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At least 217 records · Page 12
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Opportunities for the Multi Recycling of Used MOX Fuel in the US - 12122

Over the last 50 years the US has accumulated an inventory of used nuclear fuel (UNF) in the region of 64,000 metric tons in 2010, and adds an additional 2,200 metric tons each year from the current fleet of 104 Light Water Reactors. This paper considers a fuel cycle option that would be available for a future pilot U.S. recycling plant that could take advantage of the unique opportunities offered by the age and size of the large U.S. UNF inventory. For the purpose of this scenario, recycling of UNF must use the available reactor infrastructure, currently LWR's, and the main product of recycling is considered to be plutonium (Pu), recycled into MOX fuel for use in these reactors. Use of MOX fuels must provide the service (burn-up) expected by the reactor operator, with the required level of safety. To do so, the fissile material concentration (Pu-239, Pu-241) in the MOX must be high enough to maintain criticality, while, in current recycle facilities, the Pu-238 content has to be kept low enough to prevent excessive heat load, neutron emission, and neutron capture during recycle operations. In most countries, used MOX fuel (MOX UNF) is typically stored after one irradiation in an LWR, pending the development of the GEN IV reactors, since it is considered difficult to directly reuse the recycled MOX fuel in LWRs due to the degraded Pu fissile isotopic composition. In the US, it is possible to blend MOX UNF with LEUOx UNF from the large inventory, using the oldest UNF first. Blending at the ratio of about one MOX UNF assembly with 15 LEUOx UNF assemblies, would achieve a fissile plutonium concentration sufficient for reirradiation in new MOX fuel. The Pu-238 yield in the new fuel will be sufficiently low to meet current fuel fabrication standards. Therefore, it should be possible in the context of the US, for discharged MOX fuel to be recycled back into LWR's, using only technologies already industrially deployed worldwide. Building on that possibility, two scenarios are assessed where current US inventory is treated; Pu recycled in LWR MOX fuels, and used MOX fuels themselves are treated in a continuous partitioning-transmutation mode (case 2a) or until the whole current UNF inventory (64,000 MT in 2010) has been treated followed by disposal of the MOX UNF to a geologic repository (case 2b). In the recycling scenario, two cases (2a and 2b) are considered. Benefits achieved are compared with the once through scenario (case 1) where UNF in the current US inventory are disposed directly to a geologic repository. For each scenario, the heat load and radioactivity of the high activity wastes disposed to a geologic repository are calculated and the savings in natural resources quantified, and compared with the once-through fuel cycle. Assuming an initial pilot recycling facility with a capacity of 800 metric tons a year of heavy metal begins operation in 2030, ∼8 metric tons per year of Pu is recovered from the LEUOx UNF inventory, and is used to produce fresh MOX fuels. At a later time, additional treatment and recycling capacities are assumed to begin operation, to accommodate blending and recycling of used MOX Pu, up to 2,400 MT/yr treatment capacity to enable processing UNF slightly faster than the rate of generation. Results of this scenario analysis study show the flexibility of the recycling scenarios so that Pu is managed in a way that avoids accumulating used MOX fuels. If at some future date, the decision is made to dispose of the MOX UNF to a geologic repository (case 2b), the scenario is neutral to final repository heat load in comparison to the direct disposal of all UNF (case 1), while diminishing use of natural uranium, enrichment, UNF accumulation, and the volume of HLW. Further recycling of Pu at the end of the scenario (case 2a) would exhibit further benefits. As expected, Pu-241 and Am-241 are the source of long term HLW heat load and Am-241 and Np-237 are the source of long term radiotoxicity. When advanced technology is available, introduction of minor actinide recycling, in addition to Pu recycling, by the end of this scenario, or sooner, would have a major impact on final repository heat load and long term radiotoxicity of the HLW. This scenario is also compatible with a gradual introduction of a small number of FR's for Pu management. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

A novel approach to criticality accident detection for a legacy facility

For operations on the Sellafield site involving fissile material a Criticality Accident Alarm System (CAAS) has traditionally been installed unless an omission case can be made. An omission case would examine if, in the absence of all controls, a criticality could reasonably be expected or whether the potential dose to an operator would remain below a maximum acceptable dose. It is a UK legal requirement for adequate emergency arrangements to be in place for hazards on a nuclear licensed site. For facilities where criticality is credible there is a regulatory expectation that a -Criticality Warning System- should be in place. A -Criticality Warning System- is not necessarily a full CAAS and alternative means of criticality detection may be part of an adequate emergency plan. The Sellafield site has a Magnox fuel reprocessing facility which has been operated since the 1960's. Evaporation of the Medium Active (MA) liquid waste from this process is undertaken in an associated building. Following extensive review of the plant and process a number of maloperations which can route fissile material to the MA evaporator were identified. Adequate controls to prevent these scenarios were shown to be in place. In the unlikely event of a failure of controls leading to sustained loss of fissile material a criticality incident may occur and hence an emergency plan was determined to be necessary. The novel emergency arrangements which were developed utilized the existing gamma monitor network located across multiple building floors. These gamma monitors are present for radiological safety. This emergency arrangement would require rapid interpretation of the gamma alarms by an operator and the manual triggering of a criticality evacuation alarm. The gamma monitors were demonstrated to be able to detect the minimum incident of concern (10{sup 15} fissions) as well as being able to alarm during the ramp up in power of a large criticality incident (10{sup 18} fissions). It was shown that a criticality accident could readily be distinguished from a radiological event (e.g. loss of containment). To support this work, criticality excursion analysis was used to examine the likely characteristics of a criticality in the MA evaporator. This novel arrangement was justified due to a number of factors including: the limited remaining operational lifetime of this facility (<5 years), the low frequency of maloperation coupled with a failure of controls and the crucial risk reduction provided by the continued reprocessing of Magnox fuel. Furthermore, the time and cost involved with design and installation of a standard CAAS would be substantially greater than a plan based upon existing instrumentation. These arguments were accepted by the UK regulator the Office for Nuclear Regulation (ONR). (author)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Developing a new criticality methodology to support decommissioning of legacy plutonium facilities

To decommission a legacy plutonium finishing facility, a number of gloveboxes need to be disposed of. These gloveboxes have been in place since the 1960's and were operational until the 1980's. A period of initial clean out and decommissioning took place in the 1990's and the first gloveboxes were removed from the line in the early 2000's. Compliance with downstream waste handling facility requirements meant that the gloveboxes needed to be removed using intensive 'hands-on' methods in order to size reduce them to fit into 200 litre (55 gallon) drums. This is the standard approach for handling waste items known as Plutonium Contaminated Material (PCM) on the Sellafield site and was credited in the criticality safety case for the decommissioning of the gloveboxes as it allowed each waste drum to be individually assayed such that the fissile content of the waste could be known. Not only was this approach time-consuming and laborious but as it involved operators working with cutting tools in a highly contaminated area it was challenged as being overly restrictive to operations and potentially hazardous to the workers. In order to decommission the remaining nine gloveboxes in the facility, a new decommissioning approach and associated criticality methodology has been developed. Instead of intensive size reduction and waste drum assay, the gloveboxes will be removed with little or no size reduction and put into storage crates without assay. The criticality assessor will review the operational history of each glovebox, along with the decommissioning records and relevant plant data, to make an evidence-based judgement that the crated glovebox represents no greater a residual criticality risk than would have been realised from the equivalent number of PCM drums that the standard decommissioning approach would have generated. This method identifies, in conjunction with the plant personnel, a number of review points to ensure that before the glovebox is removed any bulk residue material has been recovered and no more than a safe amount of moderating material (fixative) has been added to fix any remaining fissile contamination. Finally, once these review points have been confirmed to be complete - through a stakeholder meeting with criticality safety, plant operators and project managers - the glovebox will be placed in a crate, and this crate will be declared a Low Criticality Risk crate. No formal assay will be required prior to exporting the crate to the downstream waste handling facility and no fissile mass value will be assigned to the contents of the crate. No further intrusive work will be required on the crate and it can be safely stored without any spacing requirements, pending final disposal. (author)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Improving nuclear criticality safety in maintenance and construction activities at Y-12 national security complex

Maintenance and construction activities inside aging DOE nuclear facilities often impact nuclear criticality safety controls and requirements. Gaps in communication, training, and implementation of criticality safety guidance and controls in maintenance and construction work packages were identified at the Y-12 National Security Complex in 2016 when a maintenance package in the field was discovered to not have been reviewed by criticality safety. This incident resulted in a deficiency because criticality safety review is required any time maintenance or construction is working on equipment containing fissile material or their work impacts fissile material or fissile material activities. This review is required to protect the basis and requirements of criticality safety evaluations. To address these issues, the Y-12 criticality safety group reviewed a large number of maintenance packages, created a new method of communicating criticality safety controls to maintenance and construction, and developed a criticality safety training session designed specifically for maintenance and construction work planners. This paper presents the improvements in communication and implementation made between criticality safety, maintenance, and construction at the Y-12 National Security Complex and possible applications for other DOE sites. (author)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Point Kinetics Modelling of Decay Heat and the Xenon Effect

A point kinetics model has been used to simulate the impact of decay heat and xenon-135 on the behaviour, during various power transients, of a solution containing fissile material, such as those found in Aqueous Homogeneous Reactors (AHR) for medical isotope production. The heat produced by decay of radionuclides typically constitutes around 7% of total heat output during steady-state operation. Following a drop in reactivity, however, the fission power drops faster than the decay heat, and the latter may temporarily comprise a large proportion of the total heat output. Xenon-135 is a fission product and also a product of the decay of iodine-135. It acts as a strong neutron poison whose concentration can have a large impact on the reactivity of a fissile system. Following any reactivity insertion resulting in a change in the neutron flux, the concentration of xenon-135 changes, with a corresponding impact on reactivity. The model is based on those described in [1], [2] and [3]. The model has been adapted to represent a typical fissile solution and to model the effects of decay heat and xenon-135. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2015

On the benefit of using both hand calculation methods and computer experiments algorithms to solve complex nuclear criticality safety problems - 14299

A companion paper shows the 'anomalies', and even the safety issues, that may be raised by the very usual problem of making safe a fissile mass inside a controlled geometry. These anomalies mainly come from the solving of such problems which belong to the mathematical family of 'robust inversion' including two inversion parameters (the fissile mass and the geometry) and one nuisance parameter (the moderation ratio of the fissile material). Since a 'brute force' method of resolution requires a lot of calculations and since a 'smart' method of resolution requires a lot of engineer time, it is tempting to solve these problems by a 'custom made' approach rather than to achieve a 'whole' resolution. The present paper shows how one can use a hand calculation method (derivation of the one-speed diffusion theory) to perform the whole resolution of such kind of problems with a minimum calculations and human efforts. The powerful capability of predicting the 'anomalies' highlighted in the companion paper is shown. Then the limitations of such a method are discussed: limitation in the geometry of the problem and bias in absolute values prediction. Finally, new algorithms of computer experiments design, dedicated to solve robust inversion problems and based on Kriging surrogates, are tested to solve these problems with limiting calculations and human efforts. The performance of these new algorithms may supplement the one-speed diffusion theory limitations. (authors)

42 ENGINEERING↗
DOE OSTI2015

Minimum critical masses of heterogeneous moderated plutonium and uranium metal systems and their practical application to operation limits - 14355

In criticality safety analysis it is important to recognize the significant differences in reactivity achieved when material transitions from the solid to solution regime. However, the intermediate area of mixed heterogeneous systems (i.e. collections of small pieces) has not been well characterized. To that end, a study has been completed to determine the effect of fissile material piece size/shape on minimum critical mass of a system. The study presented represents a calculation-based investigation of moderated arrays of fissionable material to expand on established plutonium and uranium solution curves of LA-10860 to add the heterogeneous regime between pure a solid and solution of fissionable material. The analysis performed consists of parametric studies utilizing the neutron transport code MCNP6 to calculate the critical mass of systems composed of moderated arrays of varying sizes of spheres, cubes, or elongated rods (height >> diameter). The boundary of the array is increased to find the critical mass for the particular piece geometry and fissile volume density. The different shapes were then compared at points with the same fissile volume density to demonstrate that the key relationship between varying piece shape is the surface area to volume ratio. The results show that as pieces get larger (i.e. more heterogeneous) the minimum critical mass starts to increase. This effect results from the moderation of neutrons becoming less and less effective due to increased self-shielding of larger individual pieces. For plutonium pieces at a surface area to volume ratio of < 13 the reactivity increase from additional neutron moderation can no longer overcome the reactivity decrease from the reduced core density (separation of material pieces). At this point the minimum critical mass is achieved when the plutonium is in a solid metal configuration. This effect is not noticed for uranium which responds much more quickly to moderation and thus almost any sized pieces will increase in reactivity when moderation is introduced. Summarizing the results of the calculation a surface area to volume ratio vs minimum critical mass curve is generated. The curve gives analysts a practical starting minimum critical mass in analysis of heterogeneous systems with known minimum material surface area to volume ratios. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Atalante research facility new mass limits in a unit containing plutonium and minor actinides - 14402

ATALANTE, located in Marcoule, is one of the main Nuclear Facilities of the French Alternative Energies and Atomic Energy Commission (CEA). This facility is mainly dedicated to research and development for nuclear energy related to the back end of the fuel cycle: spent fuel and final waste management. It is a large research complex housing chemistry and analytical laboratories and shielded cells (240 glove boxes and 11 shielded cells), equipped for basic studies through to demonstration experiments. Within the facility, one unit can contain plutonium and curium at the same time. In this case, nuclear criticality safety is achieved by exercising control over mass. A subcritical mass limit is defined for plutonium (280 g for {sup 239}Pu) and an exemption mass limit (3 g for {sup 245}Cm) is used for the actinide. In order to increase the maximum mass (in particular {sup 245}Cm) which can be handled in the unit, new mass control limits were studied, based on control over the sum of the mass ratio to the upper subcritical mass, noted m{sub lim}, for each nuclide (350 g for {sup 239}Pu; 34.8 g or 23 g for {sup 245}Cm) as shown in equation (1) below. m({sup 239}Pu)/ m{sub lim}({sup 239}Pu)+ m({sup 245}Cm)/m{sub lim}({sup 245}Cm)≤1 (1) To validate this criticality safety control method, a study was conducted to demonstrate that, for different masses of {sup 239}Pu and {sup 245}Cm meeting equation (1), the calculated effective multiplication factor is always lower than the effective multiplication factor corresponding to a sphere of 350 g of pure {sup 239}Pu, moderated, with 20 cm water reflection. The two pseudo fissile media are: - metallic plutonium (100% {sup 239}Pu, density 19.86), moderated with water, - metallic curium (100% {sup 245}Cm, density 13.566), moderated with water. The calculation model is a sphere of {sup 245}Cm metal positioned in a sphere of {sup 239}Pu metal, with optimum moderation for each fissile material, and a full water reflection (20 cm). The results show that, under certain conditions, the effective multiplication factor of the mixture can be higher than the effective multiplication factor of each pure fissile material when using the upper subcritical mass of 34.8 g for {sup 245}Cm. When using 23 g of {sup 245}Cm, the k{sub eff} of the mixture is always lower than the k{sub eff} of pure {sup 239}Pu. These calculations enable the safety documentation to be changed by introducing a new process operational procedure, based on equation (1). This change in procedure allows the operator to work with a larger quantity of curium in the unit. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Storage and concentration of plutonium and uranium solutions in LES 401 hot cell (ATALANTE) - criticality safety considerations - 14414

ATALANTE, located in Marcoule, is one of the main nuclear facilities of the French CEA (Atomic Energy and Alternative Energies Commission). The LES 401 hot cell was built in the 1980's and, since then, has been operating only with low masses of fissile material. As part of the management optimization for purified Uranium and Plutonium solutions ('Fin U Pu' Project), the new functions provided in LES 401 are now expected to be the following: - Reception of Uranium and Plutonium solutions, - Concentration of these solutions, - Storage of the concentrates and distillates. From the criticality safety point of view, the equipment was originally designed in safe geometry for Plutonium Nitrate, considering only a 2.5 cm -thick water reflector. However, the new criticality safety case had to take into account the following points: - Reception of not only Plutonium solutions, but also highly enriched Uranium solutions, - Considering flood risk as a possible contingency, - Considering precipitation as a possible contingency. These more conservative assumptions have meant introducing a mass limit for the cell, in addition to the geometry control mode. For the equipment containing very low concentrations of fissile material in normal operation, it was necessary, using the double contingency principle, to take a concentration limit into account in the specific situation of water flooding the cell. Finally, given the mass and geometry control mode, it was necessary to perform all the calculations with both fissile media (Plutonium and Uranium), as it was found that depending on the selected mass, geometry, and reflection conditions, the most penalizing media is not always the same. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Status of French regulation regulations concerning nuclear criticality safety - 14585

Hitherto, in France, licensees used to demonstrate compliance with Fundamental Safety Rule I.3.c. This Rule sets out the principles for demonstrating Nuclear Criticality Safety in all nuclear facilities (excluding reactors). Although there is a good level of consensus between the licensees and the Regulator, this Rule is now somewhat dated (it came into force in 1984) and does not cover all aspects. The French Nuclear Safety Authority (ASN) used the 2006 Law on Transparency and Nuclear Safety as the catalyst for a thorough revision and updating of its regulatory documentation. Thus, in 2010, ASN decided to set up a 'criticality working group' involving ASN, the licensees (AREVA, CEA, EDF...) and the Technical Safety Organization (IRSN). This group was tasked with drafting a regulatory text (a 'resolution') and a criticality guide. The objectives were to update Rule I.3.c, clearly defining the scope for reactors and transport packages of fissile material, to further develop areas not covered in the existing rule and to take account of the feedback and lessons learned from past events. Thus, this paper presents the status of French regulations concerning Nuclear Criticality Safety (NCS). Firstly, it focuses on the existing Rule I.3.c and then explains how this will be modified by the forthcoming improvements to the resolution and the criticality guide. It in particular addresses the scope of the new regulations (concerning reactors and the transport of fissile material), how the principle of defence in depth is to be applied to criticality safety, the double contingency principle, control modes, the reference fissile medium, the organization, the role of the Criticality Engineer, admissibility criteria, validation, etc. (authors)

42 ENGINEERING↗
DOE OSTI2014

Gamma/neutron time-correlation for special nuclear material detection – Active stimulation of highly enriched uranium

A series of simulations and experiments were undertaken to explore and evaluate the potential for a novel new technique for fissile material detection and characterization, the timecorrelated pulse-height (TCPH) method, to be used concurrent with active stimulation of potential nuclear materials. In previous work TCPH has been established as a highly sensitive method for the detection and characterization of configurations of fissile material containing Plutonium in passive measurements. By actively stimulating fission with the introduction of an external radiation source, we have shown that TCPH is also an effective method of detecting and characterizing configurations of fissile material containing Highly Enriched Uranium (HEU). The TCPH method is shown to be robust in the presence of the proper choice of external radiation source. An evaluation of potential interrogation sources is presented.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2019

Modeling a fast spectrum molten salt reactor in a systems dynamics fuel cycles code

Molten salt reactor (MSR) concepts are being actively developed by several private companies, with some touting the fuel cycle benefits of this technology. Simulating the deployment of MSRs and their effect on fuel cycle transitions is the role of a system dynamics fuel cycles tool. However, many current fuel cycles tools have difficulty modeling MSR isotopic variations because they were designed for modeling existing reactors based on discrete fuel elements with batch fueling schemes (i.e., charge and discharge). This work presents ORION fuel cycle modeling results using a specific high power density fast spectrum MSR design. For code verification, a single-stage MSR fuel cycle was set up in ORION and results were compared with the results from the SCALE-based reactor physics model (ChemTriton). A transition scenario was set up in ORION from the current light water reactor fleet to a future fleet of MSRs to study the characteristics of an MSR that affect its performance during transition. Several transition scenarios were set with holds and delays in place that are representative of a sodium fast reactor as modeled in Evaluation Group 23 for the Fuel Cycle Options Campaign Evaluation and Screening Study. Several MSR characteristics lend to its high performance and ease in transition from the current fuel cycle: fissile material is only loaded at initial startup; all fissile material is available as it is bred due to online fuel processing; and fissile material recirculating through the core is briefly held in tanks for processing and has very minimal out-of-core time (on the order of a minutes to a few days). These characteristics are in contrast to a solid-fueled reactor, which requires discharge, cooling, and fabrication of fuel all of which could take a few months to a few years and repeated loading of fresh fuel assemblies. In conclusion, a solid-fueled reactor only discharges a fraction of a core every year.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Strategies for thorium fuel cycle transition in the SD-TMSR

Liquid-fueled Molten Salt Reactor (MSR) systems represent advances in safety, economics, and sustainability. The MSR has been designed to operate with a Th/ 233 U fuel cycle with 233 U used as startup fissile material. Since 233 U does not exist in nature, we must examine other available fissile materials to start up these reactor concepts. This work investigates the fuel cycle and neutronics performance of the Single-fluid Double-zone Thorium-based Molten Salt Reactor (SD-TMSR) with different fissile material loadings at startup: High Assay Low Enriched Uranium (HALEU) (19.79%), Pu mixed with HALEU (19.79%), reactor-grade Pu (a mixture of Pu isotopes chemically extracted from Pressurized Water Reactor (PWR) spent nuclear fuel (SNF) with 33 GW d/tHM burnup), transuranic elements (TRU) from Light Water Reactor (LWR) SNF, and 233 U. The MSR burnup routine provided by SERPENT-2 is used to simulate the online reprocessing and refueling in the SD-TMSR. The effective multiplication factor, fuel salt composition evolution, and net production of 233 U are studied in the present work. Additionally, the neutron spectrum shift during the reactor operation is calculated. The results show that the continuous flow of reactorgrade Pu helps transition to the thorium fuel cycle within a relatively short time (≈ 4.5 years) compared to 26 years for 233 U startup fuel. Finally, using TRU as the initial fuel materials offers the possibility of operating the SD-TMSR for an extended period of time (≈ 40 years) without any external feed of 233 U.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Conceptual design study of neutron detectors for safeguards measurement of an irradiated pebble

Nuclear material control and accounting (MC&A) of pebble-bed reactors (PBRs) is challenging because a PBR utilizes hundreds of thousands of identical, unmarked pebbles that are continuously recirculated through the core. To develop tools that enable the implementation of international safeguards, especially in the context of MC&A of spent pebbles, we designed and simulated three neutron detection concepts to determine fissile content in individual pebbles: a differential die-away (DDA) detector, a californium interrogation prompt neutron (CIPN) detector, and a passive neutron albedo reactivity (PNAR) detector using Monte Carlo calculations. Burnup calculations were performed on the spent pebbles from the PBMR-400 classic PBR. The varying neutron and gamma source terms, and isotopic compositions in the spent pebbles calculated at various burnup levels were used in the neutron detector models. DDA was found to be sensitive to the number of passes a pebble has had through the core and to the fissile content contained in a spent pebble. Optimization in the DDA design further increased the neutron count rates and thus reduced counting uncertainty. Meanwhile, passive neutron counting using the same detector body could distinguish pebbles with different numbers of passes, but its response was dominated by neutron-emitting actinides and was not sensitive to fissile content. On the other hand, the PNAR technique was not viable for a single pebble but performed reasonably for a 27-pebble array, which suggested potential use for verification measurements of containers filled with 27 or more spent pebbles.

CIPN↗
DOE OSTI2025

A Semi-Empirical Density Law for Ternary, Homogeneous PuCl 3 /HCl/H 2 O Solutions

Nuclear material operations pose unique hazards that are not encountered in other chemical, energy, or manufacturing industries. One of these hazards is the potential for a nuclear criticality accident when handling fissile isotopes such as 235 U and 239 Pu. These hazards are particularly high when fissile material is dissolved in solution as the neutron behaviors of the system can change rapidly with the physical and chemical changes accessible in solution. Current estimates of solution density used for criticality safety are outdated and hinder fissile material handling. Developing new estimates for these safety calculations requires experimental characterization and the derivation of empirical density models. We have derived a density law describing PuCl 3 /HCl/H 2 O solutions from experimental data characterizing solution density. Density data was treated using a Pitzer-derived eight-parameter equation, defining density as a function of analyte concentrations, temperature, and interactions between these variables. The model is predictive across the concentration and temperature ranges from which it was derived. The potential effects of varying oxidation states of plutonium, which are easily accessible in aqueous media, on the bulk solution density of the ternary system were also investigated. The resulting Pitzer-derived density law was applied to a nuclear criticality safety model, and the impact of the experimental characterization of solution density relative to previous estimates was demonstrated to be significant and suggest that the current approach to estimating density in nuclear criticality safety calculations may lead to overly conservative controls.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2011

Development of the quantum theory of T-odd asymmetries for prescission and evaporated third particles in ternary nuclear fission induced by cold polarized neutrons

A comparative analysis of the results obtained by experimentally and theoretically studying T-odd asymmetries for various third particles in the true and delayed ternary nuclear fission induced by cold polarized neutrons was performed. It was confirmed that the appearance of these asymmetries was associated with the effect of rotation of a polarized system undergoing fission on the angular distributions of prescission and evaporated third particles with respect to the direction along which the emerging fission fragments flew apart, this effect being determined by the Coriolis interaction of the rotational and the internalmotion of the fissioning system. A quantum-mechanical description of particle motion in a rotating coordinate system was generalized to the case where gamma-ray emission was present. It was shown that the separation of the motions of an axially symmetric fissile system into a rotational and an internal motion was valid in the external region as well, where ternary-fission products had already been formed, if it was considered that the motion of fission fragments was tightly connected with the system symmetry axis, which rotated in the laboratory frame. It was found that the dependence of the fissile-system moment of inertia appearing in the Coriolis interaction Hamiltonian on the distance between the fission fragments flying apart generated an additional phase in the amplitude of the radial distribution of fission fragments. It was shown that this phase might change sizably the contribution of the interference between fission amplitudes of neutron resonances excited in a fissile compound nucleus to the absolute values of T -odd asymmetries, especially for third particles such as neutrons and photons, which interacted only slightly with fission fragments.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

Role of dynamical effects in the formation of T-Odd asymmetries for products of polarized-neutron-induced ternary fission of nuclei

Basic dynamical effects that accompany the cold-polarized-neutron-induced binary and ternary fission of actinide nuclei and which determine the properties of T -odd asymmetries in angular distributions of various prescission and evaporated light third particles emitted in true and delayed ternary fission are analyzed on the basis of quantum-mechanical fission theory. It is emphasized that effects associated with the conservation of axial symmetry of the fissioning system under study at all stages of its evolution from the formation of neutron resonance states of the fissile compound nucleus to the separation of its fission fragments, including the appearance of zero wriggling vibrations of the cold compound nucleus in the vicinity of its scission point, are of particular importance, the influence of quantum collective rotation of the polarized fissile system on the asymmetry of the angular distribution of both fission fragments and third particles being taken into account. It is shown that the difference in the behavior of the coefficients characterizing the T -odd asymmetries under analysis for the target nuclei being studied can be explained, upon taking into account the interference between the fission amplitudes for the neutron resonance states of fissile compound nuclei, by the difference in the contributions of even and odd components of the amplitudes of angular distributions of third particles to the coefficients in question.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Sequential character of low-energy ternary and quaternary nuclear fission

An analysis of low-energy true ternary (quaternary) nuclear fission leads to the conclusion that these fission modes have a sequential two-step (three-step) character such that the emission of a third particle (third and fourth particles) and the separation of fission fragments occur at distinctly different instants, in contrast to the simultaneous emergence of all fission products in the case of onestep ternary (quaternary) fission. This conclusion relies on the following arguments. First, the emission of a third particle (third and fourth particles) from a fissile nucleus is due to a nonevaporative mechanism associated with a nonadiabatic character of the collective deformation motion of this nucleus at the stages preceding its scission. Second, the axial symmetry of the deformed fissile compound nucleus and the direction of its symmetry axis both remain unchanged at all stages of ternary (quaternary) fission. This circumstancemakes it possible to explain themechanism of the appearance of observed anisotropies and T — odd asymmeries in the angular distributions of products of ternary (quaternary) nuclear fission. Third, the T —odd asymmetry discovered experimentally in ternary nuclear fission induced by cold polarized neutrons obeys the T —invariance condition only in the case of a sequential two-step (three-step) character of true ternary (quaternary) nuclear fission. At the same time, this asymmetry is not a T —invariant quantity in the case of the simultaneous emission of products of true ternary (quaternary) nuclear fission from the fissile compound nucleus.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗