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At least 163 records · Page 9

Uranium Mill Tailings Radiation Control Act Title II DOE Due Diligence and Lessons Learned from a Previous Site Transfer - 20352

Title II of the Uranium Mill Tailings Radiation Control Act (UMTRCA) established that a government agency will provide perpetual care for closed uranium and thorium ore-processing sites that were operating under an NRC source material license in 1978 or were licensed thereafter. Commercial owners (licensees) operating under an NRC or agreement state specific license when UMTRCA was passed are responsible for conducting reclamation of any byproduct material remaining from uranium-ore processing operations in accordance with an NRC or agreement state approved reclamation plan. Reclamation includes both surface and groundwater remedies. Upon completion of reclamation and approval by NRC, the site is required to be transferred to either the host state or the DOE for long-term surveillance and maintenance. Since UMTRCA's enactment, six Title II sites have been transferred to DOE; an additional 24 Title II sites are anticipated to be transferred before 2050. DoE's role mandated under UMTRCA Title II as the long-term care custodian is to perform 'monitoring, maintenance, and emergency measures necessary to protect the public health and safety.' UMTRCA requires that the licensee pay a long-term surveillance charge 'sufficient to cover the annual costs of site surveillance.' However, at some sites such as the Bluewater, New Mexico, Disposal Site, this mandate has required additional effort and expense by DOE, beyond the originally anticipated and intended scope within UMTRCA, but within the authority of DOE under UMTRCA. In 1997, the Bluewater site became the second UMTRCA Title II site to be transferred to DOE. The site was the location of a uranium mill operated from 1953 until 1982. The specific licensee began site reclamation in 1991, and by 1995 all tailings and contaminated materials were encapsulated in two tailings disposal cells and other disposal areas. In addition to surface contamination, milling activities impacted groundwater in the two upper aquifers. In 1989, the specific licensee attempted active groundwater remediation; however, no significant reduction in contaminant concentrations was observed. As a result, the specific licensee applied to NRC for alternate concentration limits (ACLs) in 1990, which were approved in 1996 as being protective, after additional corrective actions were performed. Since transfer of the Bluewater site to DOE, unforeseen challenges have occurred, requiring additional actions. The first challenge is the occurrence of surface depressions located on the northern section of the main tailings disposal cell. The depressions were first observed during DoE's initial inspection in 1998; however, evidence of these can be observed on satellite images taken prior to transfer. Since being first observed, the depressions have continued to grow both in depth and areal extent. Due to the design of the main tailings disposal cell, the depressions impede storm water from being effectively shed off the 101- hectare (250-acre) top slope of the main tailings disposal cell. Instead, storm water accumulates in the depressions, forming a large ephemeral pond that has stored up to 16.3 x 10{sup 6} liters (4.3 million gallons) of stormwater. The ponding poses a potential risk to the integrity of the main tailings disposal cell in the case of a large storm event, with the potential to cause the pond to overtop and erode the cover material and underlying waste. DOE has taken a number of short-term actions to monitor, measure, and reduce the ponding and is currently working with the US Army Corps of Engineers to design and construct a repair. Additional challenges are associated with groundwater at the Bluewater site. Nine wells were present on the 1335-hectare (3300-acre) site upon transfer. Groundwater compliance was called into question after the State of New Mexico reduced its uranium groundwater standard from 5.0 to 0.03 milligrams per liter in 2004, and when an ACL for uranium was exceeded in a site monitoring well in 2010. Acquiring historical groundwater data and subsequent evaluations as well as additional DOE groundwater monitoring led to installing 10 new monitoring wells and performing additional site hydrogeology recharacterization. DOE continues to evaluate groundwater conditions at the site and works with the NRC to determine regulatory requirements and a path forward. As a result of lessons learned at the Bluewater site and other Title II sites, improved processes have been implemented at a programmatic level to increase due diligence before site transfer and prevent similar issues from occurring at other UMTRCA Title II sites under long-term management. DoE's due diligence process is documented in the Process for Transition of UMTRCA Title II Disposal Sites to DOE for Long-Term Surveillance and Maintenance and is designed to ensure that DOE has no technical or compliance concerns with regulatory decisions that might compromise protectiveness following site transfer to DOE. Implementation of the due diligence process has increased DoE's role prior to site transfer and has been effective in identifying potential issues. Actions by NRC and specific licensees, in response to enhanced due diligence efforts by DOE, are expected to minimize, if not totally prevent, the need for unanticipated actions by DOE pertaining to the surface and groundwater remedies after site transfer. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Grimsel Test Site - A Successful International Underground Research Laboratory for Many Decades - 20429

For more than 35 years, Nagra and its partners from around the world have been conducting underground research projects at the Grimsel Test Site (GTS, www.grimsel.com) to contribute to the development and confirmation of safe geological disposal concepts and for the characterization of suitable host rock formations. Over the years, the results of this internationally recognized research program have been, and continue to be, incorporated directly into exploration programs, modelling, safety, and engineering feasibility studies on options for deep geological repositories. Each project of the GTS program involves field-testing, laboratory studies, design and modelling tasks, and integrates all scientific and technical aspects. Each project phase is planned with a duration of three to five years to facilitate practical and administrative aspects and allow flexibility for updating the overall project plans with the latest findings. Scientific and engineering interaction among the different projects is ensured via an international steering committee meeting. Hosting an IAEA level C radiation- controlled zone, which allows use of radionuclides, including actinides such as thorium, uranium, neptunium, plutonium and americium, in in-situ experiments is one of the reasons why GTS also developed as a center of excellence for work with radioactive tracers under realistic in-situ boundary conditions. Last year, a new five-year program (2019 to 2023) started which includes projects with a planning horizon of decades. The new five-year program includes a new phase of in-situ experiments using radionuclides such as migration experiments in the Colloid Formation and Migration project (CFM), the Long-Term Diffusion experiment (LTD) and the newly established C-14 and I-129 Migration in cement project (CIM). The 'High Temperature effects on Bentonite' (HotBENT) project is starting in the current phase and is studying the effects of elevated temperatures (>175 deg. C) on bentonite materials. As a generic underground research laboratory (URL) it is expected that the GTS will provide in the coming years a platform for international collaboration, knowledge development and knowledge transfer for the next generation of scientists and engineers in the area of radioactive waste disposal and geosciences. A key role regarding knowledge transfer and training is provided by the well-established Grimsel Training Center (GTC), which (beside many URL related issues) also covers many general aspects of radioactive waste management. In this paper we provide an overview of the current program at the GTS, focusing on the experiments that study the migration of radionuclides through engineered barrier materials and the geosphere. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Preliminary Evaluation of Loading DOE Standard Canisters in the INL CPP-603 Irradiated Fuel Storage Facility - 20543

This paper looks at the equipment and operations necessary to load United States Department of Energy (DOE)-owned Spent Nuclear Fuel (SNF) into DOE Standard Canisters in the CPP-603 Irradiated Fuel Storage Facility (IFSF) in the Idaho Nuclear Technology and Engineering Center (INTEC) area at Idaho National Laboratory (INL). Two types of fuels are looked at in this evaluation: Advanced Test Reactor (ATR) fuel (uranium-aluminide fuel with aluminum cladding) and Peach Bottom fuel (thorium-uranium carbide fuel in a graphite matrix). The fuel ready for loading would come from fuel storage canisters in the CPP-603 facility. The paper describes the facility, the fuel types, the DOE Standard Canisters, and existing equipment; lists the needed loading operations; reviews facility features and equipment to perform the operations; and then lists the decisions, analyses, designs, demonstrations, and modifications that will be needed to perform the loading of DOE-owned SNF into DOE Standard Canisters in the CPP-603 IFSF. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Strategies for the Photoreduction of Tc-99 Pertechnetate to Low-Valent Tc by Keggin Polyoxometalates

Technetium-99 (half-life of 2.1 × 10 5 yrs, β max = 0.29 MeV) is a hazardous radiological contaminant, which, in its predominant form of pertechnetate (TcO 4 – ), is highly mobile in the environment. Most strategies for the removal of pertechnetate from the environment involve uptake and/or absorption of pertechnetate using resins, clays, cationic metal-organic frameworks and even thorium borate ceramic like materials. Alternative approaches have involved the reduction and subsequent sequestration of lower valent technetium species using iron, sulfides, or iron sulfides. Here, our lab has explored this strategy using the lacunary alpha-2 Wells–Dawson polyoxometalate (α 2 -[P 2 W 17 O 61 ] 10– ) to both reduce and sequester lower valent technetium, and we have reported on the ligand features that stabilize the reduced species. In this work we investigate the potential of “plenary” Keggin POMs (XW 12 O 40 n – ) (X = P, Si, Al, n = 3, 4, 5, respectively) to both reduce TcO 4 – and stabilize the reduced Tc species. Specifically, we report on the mechanism by which the reduction of technetium occurs and find that PW 12 , SiW 12 , and AlW 12 promote the reduction of TcO 4 – to lower valent states. X-ray absorption spectroscopy was used to confirm a combination of Tc IV {in the form of TcO 2 · 2H 2 O and Tc 2 (µ-O) 2 4+ } and Tc V , which is subsequently complexed into a POM defect as a Tc V =O species.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Helium Release During Fracture and Granular Fragmentation of Rocks

Geogenic Helium-4 ( 4 He) in-situ increases locally in regions of large deformation generated naturally or anthropogenically. This gas release by deformation is a potential geochemical precursor signal for subsurface deformation. To evaluate the applicability of 4 He degassing for correlating deformation in different lithologies, we conducted high force crush tests, up to 97,800 N axial load, to assess the total 4 He released during fragmentation of the rocks. We observed that the highest 4 He released occurred in the sedimentary rocks and that release correlated strongly with lithologic age and U/Th content. Microstructural changes of the pre- and post-test rocks indicate that the degree of grain size reduction relates directly to the total 4 He released during crushing. The range of in-place 4 He was calculated based XRF measurements of uranium and thorium in each lithology, with the results indicating that the majority of the trapped 4 He was not released. However, the 4 He released by deformation depended upon how the each rock deformed during deformation and the degree of grain size reduction. We postulate that 4 He precursor signals can be used to understand subsurface deformation only if geomechanical and geochemical conditions for 4 He enrichment in a lithology are met.

Deformation signals↗

Actinium purification from elemental components of stainless steel and aluminum

Radioisotopes of actinium are valuable because of their potential use in the medical industry. Actinium-225 shows promise for treating disease—like cancer—via targeted alpha therapy (TAT), and the longer-lived actinium-227 is the parent of two radionuclides with potential application in TAT radiopharmaceuticals, thorium-227 and radium-223. Continuing progress in the development of these medical applications requires robust and diverse methods for extracting and purifying actinium from a wide range of matrices. Further, to define the strengths and limitations of separation methods commonly employed in actinium processing, we characterized the performance of ion exchange and extraction chromatographic steps for removing contaminants found in stainless steel (chromium, iron, nickel, and silicon) and aluminum from actinium-227. While AG-MP1 anion exchange resin with HCl (aq) successfully removed iron, cation exchanger AG 50W-X8 removed most other contaminants. The most persistent contaminant was aluminum, which was removed using a DGA Normal extraction resin. These results are presented within the context of applying that methodology toward actinium purification strategies.

36 MATERIALS SCIENCE↗

230Th/234U radiochronometry for uranium materials by alpha spectrometry for nuclear forensics analysis

We have developed an alpha spectrometry method for 230Th/234U radiochronometry to determine the model separation age for uranium materials. This method offers an alternative to the more commonly used, but significantly more resource intensive multi-collector inductively coupled plasma mass spectrometry (MC-ICP-MS) while providing similar accuracy and precision. The development included radiochemical separation of uranium and thorium, calibration of the 232U tracer and the alpha spectrometry measurement. The method was validated by analyzing a certified reference material, CRM 125-A, which is a uranium oxide pellet assay, isotopic, and radiochronometric standard. The results were compared to those obtained by our routine MC-ICP-MS 230Th/234U radiochronometry analysis method to assess the performance parameters for alpha spectrometry compared to the state-of-the-art technique.

Macsik, Zsuzsanna [Los Alamos National Laboratory ↗

Measurement of low energy characteristic X-rays emitted present in U-10Mo HALEU alloy fuel

Advance high density monolithic alloy HALEU fuels have been scanned with a silicon drift detector scanning system to measure low-energy characteristic X-rays emitted by uranium daughters, as well as zirconium present as a foil diffusion barrier. Four peaks were measured across the surfaces of three plates, and are consistent with other thorium and protactinium peaks present in depleted uranium samples that are induced by the decay of 235 U and 238 U. Of the four peaks detected, three are dominated by the decay of 235 U, and one peak at 15.70 keV covers two characteristic X-rays from the zirconium diffusion barrier (Kα 1 ) and from a 231 Th 15.62 keV X-ray induced by 235 U α-decay. The low energies of these lines suggest that they are mostly emitted from the fuel meat surface, and the poor counting statistics caused by strong self-attenuation make these data a poor candidate for 235 U/ 238 U ratio analysis. We suggest, then, that these X-rays could be used for passive measurement of variations in the cladding thickness of these fuels to complement manual thickness measurement methods.

HALEU↗

Using yttrium as an indicator to estimate total rare earth element concentration: a case study of anthracite-associated clays from northeastern Pennsylvania

Abstract This study demonstrated using yttrium (Y) as an indicator to estimate the total rare earth element and Y contents (REY) in coal-associated samples and to facilitate selection of samples with high REY assays in a fast and inexpensive manner. More than 10 anthracite-associated samples were collected from each of three Pennsylvanian sites (sites B, J and C) based on Thorium gamma ray logging suggesting high REY content. Several samples from each site were analyzed by ICP-MS to determine the rare earth distribution patterns and to establish the site-specific linear equations of Y and REY. The Y contents of the remaining samples were measured by a portable X-ray fluorescence analyzer, and the REY values were estimated based on the site-specific linear equation developed earlier. R-squared values above 0.70 were obtained for all the estimation equations from all three sites on both a whole sample basis and an ash basis. Previously, ash content has been widely used as an indicator of high REY content. This may not be applicable for a specific site. Site B in this study is an example where ash contents could not be statistically correlated with REY, so using Y for estimation is more applicable. The demonstrated sample screening process is suitable for samples from sites that share more similar distribution patterns (either MREY or LREY or HREY) as well as for samples from sites that share multiple distribution patterns (LREY/MREY/HREY) depending on the desirable accuracy. The demonstrated process lowers the analytical cost from $70 to 80 dollars per sample to $10–15 per sample while significantly reducing the processing time and acid consumption for ICP digestion. This is particularly true when a relatively large sample size is involved, for example, 100 samples from one site analyzed by ICP-MS/OES.

Yang, Xiaojing↗

Membrane-based solvent extraction for the recovery of rare earths from phosphate mining process streams

This study reports on the capture of rare earth elements (REEs) from phosphate industry process streams, including phosphoric acid (PA) sludge and phosphogypsum (PG), using a membrane solvent extraction (MSX) process. While MSX has been proven effective for a relatively concentrated feed, its effectiveness for dilute REEs solutions remains unexplored. Investigated PA-sludge and PG particles contain total REEs concentrations of ∼1100 and ∼320 ppm, respectively. Acid leaching, implemented to dissolve the REEs, significantly dilutes the REEs concentration to ∼210 ppm for PA-sludge leachate and ∼60 ppm for PG leachate. These low concentrations, compounded by the higher levels of non-REE ions and radioactive species, uranium (U) and thorium (Th), poses challenges to the MSX process. Here, we demonstrated that N,N,N′,N′-tetraoctyl-diglycolamide (TODGA) selectively binds REEs from a >3 M nitric-acid leachate while effectively rejecting U and Th. Concentrations of light REEs in strip solution were doubled compared to the feed, while heavy REEs were preferentially extracted. Furthermore, >99% purity gypsum, free of U and Th, was precipitated during the acid leaching process, aiding separation by removing significant amounts of non-REEs species (e.g., calcium) prior to the MSX process. Molecular simulations support the experimental data, suggesting preferential separation of heavy over light REEs. Based on these results, a cost-effective integrated process including pretreatment, acid leaching, MSX, and wastewater treatment is proposed for the co-recovery of REEs, phosphoric acid, gypsum, and U. This study shows MSX as a technically and economically feasible process for the recovery of REEs from low-concentration process streams, offering advantages over conventional solvent extraction.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Extraction chromatography of 225 Ac and lanthanides on N,N-dioctyldiglycolamic acid /1-butyl-3-methylimidazolium bis(trifluoromethylsulfonyl)imide solvent impregnated resin

The alpha-emitter 225 Ac (t 1/2 = 9.92 d) is currently under development for targeted alpha-particle therapy of cancer, and accelerator production of 225 Ac via proton irradiation of thorium targets requires robust separations of 225 Ac from chemically similar fission product lanthanides. Additionally, the lanthanide elements represent critical components in modern technologies, and radiolanthanides such as 140 Nd (t 1/2 = 3.37 d) also have potential application in the field of nuclear medicine. The ionic liquid, 1-butyl-3-methylimidazolium bis(trifluoromethylsulfonyl)imide ([Bmim][NTf 2 ]), combined with the diglycolamide extractant, N,N-dioctyldiglycolamic acid (DODGAA), was adsorbed on macroporous resin support to produce a solvent impregnated resin (SIR) that was investigated for separations of 225 Ac and lanthanides. The equilibrium distribution coefficients (Kd) of the rare earth elements (Sc(III), Y(III), Ln(III)), 225 Ac(III), Th(IV), and U(VI) on the prepared DODGAA/[Bmim][NTf 2 ]-SIR were determined from batch adsorption experiments in HCl and HNO 3 media. The DODGAA/[Bmim][NTf2]-SIR exhibited preferential uptake of the heavier lanthanide elements while allowing for the separation of the lighter lanthanides. Column separations utilizing the DODGAA/[Bmim][NTf 2 ]-SIR were effective at separating the lighter lanthanides from each other, and separating 225 Ac from a mixture of lanthanides, 213 Bi, and 225 Ra without the need for additional complexing agents.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Environmental evaluation of radioactivity levels and associated radiation hazards in groundwater around the WIPP site

Groundwater may contain radioactive substances which can be dangerous to human health. Concentrations of natural radionuclides polonium (Po), thorium (Th), uranium (U), and radium (Ra) isotopes were measured in groundwater samples collected from different locations in the vicinity of the Waste Isolation Pilot Plant (WIPP) site in Carlsbad, New Mexico. The average values of gross activity concentrations of 210 Po, 228 Th, 238 U, 234 U, 226 Ra and 228 Ra isotopes were determined to be 1.62 Bq L -1 in shallow groundwater and 5.88 Bq L -1 in deep groundwater, respectively. The total radioactivity in deep groundwater was higher than that in shallow groundwater, and most of the radioactivity in the water is from 226 Ra. Furthermore, the effective doses for ingestion of natural radionuclides were about 0.333 mSv y–1 for shallow groundwater and about 1.338 mSv y –1 for deep groundwater samples, which are higher than the World Health Organization (WHO, 2017) guideline level (0.1 mSv y –1 ) for drinking water. Ra dominated the total ingestion dose, contributing 93.06 % and 75.40 % of the total effective doses to the deep and shallow groundwater, respectively. The ingrowth and decay of natural radionuclides suggested that 228 Ra/ 226 Ra ratio can be a useful indicator of the source of radioactive contamination. The radioactivity data obtained from the investigated groundwater samples can be used to establish a baseline for radioactivity levels in groundwater around the WIPP site.

63 RADIATION, THERMAL, AND OTHER ENVIRON. POLLUTAN↗

Modeling of fission and activation products in molten salt reactors and their potential impact on the radionuclide monitoring stations of the International Monitoring System

Molten Salt Reactors (MSRs) are one of six Generation IV reactor designs currently under development around the world. Because of the unique operating conditions of MSRs, which include molten fuel and the continuous removal of gaseous fission products during operation, work was performed to analyze the potential impact of emissions on the International Monitoring System (IMS) of the Comprehensive Nuclear-Test-Ban Treaty (CTBT). Simulations were performed to predict the production of IMS-relevant radionuclides in four MSR designs operating under two scenarios: (1) a sealed reactor with releases only during operational shutdown, and (2) continuous reprocessing or sparging of the fuel salt. From these production estimates the radioxenon and radioiodine signatures were extracted and compared to three current reactor designs (Pressurized Water Reactor (PWR), BWR, RBMK). In cases where continuous reprocessing of the fuel salt occurred, both the radioxenon and radioiodine signatures were nearly indistinguishable from a nuclear explosion. Estimates were also made of the potential emission rate of radioxenon for three reactor designs and it was found that MSRs have the potential to emit radioxenon isotopes at a rate of 10^15-8×10^16 Bq/d for 133Xe if no abatement is used. An assessment was made of activation products using a candidate fuel salt (FLiBe) mixed with corrosion products for the Thorium Molten Salt Reactor (TMSR-LF1).

Johnson, Christine M.↗

Comment on “Investigations on ion irradiation induced modifications in Th 5f occupancy in thoria,” by P. Jegadeesan, S. Amirthapandian, Gurpreet Kaur, D. Sanjay Kumar, Arindam Das, D.K. Avasthi, K. Ananthasivan, B.K. Panigrahi

The primary purpose of this letter is to address some issues raised in the paper by Jegadeesan et al. There are three salient points. (1) The Branching Ratio (BR) analysis of thorium dioxide is complicated by peaks from Extended X-Ray Absorption Fine Structure (EXAFS). (2) Uranium metal is not an LS coupling but rather a delocalized 5f case. (3) Radiation damage effects in the BR measurements are small but strong effects can be seen in Resonant Photoemission.

36 MATERIALS SCIENCE↗

Ab-Initio molecular dynamics simulations of binary NaCl-ThCl 4 and ternary NaCl-ThCl 4 -UCl 3 molten salts

Molten salt reactor (MSR) with Thorium (Th) fuel cycle has attracted growing attention due to its merits such as safety, low radioactive waste production, and non-proliferation. The design and safe operation of MSR rely on a thorough understanding of the thermophysical properties of molten salts over a wide range of composition and temperature. However, experimental data of Th-based molten salts are limited due to the inherent challenges of dealing with corrosive molten salts and the radioactive nature of actinides. Here, in this work, thermophysical properties including density, heat capacity, thermal expansion coefficient, and mixing energy of binary NaCl-ThCl 4 and ternary NaCl-ThCl 4 -UCl 3 molten salts are explored using ab-initio molecular dynamic simulations (AIMD) with the dDsC dispersion correction. The calculated mixing energy of binary NaCl-ThCl 4 exhibits a minimum close to the eutectic composition. The heat capacity of the mixtures is linearly dependent on the mole fraction of each component. 7-fold and 6-fold coordinated Th complexes are dominant in the mixtures. 8-fold coordinated Th complex increases with ThCl 4 fraction due to the formation of network structures. The average coordination number of Th exhibits a minimum near the eutectic composition. The minimum mixing energy for the ternary mixtures is observed in systems with a composition close to [NaCl] 0.5 [ThCl 4 ] 0.25 [UCl 3 ] 0.25 . The density positively deviates from the ideal solution in mixtures near this composition. These results are important to fill the data and knowledge gap of ThCl 4 molten salts.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Absolute Branching Ratios of Prominent $γ$ Rays Following $α$ Decay of 233 U

The absolute $γ$-ray branching ratios (emission probabilities) for 19 prominent peaks in the $α$ decay of 233 U are reported with energies ranging from 97 to 366 keV. The branching ratios were obtained from $α-γ$ coincidence measurements, normalized to the $α$ singles. The high-purity 233 U material (> 99.98%) was separated from its decay products before preparation of a 4.74-mg/cm 2 uranyl nitrate foil with a 6-mm diameter. Furthermore, this is the first report on the absolute normalization of the branching ratios since 1984 and it is the only measure by $α-γ$ coincidences, where the $α-γ/α$ ratio removes several potential systematic biases. Overall, the results are consistent with the adopted literature, and they validate the quantities most relevant to the non-destructive assay of fissile 233 U material, which is relevant to the thorium fuel cycle and associated safeguards.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

A safeguards approach for molten salt reactors: Direct gamma-spectrometry of 233 U

Multidimensional gamma-spectrometry is a potential solution for safeguarding future thorium-fueled molten salt reactors (MSRs). A high-sensitivity system developed at Pacific Northwest National Laboratory (PNNL), the Advanced Radionuclide Gamma spectrOmeter (ARGO), has demonstrated direct 233 U assay in freshly irradiated fuel salt samples using a dual-detector configuration and the 97.1 keV gamma-emission. These measurements were made using 1.4 x 10 6 Bq of 233 U activity (at 1.4 x 10 8 fissions), at relatively high dead times (10.4%), and using short count durations of ~800 s. Other major emissions were also evaluated (42.5 keV, 54.7 keV, 146.4 keV), in additional to limited (1.96%) Compton suppression, but provided less consistent determination of the 233 U activity. Accurate 233 U measurement was also demonstrated using its four coincidence gamma-emissions (42.5 keV, 54.7 keV; 42.5 keV, 278.1 keV; 146.4 keV, 174.2 keV; 146.4 keV, 219 keV). Further improvements in coincidence measurements may be achievable with improvements in the detector configuration and nuclear data. Whilst this demonstration utilized the ARGO laboratory system, a similar dual-detector configuration could be utilized for online safeguards measurements.

Gamma-spectrometry↗

Distinguishing fissile uranium isotopes using an active well neutron coincidence counter

Proposed thorium-based nuclear fuel cycles are likely to require quantification and verification of 233 U within nuclear material. Because of their similar fission cross sections, active neutron nondestructive assay (NDA) systems may respond similarly to 233 U and 235 U. Traditional safeguards equipment has been optimized for 235 U and 238 U quantification associated with conventional uranium/plutonium fuel cycles and may not be directly applicable to 233 U quantification when mixed with other actinides. This work used models of the large volume active well coincidence counter (LV-AWCC) at Oak Ridge National Laboratory to evaluate the performance of this neutron NDA system to differentiate fissile uranium isotopes. The models were developed to simulate NDA system performance in response to a number of triangular radiation signature training device sources within the central cavity or well. This work predicted that the LV-AWCC can effectively differentiate 233 U from 235 U in certain modes of operation. In active mode, the LV-AWCC with the cadmium liner results in different doubles count rates between the fissile isotopes for a given fissile uranium mass. Without the cadmium liner, the uranium isotopes provide a statistically indistinguishable doubles count rate response for the fissile masses considered in this work (up to approximately 150 g). The cadmium liner serves to harden the neutron interrogation spectrum, which better exploits the notable difference in the 233 U and 235 U fission cross sections at approximately 1 eV. In passive mode, the two fissile isotopes exhibit different doubles and singles count rates regardless of liner presence because the passive source strength of 233 U is approximately 2 orders of magnitude stronger than that of 235 U due to the shorter half-life and correspondingly higher (α, n) yield. We conclude that using neutron interrogation in the LV-AWCC, two measurements are needed to quantify 233 U content in mixed uranium items. The first measurement is used to determine the total fissile uranium mass using a mode that cannot distinguish fissile isotopes (i.e., where a similar response is observed for both fissile uranium isotopes such as active doubles without cadmium or using a thermal neutron interrogation source). In conclusion, the second measurement is used to determine the 233 U content by using a differentiating technique (e.g., passive doubles, passive doubles to singles ratio, active doubles with cadmium).

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗