skip to main content
OSTI.GOV title logo U.S. Department of Energy
Office of Scientific and Technical Information

Title: Assessment of deep burnup concept based on graphite moderated.


No abstract prepared.

; ; ; ; ; ;
Publication Date:
Research Org.:
Argonne National Lab. (ANL), Argonne, IL (United States)
Sponsoring Org.:
OSTI Identifier:
Report Number(s):
TRN: US0900500
DOE Contract Number:
Resource Type:
Resource Relation:
Conference: PHYSOR 2006 - American Nuclear Society Topical Meeting on Reactor Physics; Sep. 10, 2006 - Sep. 14, 2006; Vancouver, Canada
Country of Publication:
United States

Citation Formats

Kim, T. K., Taiwo, T. A., Yang, W. S., Hill, R. N., Venneri, F., Nuclear Engineering Division, and General Atomics. Assessment of deep burnup concept based on graphite moderated.. United States: N. p., 2006. Web.
Kim, T. K., Taiwo, T. A., Yang, W. S., Hill, R. N., Venneri, F., Nuclear Engineering Division, & General Atomics. Assessment of deep burnup concept based on graphite moderated.. United States.
Kim, T. K., Taiwo, T. A., Yang, W. S., Hill, R. N., Venneri, F., Nuclear Engineering Division, and General Atomics. Sun . "Assessment of deep burnup concept based on graphite moderated.". United States. doi:.
title = {Assessment of deep burnup concept based on graphite moderated.},
author = {Kim, T. K. and Taiwo, T. A. and Yang, W. S. and Hill, R. N. and Venneri, F. and Nuclear Engineering Division and General Atomics},
abstractNote = {No abstract prepared.},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Sun Jan 01 00:00:00 EST 2006},
month = {Sun Jan 01 00:00:00 EST 2006}

Other availability
Please see Document Availability for additional information on obtaining the full-text document. Library patrons may search WorldCat to identify libraries that hold this conference proceeding.

Save / Share:
  • A systematic assessment of the General Atomics (GA) proposed one-pass and two-pass deep-burn concepts based on the modular helium-cooled reactor design (DB-MHR) using non-uranium fuel has been performed. Sensitivity studies are done to investigate the impact of core design parameters and concept on the transmutation performance (maximum of 60% destruction). The repository loading benefits arising from the DB-MHR and LWR Inert Matrix Fuel (IMF) concepts are also estimated and compared ({approx}2.0 and 1.6, respectively). (authors)
  • The United States operated a number of graphite moderated, water cooled, pressure tube (or channel type) reactors at the Hanford Site in Washington State. The last of these reactors to operate was the Hanford N Reactor which was designed to produce steam for the generation of electricity for civilian energy use as well as special nuclear material. In 1982, a review team recommended that a probabilistic risk assessment (PRA) be conducted for the reactor. The basis for this recommendation was that the reactor represented a unique facility and therefore the safety insights gained from operations at other facilities were notmore » directly transferable to it. Furthermore, the reactor was going through a series of upgrades and it was felt that the selection of competing improvements would benefit from a risk perspective to help judge the relative value of these upgrades. The initial Level 1 PRA was initiated just prior to the time of the Chernobyl accident in 1986. After the accident, a limited scope PRA was also conducted. The limited scope PRA was completed in July 1987. The original Level 1 PRA was brought to completion in August 1988 and was followed by complete Level 3 analyses including an evaluation of external initiating events. This paper will focus on the Level 1 effort. The objectives of the Level 1 PRA for N Reactor were: identification of the plant features that are the most important contributors to fuel-damage frequency; development of plant logic models for use in analyzing safety and reliability issues; identification of dominant accident sequences and their frequencies for input into the Level 2/3 PRA and external events analyses; and establishment of a Hanford team with demonstrable expertise in performing risk and reliability analyses.« less
  • Design studies, based primarily on neutronics analysis, have been conducted on a thermionic reactor concept that uses a combined beryllium and zirconium hydride moderator to facilitate the incorporation of heat pipe cooling into compact thermionic fuel element (TFE) based designs useful in the tens of kilowatts electrical power regime. The goal of the design approach is to achieve a single point failure free system with technologies such as TFEs, high-temperature heat pipes, and ZrH moderation, which have extensive test data bases and have been shown to be capable of long lifetimes. Beryllium is used to thermally couple redundant heat pipesmore » to TFEs and ZrH is added to reduce critical size. Neutronic analysis undertaken to investigate this design approach shows that greater reactivity can be achieved for a given geometry with a combination of the two moderator materials than with ZrH alone and that the combined moderator is much less sensitive to hydrogen loss than more traditional ZrH-moderated thermionic reactor designs. These and other analytical approaches have demonstrated the credibility of a heat pipe cooled thermionic reactor concept that has a reactor height and diameter of 60 cm and a reactor mass of 400 kg for 30-kWe power output. 14 refs., 8 figs.« less