A high-power neutron-producing spallation source is one option for meeting the mission of a fusion prototypic neutron source (FPNS). The Los Alamos Neutron Science Center at Los Alamos National Laboratory is uniquely suited to host such a source. A target concept has been developed that satisfies the initial goals established for an FPNS, as has a beam transport design that satisfies beam-on-target requirements for the concept. We discuss the potential impact of pulsed beam operation on radiation-induced changes in microstructure, as well as the ingrowth of calcium in steel alloys at a rate well beyond that expected in a fusion reactor first wall. A preliminary thermomechanical assessment shows the need to modify the target design to reduce temperatures and stresses in the tungsten target.

The Fusion Prototypic Neutron Source (FPNS) is considered to be a testbed for scientific understanding of material degradation in future nuclear fusion reactors (Zinkle and Moeslang, 2013; Summary Report on the FPNS Workshop, 2018; Pitcher a al., 2019). The primary mission of FPNS is to provide a damage rate in iron samples of 8-11 dpa/calendar year with He/dpa ratio of similar to 10 appm in irradiation volume of 50 cm(3) or larger with irradiation temperature 300-1000 degrees C and flux gradient less than 20%/cm in the plane of the sample. The Los Alamos Neutron Science Center (LANSCE) is an attractive candidate for the FPNS project. The Accelerator Facility was designed and operated for an extended period as a 0.8-MW Meson Factory. The existing setup of the LANSCE accelerator complex can nearly fulfill requirements of the fusion neutron source station. The primary function of the upgraded accelerator systems is the safe and reliable delivery of a 1.25-mA continuous proton beam current at 800-MeV beam energy from the switchyard to the target assembly to create 1 MW power of proton beam interacting with a solid tungsten target. The present study describes existing accelerator setup and further development required to meet the needs of FPNS project.

Review the moderators in use at the present operating spallation neutron sources, and give an overview of improvements being made to enhance the neutronic performance of these moderators. Characterize the present thinking on “moderator types” for the next-generation, high-power (0.6-5 MW) pulsed spallation sources. Review ideas for improving the neutronic pevformance of moderators. Survey development activities underway and being contemplated for new moderator types, and facilities for measuring the neutronic pe$ormance of moderators. Look at activities aimed at providing new scattering kernels and improving existing ones. Discuss engineering issues that must be faced in practical moderator design.

Conceptual design of the proposed Materials Test Station (MTS) at the Los Alamos Neutron Science Center (LANSCE) is now complete. The principal mission is the irradiation testing of advanced fuels and materials for fast-spectrum nuclear reactor applications. The neutron spectrum in the fuel irradiation region of MTS is sufficiently close to that of fast reactor that MTS can match the fast reactor fuel centerline temperature and temperature profile across a fuel pellet. This is an important characteristic since temperature and temperature gradients drive many phenomena related to fuel performance, such as phase stability, stoichiometry, and fission product transport. The MTS irradiation environment is also suitable in many respects for fusion materials testing. In particular, the rate of helium production relative to atomic displacements at the peak flux position in MTS matches well that of fusion reactor first wall. Nuclear transmutation of the elemental composition of the fusion alloy EUROFER97 in MTS is similar to that expected in the first wall of a fusion reactor.

Los Alamos National Laboratory's proposed signature facility, MaRIE, will provide scientists and engineers with new capabilities for modeling, synthesizing, examining, and testing materials of the future that will enhance the USA's energy security and national security. In the area of fusion power, the development of new structural alloys with better tolerance to the harsh radiation environments expected in fusion reactors will lead to improved safety and lower operating costs. The Fission and Fusion Materials Facility (F3), one of three pillars of the proposed MaRIE facility, will offer researchers unprecedented access to a neutron radiation environment so that the effects of radiation damage on materials can be measured in situ, during irradiation. The calculated radiation damage conditions within the F3 match, in many respects, that of a fusion reactor first wall, making it well suited for testing fusion materials. Here we report in particular on two important characteristics of the radiation environment with relevancy to radiation damage: the primary knock-on atom spectrum and the impact of the pulse structure of the proton beam on temporal characteristics of the atomic displacement rate. With respect to both of these, analyses show that F3 has conditions that are consistent with those of a steady-state fusion reactor first wall.

Nearly all risks to future generations arising from long-term disposal of used nuclear fuel are attributable to the transuranic elements and long-lived fission products, about 2% of its content. The transuranic elements of concern are plutonium, neptunium, americium, and curium. Long-lived (>100,000-year half-life) isotopes of iodine and technetium are also created by nuclear fission of uranium. We can reduce the problem transuranics through accelerator-based transmutation. Accelerator Driven Systems (ADS) have been proposed for over two decades as one technique to transmute used nuclear fuel. This paper covers the history and some new possible applications of accelerator driven systems.

Radionuclide inventory calculations support design and accident analyses for the Materials Test Station (MTS). MTS is a spallation source facility being designed to irradiate reactor fuels and materials in a fast neutron spectrum. Calculated radionuclide inventories are used to provide decay heat input to cooling system design, decay radiation source terms for hot cell design, and material-at-risk input to accident analyses. CINDER'90 is a transmutation code that uses MCNPX-calculated spallation productyields and neurron fluxes to calculate residual nuclide concentrations based on irradiation history. The code also calculates decay heat and photon spectra for the resulting radionuclide inventories. A total activity of 2 X 10(17) Bq is created during MTS operation. Decay heat is an important factor since in loss of primary cooling scenarios, this heat must be removed. The major sources at shutdown are 3000 W for the tungsten target plates and 6000 W for fuel pins being irradiated. Decay photon spectra result in unshielded dose rates that hot cell design must accommodate on the order of 1000 Sv/h. The MTS design includes lead-bismuth eutectic (LBE) coolant. For accident analysis (210)Po activity in the LBE is a significant concern. The calculated (210)Po activity following 2.5 yr of operation is 2 X 10(14) Bq. Radionuclide inventory calculations are important for MTS design. The CINDER'90 code is a valuable tool for this purpose.

A radiation damage database at NCSU is described that contains, at present, cross sections for damage energy, displacements, helium, and hydrogen. For neutrons and protons at 20-3200 MeV energies, the targets include 23 target elements from Mg to U and eight practical alloys. The underlying calculations employ Bertini, ISABEL, and CEM2k intranuclear cascade models as provided within MCNPX, three level-density formulations (GCCI, HETC, and Julich), and multistage pre-equilibrium model (MPM) on and off. The database also includes radiation damage cross sections for lower-energy neutrons and protons, as obtained from LA150, SPECTER, ENDF, and SRIM. Examples of the database contents are provided with regard to damage energy and displacement cross sections for neutrons on A[, Fe, and W. Also, neutron flux and displacement rate calculations are described for neutrons on Fe and the 316 stainless steel mercury container vessel at SNS. In addition, discrepancies between the codes are discussed.

The United States Department of Energy is developing technologies needed to reduce the quantity of high-level nuclear waste bound for deep geologic disposal. Central to this mission is the development of high burn-up fuel with significant inclusion of plutonium and minor actinides. Different fuel forms (e.g., nitrides, oxides, and metal matrix) and composition are under study. The success of these cannot be judged until they have been irradiated and tested in a prototypic fast neutron spectrum environment. In 2005, the US Congress authorized funding for the design of the materials test station (MTS) to perform candidate fuels and materials irradiations in a neutron spectrum similar to a fast reactor spectrum. The MTS will use a 1-MW proton beam to generate neutrons through spallation reactions. The peak neutron flux in the irradiation region will exceed 1.2×1019nm−2s−1 and the fast neutron fluence will reach 2×1026nm−2 per year of operation. Site preparation and test station fabrication are expected to take four years.

During the design of the Manuel Lujan, Jr., Neutron Scattering Center target, a simplified Monte Carlo model was used to estimate target system performance and to aid engineers as decisions were made regarding the construction of the target system. Although the simplified model ideally would perfectly reflect the as-built system performance, assumptions were made in the model during the design process that may result in deviations between the model predictions and the as-built system performance. Now that the Lujan Center target system has been completed, a more detailed, as-built, model of the target system has been completed. The purpose of this work is to investigate differences between the predicted target system performance of the simplified model and the as-built model from the standpoint of time-averaged moderator brightness. Calculated discrepancies between the two models have been isolated to a few key issues. Figure 1 shows MCNP geometric plots of the simplified and as-built models. Major differences between these two models include details in the moderator designs (plena) and piping, full versus partial moderator canisters (only in the direction of the extracted neutron beam for the simplified model), and reflector details including cooling pipes and engineering tolerance gaps. In addition, Fig. 1more » demonstrates that the detailed model includes shielding and additional material beyond that which was modeled by the original simplified model.« less