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Materials Science & Technology
The objectives of MSTD are: promote the advancement of materials science in Nuclear Science Technology; support the multidisciplines which constitute it; encourage research by providing a forum for the presentation, exchange, and documentation of relevant information; promote the interaction and communication among its members; and recognize and reward its members for significant contributions to the field of materials science in nuclear technology.
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ANS Student Conference 2025
April 3–5, 2025
Albuquerque, NM|The University of New Mexico
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Norway’s Halden reactor takes first step toward decommissioning
The government of Norway has granted the transfer of the Halden research reactor from the Institute for Energy Technology (IFE) to the state agency Norwegian Nuclear Decommissioning (NND). The 25-MWt Halden boiling water reactor operated from 1958 to 2018 and was used in the research of nuclear fuel, reactor internals, plant procedures and monitoring, and human factors.
C. E. Ragan III, G. F. Auchampaugh, A. Hemmendinger, M. G. Silbert
Nuclear Science and Engineering | Volume 61 | Number 1 | September 1976 | Pages 33-39
Technical Paper | doi.org/10.13182/NSE76-A28458
Articles are hosted by Taylor and Francis Online.
A benchmark measurement of the neutron leakage spectrum from a pulsed 38-kg uranium (93.5% 235U) sphere has been made using time-of-flight techniques. The sphere had a multiplication of ∼11 for 14-MeV neutrons, and a neutron hold-up time of ∼40 nsec. The centrally located source of 14.1 ± 0.8-MeV neutrons, produced by bombarding a tritium gas target with pulses of low-energy deuterons, was isotropic to ±7.7%. Neutrons in the 0.180- to 16.0-MeV energy range were detected at the end of a 39-m flight path by an Ne-213 liquid scintillator employing pulse-shape discrimination. The detector efficiency was measured over this same energy range using monoenergetic neutrons from the T(p,n) T(d,n), and D(d,n) reactions. The measured neutron flux as a function of energy is compared with the results of Monte Carlo calculations performed with the MCN code. Uranium cross sections from ENDF/B-IV and an older set from Lawrence Livermore Laboratory were used in these calculations. The results calculated using the ENDF/B-IV cross sections are in good agreement with the measurements, especially in the 1- to 6-MeV energy region where the uncertainties in both the calculated and experimental results are the smallest.