Uncertainty quantification of light-water-moderated and light-water-reflected cores with highly-enriched uranium fuel at Kyoto University Critical Assembly

Over its 1968–1988 life, PSCo relicensed the Fort St. Vrain (FSV) High-temperature Gas Reactor (HTGR) for light water reactor (LWR) technology requirements. Estimates of the financial losses associated with the plant range from $500 million to $2 billion in 1980 dollars. Colorado ratepayers, the shareholders of Gulf General Atomics and its corporate successors — General Atomics, GA Technologies or just GA and Public Service Company of Colorado (PSCo) bore these losses. Two critical plant issues required solution for the plant’s economic success — (1) the high-cost of 93% enriched uranium fuel and (2) low unit availability. While fuel costs were beyond utility control, low availability was controllable, yet remained unresolved. Commercially isolated for twenty years, PSCo shut the plant down in 1988. Economic success of future HTGRs depends upon avoiding similar complications. This paper examines the issues that made FSV uneconomic, including those fundamental to HTGR technology and others attributable to the utility operator and its culture. Knowing the history of FSV and HTGR design, designers should anticipate reasonable challenges. Preparations will help manage future HTGR risks, costs, and assure operating success. Regulators and industry can assure more effective, economic operations in the next round of HTGR designs.

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Temperature Coefficient of Reactivity in Light-Water Moderated and Reflected Cores Loaded with Highly-Enriched-Uranium Fuel

Journal of Nuclear Science and Technology ◽

10.1080/18811248.1987.9735861 ◽

1987 ◽

Vol 24 (8) ◽

pp. 653-667 ◽

Cited By ~ 7

Author(s):

Masaaki MORI ◽

Seiji SHIROYA ◽

Keiji KANDA

Keyword(s):

Temperature Coefficient ◽

Light Water ◽

Uranium Fuel ◽

Enriched Uranium ◽

Coefficient Of Reactivity

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Conceptual design of critical assembly using Low enriched uranium fuel and moderated light wate

Nuclear Science and Technology ◽

10.53747/jnst.v6i1.142 ◽

2021 ◽

Vol 6 (1) ◽

pp. 14-31

Author(s):

Kien Cuong Nguyenn ◽

Hai Dang Vo Doan

Keyword(s):

Conceptual Design ◽

Nuclear Power ◽

Nuclear Data ◽

Critical Assembly ◽

Design Calculation ◽

Reactor Physics ◽

Uranium Fuel ◽

Calculation Results ◽

Enriched Uranium ◽

Accident Conditions

Critical assembly is a very important facility to serve for fundamental reactor physics research, application of neutron source, training and education. In nuclear engineering, critical assembly is a facility for carrying out measurement of reactor physics parameters, creating benchmark problem, validation of neutron physics calculation tool in computer codes and nuclear data. Basing on concept using commercial Nuclear Power Plant (NPP) fuels such as PWR (AP-1000) and VVER-1000 fuel rods with limited 2 meter in length and fully controlled by water level, the conceptual design of the critical assembly has been carried out in neutronic, thermal hydraulics and safety analysis. Ten benchmark critical core configurations of critical assembly are established and investigated to show safety during normal opeartion and accident conditions. Design calculation results show that NPP fuels are fully adequate for critical assembly operating under nominal power 100W and having average neutron flux about 3×109 neutron/cm2.s.

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