Development of Fuel Materials for Nuclear Thermal Propulsion

The topic selected was Development of Fuel Materials for Nuclear Thermal Propulsion. Advancements in developing high power density fuel materials for nuclear thermal rockets (NTR) heightens the drive for future space exploration. A current aim for future manned missions is to decrease launch costs to be comparable to that of chemical rockets. Rockets with high specific impulse technologies makes this possible as it reduces the propellant weight need to put a payload in orbit.  This also infers a lower cost of launching material from the Earth.

Low-Enriched Cermet-Based Fuel Options for a Nuclear Thermal Propulsion Engine

The aim of the research was to provide an efficient way of designing a compact optimal nuclear rocket engine that was tungsten cermet low enriched uranium (LEU) fueled. In NASA’s prospective mission to Mars, two space propulsion systems were considered: nuclear thermal propulsion (NTP) and advanced thermal propulsion systems with aerocapture. NTP became NASA’s primary choice because of lower propellant requirements. NTP engines were also preferred for near term deep space missions due to their high thrust and high specific impulse. The goal was to create a NTP rocket engine that surpassed NASA’s requirements:

• 25 klb_f thrust engine
• Thrust to weight ratio > 3.5
• I_sp > 900s

Pewee, NASA’s third cycle of the Rover program, was at the core for the research conducted in this paper. Previously, work done with NTP systems involved the use of highly enriched uranium (HEU); this is now a concern to use with these systems. Newer technology utilizes LEU as fuel and must meet specific needs. The research addresses these needs, and highlighted newer findings done with materials and manufacturing. Experiments conducted focused on neutronic and thermal sensitivity and tradeoff studies. Sensitivity studies conducted tested the core’s radial and axial dimensions, reflector thickness, moderator-to-fuel element (ME-to-FE) ratio, fuel enrichment, tungsten enrichment and mass flow rate. In attempt of optimizing the design, neutronic and thermal-hydraulic analyses were conducted. Tendencies were highlighted as well as trade-offs between contrasting variables.

The two main systems of analyses included the Monte-Carlo and T/H module (THERMO). Neutronic analyses were performed with a Monte-Carlo Serpent code. Thermal-hydraulic (T/H) calculations were done by THERMO. It simulated axial propellant flow and radial heat conduction. The research also aimed at surpassing NASA’s requirements in a referenced Mars’ mission; it made use of several assumptions:

• Tungsten cermet-based LEU fuel
• Hexagonal nineteen-channel fuel elements

The study was heavily focused on NASA’s Mars Design Reference Architecture 5.0.  Several constraints were established for the duration of this study: specific impulse, total thrust and thrust-to-weight ratio. Calculations were executed using an I_sp of 900s, total thrust of 75 klb_f (25 klb_f for each of the three engines), and a thrust-to-weight ratio of 3.5. An additional constraint included was that a single engine will not exceed a diameter of 4.6 m.

Materials

The first step of the design phase included selecting appropriate material. The fuel material, FE, chosen for the nuclear thermal reactor (NTR) consisted of 19.75%enriched UO₂ particles. The UO₂ particles were set in a 95% enriched tungsten matrix with 6 mol% ThO₂ as a stabilizer. The tungsten/UO₂ combination yielded a melting point of 3695 K. Lower fission was compensated by the system being more thermally oriented. Moderating elements (ME) were used to balance fast neutrons to thermal energies. The H₂ propellant was heated to 600 K by passing it through the ME twice before reaching the FE.

Fig. 1. Radial Schematics of Moderating Element [1]

The materials chosen in the radial design were all crucial in optimizing the overall performance of the NTR. The graphite sleeve decelerated neutrons while the ring of ZrC protected the Be radial reflector from high temperatures. The reflectors also modified flux/power distribution and reduced the maximum fuel temperature. Reflectors helped increase the thrust-to-weight ratio as it allowed for a more compact core. Sixteen control drums, partially coated with a neutron absorber, were located around the radial reflector region.

Fig. 2. Radial Core Configuration [1]

Figure 3 indicated the materials used in various regions of the axial engine. The aim of the axial engine was to maintain a controlled hydrogen flow through the system.

Fig. 3. Axial System Configuration [1]

Thermal Hydraulic Approach

One of the methods utilized in the study was the T/H approach. Adaptation of this method was based on fundamental conduction and convection transfer. Analysis of temperature distribution required obtaining values of thermal conductivities of the materials. A 70% porous ZrC material was developed by Los Alamos in attempt of reducing the overall heat in the ME. Power generated and produced was conducted to the coolant; heat was transferred to the coolant via convection. A radial conduction model was established to gauge the temperature distribution within both the FE and ME. Several assumptions were made such as no heat conduction in the axial direction and uniform power distribution. A 1-D model was used to evaluate radial temperature distribution in the annular fuel and annular moderating elements. A 1-D conduction system was also used for the analysis of the ME. Although axial conduction was ignored in the study, axial convection and pressure drop was modelled. The analysis incorporated friction, form, acceleration and gravity pressure losses. The Newton-Raphson iterative method was used to observe pressure losses.

Previous T/H analysis resulted in obtaining the chamber temperature and pressure at the nozzle inlet. The throat conditions were approximated via the core outlet conditions. Experimental data equated with theoretical as the divergent part of the nozzle depicted pressure decrease and velocity increase under supersonic conditions. The area ratio was used to calculate pressure, P_x, and temperature, T_x, values at various locations, x, along the nozzle. Again, by iterations, exit temperature, pressure and velocity values were obtained when T₂ = T_x, P₂ = P_x and v₂ = v_x.

Serpent was coupled with a T/H module (THERMO) for coupled analysis. The code utilized fixed point iterations with an initial guess of temperature and core density distribution. Serpent then transferred the power distribution to the THERMO module which updated temperature and density distributions. The average conditions were then sent to the solver which calculated the thrust and I_sp.

Neutronic Sensitivity Studies

Tungsten enrichment was set at 95% for this study. ¹⁸⁴W was selected because it was the most natural abundant form of tungsten’s five isotopes, and had the lowest thermal neutron capture cross section. The study also utilized ²³⁵U with an enrichment of 19.75%. As previously indicated, one of the aims was to create a critical reactor/core. Figures 4 and 5 below showed the relationship between k_eff vs ¹⁸⁴W enrichment and k_eff vs ²³⁵U.

Fig. 4. K_eff as a function of ¹⁸⁴W enrichment [1]

Fig. 5. K_eff as a function of ²³⁵U enrichment. [1]

ZrH₂ and LiH were considered in the selection of a suitable moderator; however, ZrH₂ was ultimately chosen because it required no enrichment, had a higher weight density than LiH and would produce a more compact core.

The first case of trade-offs considered in the study arose with regards to the active core and reflector length. The relationship between k_eff and the active core height was investigated with ME-FE ratios of 1.85 and 3. A trade-off between the reduction of core mass and addition of the reflector mass was used to determine the optimum length of the reflector. The second case of reflector trade-off arose between reflector savings and increased reflector size.

T/H Sensitivity Studies

Thermal hydraulic analysis tested the constraints on a full a core. The core was studied as one T/H channel separated into 30 axial layers. An automated tool was created with parameters: active core height, ME-FE ratio and total number of elements within the core. Several core variations were analyzed before focusing on axial power distribution. Core variations were made until the fuel attained a pre-determined maximum center line fuel of 3100 K. This approach was applied to arbitrary core configurations: large core – 1327 elements, medium core – 1069 elements and small core – 835 elements. T/H analyses concluded the following results:

-          Longer cores with lower ME-FE ratio produced more thrust and I_sp.

-          The total number of fuel elements resulted in greater heat flux and hence higher I_sp.

-          Larger cores produced more power and thrust.

Design and Analysis of a Single Stage to Orbit Nuclear Thermal Rocket Reactor Engine

This paper described several NTR reactor engines to be used for the single stage to orbit launch. T/H and rocket engine analyses were utilized and resulted in specific impulses, I_sp greater than 800 s. Each reactor contained a 0.4 m diameter core with hexagonal tungsten cermet FE. Radial and axial beryllium reflectors as well as eight boron carbide control drums surrounded the core. Current research at the Center for Space Nuclear Research (CSNR) aims at manufacturing a new type of NTR fuel.

A T/H model was used to obtain the hydrogen exhaust temperature values from the NTR core. Core geometries were then refined using the Monte Carlo N-Particle (MCNP5) model.

Thermal Hydraulic Analysis

A T/H model was used to ensure the fuel did not reach melting temperature and that it reached the desired exit temperature. The tungsten cermet fuel used contained 60 vol% W-25Re alloy and 40 vol% UN enriched to 97% ²³⁵U.

Fig. 6. Dimensions of the Hexagonal Tungsten-Cermet FE. [2]

Large amounts of matrix material (MM) provided a high thermal conductivity across the cermet FE; however, the size of the reactor core was increased. The heat transfer model used in this research was derived from the simple annulus model. In developing the model, the coolant temperature was found to be a function of axial position.

Fig. 7. Simple Annulus Approximation Used. [2]

Maximum fuel temperature occurred at the outer rim of the annulus and it was mandatory that it remain below the melting temperature, 3350 K. The temperature rise across the fuel between the bulk coolant and the outer edge of the annulus was determined by implementing thermal resistances for the coolant and fuel. $\partial$

T from the edge of the annulus to the coolant channel wall was independent of the fuel radius.

Rocket Engine Performance Analysis

Neutronics results indicated that at 97% enrichment UN, the reactors’ optimum critical diameter was around 0.4 m. A cylindrical reactor core with a diameter of 0.4 m was formed by filling a 0.4 m diameter cylinder with hexagonal tungsten cermet fuel elements. The reactor core contained 367 FE; areas which were not able to hold full FE were filled with graphite spacers. The optimal core length was found by varying the reactor’s length between 0.4 and 1.20 m. The I_sp produced by different NTRs were then determined by adjusting the mass flow rate of the hydrogen.

Fig. 8. Reactor power as a function of specific impulse and core length. [2]

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Further analysis inferred that by decreasing reactor power and increasing the core length, the specific impulse increased; however, decreasing the reactor power decreased the thrust provided.

Fig. 4. Thrust as a function of specific impulse, core length and reactor power. [2]

Rocket Neutronics Analysis

A neutronics analysis was used to determine the required fuel loading, reflector thickness, and control drum configuration. The MCNP transport code determined the multiplication factor for each NTR core configuration.

Personal Assessment

Low-Enriched Cermet-Based Fuel Options for a Nuclear Thermal Propulsion Engine

The paper presented a wholistic background which led up to the details of the current research. The goal of the paper was stated early on which made it clear to understand the sequence in which various topics were addressed. Detailed descriptions of the engine and core configurations were given that led to reader to understanding the why aspect of having these various structures. Calculations were brief and concise enabling for a mathematical conceptualization of the analyses. Relationships between variables and constraints were clearly identified and analyzed.

Though the actual writeup of the paper was near perfect there were several flaws in the experimental procedure that could not be overlooked.  Early on it was highlighted that lithium-6 underwent neutron capture, but the research did not account for the production of tritium during the (n, $\alpha$

) reaction. Why was this not accounted for? This inferred this part of the experiment was led to be more idealistic than realistic. With regards to the thermal conductivity calculations, ZrC “shows a nearly constant value”. This is true; however, it does vary over a small range. Should this not be accounted for?  Another concern would be the axial convection and pressure drop models utilized. Axial heat conduction was ignored; only convection was accounted for. Likewise, with the radial configuration, only conduction was accounted for and convection was ignored. All forms of heat transfer should be accounted for during these types of procedures.

Several adjustments would be made if I were to conduct this experiment:

• More precise and realistic mathematical models should be used rather than just having models based on estimates.
• A step forward would be to conduct thorough investigation of the expander cycle and other external systems. Mass of the propulsion system was assumed to be proportional to the power of the reactor in this experiment when this will ideal play an important role in going forward. Is the external system too large or too small to continue with this exact method research?

Design and Analysis of a Single Stage to Orbit Nuclear Thermal Rocket Reactor Engine

The paper provided a comprehensive summary of the steps taken to achieve a suitable fuel source. The research revived work done and applied a modern concept to analyze and further advance the study done more than was previously possible. The paper was written such that a person with little understanding of the topic can fully get a grasp of the work being done. This is crucial into influencing future potential engineers as the previous paper fell short in this aspect.  

Again, there are several points that must be pointed out. Most of references that this paper is centered on is fairly outdated. Though still relevant, it would be nice to incorporate more recent research done in the field as there has been substantial advancements in the field. A good portion of section 2 focused on detailed steps taken into deriving the model. This was rather cumbersome and felt like an unnecessary calculus tutorial. Briefing, by just highlighting the major steps and stating assumptions would be essential enough.

The major concern that I have with this paper entails both my ‘why question’ and adjustments I would make in furthering this research. Why is enriched 97% ²³⁵U being used? The time of highly enriched uranium has gone as it is now a political climate barrier. I understand that the sole purpose of the research is to simply put a payload into orbit and not one to deal with deep space missions. That however does not lessen any of my concerns with this research as it still impacts our Earthly atmosphere regardless. A step forward would be firstly stepping backward and reconsider the UN composition being utilized. Incorporating research similar to that of the previous paper may go a long way as several groups can work on various designs based off one solid foundation.

References

• Gates, J. T., et al. “Low-Enriched Cermet-Based Fuel Options for a Nuclear Thermal Propulsion Engine.” Nuclear Engineering and Design, vol. 331, 2018, pp. 313-330. [1]
• Labib, Satira, and Jeffrey King. “Design and Analysis of a Single Stage to Orbit Nuclear Thermal Rocket Reactor Engine.” Nuclear Engineering and Design, vol. 287, 2015, pp. 36-47. [2]