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Hendricks, John S

Publications and source records attributed to Hendricks, John S.

7 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

The New MCNP6 Depletion Capability

The first MCNP based inline Monte Carlo depletion capability was officially released from the Radiation Safety Information and Computational Center as MCNPX 2.6.0. Both the MCNP5 and MCNPX codes have historically provided a successful combinatorial geometry based, continuous energy, Monte Carlo radiation transport solution for advanced reactor modeling and simulation. However, due to separate development pathways, useful simulation capabilities were dispersed between both codes and not unified in a single technology. MCNP6, the next evolution in the MCNP suite of codes, now combines the capability of both simulation tools, as well as providing new advanced technology, in a single radiation transport code. We describe here the new capabilities of the MCNP6 depletion code dating from the official RSICC release MCNPX 2.6.0, reported previously, to the now current state of MCNP6. NEA/OECD benchmark results are also reported. The MCNP6 depletion capability enhancements beyond MCNPX 2.6.0 reported here include: (1) new performance enhancing parallel architecture that implements both shared and distributed memory constructs; (2) enhanced memory management that maximizes calculation fidelity; and (3) improved burnup physics for better nuclide prediction. MCNP6 depletion enables complete, relatively easy-to-use depletion calculations in a single Monte Carlo code. The enhancements described here help provide a powerful capability as well as dictate a path forward for future development to improve the usefulness of the technology.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

MCNP6 Fission Multiplicity with FMULT Card

With the merger of MCNPX and MCNP5 into MCNP6, MCNP6 now provides all the capabilities of both codes allowing the user to access all the fission multiplicity data sets. Detailed in this paper is: (1) the new FMULT card capabilities for accessing these different data sets; (2) benchmark calculations, as compared to experiment, detailing the results of selecting these separate data sets for thermal neutron induced fission on U-235.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2011

Calculation of the performance of {sup 3}He alternative detectors with MCNPX

This paper describes the techniques that are available to calculate the performance of {sup 3}He alternative detectors using MCNPX. Calculations of the performance of safeguards detectors that use {sup 3}He have been successfully carried out for many years. In the case of coincidence or multiplicity counting, specific tallies have been implemented to calculate the Singles, Doubles and Triples counting rates. The implementation of the method was done in such a way that it equates every capture in some nuclide in the detection zone with the production of an electronic pulse from the detector. This is a very good approximation for {sup 3}He detectors and BF{sub 3} detectors. However it is not appropriate for detectors such as boron-lined proportional counters, in which the fraction of capture events leading to an electronic pulse above threshold is very dependent on the geometric arrangement, in particular the thickness and composition of the boron-containing layer. This paper gives calculations of the ideal pulse height distributions to be expected from different detector types and gives values for the probability, as a function of detector energy threshold, that a neutron capture reaction will cause an electronic pulse from the detector. This is termed the electronic efficiency. This electronic detection efficiency depends very little on the energy of the captured neutron, which in most practical cases are heavily weighted towards thermal energies. It does not depend on the position of the source neutron or moderation in the sample. For cases of interest to nuclear safeguards, measurement of uranium and plutonium in specially designed detectors, the spectrum of detected neutrons is fairly constant and thus the electronic detection efficiency becomes a detector constant. The paper discusses how the electronic detection efficiency needs to be included in the calculation of Singles, Doubles and Triples, and describes proposals to improve the tallying capability of MCNPX for such cases. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗