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Title: Preliminary Study on Utilization of Carbon Dioxide as a Coolant of High Temperature Engineering Test Reactor with MOX and Minor Actinides Fuel

Abstract

High temperature engineering test reactor (HTTR) is an uranium oxide (UO2) fuel, graphite moderator and helium gas-cooled reactor with 30 MW in thermal output and outlet coolant temperature of 950 deg. C. Instead of using helium gas, we have utilized carbon dioxide as a coolant in the present study. Beside that, uranium and plutonium oxide (mixed oxide, MOX) and minor actinides have been employed as a new fuel type of HTTR. Utilization of plutonium and minor actinide is one of the support system to non-proliferation issue in the nuclear development. The enrichment for uranium oxide has been varied of 6-20% with plutonium and minor actinides concentration of 10%. In this study, burnup period is 1100 days. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. Reactor core calculation was done by using CITATION module. The result shows that HTTR can achieve its criticality condition with 14% of {sup 235}U enrichment.

Authors:
; ;  [1]
  1. Bosscha Laboratory, Department of Physics, Nuclear Physics and Biophysics Research Group, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, INDONESIA Jl. Ganesa 10 Bandung 40132 (Indonesia)
Publication Date:
OSTI Identifier:
21410478
Resource Type:
Journal Article
Journal Name:
AIP Conference Proceedings
Additional Journal Information:
Journal Volume: 1244; Journal Issue: 1; Conference: ICANSE 2009: 2. international conference on advances in nuclear science and engineering 2009, Bandung (Indonesia), 3-4 Nov 2009; Other Information: DOI: 10.1063/1.3462747; (c) 2010 American Institute of Physics; Journal ID: ISSN 0094-243X
Country of Publication:
United States
Language:
English
Subject:
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; CARBON DIOXIDE; COMPUTERIZED SIMULATION; COOLANTS; CRITICALITY; ENRICHMENT; FISSION; GRAPHITE; HELIUM; HTTR REACTOR; ISOTOPE SEPARATION; NUCLEAR DATA COLLECTIONS; NUCLEAR FUELS; PLUTONIUM OXIDES; PROLIFERATION; REACTOR CELLS; REACTOR CORES; REACTOR FUELING; URANIUM 235; URANIUM DIOXIDE; ACTINIDE COMPOUNDS; ACTINIDE NUCLEI; ALPHA DECAY RADIOISOTOPES; CARBON; CARBON COMPOUNDS; CARBON OXIDES; CHALCOGENIDES; ELEMENTS; ENERGY SOURCES; ENRICHED URANIUM REACTORS; EVEN-ODD NUCLEI; EXPERIMENTAL REACTORS; FLUIDS; FUELS; GAS COOLED REACTORS; GASES; GRAPHITE MODERATED REACTORS; HEAVY NUCLEI; HELIUM COOLED REACTORS; HTGR TYPE REACTORS; INTERNAL CONVERSION RADIOISOTOPES; ISOMERIC TRANSITION ISOTOPES; ISOTOPES; MATERIALS; MINERALS; MINUTES LIVING RADIOISOTOPES; NONMETALS; NUCLEAR REACTIONS; NUCLEI; OXIDES; OXYGEN COMPOUNDS; PLUTONIUM COMPOUNDS; RADIOISOTOPES; RARE GASES; REACTOR COMPONENTS; REACTOR MATERIALS; REACTORS; RESEARCH AND TEST REACTORS; SEPARATION PROCESSES; SIMULATION; SPONTANEOUS FISSION RADIOISOTOPES; TRANSURANIUM COMPOUNDS; URANIUM COMPOUNDS; URANIUM ISOTOPES; URANIUM OXIDES; YEARS LIVING RADIOISOTOPES

Citation Formats

Fauzia, A F, Waris, A, and Novitrian,. Preliminary Study on Utilization of Carbon Dioxide as a Coolant of High Temperature Engineering Test Reactor with MOX and Minor Actinides Fuel. United States: N. p., 2010. Web. doi:10.1063/1.3462747.
Fauzia, A F, Waris, A, & Novitrian,. Preliminary Study on Utilization of Carbon Dioxide as a Coolant of High Temperature Engineering Test Reactor with MOX and Minor Actinides Fuel. United States. https://doi.org/10.1063/1.3462747
Fauzia, A F, Waris, A, and Novitrian,. 2010. "Preliminary Study on Utilization of Carbon Dioxide as a Coolant of High Temperature Engineering Test Reactor with MOX and Minor Actinides Fuel". United States. https://doi.org/10.1063/1.3462747.
@article{osti_21410478,
title = {Preliminary Study on Utilization of Carbon Dioxide as a Coolant of High Temperature Engineering Test Reactor with MOX and Minor Actinides Fuel},
author = {Fauzia, A F and Waris, A and Novitrian,},
abstractNote = {High temperature engineering test reactor (HTTR) is an uranium oxide (UO2) fuel, graphite moderator and helium gas-cooled reactor with 30 MW in thermal output and outlet coolant temperature of 950 deg. C. Instead of using helium gas, we have utilized carbon dioxide as a coolant in the present study. Beside that, uranium and plutonium oxide (mixed oxide, MOX) and minor actinides have been employed as a new fuel type of HTTR. Utilization of plutonium and minor actinide is one of the support system to non-proliferation issue in the nuclear development. The enrichment for uranium oxide has been varied of 6-20% with plutonium and minor actinides concentration of 10%. In this study, burnup period is 1100 days. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. Reactor core calculation was done by using CITATION module. The result shows that HTTR can achieve its criticality condition with 14% of {sup 235}U enrichment.},
doi = {10.1063/1.3462747},
url = {https://www.osti.gov/biblio/21410478}, journal = {AIP Conference Proceedings},
issn = {0094-243X},
number = 1,
volume = 1244,
place = {United States},
year = {Tue Jun 22 00:00:00 EDT 2010},
month = {Tue Jun 22 00:00:00 EDT 2010}
}