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All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Equivalent Fundamental-Mode Source Simulations for Spherical Uranium and Plutonium Systems [Abstract]

The equivalent fundamental-mode source refers to a “source” which is identically distributed in space, energy, and angle as that of a fundamental-mode fission source distribution. It has previously been shown that the system fixed source multiplication ($M_{fs}$) relates to the effective multiplication factor ($\kappa_{eff}$) as $M_{fs} = \frac{g^*}{1-k_{eff}}$ The term $\mathcal{g}$* therefore relates the system source distribution to the system equivalent fundamental-mode source. As previously shown, the $\mathcal{g}$* term is approximately 1 for deeply subcritical systems and then diverges as the system multiplication increases. For a system with a point source located in the center of an assembly (such a sphere), $\mathcal{g}$* will be greater than 1; this happens because there is a higher probability of induced fission when all of the starter neutrons originate from the center of the sphere. For a uniform source, $\mathcal{g}$* is less than 1, but does not diverge from 1 as dramatically as seen by a point source. One reason that $\mathcal{g}$* is an important parameter is that it is used in measurement methods to infer reactor kinetics parameters ($β_{eff}$ being one example). This work will discuss multiple ways to simulate $\mathcal{g}$* using MCNP®6.2, which include use of traditional methods versus a single input file method that was recently published. A discussion related to methods to simulate $M_{fs}$ will be presented. Spherical systems of Highly Enriched Uranium (HEU) will be compared with previous works. In addition, results of plutonium including multiple reflector materials will be investigated. Simulated systems will include both parametric studies as well as configurations used in recent experiments performed at the National Criticality Experiments Research Center (NCERC). These studies will include subtleties associated with system geometry that have not been previously shown. Last, this work will investigate the relationship between $\mathcal{g}$* and reactor kinetics parameters (such as $β_{eff}$).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Lessons Learned From Ventilation and Glovebox Flooding Via Overfilling of the Wet Vacuum System in a Plutonium Facility

In the summer of 2021, operations personnel requested maintenance personnel to fill a wet vacuum (Wetvac) seal water tank in a plutonium facility at Los Alamos National Laboratory (LANL). After the filling was finished, the fill valve of the Wetvac seal water tank was left open and a ‘dead-man’ valve on an isolation line failed to close properly. After a series of other failures, the water exited the glovebox and contaminated three rooms of the first floor of the facility and an underlying area in the basement. A retrospective analysis of this event determined that there was no risk of a criticality due to the inherent design features of the ventilation system. This paper will explain the driving factors behind how this event was able to happen, as well as some of the major lessons learned that can be universally applied to other nuclear facilities. One of the main causes of the event was the delegation of work to maintenance personnel that were not qualified for the work being asked of them. Additionally, there was a lack of coordination between criticality safety engineers and facility operations equipment owners regarding equipment and connections present on the seal water tanks for the Wetvac, specifically the presence of overflow drain lines. The respective criticality safety evaluation credited an overflow drain line being present on the seal water tanks that was later found to have not been present. Lastly, there are lessons to be learned regarding equipment configuration and designing systems with safety in mind.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2024

How to Safely Build 100-plus Kilograms of Weapons-Grade Plutonium

The goal of the EUCLID (Experiments Underpinned by Computational Learning for Improvements in Nuclear Data) project was to reduce compensating errors by utilizing machine learning to both help determine which reactions contain compensating errors as well as optimizing an experiment which can be used to maximally reduce these errors. Compensating errors can adversely impact the predictive power of application simulations, and therefore it’s useful to further constrain nuclear data and reduce these errors. The EUCLID project included building two configurations at the National Criticality Experiments Research Center (NCERC). These two configurations had very different geometries (one was cube-like and one was slab-like). Previous works focus on selection of the target experiment(s), radiation transport capabilities developed in the project, the experiment optimization, and the performance of the experiments. This work will focus only on the safety aspects of performing this experiment, which utilized over 100 kg of weapons-grade plutonium.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Destructive Examination Protocol for 3013-Container-Package Storing Plutonium-Bearing Materials

The 3013-container-package consists of a convenience, inner and outer container and is used for long-term storage of plutonium-bearing materials. A destructive examination (DE) protocol has been developed to examine the container package visually and with microscopic aid to find any corrosion conditions that could result in the loss of the integrity of the container package over its lifetime. The DE protocol contains three main steps: initial container examination, helium (He) leak testing, and detailed imaging and analysis of the inner container closure weld region (ICCWR). The ICCWR has been determined to be bounding, defined as exhibiting worst case conditions for stress corrosion cracking (SCC) of the inner container. To assess SCC in the ICCWR, the inner container lid is cut into quarters and the weld and He-leak testing gasket are removed. Then a citric acid wash is performed to remove adherent chlorides from the ICCWR. The wash is then sent for analysis to determine the concentration of chlorides in the ICCWR. While the analysis for chloride concentration is being performed, the quarter sections are further sectioned into 1/8th subsections by cutting each section in half. These subsections are washed using nitric acid to remove corrosion products. Then each subsection is imaged using a Wide Angle 3-D Measurement System (WAMS). After analysis of microscope images for potential SCC, additional imaging can be performed, including subsurface imaging. After review, a determination is made of whether the container integrity may potentially affect the safe storage of the material.

container↗
DOE OSTI2022

Solubility controls on plutonium and americium release in subsurface environments exposed to acidic processing wastes

To identify the role of waste composition and sediment interactions in controlling Pu and Am mobility in contaminated sediments at Hanford, a legacy nuclear site, Pu and Am concentrations in solutions equilibrated with contaminated sediments from beneath the 216-Z-9 (Z-9) Trench were compared to the solubilities of PuO 2 materials, synthesized by methods representative of the disposed wastes, in the absence of sediments. Furthermore, this work shows that the solubilities of PuO 2 materials synthesized by different methods, and with varying particle sizes, agree with PuO 2 (am,hyd), although dissolution kinetics differed between materials. According to saturation index (SI) calculations, PuO 2 (am,hyd) is likely also controlling Pu release from sediments under conditions where phosphate concentrations are low. However, both Pu-phosphate and Am-phosphate phases, identified in SI calculations and by high resolution transmission electron microscopy, play roles in controlling release in low pH, high phosphate, shallow sediments just below the Z-9 Trench. The elevated phosphate is likely due to decomposition of tributyl phosphate from waste solutions over time. Sediments from deeper in the subsurface beneath the Z-9 Trench are less acidic and contain less phosphate, with Pu solubility likely controlled by PuO 2 (am,hyd) that precipitated following neutralization of the acidic waste stream. Controls on Am concentrations in deeper sediments are more complex and potentially involve sediment adsorption and/or release from Pu (1-x) Am x O 2 following Am in-growth. The concentrations of both Pu and Am were elevated in the colloidal fraction associated with shallow sediments, but not in PuO 2 experiments, suggesting the presence of Pu/Am pseudocolloids (e.g., Pu/Am associated with mineral colloids). However, Pu and Am association with the colloidal size fraction was not observed in deeper sediments, suggesting transport of Pu and Am to these depths beneath the Z-9 Trench was not due to colloidal transport.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2016

Low temperature dissolution flowsheet for plutonium metal

The H-Canyon flowsheet used to dissolve Pu metal for PuO 2 production utilizes boiling HNO 3 . SRNL was requested to develop a complementary dissolution flowsheet at two reduced temperature ranges. The dissolution and H 2 generation rates of Pu metal were investigated using a dissolving solution at ambient temperature (20-30 °C) and for an intermediate temperature of 50-60 °C. Additionally, the testing included an investigation of the dissolution rates and characterization of the off-gas generated from the ambient temperature dissolution of carbon steel cans and the nylon bags that contain the Pu metal when charged to the dissolver.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Corrective Action Plan for Corrective Action Unit 568: Area 3 Plutonium Dispersion Sites Nevada National Security Site, Nevada, Revision 0 with ROTC 1

The purpose of this CAP is to provide the plan for implementation of the recommended corrective action alternatives (CAAs) for CAU 568. Site characterization activities were performed in 2014, and the results are presented in Appendix A of the CAU 568 CADD. The CAAs were recommended in the CADD. The scope of work required to implement the recommended CAAs of closure in place and clean closure at 11 of the 14 CASs includes the following: The installation of physical barriers over the nine safety experiment ground zeroes to cover contamination at CASs 03-23-20 (Otero), 03-23-23 (San Juan and Pascal-C), 03-23-31 (Pascal-B, Luna, Colfax), 03-23-32 (Pascal-A), 03-23-33 (Valencia), and 03-23-34 (Chipmunk); the characterization and removal of three soil and debris piles at CAS 03-08-04, and one HCA soil pile at CAS 03-23-30; the removal of three steel well head covers (PSM) from CASs 03-23-20 (Otero), 03-23-31 (Luna), and 03-23-33 (Valencia); the removal of soil and lead PSM from two locations at CAS 03-26-04; Implementation of FFACO use restrictions at nine safety experiment ground zeroes at CASs 03-23-20, 03-23-23, 03-23-31, 03-23-32, 03-23-33, and 03-23-34; the steel well head cover at CAS 03-23-23; the areas meeting HCA conditions at CASs 03-23-19 and 03-23-31; and the Boomer crater area at CAS 03-45-01. The FFACO use restriction boundaries will be presented in the CAU 568 closure report.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2025

The Jupiter Experiments: High-240 Plutonium Metal Plates Separated by Lead and Reflected by Copper

Each layer consisted of a 6 by 6 matrix of either these fuel-filled containers or solid blocks of copper of the same outer dimensions. An example of this arrangement, along with one of the copper inner reflectors, is shown in Figure 1.4. This figure shows the same aluminum containers from Figure 1.3 but with the aluminum containers completely closed. The lifting rings shown on the copper inner reflector were only for assembly and were not present for the measurement (in which the holes were filled with copper plugs). Figure 1.5 shows the loading arrangement of the aluminum containers, along with aluminum shims around the container arrays to ensure a tight fit between these layers and the surrounding reflectors. All plates were in the same orientation; none were rotated in any fashion. Additional loading information for the PAHN plates and plate loading respective to the room entrance are provided in Figures 1.6 and 1.7, respectively.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

One-group fission cross sections for plutonium and minor actinides inserted in calculated neutron spectra of fast reactor cooled with lead-208 or lead-bismuth eutectic

The paper is dedicated to one-group fission cross sections of Pu and MA in LFRs spectra with the aim to increase these values by choosing a coolant which hardens neutron spectra. It is shown that replacement of coolant from Pb-Bi with Pb-208 in the fast reactor RBEC-M, designed in Russia, leads to increasing the core mean neutron energy. As concerns fuel Pu isotopes, their one-group fission cross sections become slightly changed, while more dramatically Am-241 one-group fission cross section is changed. Another situation occurs in the lateral blanket containing small quantities of minor actinides. It is shown that as a result of lateral blanket mean neutron energy hardening the one-group fission cross sections of Np-237, Am-241 and Am-243 increases up to 8-11%. This result allows reducing the time of minor actinides burning in FRs. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Experimental and quantum mechanical characterization of an oxygen-bridged plutonium(IV) dimer

We report the synthesis and characterization of K 4 {[PuCl 2 (NO 3 ) 3 ] 2 (μ 2 -O)}·H 2 O, which contains the first known μ 2 -oxo bridge between two Pu IV metal centers. Adding to its uniqueness is the Pu-(μ 2 -O) bond length of 2.04 Å, which is the shortest of other analogous compounds. The Pu-(μ 2 -O)-Pu bridge is characterized by the mixing of s -, d -, and p -orbitals from Pu with the p -orbitals of O; the 5 f -orbitals do not participate in bonding. Natural bond orbital analysis indicates that Pu and O interact through one 3c-2e σ Pu-O-Pu and two 3c-2e π Pu-O-Pu bonding orbitals and that the electron density is highly polarized on the μ 2 -O. Bond topology properties analysis indicates that the Pu-(μ 2 -O) bond shares both ionic and covalent character. Quantum mechanical calculations also show that the dimer has multiconfigurational ground states, where the nonet, septet, quintet, triplet, and singlet are close in energy. This work demonstrates the interplay between experimental and computational efforts that is required to understand the chemical bonding of Pu compounds.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2019

Complexation of Plutonium(IV) with trans-1,2-Diaminocyclohexane-N,N,N',N'-tetraacetic Acid (CDTA) in Acidic Solution

Understanding the interaction of Pu(IV) with complexing agents present in the nuclear fuel cycle is important for predicting the performance of used nuclear fuel separations. The complexation of Pu(IV) with trans -1,2-diaminocyclohexane-N,N,N’,N’-tetraacetic acid (CDTA) was studied in acidic solutions of 0.10 0.50 mol∙L -1 HClO 4 with 1.00 mol∙L -1 (Na,H)ClO 4 total ionic strength by a liquid–liquid extraction method using tracer quantities of 238 Pu. The acid dissociation constants of CDTA and the autoprotolysis constant of water were determined via potentiometric titrations in 0.10–2.00 mol∙L -1 NaClO 4 and 25.0 ± 0.1 °C. The variation of the dissociation constants with ionic strength was modeled with the Specific ion Interaction Theory (SIT) and the associated SIT parameters were obtained. The thermodynamic dissociation 23 constants at zero ionic strength for water and CDTA were determined from this analysis as PK$^o_w$ = 14.00 ± 0.03, PK$^o_{a2}$ = 1.52 ± 0.04, PK$^o_{a3}$ = 2.78 ± 0.05, PK$^o_{a4}$ = 4.17 ± 0.04, PK$^o_{a5}$ = 6.75 ± 0.02, and PK$^o_{a6}$ = 10.64 ± 0.04 at 25.0 ± 0.1 °C. The results of the liquid–liquid extraction experiments indicated the formation of a 1:0:1 complex, PuCDTA o , and the presence of additional protonated species, Pu(HCDTA) + and Pu(H 2 CDTA )2+ , at these acidities. The corresponding stability constants in 1.00 mol·L -1 (Na,H)ClO 4 and 23 ± 1 °C were determined to be log 10 β 101 =24.2 ± 0.3, log 10 β 111 =25.4 ± 0.2, and log 10 β 121 = 25.8 ± 0.1.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Distribution of neptunium and plutonium in New Mexico lichen samples ( Usnea arizonica ) contaminated by atmospheric fallout

In this study, the concentrations of 237 Np, 239 Pu and 240 Pu were determined in lichen samples ( Usnea arizonica ) that were collected from ten locations in New Mexico between 2011 and 2013 using isotope dilution inductively-coupled plasma mass spectrometry (ID-ICP-MS). The observed isotopic ratios for 237 Np/ 239 Pu and 240 Pu/ 239 Pu indicate trace contamination from global and regional fallout (e.g. Trinity test and atmospheric testing at the Nevada Test Site). The fact that actinide contamination is detected in recent lichen collections suggests continuous re-suspension of fallout radionuclides even 50 years after ratification of the Limited Test Ban Treaty.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2017

A multiple ion counter total evaporation (MICTE) method for precise analysis of plutonium by thermal ionization mass spectrometry

This study presents a total evaporation method for the analysis of sub-picogram quantities of Pu, utilizing an array of multiple ion counters. Data from three standards are presented to assess the utility of the technique. An external precision of 1.5% RSD (2σ) was achieved on aliquots approaching 100 fg for the minor 240 Pu isotope. Accurate analysis of <1 femtogram of 240 Pu, is achievable, with an external reproducibility of better than 10% RSD (2σ). Finally, this new technique represents a significant advance in the total evaporation method and will allow routine measurement of femtogram sized Pu samples by thermal ionization mass spectrometry.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2018

A case study in plutonium radiochronometry using multiple isotope systems

In this work, radiochronometry analyses of two Pu metals were performed using the 237 Np– 241 Am– 241 Pu, 234 U– 238 Pu, 235 U– 239 Pu, and 236 U–240Pu decay series. For one sample, all radiochronometers yield concordant model dates in 1959–1960, suggesting that the aqueous processing method used to purify Pu effectively removed U, Am and Np decay products from the bulk Pu. The second sample yields discordant model dates that are also older than the known production date in 1982. The excess U, Am and Np present in the sample imply that the sample was purified during at least two different episodes, using a combination of aqueous methods and molten salt extraction.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Measuring key Sm isotope ratios in irradiated UO 2 for use in plutonium discrimination nuclear forensics

As part of an experimental validation of a nuclear forensics methodology for Pu source reactor-type discrimination, destructive analysis has been performed here on two irradiated UO 2 pellets with different irradiation histories. Analysis has focused on measuring key Sm fission product isotope ratios used in a previously published maximum likelihood methodology to determine the most likely irradiation history of the pellets. A total of 21 Sm isotope ratios were measured within the irradiated pellets, and generally agreed within 20% of the irradiations as simulated using the Monte Carlo Radiation Transport and material depletion code, MCNP6. Results indicate the chosen approach can accurately measure the isotope ratios within 5% experimental error.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2020

Stability of plutonium oxide nanoparticles in the presence of montmorillonite and implications for colloid facilitated transport

The physical and chemical stability of PuO 2 nanoparticles (intrinsic colloids) in groundwater will control their transport and may affect the performance of high-level radioactive waste repositories. In this study, we examined the chemical stability of two types of PuO 2 nanoparticles at both 25 and 80 °C. The “alkaline” PuO 2 nanoparticles were prepared by neutralizing an acidic Pu(IV) solution with dilute NaOH to pH ~9.5. The “acidic” PuO 2 nanoparticles were precipitated from 0.1 M nitric acid by heating a Pu(IV) solution at 60–80 °C for 30 min. The chemical stability of these PuO 2 nanoparticles was tested in the presence of montmorillonite, a common mineral in the environment and potentially relevant backfill material in some nuclear waste repository designs. The “alkaline” PuO 2 nanoparticles were found to be unstable over a timescale of months at both 25 and 80 °C, with elevated temperature enhancing their dissolution rates and sorption to montmorillonite. PuO 2 nanoparticle dissolution rates decreased with increasing Pu concentration, consistent with solution saturation. The “acidic” PuO 2 nanoparticles appeared to remain stable for much longer than the “alkaline” PuO2 nanoparticles. The “alkaline” PuO 2 nanoparticle dissolution rate constants were as high as 10 -11.9±0.4 mol m -2 s -1 and 10 -11.2±0.2 mol m -2 s -1 at 25 and 80 °C, respectively, while the “acidic” PuO 2 nanoparticle dissolution rate at 80 °C was 10 -13.5 mol m -2 s -1 . Based on transmission electron microscopy, the “alkaline” and “acidic” PuO 2 nanoparticles were of similar size (2.5–4.5 and 2–3 nm nanoparticles, respectively). However, the “acidic” PuO 2 nanoparticles formed more ordered nanoparticle aggregates. Our results suggest that the specific conditions experienced during PuO 2 nanoparticle formation could significantly affect the stability of PuO 2 in the presence of competing sorption processes and, in turn, the relative importance of intrinsic versus pseudocolloid transport in the environment.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2023

Development of the generation III (ATR generation I) Plutonium-238 production target design

A domestic supply chain to produce 238 Pu fuel for radioisotope thermoelectric generators is critical for enabling future space exploration. A multi-laboratory effort has worked to establish a common target design to efficiently produce 238 Pu in two research reactors. This approach ensures that the annual production goals set forth by NASA are met, while also establishing redundant production capabilities. Here, this paper describes the effort to develop the common target design as well as considerations for future applications for the irradiation platform.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗