Abstract
The ORIGEN-JUeL-II in its basic calculating method is derived from the ORNL point depletion code ORIGEN. The advanced JUeL-II-version can make use of irradiation-histogram data of fuel batches given in the form of multigroup cross sections and multigroup neutron fluxes, both varying with time during the irradiation period of the fuel. A high flexibility in processing nuclear data enables the user to determine, to which extend will substitute the nuclear data of the ORIGEN-libraries by his own, externally generated multigroup cross sections. He may decide, how much to go into detail, describing the time-dependent change not only of the height but also of the spectrum of the neutron flux. As results the user may obtain the radioactivity of the fuel, the decay power, hazards of incorporation, {gamma}-spectra and neutron sources at any point in time during the in-core irradiation as well as the properties of spent fuel during the periods of fuel storage, obtaining an improved degree of accuracy. The JUeL-II-specific application mode of the code is restricted to nuclear facilities having a thermal neutron spectrum. (orig.)
Citation Formats
Ruetten, H J.
The depletion computer code ORIGEN-JUEL-II.
Germany: N. p.,
1993.
Web.
Ruetten, H J.
The depletion computer code ORIGEN-JUEL-II.
Germany.
Ruetten, H J.
1993.
"The depletion computer code ORIGEN-JUEL-II."
Germany.
@misc{etde_10119786,
title = {The depletion computer code ORIGEN-JUEL-II}
author = {Ruetten, H J}
abstractNote = {The ORIGEN-JUeL-II in its basic calculating method is derived from the ORNL point depletion code ORIGEN. The advanced JUeL-II-version can make use of irradiation-histogram data of fuel batches given in the form of multigroup cross sections and multigroup neutron fluxes, both varying with time during the irradiation period of the fuel. A high flexibility in processing nuclear data enables the user to determine, to which extend will substitute the nuclear data of the ORIGEN-libraries by his own, externally generated multigroup cross sections. He may decide, how much to go into detail, describing the time-dependent change not only of the height but also of the spectrum of the neutron flux. As results the user may obtain the radioactivity of the fuel, the decay power, hazards of incorporation, {gamma}-spectra and neutron sources at any point in time during the in-core irradiation as well as the properties of spent fuel during the periods of fuel storage, obtaining an improved degree of accuracy. The JUeL-II-specific application mode of the code is restricted to nuclear facilities having a thermal neutron spectrum. (orig.)}
place = {Germany}
year = {1993}
month = {Mar}
}
title = {The depletion computer code ORIGEN-JUEL-II}
author = {Ruetten, H J}
abstractNote = {The ORIGEN-JUeL-II in its basic calculating method is derived from the ORNL point depletion code ORIGEN. The advanced JUeL-II-version can make use of irradiation-histogram data of fuel batches given in the form of multigroup cross sections and multigroup neutron fluxes, both varying with time during the irradiation period of the fuel. A high flexibility in processing nuclear data enables the user to determine, to which extend will substitute the nuclear data of the ORIGEN-libraries by his own, externally generated multigroup cross sections. He may decide, how much to go into detail, describing the time-dependent change not only of the height but also of the spectrum of the neutron flux. As results the user may obtain the radioactivity of the fuel, the decay power, hazards of incorporation, {gamma}-spectra and neutron sources at any point in time during the in-core irradiation as well as the properties of spent fuel during the periods of fuel storage, obtaining an improved degree of accuracy. The JUeL-II-specific application mode of the code is restricted to nuclear facilities having a thermal neutron spectrum. (orig.)}
place = {Germany}
year = {1993}
month = {Mar}
}