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What follows are my personal notes developed as I’ve struggled to understand the Statistical Theory of Fission Chains. The theoretical work detailed here is principally the work of R. P. Feynman, G. I. Bell, K. Boehnel, W. Hage, D. M. Cifarelli, M. Prasad, and N. Snyderman, plus a little of my own dablings around the edges.
The 95 Zr manuscript describing our experimental approach to measure branching ratios of long-lived fission products has been completed and successfully published in NIM A. This article is an important milestone in this effort and will serve as a great reference in any future reports describing results of other measurements we plan to perform. We have also expended our team, and Daniel Hoff joined us as a postdoc. We have been training him on how to use our analysis and simulation tools, as well as bringing him up to speed on the technical details of our experiment.
This presentation includes an overview of the process, example inputs, and example results for the Fission Meter and its response.
NASA’s Space Technology Mission Directorate (STMD) intends to demonstrate a 10 kWe fission surface power (FSP) system on the surface of the Moon as part of a Capability Demonstration Mission. STMD tasked a joint NASA-DOE team to ‘outline an FSP reactor concept that can be readied for 2027 launch,’ and serves ‘as a pathfinder for future power modules for Mars.’ Tables 1 and 2 respectively list FSP performance requirements and design guidance provided for the study. DOE is the leading agency for the reactor module, reactor controls and radiation shielding, and shares joint responsibility for the overall FSP system. This report summarizes an analysis of alternative reactor classes and shielding configurations carried out by the DOE national laboratories. Representative reactor and shielding configurations were based on heritage and are in alignment with the industry initiatives. This report presents no recommendations on the appropriate choice of fuel or class of reactor that NASA should pursue. Instead the focus here is to provide a summary and findings of analyses carried out by DOE national laboratory personnel.
Oak Ridge National Laboratory is developing advanced low-Cr oxide dispersion strengthened (ODS) FeCrAl alloys for accident tolerant fuel (ATF) cladding. This report presents the characterization of two new ODS FeCrAl tubes by using powder commercially mechanically alloyed by Zoz GmbH, which underwent high-precision tube rolling with two different annealing schedules. This work shows that the ODS FeCrAl tube without recrystallization exhibits consistent mechanical behavior in comparison with a previously produced tube. The second tube, which underwent a full recrystallization step before the final tube rolling step, showed some circumferential cracking and exhibited limited ductility in the axial tube tensile tests. Based on current and prior work on the ODS FeCrAl alloy system, this report compares the current state of the scientific literature on the alloy class with that of wrought FeCrAl without oxide additions. Then, this report makes recommendations for future research directions that should be undertaken to increase the technology readiness level for ODS FeCrAl as a viable ATF candidate material. This report has been submitted as fulfillment of milestone M3FT-20OR020202064 titled, “STEP Report on Advanced ODS FeCrAl Alloys for Fission Applications” for the US Department of Energy Office of Nuclear Energy’s Advanced Fuel Campaign of the Fuel Cycle R&D program.
This work supports the recovery of high-assay low-enriched uranium (HALEU, uranium enriched in 235 U to >5% and <20%) from used nuclear fuel through a process based on the zirconium extraction (ZIRCEX) process. The ZIRCEX process starts with the dry head-end removal of zirconium or aluminum cladding from the nuclear fuel via chlorination or hydrochlorination, which results in the separation of cladding from the fuel as the metal chloride (e.g., aluminum trichloride or zirconium tetrachloride). The hydrochlorination process is included in the Material Recovery Pilot Plant (MRPP), a ZIRCEX-based demonstration plant at Idaho National Laboratory (INL). The experiments outlined in this report assessed fission product (FP) volatility and transport under the hydrochlorination conditions used to separate cladding from used nuclear fuel.
We continued our work on preparation for the upcoming branching ratio measurements. The CARIBU facility has not been delivering radioactive beams since January this year due to an issue with the magnet that is used for mass separation. The parts needed to be replaced to fix the issue has been since ordered, shipped from Germany, and installed at CARIBU. The CARIBU is now operational and the first experiments that are ran will be those that had to be cancelled in January due to the facility complications, and we are in a position to schedule our next sample collection. In the meantime, we have been continuing work on data analysis of fission product mass measurements in the Canadian Penning Trap. This analysis is now complete. We continue our outreach to recruit a graduate student who would be interested in working with us on this project. In the past month, we have a conversation with a potential student at UC Berkeley and Texas A&M. There students are still to decide if they would accept the offer to these graduate schools.
The DOE Nuclear Criticality Safety Program has funded a multi-year effort to measure the 240 Pu(n,f) Prompt Fission Neutron Spectrum at Chi-Nu. This project is a joint LANL-LLNL effort, involving the preparation of 12 240 Pu foils and construction of a Parallel-Plate Avalanche Counter at LLNL, with subsequent measurements of the 240 Pu(n,f) PFNS at the Chi-Nu beamline at the Los Alamos Neutron Science Center’s Weapons Neutron Research facility (LANSCE/WNR). We expect the PPAC to be ready in time to perform the measurements this run cycle.
The report details the joint LANL/LLNL measurement of the 240 Pu( n,f ) prompt fission neutron spectrum (PFNS) for incident neutron energies from 1–20 MeV. A summary of the experimental setup, including details of the parallel-plate avalanche counter (PPAC) used for this experiment, beam time awarded and received, and adaptations made for specifically for this experiment are described in Sec. II. Some preliminary results obtained with a portion of the total available statistics from the executed experiment are then described in Sec. III with a discussion of potential benefits and losses if more data were collected. Finally, some primary tasks remaining to convert these data to final results are discussed in Sec. IV.
The Ultracold Neutron (UCN) source at LANSCE is currently the source of the highest density of UCN in the world. We intend to use innovative new solutions to boost our UCN density in order to maintain our leadership in this competitive field. Prior studies suggest that a fission plate neutron multiplier made from low enriched uranium (LEU) installed between our spallation target and UCN source moderator can boost the density by up to a factor of ten. Here we present a prototype of this idea that uses an existing one liter volume of LEU that can boost the UCN density by roughly a factor of 2. Additional simulation studies presented here also suggest that we may find more gains by replacing some of the moderator or reflector geometries with beryllium.
Molten Salt Reactors (MSRs) are one of six Generation IV reactor designs currently under development around the world. Because of the unique operating conditions of MSRs, which include molten fuel and the continuous removal of gaseous fission products during operation, work was performed to analyze the potential impact of emissions on the International Monitoring System (IMS) of the Comprehensive Nuclear-Test-Ban Treaty (CTBT). Simulations were performed to predict the production of IMS-relevant radionuclides in four MSR designs operating under two scenarios – (1) a sealed reactor with releases only during operational shutdown and (2) continuous reprocessing or sparging of the fuel salt. From these production estimates the radioxenon and radioiodine signatures were extracted and compared to three current reactor designs (PWR, BWR, RBMK). In the cases where continuous reprocessing of the fuel salt occurred both the radioxenon and radioiodine signatures are nearly indistinguishable from a nuclear explosion. Estimates were also made of the potential emission rate of radioxenon for three reactor designs and it was found that MSRs have the potential to emit radioxenon isotopes at a rate of 1015 – 8×1016 Bq/d for 133Xe if no abatement is used. An assessment was also made of activation products using a candidate fuel salt (FLiBe) mixed with corrosion products for the Thorium Molten Salt Reactor (TMSR-LF1)
We have addressed another round of referees’ comments and resubmitted our mass measurement of the isomer and ground state of 128 Sb to Phys. Rev. Letters. One of the referees recommended publication so we anticipate this will be the last round with the referees. This paper focuses on the astrophysical implications of our measurement, as we can directly measure the excitation energy of the isomer along with other observables like the isomer to ground state ratio produced in the fission of 252 Cf. With our summer student Asha, we were able to determine that there was an error in our Gas Counter simulations, which has been corrected and is now implemented in our 111 Ag simulations. This report describes the progress on this effort from July until August 2023. We have paused the work on this project in the last 7 Weeks while the PI was on leave. We are resuming the work on this effort in early October.
The linear and nonlinear optical properties of glass and sapphire are investigated to determine the material’s suitability to support optical instrumentation for advanced fission reactors. A post-irradiation examination experiment consisting of a Z-scan, measuring nonlinear optical absorption and refraction, and a spectrometer, measuring linear optical absorption was developed and used. Radiation-induced negative nonlinear optical absorption was observed for the first time in fused silica, sapphire, and borosilicate glass under both types of thermal annealing conditions and the respective effects of thermal annealing are reported.
The projection is proceeding on schedule. The remaining budget for the year is about ~$\$$30k and that should be sufficient to accomplish tasks scheduled for the remaining quarter. The anticipated remaining deliverable is a report with results for 111 Ag long-lived fission product.
Overall fission product (FP) release will be an important consideration for the licensing and deployment of advanced reactors utilizing tristructural isotropic (TRISO) fuels. This work focuses on enhancing and applying the BISON models needed to predict FP transport within TRISO particles and particle failure probability, both of which factor directly into release predictions. Specifically, this report details (1) the development of the models needed to predict palladium (Pd) conservation at the engineering scale and the application of those models to characterize Pd fluxes for input into a mechanistic multiscale model for Pd penetration; (2) the refinement of sorption mass transfer models and the development of models for trapping in porous layers, which were applied and compared to particle scans from AGR-2 to provide proof of concept for a method of particle-scale validation that may reduce uncertainties compared to compact-scale validation using data from integral effects tests; (3) the development of a failure-statistics-informed, mesh-independent methodology for applying smeared cracking, enabling further study of the localized multiphysics behaviors associated with cascading particle failure mechanisms; and (4) the preliminary characterization of those coupled multiphysics particle failure behaviors using smeared, nonretentive diffusivities to provide a baseline for future study and to guide ongoing engineering applications.
As a part of DOE NE High Burnup Spent Fuel Data Project, Oak Ridge National Laboratory (ORNL) is performing destructive examinations (DEs) of high burnup (HBU) (>45 GWd/MTU) spent nuclear fuel (SNF) rods from the North Anna Nuclear Power Station operated by Dominion Energy. The SNF rods, called sister rods or sibling rods, are all HBU and include four different kinds of fuel rod cladding: standard Zircaloy-4 (Zirc-4), low-tin (LT) Zirc-4, ZIRLO, and M5. The DEs are being conducted to obtain a baseline of the HBU rod’s condition before dry storage and are focused on understanding overall SNF rod strength and durability. Composite fuel and defueled cladding will be tested to derive material properties. Although the data generated can be used for multiple purposes, one primary goal for obtaining the post-irradiation examination data and the associated measured mechanical properties is to support SNF dry storage licensing and relicensing activities by (1) addressing identified knowledge gaps and (2) enhancing the technical basis for post-storage transportation, handling, and subsequent disposition. This appendix documents the status of the ORNL Phase 1 DE activities related to fission gas sample analysis, fuel burnup analysis, and fuel isotopic analysis of selected sister specimens in Phase 1 of the sister rod test program.
A quasi-monoenergetic neutron beam was used to measure the neutron-induced fission cross-section ratios for 238 U(n,f)/ 235 U(n,f) (Table 1) and 239 Pu(n,f)/ 235 U(n,f) (Table 2). These results are plotted in comparison with data from the fissionTPC and ENDF/B-VIII.0 in Figs.1-3. For each cross section ratio, the total, systematic, and statistical uncertainties are listed, along with the total beamtime required.
Accurate prediction of nuclide compositions and neutron and gamma sources and spectra for fresh and spent nuclear fuel through computational modeling and simulations is the basis for safeguards instruments design, optimization, and calibration, and for validation of the measured responses. The accuracy of these simulations can be significantly impacted by uncertainties in input parameters to the applied computer code. Input parameters are, for example, dimensions, material compositions, and temperatures of the model. Additionally, important but sometimes neglected input parameters for simulations of nuclear systems are the nuclear data that include nuclear reaction cross sections, fission product yields, and decay data. It has been shown that uncertainties of calculated results are dominated by uncertainties of these nuclear data.