Engineering topics

Bays, Samuel

Publications and source records attributed to Bays, Samuel.

7 records

DOE OSTI2012

Simple Coupling of Reactor Physics Effects and Uncertain Nuances

The "Simple Coupling of Reactor Physics Effects and Uncertain Nuances" (SCORPEUN) code is a simple r-z 1-group neutron diffusion code where each r-mesh is coupled to a single-flow-channel model that represents all flow-channels in that r-mesh. This 1-D model assesses q=m*Cp*deltaT for each z-mesh in that channel. This flow channel model is then coupled to a simple 1-D heat conduction model for ascertaining the peak center-line fuel temperature in a hypothetical pin assigned to that flow channel. The code has property lookup capability for water, Na, Zirc, HT9, metallic fuel, oxide fuel, etc. It has linear interpolation features for microscopic cross-sections with respect to coolant density and fuel temperature. ***This last feature has not been fully tested and may need development***. The interpolated microscopic cross-sections are then combined (using the water density from the T/H calculation) to generate macroscopic diffusion coefficient, removal cross-section and nu-sigmaF for each r-z mesh of the neutron diffusion code.

Bays, Samuel↗

DOE OSTI2012

Report on Potential Advantages and Uncertainties of the Thorium Seed Blanket Unit Fuel Concept

Thorium is a naturally occurring radioactive element with potential application in the nuclear fuel cycle. Thorium has no fissile isotopes, meaning that unlike uranium it cannot be enriched to produce a viable nuclear fuel. Natural thorium is almost entirely comprised of the isotope 232 Th and a microscopic amount of 228 Th that is created from the radioactive decay of 232 Th. Small amounts of other thorium isotopes can be detected from the decay of uranium that often occurs with thorium in nature. During the early stages of the development of nuclear power in the United States it was believed that uranium supplies would be insufficient and that breeding would be required to fuel the envisioned nuclear infrastructure. The use of 232 Th in the nuclear fuel cycle would play a role equivalent to 238 U. Both elements are fertile and can absorb a neutron to transmute into a fissile isotope capable of sustaining a chain reaction. In the case of 238 U the fissile isotope produced is 239 Pu. 232 Th produces the less well-known fissile isotope 233 U. This report provides a brief overview of light water reactor thorium technology, and an assessment of the technology proposed by Lightbridge based on available open literature and information provided by Lightbridge Corporation. It serves to identify gaps in the current understanding of this technology, and to identify areas where further work may be required or beneficial. Analysis of the fuel cycle was not included in the scope of this report. Based on the data available and the scope of this assessment, this document cannot draw definitive conclusions about this particular technology, and should not be understood to be comprehensive or to make a recommendation, either positive or negative, on the suitability of this fuel cycle for commercial use.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2011

Benchmark of Advanced Burner Test Reactor Model Using MCNPX 2.6.0 and ERANOS 2.1

Significant research is currently being performed whereby fast reactor cores have been designed to burn transuranic materials reducing the volume and long-term radiotoxicity of spent nuclear fuel. These core and depletion models depend on various computer codes. This research used MCNPX 2.6.0 and ERANOS 2.1 to model a standard 250MWt Advanced Burner Test Reactor (ABTR) core. The intent was to benchmark criticality and burnup results from a stochastic Monte Carlo code and a deterministic depletion code using a standard ABTR model created by Argonne National Laboratory. Because each of these codes solve the transport and burnup problem differently, there is a need to benchmark the core models in order to verify results and identify root causes for significant differences in results between codes. Flux calculations in ERANOS were performed using diffusion theory, Legendre polynomial approximations (using the VARIANT module) and discrete ordinates methods. The k-effective for the higher-order transport models remained within 1000 pcm of the MCNPX model. The difference between the total heavy nuclide mass balance in ERANOS using the various flux calculations and the MCNPX depletion model was less than 0.4% out to a burnup of 1095 days (67.45 GWd/MTHM). For individual heavy nuclides, the depletion models closely matched (< 5.0 % difference) throughout the depletion for isotopes of Uranium, Neptunium and Plutonium and most of the higher transuranics. Notable exceptions were 242Am, 242Cm, 243Cm and 246Cm where differences ranged from 0.1 – 0.2% after 26 days and increased to 11 - 136% at 1095 days.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2011

Actinide destruction and power peaking analysis in a 1000 MWt advanced burner reactor using moderated heterogeneous target assemblies

The purpose of this research was to determine the effect of moderated heterogeneous subassemblies located in the core of a sodium-cooled fast reactor on minor actinide (MA) destruction rates over the lifecycle of the core. Additionally, particular emphasis was placed on the power peaking of the pins and the assemblies with the moderated targets as compared to standard unmoderated heterogeneous targets and a core without MA targets present. Power peaking analysis was performed on the target assemblies and on the fuel assemblies adjacent to the targets. The moderated subassemblies had a marked improvement in the overall destruction of heavy metals in the targets. The design with acceptable power peaking results had a 12.25% greater destruction of heavy metals than a similar ex-core unmoderated assembly. The increase in minor actinide destruction was most evident with americium where the moderated assemblies reduced the initial amount to less than 3% of the initial loading over a period of five years core residency. In order to take advantage of the high minor actinide destruction and minimize the power peaking effects, a hybrid scenario was devised where the targets resided ex-core in a moderated assembly for the first 506.9 effective full power days (EFPDs) and were moved to an in-core arrangement with the moderated targets removed for the remainder of the lifecycle. The hybrid model had an assembly and pin power peaking of less than 2.0 and a higher heavy metal and minor actinide destruction rate than the standard unmoderated heterogeneous targets either in-core or ex-core. The hybrid model has a 54.5% greater Am reduction over the standard ex-core model. It also had a 27.8% greater production of Cm and a 41.5% greater production of Pu than the standard ex-core model. The radiotoxicity of the targets in the hybrid design was 20% less than the discharged standard ex-core targets.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2010

Transportation Implications of a Closed Fuel Cycle.

Transportation for each step of a closed fuel cycle is analyzed in consideration of the availability of appropriate transportation infrastructure. The United States has both experience and certified casks for transportation that may be required by this cycle, except for the transport of fresh and used MOX fuel and fresh and used Advanced Burner Reactor (ABR) fuel. Packaging that had been used for other fuel with somewhat similar characteristics may be appropriate for these fuels but would be inefficient. Therefore, the required neutron and gamma shielding, heat dissipation, and criticality were calculated for MOX and ABR fresh and spent fuel. Criticality would not be an issue, but the packaging design would need to balance neutron shielding and regulatory heat dissipation requirements.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2010

FUEL CYCLE ISOTOPE EVOLUTION BY TRANSMUTATION DYNAMICS OVER MULTIPLE RECYCLES

Because all actinides have the ability to fission appreciably in a fast neutron spectrum, these types of reactor systems are usually not associated with the buildup of higher mass actinides: curium, berkelium and californium. These higher actinides have high specific decay heat power, gamma and neutron source strengths, and are usually considered as a complication to the fuel manufacturing and transportation of fresh recycled transuranic fuel. This buildup issue has been studied widely for thermal reactor fuels. However, recent studies have shown that the transmutation physics associated with "gateway isotopes" dictates Cm-Bk-Cf buildup, even in fast burner reactors. Assuming a symbiotic fuel relationship with light water reactors (LWR), Pu-242 and Am-243 are formed in the LWRs and then are externally fed to the fast reactor as part of its overall transuranic fuel supply. These isotopes are created much more readily in a thermal than in fast spectrum systems due to the differences in the fast fission (i.e., above the fission threshold for non-fissile actinides) contribution. In a strictly breeding fast reactor this dependency on LWR transuranics would not exist, and thus avoids the introduction of LWR derived gateway isotopes into the fast reactor system. However in a transuranic burning fast reactor, the external supply of these gateway isotopes behaves as an external driving force towards the creation and build-up of Cm-Bk-Cf in the fuel cycle. It was found that though the Cm-Bk-Cf concentration in the equilibrium fuel cycle is dictated by the fast neutron spectrum, the time required to reach that equilibrium concentration is dictated by recycle, transmutation and decay storage dynamics.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2010

On the Criticality Safety of Transuranic Sodium Fast Reactor Fuel Transport Casks

This work addresses the neutronic performance and criticality safety issues of transport casks for fuel pertaining to low conversion ratio sodium cooled fast reactors, conventionally known as Advanced Burner Reactors. The criticality of a one, three, seven and 19-assembly cask capacity is presented. Both dry “helium” and flooded “water” filled casks are considered. No credit for fuel burnup or fission products was assumed. As many as possible of the conservatisms used in licensing light water reactor universal transport casks were incorporated into this SFR cask criticality design and analysis. It was found that at 7-assemblies or more, adding moderator to the SFR cask increases criticality margin. Also, removal of MAs from the fuel increases criticality margin of dry casks and takes a slight amount of margin away for wet casks. Assuming credit for borated fuel tube liners, this design analysis suggests that as many as 19 assemblies can be loaded in a cask if limited purely by criticality safety. If no credit for boron is assumed, the cask could possibly hold seven assemblies if low conversion ratio fast reactor grade fuel and not breeder reactor grade fuel is assumed. The analysis showed that there is a need for new cask designs for fast reactors spent fuel transportation. There is a potential of modifying existing transportation cask design as the starting point for fast reactor spent fuel transportation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗