Study of reduced-enrichment uranium fuel possibility for research reactors

V. A. Ruppel, Yu S. Tretyakova, S. V. Lavrinenko, A. A. Matveeva, V. N. Martyshev

Research output: Contribution to journalConference article

6 Citations (Scopus)

Abstract

Having analyzed the results obtained in the work, it is possible to conclude that the flux density of fast and thermal neutrons in the shell of fuel elements in EFA in REU-zone decreased on average by 5%for UO2 fuel and by 7%for U9%Mo fuel. Change of neutrons flux density during the cycle does not exceed 4%for both fuel types. On average the fuel burnup in reactor core during the cycle for UO2 and U9%Mo increased by 2.8%. It is 1%less that in HEU-zone, which is conditioned by higher initial loading of 235U in fuel assembly with REU fuel.

Original languageEnglish
Article number01059
JournalMATEC Web of Conferences
Volume37
DOIs
Publication statusPublished - 22 Dec 2015
EventSmart Grids 2015 - Tomsk, Russian Federation
Duration: 28 Sep 20152 Oct 2015

Fingerprint

Research reactors
Uranium
Neutron flux
Reactor cores
Neutrons
Fluxes

ASJC Scopus subject areas

  • Chemistry(all)
  • Engineering(all)
  • Materials Science(all)

Cite this

Study of reduced-enrichment uranium fuel possibility for research reactors. / Ruppel, V. A.; Tretyakova, Yu S.; Lavrinenko, S. V.; Matveeva, A. A.; Martyshev, V. N.

In: MATEC Web of Conferences, Vol. 37, 01059, 22.12.2015.

Research output: Contribution to journalConference article

@article{44cbf9a2780a49049148ac2c332ce017,
title = "Study of reduced-enrichment uranium fuel possibility for research reactors",
abstract = "Having analyzed the results obtained in the work, it is possible to conclude that the flux density of fast and thermal neutrons in the shell of fuel elements in EFA in REU-zone decreased on average by 5{\%}for UO2 fuel and by 7{\%}for U9{\%}Mo fuel. Change of neutrons flux density during the cycle does not exceed 4{\%}for both fuel types. On average the fuel burnup in reactor core during the cycle for UO2 and U9{\%}Mo increased by 2.8{\%}. It is 1{\%}less that in HEU-zone, which is conditioned by higher initial loading of 235U in fuel assembly with REU fuel.",
author = "Ruppel, {V. A.} and Tretyakova, {Yu S.} and Lavrinenko, {S. V.} and Matveeva, {A. A.} and Martyshev, {V. N.}",
year = "2015",
month = "12",
day = "22",
doi = "10.1051/matecconf/20153701059",
language = "English",
volume = "37",
journal = "MATEC Web of Conferences",
issn = "2261-236X",
publisher = "EDP Sciences",

}

TY - JOUR

T1 - Study of reduced-enrichment uranium fuel possibility for research reactors

AU - Ruppel, V. A.

AU - Tretyakova, Yu S.

AU - Lavrinenko, S. V.

AU - Matveeva, A. A.

AU - Martyshev, V. N.

PY - 2015/12/22

Y1 - 2015/12/22

N2 - Having analyzed the results obtained in the work, it is possible to conclude that the flux density of fast and thermal neutrons in the shell of fuel elements in EFA in REU-zone decreased on average by 5%for UO2 fuel and by 7%for U9%Mo fuel. Change of neutrons flux density during the cycle does not exceed 4%for both fuel types. On average the fuel burnup in reactor core during the cycle for UO2 and U9%Mo increased by 2.8%. It is 1%less that in HEU-zone, which is conditioned by higher initial loading of 235U in fuel assembly with REU fuel.

AB - Having analyzed the results obtained in the work, it is possible to conclude that the flux density of fast and thermal neutrons in the shell of fuel elements in EFA in REU-zone decreased on average by 5%for UO2 fuel and by 7%for U9%Mo fuel. Change of neutrons flux density during the cycle does not exceed 4%for both fuel types. On average the fuel burnup in reactor core during the cycle for UO2 and U9%Mo increased by 2.8%. It is 1%less that in HEU-zone, which is conditioned by higher initial loading of 235U in fuel assembly with REU fuel.

UR - http://www.scopus.com/inward/record.url?scp=84976490968&partnerID=8YFLogxK

UR - http://www.scopus.com/inward/citedby.url?scp=84976490968&partnerID=8YFLogxK

U2 - 10.1051/matecconf/20153701059

DO - 10.1051/matecconf/20153701059

M3 - Conference article

AN - SCOPUS:84976490968

VL - 37

JO - MATEC Web of Conferences

JF - MATEC Web of Conferences

SN - 2261-236X

M1 - 01059

ER -