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All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Countercurrent Actinide Lanthanide Separation Process (ALSEP) Demonstration Test with a Simulated PUREX Raffinate in Centrifugal Contactors on the Laboratory Scale

An Actinide Lanthanide Separation Process (ALSEP) for the separation of trivalent actinides (An(III)) from simulated raffinate solution was successfully demonstrated using a 32-stage 1 cm annular centrifugal contactor setup. The ALSEP solvent was composed of a mixture of 2-ethylhexylphosphonic acid mono-2-ethylhexyl ester (HEH[EHP]) and N,N,N',N'-tetra-(2-ethylhexyl)-diglycolamide (T2EHDGA) in n-dodecane. Flowsheet calculations and evaluation of the results were done using the Argonne’s Model for Universal Solvent Extraction (AMUSE) code using single-stage distribution data. The co-extraction of Zr(IV) and Pd(II) was prevented using CDTA (trans-1,2-diaminocyclohexane-N,N,N',N'-tetraacetic acid) as a masking agent in the feed. For the scrubbing of co-extracted Mo; citrate-buffered acetohydroxamic acid was used. The separation of An(III) from the trivalent lanthanides (Ln(III)) was achieved using citrate-buffered diethylene-triamine-N,N,N',N",N"-pentaacetic acid (DTPA), and Ln(III) were efficiently back extracted using N,N,N',N'-tetraethyl-diglycolamide (TEDGA). A clean An(III) product was obtained with a recovery of 95% americium and curium. The Ln(III) were efficiently stripped; but the Ln(III) product contained 5% of the co-stripped An(III). The carryover of Am and Cm into the Ln(III) product is attributed to too few actinide stripping stages, which was constrained by the number of centrifugal contactors available. Improved separation would be achieved by increasing the number of An strip stages. The heavier lanthanides (Pr, Nd, Sm, Eu, and Gd) and yttrium were mainly routed to the Ln product, whereas the lighter lanthanides (La and Ce) were mostly routed to the raffinate.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Identifying Challenges in Safeguards for Metallic Fuel Fabrication Facilities

As new advanced reactors gain popularity, there is an increasing interest in metallic fuel fabrication for fast reactors. While metallic fuels themselves are not a new idea, as many of the first reactors employed metallic fuels, new designs, compositions, and fabrication methods are appearing throughout the nuclear community. As the interest grows and facilities are constructed, both domestic and international safeguards will need to be heavily involved to support safeguards-by-design (SBD) measures from the start. This work compiles a review of historical and modern fuel types and fabrication methods, fabrication processes, safeguards gaps, and potential safeguards solutions. Metallic nuclear fuel types have been around for many decades and were included in some of the first reactors including the Experimental Breeder Reactor (EBR)-I and -II, the Fermi 1 reactor, the Integral Fast Reactor (IFR), and the Dounreay Fast Reactor (DFR). These reactors used various compositions including pure uranium (U) metal, U-zirconium (Zr) alloys, plutonium (Pu)-aluminum (Al) alloys, U-fissium (Fs) alloys, U-Pu-Zr alloys, and U-molybdenum (Mo) alloys [1, 2, 3, 4, 5]. These small alloying additions are included to improve the material properties of the pure U metal. The alpha-phase U (stable below 661C) suffers elongation in one direction causing grain boundary cracking and increasing creep rate due to irradiation growth, thermal cycling, and preferential crystal orientation. It is ideal to utilize the gamma-phase U (typically stable above 769C) by adding small amounts of alloying elements such as Zr or Mo to stabilize this phase down to room temperature [3]. Additionally, some research has been focused on U with transuranic (TRU) elements present, typically coming from the used fuel recycling process. Including these elements in fast reactor fuel can aid in the reduction of nuclear waste by burning minor long-lived actinides. However, the additions of TRU elements can cause concerns to arise when trying to fabrication or safeguard metallic fuels. A typical metallic fuel element is shown in Figure 1. Sodium is added into the cladding to create a thermal bond between the fuel slug and cladding wall. The fuel slug is then inserted and the end plug is welded on to the top of the fuel element. A gas plenum is left to create a headspace for gaseous fission products to escape rather than continue to build in the fuel itself [1, 5]. Other fuel element geometries exist as well, such as the Lightbridge twisted cruciform geometry shown in Figure 2 [6]. This design allows for better cooling performance and provides room for fuel rod swelling without impacting the fuel rod diameter. There are many different fabrication methods for metallic fuels, which is one of the many benefits of these fuel types. Many of these fabrication methods are relatively easy and cost-efficient. The most popular fabrication method is injection casting, sometimes called vacuum induction melting (VIM), shown in Figure 3 [4, 8, 9, 7, 10]. This method was largely used for EBR-II fuel fabrication. The injection casting system is contained inside of a vessel consisting of a Y2O3-coated graphite crucible surrounded by an induction coil with ZrO2-coated quartz molds suspended above the crucible. The fuel feedstock is placed inside of the graphite crucible and melted using the induction furnace. The induction furnace utilizes a dual frequency with the high frequency melting the feedstock and the low frequency causing stirring of the melted feedstock to form a homogeneous mixture. The mixture is heated to approximately 1600C in an argon environment. The vessel is evacuated and then the quartz molds are lowered into the graphite crucible containing the molten metal and the vessel is repressurized to inject the metal fuel upwards into the molds. The molds are removed and then shattered to release the fuel slugs. This fabrication method was used to fabricate 39,000 metallic fuel pins for EBR-II. While injection casting has been the most common metallic fuel fabrication method throughout the decades, many other methods have been explored including low-pressure gravity casting, microwave casting, continuous casting, centrifugal casting, coextrusion, and many others [11, 12, 8, 13, 14, 15]. Some of these methods aim to mitigate challenges that arise with americium (Am) volatilization during the casting process for TRU-containing fuel feedstocks, an issue with injection casting. Coextrusion is one of the methods explored at the Idaho National Laboratory (INL) and has been utilized for the initial fabrication tests of Lightbridge's unique fuels, as well as other metallic fuels with cladding coextruded. In this process, large billets are formed and machined and then inserted into a molten salt bath for approximately 30 minutes. The billets are then loaded into the extrusion press and extruded. This process can be seen in Figure 4 [15].

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Structural and Bonding Analysis in Monomeric Actinide(IV) Oxalate from Th(IV) to Pu(IV): Comparison with the An(IV) Nitrate Series

In this work, single-crystal X-ray diffraction (SC-XRD) structures and Raman spectra of a series of new isomorphous molecular An(IV)-oxalate compounds (Th, U, Np, and Pu) are reported. These complexes are crystallized with cobalt(III) hexamine ([Co(NH 3 ) 6 ] 3+ ) as the counter cations, [Co(NH 3 ) 6 ] 2 [An(C 2 O 4 ) 5 ]·4H 2 O, revealing five bidentate nonbridging oxalate ligands in the first coordination sphere (CN = 10). The nonbridging oxalate is rather uncommon for An(IV)-oxalate systems, which are widely characterized as polymeric compounds. Density functional theory (DFT) calculations were performed to examine the bonding between An(IV) cations and oxalate ligands. For comparison, we also report results obtained for the An(IV)-hexanitrate series, [(C 2 H 5 ) 4 N] 2 [An(NO 3 ) 6 ] (with An = Th, U, Np, Pu, and Ce), which consists of O-donor ligands as well but with a larger coordination number (CN = 12). The bonding analysis confirms that the actinide–oxygen bond is predominantly ionic with a minor increase in covalency from Th to U and slight variations from U to Pu. Further comparison showed that the charge transfer increases slightly when increasing the number of anions in the coordination sphere (C 2 O$_4^{2–}$: CN = 10; NO$_3^–$: CN = 12), but covalent effects as indicated by the amount of internuclear electron density accumulation are small and similar for oxalate and nitrate.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Validation of Recent Changes in Actinide Evaluations U-235, U-233, Pu-239 [Slides]

This presentation discusses U-235 updates, including fluctuations in the fission cross section in the URR range that were refined to represent measured data following the Paradela et al. evaluation. There were small changes to the low-energy resonances to improve the fit to measured capture data from RPI (capture reduced by 5% from 0.06-7.8 eV, by 7.7% from 7.8-11 eV) and there was a very small change to the thermal nu-bar within uncertainties. Discrete level data are stored in MF=6 (format requirement, no impact on calculations) and issues leading to negative eigenvalues in cross-covariances in the RR have been sorted out. In conclusion, many improvements were made to evaluated nuclear data of actinides in the resonance region and above, such as minor changes to U-235 including assessment of changes that depend on the finalization of the O-16 evaluation. U-233 improvements show significant progress in performance and some further improvement might be needed at intermediate energies. Pu-239 ORNL/IAEA evaluation is promising; however, more work is needed.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2021

Actinide arene-metalates: ion pairing effects on the electronic structure of unsupported uranium–arenide sandwich complexes

Addition of [UI 2 (THF) 3 (μ-OMe)] 2 ·THF (2·THF) to THF solutions containing 6 equiv. of K[C 14 H 10 ] generates the heteroleptic dimeric complexes [K(18-crown-6)(THF) 2 ] 2 [U(η 6 -C 14 H 10 )(η 4 -C 14 H 10 )(μ-OMe)] 2 ·4THF (1 18C6 ·4THF) and {[K(THF) 3 ][U(η 6 -C 14 H 10 )(η 4 -C 14 H 10 )(μ-OMe)]} 2 (1 THF ) upon crystallization of the products in THF in the presence or absence of 18-crown-6, respectively. Both 1 18C6 ·4THF and 1 THF are thermally stable in the solid-state at room temperature; however, after crystallization, they become insoluble in THF or DME solutions and instead gradually decompose upon standing. X-ray diffraction analysis reveals 1 18C6 ·4THF and 1 THF to be structurally similar, possessing uranium centres sandwiched between bent anthracenide ligands of mixed tetrahapto and hexahapto ligation modes. Yet, the two complexes are distinguished by the close contact potassium-arenide ion pairing that is seen in 1 THF but absent in 1 18C6 ·4THF, which is observed to have a significant effect on the electronic characteristics of the two complexes. Structural analysis, SQUID magnetometry data, XANES spectral characterization, and computational analyses are generally consistent with U(IV) formal assignments for the metal centres in both 1 18C6 ·4THF and 1 THF , though noticeable differences are detected between the two species. For instance, the effective magnetic moment of 1 THF (3.74 μ B ) is significantly lower than that of 1 18C6 ·4THF (4.40 μ B ) at 300 K. Furthermore, the XANES data shows the U L III -edge absorption energy for 1 THF to be 0.9 eV higher than that of 1 18C6 ·4THF, suggestive of more oxidized metal centres in the former. Of note, CASSCF calculations on the model complex {[U(η 6 -C 14 H 10 )(η 4 -C 14 H 10 )(μ-OMe)] 2 } 2– (1*) shows highly polarized uranium–arenide interactions defined by π-type bonds where the metal contributions are primarily comprised by the 6d-orbitals (7.3 ± 0.6%) with minor participation from the 5f-orbitals (1.5 ± 0.5%). These unique complexes provide new insights into actinide–arenide bonding interactions and show the sensitivity of the electronic structures of the uranium atoms to coordination sphere effects.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2021

Probing electronic structure in berkelium and californium via an electron microscopy nanosampling approach

Due to their rarity and radioactive nature, comparatively little is known about the actinides, particularly those with atomic numbers higher than that of plutonium, and their compounds. In this work, we describe how transmission electron microscopy can provide comprehensive, safe, and cost-effective characterization using only single nanogram amounts of highly-radioactive, solid compounds. Chlorides of the rare elements berkelium and californium are dropcast and then converted in situ to oxides using the electron beam. The f-band occupancies are probed using electron energy loss spectroscopy and an unexpectedly weak spin-orbit-coupling is identified for berkelium. In contrast, californium follows a jj coupling scheme. These results have important implications for the chemistries of these elements and solidify the status of californium as a transitional element in the actinide series.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2022

Activation of CO 2 by Actinide Cations (Th + , U + , Pu + , and Am + ) as Studied by Guided Ion Beam and Triple Quadrupole Mass Spectrometry

Reactions of CO 2 with Th+ have been studied using guided ion beam tandem mass spectrometry (GIBMS) and with An + (An + = Th + , U + , Pu + , and Am + ) using triple quadrupole inductively coupled plasma mass spectrometry (QQQ-ICP-MS). Additionally, the reactions ThO + + CO and ThO + + CO 2 were examined using GIBMS. Modeling the kinetic energy dependent GIBMS data allowed determination of bond dissociation energies (BDEs) for D o (Th + -O) and D o (OTh + -O) that are in reasonable agreement with previous GIBMS measurements. The QQQ-ICP-MS reactions were studied at higher pressures where multiple collisions between An + and the neutral CO 2 occur. As a consequence, both AnO + and AnO 2 + products were observed for all An + except Am + , where only AmO + was observed. Here, the relative abundances of the observed monoxides compared to the dioxides are consistent with previous reports of the AnO n + (n = 1, 2) BDEs. Comparison of the periodic trends of the group 4 transition metal, lanthanide (Ln), and actinide atomic cations in reactions with CO 2 (a formally spin-forbidden reaction for most M + ground states), and O 2 (a spin unrestricted reaction) indicate that spin conservation plays a minor role, if any, for the heavier An + metals. Further correlation of Ln + and An + + CO 2 reaction efficiencies with the promotion energy (E p ) to the first electronic state with two valence d-electrons (E p (5d 2 ) for Ln + and E p (6d 2 ) for An + ) indicates that the primary limitation in the activation of CO 2 is the energetic cost to promote from the electronic ground state of the atomic metal ion to a reactive state.

bond activation↗
DOE OSTI2025

Structural and Spectroscopic Trends in Alkali Tris(Acetato)Actinyl(VI) Complexes

A series of alkali tris(acetato)actinyl(VI) compounds M[AnO 2 (CH 3 COO) 3 ] (M=Na, K, Rb, Cs and An=U, Np, Pu) is reported, applying a combination of singlecrystal X-ray diffraction, solid-state Raman spectroscopy, and U(VI) luminescence spectroscopy for their structural and spectroscopic characterization. The results show that complexes formed from the same alkali metal but different actinyl ions are mostly isostructural. The symmetric actinyl stretch (ν 1 (An=O yl )) in these isostructural complexes undergoes a redshift within the actinide series despite the shortening of the actinyl bond due to the actinide contraction. Complexes formed from the same actinyl but co-crystallized with different alkali metals show a progression toward lower space group symmetry. While the actinyl bond length remains unmodulated within the alkali series, a small redshift of ν 1 (An=O yl ) is observed, indicating minor changes in the vibrational properties of the actinyl moiety caused by the structural progression towards lower symmetry. This subtle change is also evidenced by a redshift of the U(VI) luminescence spectra within the M[UO 2 (CH 3 COO) 3 ] series.

Raman↗
DOE OSTI2020

Dynamics of Cation-Induced Conformational Changes in Nanometer-Sized Uranyl Peroxide Clusters

Conformational changes of the pyrophosphate (Pp)-functionalized uranyl peroxide nanocluster [(UO 2 ) 24 (O 2 ) 24 (P 2 O 7 ) 12 ] 48– ({U 24 Pp 12 }), dissolved as a Li/Na salt, can be induced by the titration of alkali cations into solution. The most symmetric conformer of the molecule has idealized octahedral ( O h ) molecular symmetry. One-dimensional 31 P NMR experiments provide direct evidence that both K + and Rb + ions trigger an O h -to- D 4 h conformational change within {U 24 Pp 12 }. Variable-temperature 31 P NMR experiments conducted on partially titrated {U 24 Pp 12 } systems show an effect on the rates; increased activation enthalpy and entropy for the D 4 h -to- O h transition is observed in the presence of Rb + compared to K + . Two-dimensional, exchange spectroscopy 31 P NMR revealed that magnetization transfer links chemically unique Pp bridges that are present in the D 4 h conformation and that this magnetization transfer occurs via a conformational rearrangement mechanism as the bridges interconvert between two symmetries. The interconversion is triggered by the departure and reentry of K (or Rb) cations out of and into the cavity of the cluster. This rearrangement allows Pp bridges to interconvert without the need to break bonds. Cs ions exhibit unique interactions with {U 24 Pp 12 } clusters and cause only minor changes in the solution 31 P NMR signatures, suggesting that O h symmetry is conserved. Single-crystal X-ray diffraction measurements reveal that the mixed Li/Na/Cs salt adopts D 2 h molecular symmetry, implying that while solvated, this cluster is in equilibrium with a more symmetric form. Thus, these results highlight the unusually flexible nature of the actinide-based {U 24 Pp 12 } and its sensitivity to countercations in solution.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2023

A Review of Advanced Test Reactor Fuel and Assessment of Its Compatibility with the ZIRCEX Chlorination Process

Advanced Test Reactor (ATR) fuel has been identified as a resource for high-assay low-enriched uranium (HALEU) production. A survey was performed on the published literature describing ATR fuel. The geometry of the fuel is complex; different parts of the fuel compact experience differing neutron flux and burnup. The literature is sparse, and access is controlled. Therefore, fundamental studies of fuel reprocessing must use a model fuel that represents the main chemical and structural features. Advanced chlorination, or chlorination with sulfur-chlorine bearing reagents is being investigated as way to separate the fuel from metal matrix alloys. A UAl x alloy will be fabricated with x = 3, 4, and 5. The potential chlorination of individual UAl x intermetallics will be assessed in the advanced chlorination process of Al-8001 and Al-6061 as well as a representative mixture. Initial studies will track the alloying elements of the Al, which are Si, Fe, Cu, Mn, Mg, Cr, Zn, and Ti, in addition to the U itself. Further studies will include fission product simulants. Because advanced chlorination solvents include sulfur, the chemistry of sulfur with major and minor constituents will also be investigated. The experimental work accompanied by neutronic calculations will allow the assessment of the feasibility of advanced chlorination to separate aluminum from uranium. If bench-scale testing appears promising, then small-scale tests in shielded facilities with irradiated cladding, lightly irradiated fuel, and spent nuclear fuel are recommended to track the complete inventory of fissile actinides, fission product impurities, and reagent solids and liquids.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

${S}^5$: The destruction of a bright dwarf galaxy as revealed by the chemistry of the Indus stellar stream

The recently discovered Indus stellar stream exhibits a diverse chemical signature compared to what is found for most other streams due to the abundances of two outlier stars, Indus$\_$0 and Indus$\_$13. Indus$\_$13, exhibits an extreme enhancement in rapid neutron-capture ($r$-)process elements with $\mathrm{[Eu/Fe]} = +1.81$. It thus provides direct evidence of the accreted nature of $r$-process enhanced stars. In this paper we present a detailed chemical analysis of the neutron-capture elements in Indus$\_$13, revealing the star to be slightly actinide poor. The other outlier, Indus$\_0$, displays a globular cluster-like signature with high N, Na, and Al abundances, while the rest of the Indus stars show abundances compatible with a dwarf galaxy origin. Hence, Indus$\_0$ provides the first chemical evidence of a fully disrupted dwarf containing a globular cluster. We use the chemical signature of the Indus stars to discuss the nature of the stream progenitor which was likely a chemically evolved system, with a mass somewhere in the range from Ursa Minor to Fornax.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2020

TCCR Operational Summary and Optimization for Tank 9 Processing - 20314

Savannah River Remediation (SRR) manages and operates the liquid waste facilities at Savannah River Site (SRS) for the Department of Energy (DOE). Stored liquid waste is a complex mixture of insoluble solids (sludge) and soluble salts in an alkaline solution. SRR has deployed the Tank Closure Cesium Removal (TCCR) system, a tank-side ion exchange process, to remove radioactive cesium from salt waste and enable onsite disposal of the resulting decontaminated salt solution as low-level waste at the Saltstone facilities. The TCCR system consists of two prefilters, four ion exchange (IX) columns, one resin trap, and a ventilation system. The IX process uses a form of inorganic crystalline silicotitanate (CST), which has a high affinity for cesium and other alkali metals, strontium, and actinides. This process is currently deployed utilizing salt feed from Tank 10, with future plans to dissolve solid salt in Tank 9 and transfer the salt solution to Tank 10 for processing through TCCR. The feed for TCCR must be created from salt-cake in Tank 10 through a dissolution process. Once enough salt has been dissolved, a qualification process is entered. This process characterizes the feed and ensures the cesium loading on the columns will not cause boiling of waste within the columns during or after processing. Once the batch has been qualified, salt waste is fed to the TCCR system through a transfer pump in the center of the tank. The waste is filtered through a set of two shielded, dead-end prefilters that prevent solids buildup in the columns. The filtered salt solution then travels to the shielded IX columns, which can be operated individually or in series, where the cesium is sorbed on the CST media. The decontaminated salt solution (DSS) then travels through a resin trap and out of the module to Tank 11. TCCR has successfully processed approximately 795,000 L of Tank 10H radioactive salt waste over two batches to date. There has not yet been a system induced shutdown. The prefilters performed as expected with only minor degradation in recovery of differential pressure after a backflush sequence. The time between backflushes decreased as each batch reached the end of processing. The hydraulics in the IXCs mostly performed as expected at all flow rates, except for one IXC that will be further investigated during Batch 3 processing. The TCCR system has shown some opportunities for more efficient processing during the length of the demonstration so far. For future processing of material from Tank 9H through Tank 10H and the TCCR unit, TCCR 1A will implement changes to the prefilters and the IXCs. The prefilters will have an increased surface area and a new filter media in an effort to increase time between filter swaps and improve backwashing cleaning capability. The IXCs will have a reduced diameter to allow for increased heat transfer out of the column and increased loading of Cs-137. Additionally, a new form of CST with an increased kinetic performance is being investigated for use during TCCR 1A operation. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Biogeochemistry of Pond B (Savannah River Site, South Carolina, USA): Sediment Core, Total extraction data, Pond B Savannah River Site July 2019. Subsurface Biogeochemistry of Actinides SFA

Pond B at Savannah River Site (SRS, South Carolina) is a monomictic reservoir that received SRS R reactor cooling water from 1961–1964. Previous studies conducted between the 1980s–1990s on the water column and sediments of Pond B measured trace amounts of Pu (33 MBq 238Pu and 430 MBq 239,240Pu), 241Am, and 137Cs. Since then, the pond has been relatively isolated and the radionuclide concentrations have not been monitored over time. Herein, about 30 years after the last publication on Pond B, we are re-evaluating the geochemistry and radionuclide distribution within Pond B at four locations along a horizontal transect from the inlet to outlet.This study investigated the distribution of anthropogenic radionuclides Pu-239 and Cs-137 along with total organic carbon, iron, and trace element in contaminated sediments of Pond B at the Savannah River Site (SRS). Pond B received reactor cooling water from 1961 to 1964, and trace amounts of Pu-239 and Cs-137 during operations. Our study collected sediment cores to determine concentrations of Pu-239, Cs-137, and major and minor elements in solid phase, pore water and an electrochemical method was used on wet cores to determine dissolved elemental concentrations.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2020

Gas-solid reaction pathway for chlorination of rare earth and actinide metals using hydrogen and chlorine gas

Hydriding and chlorination of buttons of metallic cerium have been tested as a head-end step to separating metal impurities from actinides. The objective here was to achieve complete conversion to chlorides while maintaining high surface area via small particles and open porosity. Hydriding by reaction with 100% H 2 gas at 573 K was successful at reducing cm-scale metal buttons into small particles/fragments. This step was followed by reaction with flowing 99.5% anhydrous Cl 2 gas at temperatures ranging from 423 to 573 K for 60 min. In experiments reported here, up to 92% conversion to cerium chloride was achieved from reacting with the anhydrous chlorine gas. The final chloride product has high porosity coupled with small particle sizes (microns to 1600 μm), which makes it suitable for subsequent purification processes. Higher chlorine concentrations in the gas favor higher conversion to chloride. The optimal temperature to achieve high chlorination is 523 K. Oxygen is a minor impurity in the chlorine gas which appears to prevent complete chlorination from being achieved.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2020

Challenges and Solutions in Operation and Lay-Up of the Modular Caustic- Side Solvent Extraction Unit - 20078

The Modular Caustic-Side Solvent Extraction (CSSX) Unit (MCU) was designed and constructed to provide interim salt waste processing so that the Liquid Waste Disposition Program (LWDP) could continue waste removal and tank closure until start-up of the Salt Waste Processing Facility (SWPF). MCU processes Clarified Salt Solution (CSS), or salt solution (SS) that has undergone actinide removal/filtration, to produce two waste streams: the cesium laden Strip Effluent (SE), which is incorporated into glass at the Defense Waste Processing Facility (DWPF), and the cesium depleted Decontaminated SS (DSS), which is incorporated into grout at the Saltstone Production Facility (SPF). MCU completed start-up in 2008 with an initial operating life of 3 years and design life of 5 years. Prior to reaching the end of the design life, multiple critical components were upgraded and/or replaced to mitigate risks associated with any delays in SWPF start-up. During the remainder of operation (until June 2019), MCU was challenged to maximize processing with minimal additional modifications to repair/improve aging infrastructure, while maintaining low levels of risk to workers and the environment. Notable challenges during the extended operating life of the facility involved both mechanical and operational issues. Two of the most significant issues were the increased frequency of Process Vessel Ventilation (PVV) High Efficiency Particulate Air (HEPA) filter change-outs and the biofouling of the SE coalescer (SEC). The major impacts of both issues were that replacement of these components 1) resulted in high exposure to workers, and 2) required substantial downtime, thus hindering achievement of processing goals. PVV HEPA filter replacements are required based on dose rate and differential pressure (dP) limits. After introducing a higher-curie feed to MCU, the operating time between change outs was reduced by approximately half. SEC replacements are required based on dP limits. After restarting from an extended outage, the operating life of the SEC was significantly reduced by more than half due to biofouling. As a result of extensive troubleshooting and process improvements, the facility was able to recover performance and extend the operating life of the PVV HEPA filters and the SEC media. Through troubleshooting both of these issues, the facility was also able to learn from prior experience and adapt in order to minimize down time and maintain throughput. Although the facility took advantage of short periods of downtime to perform nonintrusive troubleshooting and minor corrective/preventative maintenance activities, longer outages were periodically required for corrective maintenance involving process cell entry. During these outages, remote tools and specialized shield plates were employed to minimize dose to workers. Mock-ups of non-routine maintenance activities were critical to early identification of potential issues and improvements, which shortened outage duration. Consistent with the facility's operational strategy, evolutions needed for MCU layup were identified and sequenced to ensure safe conditions within the facility, meet pre-defined criteria for lay-up configuration, and protect workers, while minimizing the impact to SWPF integration. De-inventory and flushing of process areas was prioritized so that the total residual inventory was reduced as efficiently as possible. This paper discusses lessons learned and best practices from troubleshooting unique issues, performing complex maintenance activities, and implementing a layup strategy that best supports the overall system mission. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2026

Integration of Online Cross-Section Generation Capability with Depletion and Transient Solvers in Griffin

Griffin is a Multiphysics Object-Oriented Simulation Environment (MOOSE)-based reactor multiphysics analysis application jointly developed by Argonne and Idaho National Laboratories under the DOENE Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. In FY25, an online crosssection generation capability based on the Self-Shielding Application Programming Interface (SSAPI) was demonstrated for TRISO-fueled reactor problems under steady-state conditions. This fiscal year, that capability was extended to support depletion and transient multiphysics calculations, enabling high-fidelity analyses that generate self-shielded cross sections on the fly from the actual evolving composition and temperature states rather than from pre-tabulated libraries. For depletion, a two-way coupling was established in which SSAPI computes compact-averaged self-shielded cross sections that the depletion solver then uses to advance the Bateman equations, with the updated compositions returned to SSAPI at each step; the depletion module was refactored to support both library-based and SSAPI-based cross sections, and additional logic was added to track daughter isotopes and to exclude minor isotopes for efficiency. For transient analysis, the SSAPI multigroup library was extended with the kinetics data required for time-dependent calculations, the Improved Quasi-Static (IQS) scheme was coupled with SSAPI, and several supporting capabilities were implemented, including a self-shielding treatment that lets control rods and drums move within a self-shielded model, which had previously been impossible and had ruled out rod- and drum-movement transients with on-the-fly cross sections altogether, a new mixing scheme for delayed-neutron precursor decay constants, a checkpoint-based restart workflow, and performance improvements such as pointwise cross-section interpolation and the bypassing of unnecessary Dancoff factor calculations. The implemented capabilities were verified against Serpent Monte Carlo solutions. For depletion, a prismatic pin-cell problem based on a Next Generation Nuclear Plant (NGNP) Very High Temperature Reactor benchmark showed excellent agreement, with eigenvalue differences within 200 pcm over the entire burnup range (up to 140 MWD/kgU) and fission-product and actinide inventories agreeing to within 0.8% and 2.5%, respectively; a heat-pipe microreactor assembly problem with a much higher fuel loading confirmed the same behavior and quantified the bias introduced when the multigroup equivalence effect is neglected. For transient analysis, a pin-cell problem with a step reactivity insertion and temperature feedback reproduced the analytically expected asymptotic power and showed close agreement between the direct and IQS solutions, and a two-dimensional microreactor core problem with control-drum rotation exercised the new moving-drum self-shielding treatment and demonstrated successful coupling of the online crosssection generation with both the direct and IQS transient methods. The capability was further exercised on a full-core pebble-bed problem, in which Griffin was coupled with the System Analysis Module (SAM) to simulate load-following operation of the gPBR with the Doppler feedback resolved at the TRISO fuel kernel temperature. These developments in Griffin provide a convenient, high-fidelity approach to cross-section generation for advanced thermal reactors with geometrically complex and highly heterogeneous configurations, including TRISO-fueled prismatic and pebble-bed systems, and support steady-state, depletion, and transient multiphysics calculations. They also enable self-shielded cross sections to be evaluated directly at the actual coupled state of the system, thereby establishing a foundation for high-fidelity, fully coupled multiphysics analysis of advanced reactors

Park, H.↗
DOE OSTI2020

Measurement of the 239 Pu$(n, f$) prompt fission neutron spectrum from 10 keV to 10 MeV induced by neutrons of energy 1–20 MeV

Although the prompt fission neutron spectrum (PFNS) is an essential component of neutron-driven systems that has been measured for decades, there are still multiple glaring unknowns regarding the PFNS of major actinides in the fission neutron incident energy range, specifically with regard to multichance fission and pre-equilibrium neutron emission processes. The only impactful experimental 239 Pu PFNS measurements included in recent nuclear data evaluations were measured over a limited outgoing neutron energy range at thermal and 1.5-MeV average incident neutron energy, while other potentially impactful measurements have been shown to contain errors that resulted in either large uncertainty increases or in complete exclusion from nuclear data evaluation. Here, we report a measurement of the 239 Pu PFNS over a wide range of incident neutron energy (1–20 MeV) and three orders of magnitude in outgoing neutron energy (0.01–10 MeV) resulting from the Chi-Nu experiment at the Los Alamos Neutron Science Center. These results are the combination of separate PFNS measurements in the same experimental area, one using a Li-glass and the other a liquid scintillator detector array. Covariances between all PFNS data points from each detector and within each incident energy range were generated between all other data in both detector arrays and within all other incident neutron energy bins, yielding a single covariance matrix for all 1300 PFNS data points reported here. These covariances are based on a thorough assessment of systematic bias and uncertainties associated with the measurement, PFNS extraction technique, combination of data from each detector type, and other aspects of the analysis. The existence of covariances between PFNS data points in different incident neutron energy ranges yielded covariances between average PFNS energy values at each incident energy to be reported here as well, which allowed for firm statements to be made regarding a shape of a purely experimental mean PFNS energy trend for the first time. Although minor PFNS shape differences exist between the results reported here and recent nuclear data evaluations, the ENDF/B-VIII.0 and JEFF-3.3 PFNS evaluations agree reasonably well with the present results from 1-to 10-MeV incident neutron energy, which spans the well-measured 1.5-MeV incident neutron energy PFNS from Lestone and Shores as well as the onset of second-chance fission. However, while the pre-equilibrium component of the PFNS above 12-MeV incident neutron energy roughly agrees in position and magnitude with ENDF/B-VIII.0 and JEFF-3.3, clear differences relating to the relative magnitude of third-chance fission PFNS features are present in the PFNS shape and in the mean PFNS energy trends.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗