Hans Ludewig

Design of particle bed reactors for the space nuclear thermal propulsion program

Abstract This paper describes the design for the Particle Bed Reactor (PBR) that we considered for the Space Nuclear Thermal Propulsion (SNTP) Program. The methods of analysis and their validation are outlined first. Monte Carlo methods were used for the physics analysis, several new algorithms were developed for the fluid dynamics, heat transfer and transient analysis; and commercial codes were used for the stress analysis. We carried out a critical experiment, prototypic of the PBR to validate the reactor physics; blowdown experiments with beds of prototypic dimensions were undertaken to validate the power-extraction capabilities from particle beds. In addition, materials and mechanical design concepts for the fuel elements were experimentally validated. Four PBR rocket reactor designs were investigated parametrically. They varied in power from 400 MW to 2000 MW, depending on the missions goals. These designs all were characterized by a negative prompt coefficient, due to Doppler feedback, and a moderator feedback coefficient which varied from slightly positive to slightly negative. In all practical designs, we found that the coolant worth was positive, and the thrust/weight ratio was greater than 20.

A High-Power Target Experiment

We describe an experiment designed as a proof-of-principle test for a target system capable of converting a 4-MW proton beam into a high-intensity muon beam suitable for incorporation into either a neutrino factory complex or a muon collider. The target system is based on exposing a free mercury jet to an intense proton beam in the presence of a high-strength solenoidal magnetic field.

Research Activities on Neutrorics under ASTE Collaboration at AGS/BNL

A series of experiments on a mercury spallation target using high-peak-power GeV proton-beam from the Alternating Gradient Synchrotron (AGS) of Brookhaven National Laboratory (BNL) has been performed under an international collaboration among the laboratories in Japan, U.S. and Europe, namely the ASTE (AGS Spallation Target Experiment) collaboration. This paper reviews the current status of the experiments on neutronic performance of the mercury target.

High performance nuclear thermal propulsion system for near term exploration missions to 100 A.U. and beyond

Abstract A new compact ultra light nuclear reactor engine design termed MITEE (MIniature Reac Tor EnginE) is described. MITEE heats hydrogen propellant to 3000 K, achieving a specific impulse of 1000 seconds and a thrust-to-weight of 10. Total engine mass is 200 kg, including reactor, pump, auxiliaries and a 30% contingency. MITEE enables many types of new and unique missions to the outer solar system not possible with chemical engines. Examples include missions to 100 A.U. in less than 10 years, flybys of Pluto in 5 years, sample return from Pluto and the moons of the outer planets, unlimited ramjet flight in planetary atmospheres, etc. Much of the necessary technology for MITEE already exists as a result of previous nuclear rocket development programs. With some additional development, initial MITEE missions could begin in only 6 years.

Thorium fuel cycles with externally driven systems

Abstract Externally driven subcritical systems are closely associated with thorium, partially because thorium has no naturally occurring fissile isotopes. Both accelerator-driven systems (ADSs) and fusion-driven systems have been proposed. This paper highlights key literature related to the use of thorium in externally driven systems (EDSs) and builds upon this foundation to identify potential roles for EDSs in thorium fuel cycles. In fuel cycles with natural thorium feed and no enrichment, the potential roles are (1) a once-through breed-and-burn fuel cycle and (2) a fissile breeder (mainly 233U) to support a fleet of critical reactors. If enriched uranium is used in the fuel cycle in addition to thorium, EDSs may be used to burn transuranic material. These fuel cycles were evaluated in the recently completed U.S. Department of Energy Evaluation and Screening of nuclear fuel cycle options relative to the current once-through commercial nuclear fuel cycle in the United States. The evaluation was performed with respect to nine specified high-level criteria, such as waste management and resource utilization. Each of these fuel cycles presents significant potential benefits per unit energy generation compared to the present once-through uranium fuel cycle. A parametric study indicates that fusion-fission–hybrid systems perform better than ADSs in some missions due to a higher neutron source relative to the energy required to produce it. However, both potential externally driven technology choices face significant development and deployment challenges. In addition, there are significant challenges associated with the use of thorium fuel and with the transition from a uranium-based fuel cycle to a thorium-based fuel cycle.

Super-Invar as a target for a pulsed high-intensity proton beam

We describe measurements performed on samples consisting of the alloy Super-Invar, which is a candidate material for a robust solid target used in conjunction with an intense pulsed proton beam. A low coefficient of thermal expansion is the characteristic property which makes Super-Invar an attractive target candidate. We have irradiated our samples at the Brookhaven Linac Isotope Producer facility. Tests for variations of the thermal expansion coefficient as a function of inflicted radiation damage are described. The high radiation dose is severely detrimental to its low coefficient of thermal expansion.

Collimator systems for the SNS ring

The requirements and performance goals for the collimators are to reduce the uncontrolled beam loss by 2/spl times/10/sup -4/, absorb 2 kW of deposited heat, and minimize production and leakage of secondary radiation. In order to meet these requirements a self-shielding collimator configuration consisting of a layered structure was designed. The front layers (in the direction of the proton beam) are relatively transparent to the protons, and become progressively less transparent (blacker) with depth into the collimator. In addition, a high density (iron) shield is added around the outside. The protons will be stopped in the center of the collimator, and thus the bulk of the secondary particles are generated at this location. The conceptual design described, the method of analysis discussed, and preliminary performance parameters outlined.

High-Performance Ultra-light Nuclear Rockets for Near-Earth Objects Interaction Missionsa

ABSTRACT: The performance capabilities and technology features of ultra compact nuclear thermal rockets based on very high power density (30 Megawatts per liter) fuel elements are described. Nuclear rockets appear particularly attractive for carrying out missions to investigate or intercept near‐Earth objects (NEOs) that potentially could impact on the Earth. Many of these NEO threats, whether asteroids or comets, have extremely high closing velocities, i.e., tens of kilometers per second relative to the Earth. Nuclear rockets using hydrogen propellant enable flight velocities 2 to 3 times those achievable with chemical rockets, allowing interaction with a potential NEO threat at a much shorter time, and at much greater range. Two versions of an ultra compact nuclear rocket based on very high heat transfer rates are described: the PBR (Particle Bed Reactor), which has undergone substantial hardware development effort, and MITEE (MIniature ReacTor EnginE) which is a design derivative of the PBR. Nominal performance capabilities for the PBR are: thermal power ≃1000 MW thrust ≃45,000 lbsf, and weight ≃500 kg. For MITEE, nominal capabilities are: thermal power 100 MW; thrust ≃4500 lbsf, and weight ≃50 kg. Development of operational PBR/MITEE systems would enable spacecraft launched from LEO (low‐Earth orbit) to investigate intercept NEOs at a range of ∼100 million kilometers in times of ∼30 days.

Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety

This report summarizes the results of an expert-opinion elicitation activity designed to qualitatively assess the status and capabilities of currently available computer codes and models for accident analysis and reactor safety calculations of advanced sodium fast reactors, and identify important gaps. The twelve-member panel consisted of representatives from five U.S. National Laboratories (SNL, ANL, INL, ORNL, and BNL), the University of Wisconsin, the KAERI, the JAEA, and the CEA. The major portion of this elicitation activity occurred during a two-day meeting held on Aug. 10-11, 2010 at Argonne National Laboratory. There were two primary objectives of this work: (1) Identify computer codes currently available for SFR accident analysis and reactor safety calculations; and (2) Assess the status and capability of current US computer codes to adequately model the required accident scenarios and associated phenomena, and identify important gaps. During the review, panel members identified over 60 computer codes that are currently available in the international community to perform different aspects of SFR safety analysis for various event scenarios and accident categories. A brief description of each of these codes together with references (when available) is provided. An adaptation of the Predictive Capability Maturity Model (PCMM) for computational modeling and simulation is described for use in this work. The panels assessment of the available US codes is presented in the form of nine tables, organized into groups of three for each of three risk categories considered: anticipated operational occurrences (AOOs), design basis accidents (DBA), and beyond design basis accidents (BDBA). A set of summary conclusions are drawn from the results obtained. At the highest level, the panel judged that current US code capabilities are adequate for licensing given reasonable margins, but expressed concern that US code development activities had stagnated and that the experienced user-base and the experimental validation base was decaying away quickly.

Calculation of the prompt neutron lifetime in the NBSR

Abstract The prompt neutron lifetime was calculated for the NBSR, a heavy water-cooled and -moderated research reactor at the National Institute of Standards and Technology. The method is based on the fact that the decay of a pulse of fast neutrons is related to the prompt neutron lifetime (and the multiplication constant for the reactor and the delayed neutron fraction). A Monte Carlo simulation of the decay is then used to calculate the prompt neutron lifetime at two points in the fuel cycle. At the start-up of a new cycle, the prompt neutron lifetime was calculated to be 774 ± 35 μs, and at the end of a cycle, it was calculated to be 819 ± 48 μs.