct 1973 00:00:00 GMT</pubDate>
      <guid>https://trid.trb.org/View/30</guid>
    </item>
    <item>
      <title>ECCSA-1: A DIGITAL COMPUTER PROGRAM FOR THERMAL AND HYDRAULIC ANALYSIS OF CORE CHANNELS IN THE EVENT OF A NUCLEAR-REACTOR LOSS-OF-COOLANT ACCIDENT</title>
      <link>https://trid.trb.org/View/40</link>
      <description><![CDATA[ECCSA-I is a digital computer program specifically designed to predict the thermal and hydraulic performance of reactor core channels during a loss-of-coolant accident and subsequent emergency core cooling injection.  The program represents reactor fuel rods and associated flow channels by means of an equivalent single channel.  Thermal and hydraulic behavior in the channel is predicted on a local basis at various axial positions during the blowdown, core heat-up, and emergency core cooling injection phases of an accident.  Blowdown pressure and flow information are used as input data for ECCSA-I.  The development and content of the program is described, including major assumptions, mathematical models, and computer equation.  A source deck and sample problem for the CDC-6400 computer will be available through the ANL Code Center.  Typical results are presented for analyses of BWR, PWR, and LOFT loss-of-coolant accidents.  ( auth )]]></description>
      <pubDate>Tue, 30 Oct 1973 00:00:00 GMT</pubDate>
      <guid>https://trid.trb.org/View/40</guid>
    </item>
    <item>
      <title>A NUCLEAR PARTICLE PRIMER</title>
      <link>https://trid.trb.org/View/49</link>
      <description><![CDATA[A survey of aerosol science is presented which has the objective of providing a short list of phenomena which could account ( to one significant figure ) for the aerosol concentration in a water-cooled reactor containment as a function of time after a very severe accident.  Three categories of aerosol science are discussed: formation of particles, mechanics of particles, and surface chemistry of particles.  Particles are expected to form primarily by condensation of vapor from an overheated core.  There appears to be sufficient theory available to estimate the effects of core surface temperature and sweeping gas temperature and velocity on the primary particle size. Condensation of water on the filter is probably the most serious potential problem.  Particle removal by natural effects in the containment will occur as a result of many phenomena.  The most important are probably coagulation and settling, and diffusiophoresis and thermophoresis ( which sweep particles toward surfaces onto which water is condensing ).  Water can condense in quantity on particles, make them larger and make them settle faster.  Estimates of the effect of this phenomena on concentration in the containment vs time are presented, the effect can probably be very important.  34 references. ( auth )]]></description>
      <pubDate>Tue, 30 Oct 1973 00:00:00 GMT</pubDate>
      <guid>https://trid.trb.org/View/49</guid>
    </item>
    <item>
      <title>REACTOR PHYSICS ON THE N.S. SAVANNAH CORE</title>
      <link>https://trid.trb.org/View/3504</link>
      <description><![CDATA[This paper describes the nuclear model and the methods of analysis used in the nuclear design of the reactor core for the NS Savannah.]]></description>
      <pubDate>Fri, 24 Aug 1973 00:00:00 GMT</pubDate>
      <guid>https://trid.trb.org/View/3504</guid>
    </item>
    <item>
      <title>NUCLEAR MERCHANT SHIP REACTOR PROJECT EXTENDED ZERO POWER TESTS, NS SAVANNAH CORE 1. FINAL REPORT. VOLUME 1</title>
      <link>https://trid.trb.org/View/3814</link>
      <description><![CDATA[Experiments were performed on the N.S. Savannah Service Core 1 to supply information on future shipboard operation of the Nuclear Merchant Ship Reactor. Testing of instrumentation equipment shipped to The Babcock and Wilcox Company from New York Shipbuilding Corporation was satisfactory. Correction factors for shipboard use were determined. Fast neutron flux, neutron flux distribution, and stuck control rod studies were carried out successfully, as was a three-dimensional calculation (TKO) to match two critical rod patterns.]]></description>
      <pubDate>Mon, 09 Jul 1973 00:00:00 GMT</pubDate>
      <guid>https://trid.trb.org/View/3814</guid>
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