Program - Brookhaven National Laboratory

A. Wallner, VERA Laboratory, Faculty of Physics, University of Vienna, Austria Department of Nuclear
Physics, Research School of Physics and Engineering, The Australian National University, Canberra,
Australia. K. Buczak, C. Lederer, VERA Laboratory, Faculty of Physics, University of Vienna, Austria.
I. Dillmann, GSI Helmholtzzentrum für Schwerionenforschung GmbH, Planckstrasse 1, D-64291
Darmstadt, Germany. F. Käppeler, Karlsruhe Institute of Technology (KIT), Campus Nord, Institut für
Kernphysik, Karlsruhe, Germany. A. Mengoni, Agenzia nazionale per le nuove tecnologie, l’energia e lo
sviluppo economico sostenibile (ENEA), Bologna, Italy.
Accelerator mass spectrometry (AMS) is a relatively new technique for measuring cross sections independent
of the half-lives of reaction products. For specific reactions the sensitivity of AMS offers a unique tool
to pin down uncertainties, thus elucidating current open questions in nucleosynthesis, e.g. within the sand
p-process path. The combination of sample activation and subsequent AMS measurement was applied
for a range of measurements where off-line decay counting is difficult or impossible due to long half-lives
of reaction products or due to the absence of suitable γ-ray transitions. In particular, a series of samples
was irradiated at Karlsruhe Institute of Technology (KIT) with neutrons simulating a Maxwell-Boltzmann
distribution of 25 keV, and also with quasi-monoenergetic neutrons of energies up to 500 keV. The AMS
measurement of such neutron (and charged-particle) induced cross sections were performed at the VERA
(Vienna Environmental Research Accelerator) facility. This approach directly counts the produced atoms
in the sample after the neutron activation rather than measuring the associated γ-radiation or the emitted
particles during the irradiation. Recent AMS data for a series of neutron-capture reactions in the mass
range between Be and U allow a systematic comparison with existing data from complementary techniques.
Our results are based on individual AMS reference materials and the neutron fluence was determined relative
to 197 Au(n,γ). This approach allows to generate cross section data with uncertainties between 2 and
5 %. The new data include neutron-capture measurements for 9 Be, 13 C, 35 Cl, 40 Ca, 54 Fe and 235,238 U, and
for 14 N(n,p). We will also demonstrate the reproducibility and limits of this approach.
PR 9
Incineration of Weapons Grade Plutonium in the GT-MHR Reactor
E.F. Mitenkova, D.A. Blokhin, N.V. Novikov, Nuclear Safety Institute of Russian Academy of Sciences
(IBRAE), B.Tulskaya str. 52, Moscow, Russia. A.I. Blokhin, A.I. Leypunsky Institute of Physics and
Power Engineering, Bondarenko sq., 1, 249033 Obninsk, Russia.
In this paper a simulation research of the mixed oxide fuel burnup in the high temperature reactor (HTR)
are discussed. The possibility of fueling the GT-MHR [1] reactor with weapons grade plutonium (WGPu
- weapons grade plutonium) was considered. This reactor design is based on the deep burnup of initially
loaded plutonium fuel at its single use in core and the subsequent disposal of the spent fuel without additional
processing. The fuel in the GT-MHR reactor is used in the form of WGPu plutonium oxide spheres
of 200 µm diameter coated by three layers with pyrocarbon and silicon carbide (multi-layer ceramic coating
TRISO fuel). The isotopic composition of the spent fuel and an accumulation of minor actinides from Th-
230 to Cf-254 were estimated at various operational times. It is shown the amount of transuranic isotopes
depends on Pu-239 burnup level significantly. The advantage of using MOX plutonium fuel in the nuclear
waste transmutation context is mainly related to the possibility to incinerate Pu-239 efficiently. The influence
of actinides and all significant fission products on the reactor characteristics was considered. The
radiotoxicity, gamma-ray dose rates and fluxes, decay-heat values generated by the spent fuel at various
irradiation and cooling times were also investigated. To clarify these questions the computational methods
and cross-section nuclear data libraries were developed. All burnup calculations were performed using
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