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At least 829 records · Page 46

Energy dependence of $\textit{p}$ + 232 Th fission mass distributions: Mass-asymmetric standard I and standard II modes, and multichance fission

Background: The predominant mass-asymmetric fission of actinide nuclides occurs mainly through the so-called standard I and standard II modes. Though understood to be caused by shape-dependent shell structures encountered between the fission barrier deformation and scission, the most relevant shell gaps are still not firmly established. The standard I mode had been associated with the spherical doubly magic 132 Sn, and thus the $\textit{Z}$ = 50 proton shell, but recently it has been proposed that standard I and standard II are associated with quadrupole and octupole deformed gaps at $\textit{Z}$ = 52 and 56, respectively. Purpose: In this work, we investigate how the relative probabilities of the standard I and standard II modes vary with excitation energy near threshold, probing where the two modes bifurcate. Methods: The Australian National University Heavy Ion Accelerator Facility and CUBE fission spectrometer have been used to measure fission mass distributions for the $\textit{p}$ + 232 Th reaction (forming 233 Pa) at closely spaced bombarding energy intervals from 6.5 to 28 MeV. Results: A model-independent analysis of the energy dependence of the shape of the mass-asymmetric peak shows a strong dependence of the standard I and standard II relative probability on excitation energy near threshold. Conclusions: The results are consistent with the standard II mode having a lower fission barrier than standard I in 233 Pa, with the latter increasing continually in relative probability above its barrier energy. It is concluded that multichance fission, in particular last chance fission, plays a strong role in determining the observed energy dependence of all fission modes.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Structure of high-lying excited states in 47 Ca with ℓ ≥ 3 populated in fast-beam 𝛾-ray-tagged one-neutron pickup reactions

Located just one neutron removed from doubly magic, stable 48 Ca, the adjacent isotope 47 Ca has been studied extensively with neutron-removing transfer reactions and interpreted within a shell-model picture. Neutron-adding transfer reactions onto the stable 46 Ca, however, have not been performed in nearly 60 years and never with the aid of high-resolution 𝛾-ray spectroscopy. Here, we report on the 12 C⁡( 46 , Ca 47 Ca+𝛾)⁢𝑋 one-neutron pickup reaction from 12 C onto 46 Ca at more than 50 MeV/nucleon, a regime of orbital angular momentum mismatch that selectively populates 47 Ca final states based on ℓ ≥ 3 neutron configurations. Newly observed excited states and 𝛾-ray transitions are discussed in comparison to 𝑓⁡𝑝 shell-model calculations in a variety of configuration spaces. Furthermore, it is shown that the nature of the observed spectrum of final states can be well interpreted, albeit with a significant fragmentation of the 𝜈⁢0⁢𝑓 5/2 strength observed in the measurement that is not borne out in the structure calculations.

39 ≤ A ≤ 58↗

Crossing N = 28 Toward the Neutron Drip Line: First Measurement of Half-Lives at FRIB

Here, new half-lives for exotic isotopes approaching the neutron drip-line in the vicinity of N~28 for Z=12–15 were measured at the Facility for Rare Isotope Beams (FRIB) with the FRIB decay station initiator. The first experimental results are compared to the latest quasiparticle random phase approximation and shell-model calculations. Overall, the measured half-lives are consistent with the available theoretical descriptions and suggest a well-developed region of deformation below 48 Ca in the N=28 isotones. The erosion of the Z=14 subshell closure in Si is experimentally confirmed at N=28, and a reduction in the 38 Mg half-life is observed as compared with its isotopic neighbors, which does not seem to be predicted well based on the decay energy and deformation trends. This highlights the need for both additional data in this very exotic region, and for more advanced theoretical efforts.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Design, Fabrication and Start-up of Test Facilities for the LAWPS Full-Scale IX Column Test and Engineering Scale Test - 17418

The U.S. Department of Energy (DOE), Office of River Protection's (ORP) primary mission is to retrieve and treat Hanford's tank waste and close the Tank Farms to protect the Columbia River. Mixed radioactive waste is stored in 177 underground tanks at the Hanford Site as reported in DOE/ORP 2003-02, Environmental Impact Statement for Retrieval, Treatment, and Disposal of Tank Waste and Closure of the Single Shell Tanks at the Hanford Site, Richland WA, Inventory and Source Term Data Package. The 177 underground tanks are estimated to contain about 56 million gallons of waste. The Low Activity Waste Pretreatment System (LAWPS) Project provides for the early production of immobilized low activity waste (ILAW) by feeding LAW directly from Tank Farms to the Hanford Waste Treatment Plant's (WTP's) LAW Facility, bypassing the Pretreatment Facility. Prior to the transfer of feed to the WTP LAW Vitrification Facility, tank supernatant waste will be pretreated in the LAWPS to meet the WTP LAW waste acceptance criteria. The LAWPS will also facilitate the return of secondary liquid wastes from the WTP LAW Vitrification Facility to the Tank Farms. In November of 2015 DOE's Tank Operations Contractor, Washington River Protection Solutions (WRPS), awarded a contract to the AECOM Team to design, fabricate and operate two test platforms to generate performance data at two different scales and demonstrate technology maturation. The first platform models the ion exchange (IX) columns at full scale, and the second platform models the integrated flowsheet (cross flow filter (CFF) and IX) at 1/9. scale. Both test platforms underwent shakedown and start-up testing in August 2016 and moved into actual testing and NQA-1 data generation shortly thereafter. This paper covers the engineering design, procurement, construction, installation, and acceptance testing of the test platforms (which are now operational) under an aggressive fast track schedule, and initial test data as available. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

System Configuration Evaluation for Process Settling of Hanford Waste Solid Particles

Direct Feed High-Level Waste (DFHLW) is a potential flowsheet operations approach to initiating high-level waste (HLW) vitrification prior to completion of the Hanford Waste Treatment and Immobilization Plant (WTP) Pretreatment Facility. A settle/decant process has been proposed to concentrate solids prior to delivery to the WTP HLW Facility during DFHLW operations, wherein the solids in a settled layer would be remixed with the supernatant liquid remaining after decanting operations to provide the feed at required solids concentrations. Settling would be used in lieu of purpose-built filtration or other solids separation equipment. Pacific Northwest National Laboratory (PNNL) is providing baseline technical support to the Washington River Protection Solutions (WRPS) Flowsheet Integration group. To support planning for DFHLW, WRPS previously requested that PNNL evaluate the current data set available to predict the time needed for HLW solids to settle and the solids concentration and strength of that settled layer, to identify gaps in the understanding and predictive capability of HLW solids waste settling times, and to provide scoping estimates of the potential settling times. Eight technical gaps were identified for predicting settling times and characteristics of the formed sediment layers. In addition to the data gaps, an overarching observation was made that there is significant variation in behavior of settling rate and settled layer data. The settling time required to concentrate solids via a settle/decant process was determined from the limited data to have a difference of potentially more than a factor of 5,000 in the estimated settling times, varying from 0.2 to 1,060 days for example depending on process vessel depth and final sediment solids concentration. In contrast, successful processes of liquid-forward output streams resulting from in-tank settling and decanting forward liquid have been reported for operations conducted at the Hanford Site. The purpose of this current report is to further support DFHLW planning by evaluating double-shell tank (DST) and alternate vessel equipment and operational configurations to enable optimization of the settle/decant process to concentrate solids. Hanford waste processing behavior specific to liquid feed availability following a slurry transfer in a DST is summarized, including process stream characteristics and process equipment configurations. The performance of DST process equipment configurations is evaluated for possible improvements using computational fluid dynamics (CFD) and simple analytical models. Potential new vessel design(s) specific to enabling effective settle/decant processes, and cursory summary of other separate and inline solids separations processes, are also provided. The CFD results indicated that improvement in outflow solids concentration was promoted by a reduction in the slurry flow rate, angling the distributor nozzles downward, and lifting the transfer pump. The solid-liquid analysis evaluating particle trajectory confirmed that the potential for particle ingestion (in the transfer pump) was decreased with increased radial separation between the inlet and outlet (transfer pump inlet), decreased inlet flow, and decreased liquid density and viscosity for a neutrally buoyant inlet flow. An assessment was also made of the potential for inflow configuration changes to result in the discrete mounding or piling of solids within the tank. Based on the characterization of the settled waste to date, HLW sediments will be unlikely to sustain a substantial angle of repose to facilitate significant variations in the elevation of the settled solids.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

AGR-5/6/7 Release-to-Birth Ratio Data Analysis for Cycles 162B, 163A, 164A, and 164B

AGR-5/6/7 is the last of a series of Advanced Gas Reactor (AGR) experiments conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory in support of development and qualification of tristructural isotropic (TRISO) low-enriched fuel used in the high-temperature gas cooled reactor (HTGR). Each AGR test consists of multiple independently controlled and monitored capsules containing fuel compacts placed in a graphite cylinder shrouded by a steel shell. These capsules are instrumented with thermocouples (TC) that are embedded in the graphite, thus providing feedback for temperature control. AGR configuration and irradiation conditions are based on prismatic HTGR technology that is distinguished primarily through use of helium coolant, a low-power-density ceramic core capable of withstanding very high temperatures, and TRISO-coated particle fuel. The AGR tests provide valuable irradiation-performance data to support fuel process development, qualify fuel for normal operation and accident conditions, and support development and validation of fuel performance and fission-product transport models and codes. The release-to-birth ratios (R/B) for fission-gas isotopes (i.e., krypton and xenon) are calculated from release rates into the sweep-gas flow, measured by the high-purity germanium detectors used in the AGR Fission Product Monitoring System (FPMS) installed downstream from each irradiated capsule. Birth rates are calculated based on the fission power in the experiment and models of fission-product generation. Thus, the R/B is a measure of the ability of fuel kernel, particle coating layers, and compact matrix to retain fission-gas atoms, preventing their release into the sweep-gas flow, especially in the presence of initially defective particle and/or the event of particle-coating failures that occurred during the irradiation. For fission-gas isotopes, particle failure occurred when all coating layers are fractured that allow gaseous fission atoms to escape from a particle. In this case, the R/B per-exposed-kernel is used to compare fuel performance across experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Quantifying Uncertainties In Neutron-Alpha Scattering With Chiral Nucleon-Nucleon And Three-Nucleon Forces

We report that modern ab initio theory combined with high-quality nucleon-nucleon (NN) and three-nucleon (3N) interactions from chiral effective field theory (EFT) can provide a predictive description of low-energy light-nuclei reactions relevant for astrophysics and fusion-energy applications. However, the high cost of computations has so far impeded a complete analysis of the uncertainty budget of such calculations. Starting from NN potentials up to fifth order (N 4 LO) combined with leading-order 3N forces, we study how the order-by-order convergence of the chiral expansion and confidence intervals for the 3N contact and contact-plus-one-pion-exchange low-energy constants (c E and c D ) contribute to the overall uncertainty budget of many-body calculations of neutron- 4 He (n-α) elastic scattering. We compute structure and reaction observables for three-, four-, and five-nucleon systems within the ab initio frameworks of the no-core shell model and no-core shell model with continuum. Using a small set of design runs, we construct a Gaussian process model (GPM) that acts as a statistical emulator for the theory. With this, we gain insight into how uncertainties in the 3N low-energy constants propagate throughout the calculation and determine the Bayesian posterior distribution of these parameters with Markov-Chain Monte Carlo. We find rapidly converging n-α phase shifts with respect to the chiral order. With the adopted leading-order 3N force, calculations based on the NN interaction at N 4 LO of Entem, Machleidt, and Nosyk are unable to reproduce the experimental phase shifts in the 3/2 - channel within the estimated chiral truncation errors. Closer agreement with empirical data is found when using an older parametrization of the NN interaction at order N 3 LO, and the position and width of the P-wave resonances can be used to reduce the uncertainty of the 3N low-energy constants. The present results point to a lack of spin-orbit strength when the newer parametrization of the chiral NN force up to fifth order is combined with the leading-order 3N force. The inclusion of higher-order 3N-force terms may be required to recover the missing strength. GPMs can act as fast and accurate emulators of ab initio many-body calculations of low-energy scattering and reactions of light nuclei, opening the way to a robust quantification of theoretical uncertainties grounded in the description of the underlying chiral Hamiltonian.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Population of the giant pairing vibration

Background: We present that the giant pairing vibration (GPV), a correlated two-nucleon mode in the second shell above the Fermi surface, has long been predicted and expected to be strongly populated in two-nucleon transfer cross sections similar to those of the normal pairing vibration. Recent experiments have provided evidence for this mode in 14,15 C but, despite sensitive studies, it has not been definitively identified in either Sn or Pb nuclei where pairing correlations are known to play a crucial role. Purpose: Our aim is to test whether features inherent to the mixing of bound and unbound levels might account for this and to study the effect in a simple and intuitively clear approach. Method: We study the mixing of unbound levels in a set of toy models that capture the essential physics of the GPV, along with a more realistic calculation including distorted-wave Born approximation transfer amplitudes. Results: The calculated (relative) cross section to populate a simulated GPV state is effectively low, compared to the case of bound levels with no widths Conclusions: The mixing turns out to be only a minor contributor to the weak population. Rather, the main reason is the melting of the GPV peak due to the width it acquires from the low orbital angular momentum single-particle states playing a dominant role in two-nucleon transfer amplitudes. Lastly, this effect, in addition to a severe Q-value mismatch, may account for the elusive nature of this mode in (t,p) and (p,t) reactions.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Dose Rate Evaluation for the ES-3100 Package with HEU Content Using MCNP, ADVANTG, Monaco, and MAVRIC

This paper presents a comparative study of dose rate calculations for the ES-3100 package with highly enriched uranium (HEU) content for different source configurations using computer codes: (1) MCNP; (2) Automated Variance Reduction Generator (ADVANTG)/MCNP; (3) Monaco; and (4) Monaco with Automated Variance Reduction using Importance Calculations (MAVRIC). The Model ES-3100 package was developed at the Y-12 National Security Complex for domestic and international transportation of Type B fissile radioactive material. In this study, six different source configurations (i.e., solid cylinder, cylindrical hemi-shell, cylindrical shell, rectangular plate, cylindrical rod, or cylindrical segment form) having 36 kg of HEU metal inside the package containment vessel (based on configurations in the ES-3100/HEU safety analysis report for packaging) are evaluated. Dose rates at the package surfaces and 1 m from the package surfaces are calculated for these different source configurations. MCNP and Monaco cases are run without any biasing options to accelerate the convergence. Consistent Adjoint Driven Importance Sampling (CADIS) and Forward-Weighted CADIS (FW-CADIS) methods developed at the Oak Ridge National Laboratory are implemented in ADVANTG/MCNP and MAVRIC codes to accelerate the convergence. ADVANTG generates variance reduction parameters using Denovo code, and MCNP is used with the variance reduction parameters to accelerate the convergence. MAVRIC uses Denovo code to construct an importance map and a biased source distribution that are supplied to Monaco to accelerate the Monte Carlo simulation. The FW-CADIS option in ADVANTG and MAVRIC is used to accelerate the convergence in this study. The accelerated convergence cases (ADVANTG/MCNP and MAVRIC) are about 100 times faster with 100 times less particle simulation than those cases run without biasing options (analog MCNP and analog Monaco). The MCNP, ADVANTG/MCNP, Monaco, and MAVRIC calculated dose rates at the package surfaces and at 1 m from the package surfaces for the different source configurations are compared and are found to be in general agreement.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Proton Shell Evolution below Sn 132 : First Measurement of Low-Lying β -Emitting Isomers in Ag 123 , 125

The β-delayed γ-ray spectroscopy of neutron-rich 123,125 Ag isotopes is investigated at the Radioactive Isotope Beam Factory of RIKEN, and the long-predicted 1/2 - β-emitting isomers in 123,125 Ag are identified for the first time. With the new experimental results, the systematic trend of energy spacing between the lowest 9/2 + and 1/2 - levels is extended in Ag isotopes up to N = 78, providing a clear signal for the reduction of the Z = 40 subshell gap in Ag towards N = 82. Shell-model calculations with the state-of-the-art V MU plus M3Y spin-orbit interaction give a satisfactory description of the low-lying states in 123,125 Ag. The tensor force is found to play a crucial role in the evolution of the size of the Z = 40 subshell gap. The observed inversion of the single-particle levels around 123 Ag can be well interpreted in terms of the monopole shift of the π1g 9/2 orbitals mainly caused by the increasing occupation of ν1h 11/2 orbitals.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

The ν -process with Fully Time-dependent Supernova Neutrino Emission Spectra

The neutrino process that occurs in the outer stellar shells during a supernova explosion and involves neutrino–nucleus reactions produces a range of rare, stable, and radioactive isotopes. We improve previous ν-process studies by using, for the first time, the time-dependent neutrino emission spectra, as predicted from supernova simulations, rather than a simplified parametric description modeled after the neutron star cooling phase. In particular, our calculations use time-dependent neutrino spectra for all neutrino species, consider their deviation from a Fermi–Dirac distribution, and account for the neutrino emission from the neutrino burst and accretion phases. We find that the time-dependent treatment of the neutrino emission spectra results in higher yields for the selected nuclei produced by the ν process as compared to previous studies and to the approximation of assuming constant neutrino energies corresponding to the time-averaged mean energy radiated in each species. The effect is largest for nuclides produced by charged-current reactions. Our results reflect the dynamical competition between neutrino-induced reactions and the effect of the shock passage through the star. By varying the neutrino burst luminosity and the duration of the accretion phase, we study the impact of these early emission phases and their uncertainties on the ν-process nucleosynthesis. We find that the deviation of the neutrino spectra from a Fermi–Dirac distribution calculated in supernova simulations has a negligible effect on the ν-process yields.

79 ASTRONOMY AND ASTROPHYSICS↗

Gap Size Uncertainty Quantification in Advanced Gas Reactor TRISO Fuel Irradiation Experiments

The Advanced Gas Reactor (AGR)-3/4 experiment is the combination of the third and fourth tests conducted within the tristructural isotropic fuel development and qualification research program. The AGR-3/4 test consists of twelve independent capsules containing a fuel stack in the center surrounded by three graphite cylinders and shrouded by a stainless steel shell. This capsule design enables temperature control of both the fuel and the graphite rings by varying the neon/helium gas mixture flowing through the four resulting gaps. Knowledge of fuel and graphite temperatures is crucial for establishing the functional relationship between fission product release and irradiation thermal conditions. These temperatures are predicted for each capsule using the commercial finite-element heat transfer code ABAQUS. Uncertainty quantification reveals that the gap size uncertainties are among the dominant factors contributing to predicted temperature uncertainty due to high input sensitivity and uncertainty. Gap size uncertainty originates from the fact that all gap sizes vary with time due to dimensional changes of the fuel compacts and three graphite rings caused by extended exposure to high temperatures and fast neutron irradiation. Gap sizes are estimated using as-fabricated dimensional measurements at the start of irradiation and post irradiation examination dimensional measurements at the end of irradiation. Uncertainties in these measurements provide a basis for quantifying gap size uncertainty. However, lack of gap size measurements during irradiation and lack of knowledge about the dimension change rates lead to gap size modeling assumptions, which could increase gap size uncertainty. In addition, the dimensional measurements are performed at room temperature, and must be corrected to account for thermal expansion of the materials at high irradiation temperatures. Uncertainty in the thermal expansion coefficients for the graphite materials used in the AGR-3/4 capsules also increases gap size uncertainty. This study focuses on analysis of modeling assumptions and uncertainty sources to evaluate their impacts on the gap size uncertainty.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Leak Testing with Simulated Waste of Hanford Tank Farm Valves Utilized for Double Valve Isolation - 16108

The existing Hanford Tank Farms Waste Transfer System has several models of ball valves used for double valve isolation (DVI) that have been installed for 10 or more years. The safety function of the safety-significant valves is to limit leakage of waste past the valves in order to decrease the consequences of a fine spray release due to a transfer misroute. While previous leak testing of the valves has been conducted, the testing did not address concerns in the Documented Safety Analysis concerns about the impact on valve operation and service life of the potential abrasive/erosive properties of the tank waste on the valves when cycled. Pacific Northwest National Laboratory (PNNL) conducted DVI tests at their Multi-Phase Transport Evaluation Loop Facility. Testing evaluated the leak rate of safety-significant isolation valves for DVI as a function of simulated operating cycles. PNNL developed a performance-based simulant specifically for this test program. The test environment simulated the abrasive characteristics of the Hanford Tank Farms Waste Transfer System during single-shell tank retrieval and waste feed delivery to the Hanford Tank Waste Treatment and Immobilization Plant. The testing consisted of periodically measuring leak rates and maximum operational torques on the DVI valves after prescribed numbers of valve operational cycles (open and close). The test results from the simulant cycle tests provide Washington River Protection Solutions with performance characteristics that can be used to verify compliance with the Documented Safety Analysis limits for valve seat leakage. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

AGR-5/6/7 Thermal Model with Non-uniform Gas Gaps

Fuel compact temperatures are a crucial factor in assessing irradiation performance of the Tristructural isotropic fuel particles. In the absence of direct measurement, fuel compact temperatures were calculated using a three-dimensional finite element thermal model, which is subject to simulation uncertainty. The most dominant factor in uncertainty of calculated fuel temperature is the gas gap uncertainty due to the nub-to-shell clearance because of fabrication error. A revision to the thermal model was made to examine the most probable offset position for four different dates of interest. The offset position options varied in magnitude and azimuthal direction of offset of the holder top and bottom position. The best-fit offset of the holder is estimated based on the minimum root mean square error of residuals for all operational thermocouples (TCs). From these results, the following conclusions were made: 1) During earlier cycles (162A – 164A) when numerous TCs were still operational, the best-fit offset distance varied in range [0.002 – 0.0035 in.] for both the top and bottom holder while offset azimuthal direction varied widely, especially for the holder bottom. 2) Holder offset led to slightly lower Capsule 1 average temperature, but wider temperature variation (lower minimum and higher peak fuel temperatures). 3) The maximum offset of 0.006 in. to the northwest direction for both the top and bottom holder yielded a much higher temperature than the original peak fuel temperature, which increased from 1422 to 1557 ? (or 135 ? increase). This could be one of the reasons for massive particle failure near the end of Cycle 165A. 4) Even though, the highest temperature at the TC-1-7 tip reached slightly higher than 1000?, the temperature along the TC wire reached as high as 1335? assuming a holder offset of 0.006 in. to the northwest. Consequently, at this temperature nickel from this TC wire could cause particle failures in Capsule 1.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

TORCHE - Toolbox for Reactor Cross-Flow Heat Exchangers

As interest within the scientific community for the development of small modular reactors (SMRs) grows, utilizing compact and cross-flow heat exchangers present many design advantages. Cross-flow configurations of tube bundles allow water, liquid metal, or a molten salt to be contained within the tube side or shell side in an efficient spatial arrangement. The goal of this project is to provide the user with an easy-to-use set of tools for cross-flow heat exchanger design. The tools are intended to require little to no geometric set-up like meshing in a CFD code in order to enable rapid preliminary design and optimization. With this set of simple Python scripts, the user can determine the necessary quantities and correctional factors in order to assess the heat exchanger characteristics. The two characteristics used in designing novel and efficient heat exchangers incorporated into this work currently are pressure drop and heat transfer coefficients. In support of these efforts, the literature has been scoured for sources of experimental data and proposed methods for calculating pressure drop and heat transfer in cross-flow tube banks. Several empirical correlations and methodologies for calculating pressure drop and heat transfer incorporated into the Toolbox for Reactor Cross-Flow Heat Exchangers (TORCHE). TORCHE is an expansion on the cross-flow pressure drop scripts reported by Haefner et al.. Presented in this summary is a synopsis of the empirical correlations and methodologies of pressure drops and heat transfer coefficient. A validation effort of these models is also presented to provide confidence in the implementation. For this validation of TORCHE, experimental data has been gathered and extracted from graphs in the literature. Error bars were not reported in the experimental data used in this summary and therefore is not included in the analysis.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

The evidence of N = 16 shell closure and β -delayed neutron emission from 25 F

We measured the β -delayed neutron emission from 25 F for the first time at the Facility for Rare Isotope Beams (FRIB). Using combined neutron and γ -ray detector systems of the FRIB Decay Station Initiator (FDSi), we observed β -decay transitions populating neutron unbound states between 4.2 and 8 MeV in 25 Ne. The experimental results led to the revision of the β -decay half-life and β -delayed neutron-emission probability of 25 F. The β -decay strength distribution of 25 F extracted from the data agrees with the shell-model predictions using the USDB and SDPF-M effective interactions. This result indicates that the spherical neutron shell gap persists in 25 F and 25 Ne.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗

Accelerating Hanford Site Cleanup through Operations Research Modeling - 20238

The Hanford Site cleanup effort will require the integration of dozens of unique facilities and processes, many of which will be first-of-a-kind in implementation and design. Each facility will be governed by its own set of operating logic, configured with a unique array of unit operations, and subject to a set of constraints that will affect its behavior. The collection of facilities have multiple points of interface, making the operations of any one facility potentially significant to the operations of other up- or downstream processes. It is therefore highly desirable to accurately predict these operations, as it allows for Site officials to identify and preempt bottlenecks and vulnerabilities before they unexpectedly inhibit the cleanup mission. With the quantity and complexity of the processes that will be on Site, building a pen-and-paper or even a spreadsheet-assisted model of the cleanup mission quickly becomes overwhelming in scope and inaccurate in execution. The Engineering organization for the Site's Tank Operations Contract (TOC) has therefore implemented the use of operations research (OR) modeling to simulate and predict future operations of Site facilities. These models are created using a discrete event simulation tool that allows for the development of detailed, versatile, and robust models. Not only can these models account for complex logical behaviors, but they can also simulate process details down to the level of vessel sizing, labor utilization, equipment reliability, and resource availability. To date, the TOC has developed OR models for several facilities on Site, including for single-shell tank (SST) farms, double-shell tank (DST) farms, the Effluent Treatment Facility (ETF), and the waste transfer system. These models have focused on identifying bottlenecks and operational constraints, and have been used to quantify the effects of implementing process changes. This latter point is particularly valuable, as it allows for several alternatives to be studied in a virtual setting before committing resources to making a change in the field. The decision to develop OR models has gained tremendous support from the Site's stakeholders and the U.S. Department of Energy (DOE) management, and has prompted the use of the tool to support additional internal and external initiatives. Recently, an initiative was proposed to use the models to help identify and provide quantitative backing for risks and opportunities for the TOC. This application of OR could not only help inform how the TOC manages its risks (e.g. quantities and types of spare parts), but could also help drive process improvements whose benefits might otherwise be hard to quantify. The models have also been used to drive the TOC's cloud computing, artificial intelligence (AI), and machine learning (ML) initiatives. These initiatives will not only improve the ability of the TOC to more rapidly respond to the needs of its customers, but it will also aid in the ability of the TOC to analyze and improve the processes it studies. Partnership with two external software development and consulting companies (Lanner and Ynformed) has furthered not only the application of AI and ML within the TOC, but has also spurred the development of new/improved software tools and platforms used by the companies. These partnerships have proven to be mutually beneficial and productive, and have set a precedent for the types of gains that can be made by exploring such options. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Thermal Hydraulic Experimental Test Article - Fiscal Year 2023 (Final Report)

The Thermal Hydraulic Experimental Test Article (THETA) is a facility that is used to develop sodium components and instrumentation as well as acquire experimental data for validation of reactor thermal hydraulic and safety analysis codes. The facility simulates nominal conditions as well as protected/unprotected loss of flow accidents in a sodium-cooled fast reactor (SFR). High fidelity distributed temperature profiles of the developed flow field may be acquired with Rayleigh backscatter based optical fiber temperature sensors. The facility was designed in partnership with systems code experts to tailor the experiment to ensure the most relevant and highest quality data for code validation. THETA is comprised of a traditional primary coolant and secondary coolant system. The primary system is submerged in the pool of sodium and consists of a pump, electrically heated core, intermediate heat exchanger, and connected piping and thermal barriers (redan). The secondary system, located outside of the sodium pool, consists of a pump, sodium to air heat exchanger, and connected piping and valves. To date a test matrix has been completed utilizing the primary system of THETA. These tests, along with computational fluid dynamics and systems code models, determined the heat transfer across the core barrel and intermediate heat exchanger outlet was too great to effectively represent scaled thermal hydraulic phenomena of a liquid metal cooled reactor. Therefore, a significant effort was made to remove the primary system from the METL 28” test vessel #4 and clean the residual sodium from the primary system to facilitate upgrades. Thermal insulation was then incorporated in the core barrel and intermediate heat exchanger outlets. The primary system was then replaced, and a series of tests were performed to assess the performance of the thermal insulation. With primary system testing and upgrades complete, the secondary system could then be brought online. The tube side of the shell-and-tube intermediate heat exchanger was installed onto the primary system flange to begin installing the secondary system. The support structure for the secondary system was then erected on the METL mezzanine alongside the THETA primary system to facilitate installation of the secondary system components (sodium-to-air heat exchanger, flowmeter and pump). The piping and expansion tank were welded into the secondary system. Non-destructive examination of the secondary system welds was completed in order to satisfy ASME B31.3 pipe code for class M process fluids. The heating system and insulation were then added to prepare the system to be filled with sodium. The ancillary electrical equipment was installed which included the pump control box, blower VFD, pipe heater control system, etc. The secondary system will be filled, and a test matrix will be completed in early FY2024.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗