Jay S. Elson

Initial MCNP6 Release Overview

MCNP6 is simply and accurately described as the merger of MCNP5 and MCNPX capabilities, but it is much more than the sum of those two computer codes. MCNP6 is the result of five years of effort by the MCNP5 and MCNPX code development teams. These groups of people, residing in Los Alamos National Laboratory’s (LANL) X Computational Physics Division, Monte Carlo Codes Group (XCP-3), and Decision Applications Division, Radiation Transport and Applications Team (D-5), respectively, have combined their code development efforts to produce the next evolution of MCNP. While maintenance and bug fixes will continue for MCNP5 1.60 and MCNPX 2.7.0 for upcoming years, new code development capabilities only will be developed and released in MCNP6. In fact, the initial release of MCNP6 contains 16 new features not previously found in either code. These new features include the abilities to import unstructured mesh geometries from the finite element code Abaqus, to transport photons down to 1.0 eV, to transport electrons down to 10.0 eV, to model complete atomic relaxation emissions, and to generate or read mesh geometries for use with the LANL discrete ordinates code Partisn. The first release of MCNP6, MCNP6 Beta 2, is now available through the Radiation Safety Information Computational Center, and the first production release is expected in calendar year 2012. High confidence in the MCNP6 code is based on its performance with the verification and validation test suites, comparisons to its predecessor codes, the regression test suite, its code development process, and the underlying high-quality nuclear and atomic databases.

Small fission power systems for Mars

A Mars surface power system configuration with an output power of 3 kWe and a system mass of 775 kg is described. It consists of a heatpipe-cooled reactor with UN fuel coupled to a Stirling engine with a fixed conical radiator driven by loop heat pipes. Key to achieving this low mass is the use of a highly radiation-resistant multiplexer for monitoring and controlling the reactor, as well as radiation resistant generators and motors. Also key is the judicious placement of shields to prevent radiation scattered from the Martian surface and air from damaging the reactor controls. Several alternate configurations also are briefly looked at, including a moderated reactor with UZrH fuel and a reactor using 233U instead of 235U. The moderated reactor system has essentially the same mass as the baseline unmoderated UN system and yields the same radiation shielding requirements. The 233U reactor is significantly smaller and yields a system mass about 228 kg lighter than with 235U, but part of this weight reductio...

Small fission power systems for NEP

Two nuclear electric propulsion (NEP) power system configurations are presented, each with an output power of 50 kWe and a system mass of about 2500 kg. Both consist of a reactor coupled to a recuperated Brayton power conversion system with a fixed conical radiator driven by loop heat pipes. In one system the reactor is gas-cooled with the gas directly driving the Brayton power conversion system. In the other the reactor is heatpipe-cooled with a heat exchanger between the reactor and the Brayton system. Two variations are described briefly with powers of 100 k We and 150 kWe. The mass scales approximately with the square root of the power.

Neutronics and engineering design of the aqueous-slurry accelerator transmutation of waste blanket

A conceptual target and blanket design for an accelerator transmutation of waste system capable of transmuting the high-level waste stream from 2.5 light water reactors is described. Typically, four such target-blanket designs would be served by a single linear accelerator. The target consists of rows of solid tungsten rod bundles, cooled by heavy water and surrounded by a lead annulus. The annular blanket, which surrounds the target, consists of a set of actinide-oxide-slurry-bearing tubes, each 3 m long, surrounded by heavy water moderator. Heat is removed from the slurry tubes by passing the slurry through an external heat exchanger. Long-lived fission products are burned in regions that are separate from the actinides. Using the Monte Carlo codes LAHET and MCNP, a conceptual design for a beam current of 62.5 mA/target of 1.6-GeV protons has been developed. Preliminary engineering analyses on key system components have been performed. A preliminary layout of the concept and the associated primary-heat transport subsystems was developed, demonstrating a multiple-containment-boundary design philosophy.

Conceptual design of a thorium target for molten salt transmutation systems

A spallation target constructed of thorium metal has been designed for applications using molten‐salt as the target coolant. The design consists of an array of wire‐wrapped, hastelloy‐clad, thorium rods in which a parabolic void region is introduced in the upper regions. Each target rod is approximately 1 m in length, 3.1 cm in diameter, and has a clad thickness of 0.05 cm; 140 rods are arranged in a triangular lattice with a pitch of 3.2 cm, which results in a cylindrical target configuration with a radius of 20 cm and an estimated yield of 17 neutrons/proton for 800 MeV protons.

MCNPX Improvements for Threat Reduction Applications

Enhancements contained in the current MCNPX 2.6.0 Radiation Safety Information Computational Center (RSICC) release will be presented, including stopped‐muon physics, delayed neutron and photon generation, and automatic generation of source photons. Preliminary benchmarking comparisons with data taken with a muon beam at the Paul Scherrer Institute Spallation Neutron Source accelerator will be discussed. We will also describe current improvements now underway, including Nuclear Resonance Fluorescence (NRF), pulsed sources, and others. We will also describe very new work begun on a threat‐reduction (TR) user interface, designed to simplify the setup of TR‐related calculations, and introduce standards into geometry, sources and backgrounds.

Thermal-Hydraulic Design of the Accelerator Production of Tritium Tungsten Neutron Source

The thermal-hydraulic design of the accelerator production of tritium (APT) tungsten neutron source is presented. A carefully engineered thermal-hydraulic design is required to remove the deposited power effectively during normal operations and remove the decay power during plant shutdown and postulated accidents. For steady-state operations and operational and anticipated transients, the design criterion is to maintain single-phase flow conditions with a margin to onset of nucleate boiling. The margin is determined based on phenomenological and geometric uncertainties associated with the design. A large margin to thermal excursion limits, such as critical heat flux and onset of flow instability, also is maintained during normal operations. In general, a very robust thermal-hydraulic design can be accomplished using the traditional models and correlations available in the engineering literature. However, two issues require further attention: maintaining adequate flows in a parallel network of flow channels and minimizing the volume fraction of heavy water to maximize tritium production. The design uses ladderlike structures that contain clad tungsten cylinders in the rungs that have coolant supplied and removed by the vertical ladder rails. Because the power density drops in the beam direction, the thickness of the tungsten cylinders is increased with increasing beam penetration length. The cooling requirement is determined using a conservative criterion where the minimum wall subcooling inside the rungs is at least 40°C and the minimum Reynolds number is 6000. Initial flow distribution tests were conducted with a full-scale model of an APT ladder assembly based on a preliminary design. Flow distributions can be made more even by using a larger riser than downcomer and also by increasing the flow resistance across each rung. The calculations discussed assume nominal dimensions, even though the power deposition and removal use a conservative approach. The effect of manufacturing tolerances will be investigated in future research. Also, the applicability of the critical heat flux and onset of flow instability models to small coolant channels is being verified experimentally. Further design optimization will be possible when these studies are completed.

Alert-derivative bimodal space power and propulsion systems

Safe, reliable, low‐mass bimodal space power and propulsion systems could have numerous civilian and military applications. This paper discusses potential bimodal systems that could be derived from the ALERT space fission power supply concept (Ranken 1990). These bimodal concepts have the potential for providing 5 to 10 kW of electrical power and a total impulse of 100 MN‐s at an average specific impulse of 770 s. System mass is on the order of 1000 kg.