Detailed core design and flow coolant conditions for neutron flux maximization in research reactors

F. E. Teruel, Rizwan-Uddin

Research output: Chapter in Book/Report/Conference proceedingConference contribution

Abstract

Following the design of the German research reactor, FRM-II, which delivers high thermal neutron fluxes, we have already developed an asymmetric compact cylindrical core with an inner and outer reflector. The goal was to maximize the maximum thermal flux to power ratio, which is a desirable characteristic of a modern research reactor. This design, for a 10 MW power, was analyzed using MCNP, ORIGEN2 and MONTEBURNS codes, considering a homogeneous mixture for the core material. Promising results showed that with standard fuel material of low enrichment uranium, high thermal fluxes are delivered by this core in the outer reflector. In addition, the life cycle, which is a limiting parameter regarding the compactness of the core, was calculated to be 41 days. In this paper, we report the results of recent developments in the design of this core. A detailed modeling of a suitable fuel element is performed with MCNP. Neutron fluxes are recalculated to assure that the same levels are achieved with this more detailed description. In this regard, three different zones with different flux levels were identified: a high neutron flux zone (4.0E14 n·cm-2s -1), a moderate thermal neutron flux zone (2.5E14 n·cm -2s-1), and a low thermal flux zone (1.0E14 n·cm-2s-1). Heat deposition in the cladding, coolant and fuel material is also calculated to determine coolant flow rate, coolant outlet temperature and maximum fuel temperature for safe operating conditions.

Original languageEnglish (US)
Title of host publicationFourteenth International Conference on Nuclear Engineering 2006, ICONE 14
Volume2006
DOIs
StatePublished - 2006
EventFourteenth International Conference on Nuclear Engineering 2006, ICONE 14 - Miami, FL, United States

Other

OtherFourteenth International Conference on Nuclear Engineering 2006, ICONE 14
CountryUnited States
CityMiami, FL
Period7/17/067/20/06

Fingerprint

Neutron flux
Coolants
Fluxes
Research reactors
Temperature
Uranium
Life cycle
Flow rate

ASJC Scopus subject areas

  • Engineering(all)
  • Energy(all)

Cite this

Teruel, F. E., & Rizwan-Uddin (2006). Detailed core design and flow coolant conditions for neutron flux maximization in research reactors. In Fourteenth International Conference on Nuclear Engineering 2006, ICONE 14 (Vol. 2006). DOI: 10.1115/ICONE14-89547

Detailed core design and flow coolant conditions for neutron flux maximization in research reactors. / Teruel, F. E.; Rizwan-Uddin.

Fourteenth International Conference on Nuclear Engineering 2006, ICONE 14. Vol. 2006 2006.

Research output: Chapter in Book/Report/Conference proceedingConference contribution

Teruel, FE & Rizwan-Uddin 2006, Detailed core design and flow coolant conditions for neutron flux maximization in research reactors. in Fourteenth International Conference on Nuclear Engineering 2006, ICONE 14. vol. 2006, Fourteenth International Conference on Nuclear Engineering 2006, ICONE 14, Miami, FL, United States, 17-20 July. DOI: 10.1115/ICONE14-89547
Teruel FE, Rizwan-Uddin. Detailed core design and flow coolant conditions for neutron flux maximization in research reactors. In Fourteenth International Conference on Nuclear Engineering 2006, ICONE 14. Vol. 2006. 2006. Available from, DOI: 10.1115/ICONE14-89547

Teruel, F. E.; Rizwan-Uddin / Detailed core design and flow coolant conditions for neutron flux maximization in research reactors.

Fourteenth International Conference on Nuclear Engineering 2006, ICONE 14. Vol. 2006 2006.

Research output: Chapter in Book/Report/Conference proceedingConference contribution

@inbook{4c1b8c99cc2f49ae906db71d13b87b38,
title = "Detailed core design and flow coolant conditions for neutron flux maximization in research reactors",
abstract = "Following the design of the German research reactor, FRM-II, which delivers high thermal neutron fluxes, we have already developed an asymmetric compact cylindrical core with an inner and outer reflector. The goal was to maximize the maximum thermal flux to power ratio, which is a desirable characteristic of a modern research reactor. This design, for a 10 MW power, was analyzed using MCNP, ORIGEN2 and MONTEBURNS codes, considering a homogeneous mixture for the core material. Promising results showed that with standard fuel material of low enrichment uranium, high thermal fluxes are delivered by this core in the outer reflector. In addition, the life cycle, which is a limiting parameter regarding the compactness of the core, was calculated to be 41 days. In this paper, we report the results of recent developments in the design of this core. A detailed modeling of a suitable fuel element is performed with MCNP. Neutron fluxes are recalculated to assure that the same levels are achieved with this more detailed description. In this regard, three different zones with different flux levels were identified: a high neutron flux zone (4.0E14 n·cm-2s -1), a moderate thermal neutron flux zone (2.5E14 n·cm -2s-1), and a low thermal flux zone (1.0E14 n·cm-2s-1). Heat deposition in the cladding, coolant and fuel material is also calculated to determine coolant flow rate, coolant outlet temperature and maximum fuel temperature for safe operating conditions.",
author = "Teruel, {F. E.} and Rizwan-Uddin",
year = "2006",
doi = "10.1115/ICONE14-89547",
isbn = "0791837831",
volume = "2006",
booktitle = "Fourteenth International Conference on Nuclear Engineering 2006, ICONE 14",

}

TY - CHAP

T1 - Detailed core design and flow coolant conditions for neutron flux maximization in research reactors

AU - Teruel,F. E.

AU - Rizwan-Uddin,

PY - 2006

Y1 - 2006

N2 - Following the design of the German research reactor, FRM-II, which delivers high thermal neutron fluxes, we have already developed an asymmetric compact cylindrical core with an inner and outer reflector. The goal was to maximize the maximum thermal flux to power ratio, which is a desirable characteristic of a modern research reactor. This design, for a 10 MW power, was analyzed using MCNP, ORIGEN2 and MONTEBURNS codes, considering a homogeneous mixture for the core material. Promising results showed that with standard fuel material of low enrichment uranium, high thermal fluxes are delivered by this core in the outer reflector. In addition, the life cycle, which is a limiting parameter regarding the compactness of the core, was calculated to be 41 days. In this paper, we report the results of recent developments in the design of this core. A detailed modeling of a suitable fuel element is performed with MCNP. Neutron fluxes are recalculated to assure that the same levels are achieved with this more detailed description. In this regard, three different zones with different flux levels were identified: a high neutron flux zone (4.0E14 n·cm-2s -1), a moderate thermal neutron flux zone (2.5E14 n·cm -2s-1), and a low thermal flux zone (1.0E14 n·cm-2s-1). Heat deposition in the cladding, coolant and fuel material is also calculated to determine coolant flow rate, coolant outlet temperature and maximum fuel temperature for safe operating conditions.

AB - Following the design of the German research reactor, FRM-II, which delivers high thermal neutron fluxes, we have already developed an asymmetric compact cylindrical core with an inner and outer reflector. The goal was to maximize the maximum thermal flux to power ratio, which is a desirable characteristic of a modern research reactor. This design, for a 10 MW power, was analyzed using MCNP, ORIGEN2 and MONTEBURNS codes, considering a homogeneous mixture for the core material. Promising results showed that with standard fuel material of low enrichment uranium, high thermal fluxes are delivered by this core in the outer reflector. In addition, the life cycle, which is a limiting parameter regarding the compactness of the core, was calculated to be 41 days. In this paper, we report the results of recent developments in the design of this core. A detailed modeling of a suitable fuel element is performed with MCNP. Neutron fluxes are recalculated to assure that the same levels are achieved with this more detailed description. In this regard, three different zones with different flux levels were identified: a high neutron flux zone (4.0E14 n·cm-2s -1), a moderate thermal neutron flux zone (2.5E14 n·cm -2s-1), and a low thermal flux zone (1.0E14 n·cm-2s-1). Heat deposition in the cladding, coolant and fuel material is also calculated to determine coolant flow rate, coolant outlet temperature and maximum fuel temperature for safe operating conditions.

UR - http://www.scopus.com/inward/record.url?scp=33845789093&partnerID=8YFLogxK

UR - http://www.scopus.com/inward/citedby.url?scp=33845789093&partnerID=8YFLogxK

U2 - 10.1115/ICONE14-89547

DO - 10.1115/ICONE14-89547

M3 - Conference contribution

SN - 0791837831

SN - 9780791837832

VL - 2006

BT - Fourteenth International Conference on Nuclear Engineering 2006, ICONE 14

ER -