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The depletion computer code ORIGEN-JUEL-II

Technical Report:

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.)
Authors:
Publication Date:
Mar 01, 1993
Product Type:
Technical Report
Report Number:
Juel-2739
Reference Number:
SCA: 220100; 663610; 050900; PA: DEN-94:0F0657; EDB-94:024072; ERA-19:006647; NTS-94:015645; SN: 94001136104
Resource Relation:
Other Information: PBD: Mar 1993
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS; 73 NUCLEAR PHYSICS AND RADIATION PHYSICS; 11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; REACTOR PHYSICS; O CODES; FUEL MANAGEMENT; FISSION SPECTRA; COMPUTER CODES; MULTIGROUP THEORY; NEUTRON TRANSPORT THEORY; NEUTRON FLUX; CROSS SECTIONS; TIME DEPENDENCE; BURNUP; GAMMA SPECTRA; THERMAL NEUTRONS; SPENT FUEL STORAGE; NUCLEAR DATA COLLECTIONS; THERMAL FISSION; COMPUTER PROGRAM DOCUMENTATION; 220100; 663610; 050900; THEORY AND CALCULATION; NEUTRON PHYSICS; TRANSPORT, HANDLING, AND STORAGE
OSTI ID:
10119786
Research Organizations:
Forschungszentrum Juelich GmbH (Germany). Inst. fuer Sicherheitsforschung und Reaktortechnik
Country of Origin:
Germany
Language:
English
Other Identifying Numbers:
Other: ON: DE94733998; TRN: DE94F0657
Availability:
OSTI; NTIS (US Sales Only); INIS
Submitting Site:
DEN
Size:
44 p.
Announcement Date:
Jun 30, 2005

Technical Report:

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}
}