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The Spallation Neutron Source (SNS) at Oak Ridge National Laboratory reached 180 kW in August 2007, becoming the brightest pulsed neutron source in the world. At its full power of 1.4 MW, SNS will have thermal neutron fluxes approximately an order of magnitude greater than any existing pulsed spallation source. It thus brings a serious challenge to the lifetime of the moderator poison sheets. The SNS moderators are integrated with the inner reflector plug (IRP) at a cost of -$2 million a piece. A replacement of the IRP presents a significant drawback to the facility due to the a-;;tivation and the operation cost. Although there are many factors limiting the lifetime of the IRP, such as radiation damage to the structural material and helium production in beryllium, the limiting factor is the lifetime of the moderator poison sheets. The current operating target system of SNS was built with thick Gd poison sheets at a projected lifetime of 3 years. A recent design based on the MCNPX calculation proposed to replace the Gd poison sheets with even thicker Cd poison sheets, aiming to extend the poison sheet lifetime from 3 to 4 years accompanied by an approximate 5% gain of the moderator performance. An experiment was carried out to verify the calculated moderator performance at the Low Energy Neutron Source (LENS), Indiana University, where the spectra of two polyethylene moderators were measured. The moderators are Cd-decoupled and are poisoned with 0.8 mm Gd and 1.2 mm Cd, respectively. The preliminary analysis of the experiment data shows that the characteristics of the measured spectra of the Gdand Cd-poisoned moderators agree well with what the calculation predicted. A better moderator performance is observed in the Cd-poisoned moderator. The measured ratio of Cd over Gd on the moderator performance is in a reasonable agreement with the calculation. Further investigation is underway for a better understanding of the difference between the experiment and the calculation.

The Monte Carlo particle transport code MCNPX and its associated data have been the focus of a major development effort at Los Alamos for several years. The system has reached a mature state, and has become a significant tool for many intermediate and high-energy particle transport applications. A recent version has been released to the Radiation Safety Information Computational Center (RSICC). A recent report provides an overview of the code and an extensive set of references for the component physics modules used in the code. In this paper we review the status of the developmental version of MCNPX, and describe some important new enhancements, including the use of evaluated nuclear data files for proton transport; the use of photonuclear reaction data; improved elastic and inelastic react ion cross sections for nucleons, antinucleons, pions, and kaons; and two new modes of operation of the code. We also illustrate the use of the new proton and photonuclear data in two representative applications.

A new method for the calculation of radiation damage parameters in large geometries encompassing multiple materials using the MCNPX mesh tally has been developed. The method has been tested against previously published calculations of the displacement rate for protons and neutrons at the center of the SNS 316LN stainless steel target vessel nose. Displacement rates for neutrons, protons, and the total using the mesh tally method are shown to agree with previous work. The mesh tally method is also applied to the SNS aluminum moderator vessels and to the SNS inner reflector plug composed of aluminum, beryllium, and stainless steel. Results are given for displacement, helium, and silicon production rates.

EURISOL DS/Task5/TN-06-18