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DOE OSTI2018

Unified Mechanism behind the Appearance of T-Odd TRI and ROT Asymmetries in Actinide Fission Induced by Cold Polarized Neutrons

Some shortcomings of the approaches that are used to describe T-odd ROT and TRI asymmetries in true ternary fission via reactions involving the emission of prescission alpha particles and which are based on employing the classical method of trajectory calculations are analyzed. These shortcomings are caused by the disregard of the interference between the fission widths of different sJ{sub s} neutron resonance states formed in the first well of the deformation potential of fissile compound nuclei. It is shown that the method used in some studies to determine T-odd TRI-asymmetries for prescission alpha particles is at odds with basic concepts of the generalizedmodel of the nucleus and approaches to constructing collective (for example, bending) vibrations of a fissile compound nucleus. Quantum-mechanical fission theory is generalized via employing a unified mechanism of formation of T-odd TRI and ROT asymmetries for prescission alpha particles and evaporated photons (neutrons). The proposed mechanism takes correctly into account the effect of quantum rotation of a fissile compound nucleus on the angular distributions of fission fragments and alpha particles for true ternary fission, as well as on the angular distribution of prompt photons (neutrons) emitted by fragments originating from the delayed fission of the aforementioned nuclei.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2012

LWR First Recycle of TRU with Thorium Oxide for Transmutation and Cross Sections

Thorium has been considered as an option to uranium-based fuel, based on considerations of resource utilization (thorium is approximately three times more plentiful than uranium) and as a result of concerns about proliferation and waste management (e.g. reduced production of plutonium, etc.). Since the average composition of natural Thorium is dominated (100%) by the fertile isotope Th-232, Thorium is only useful as a resource for breeding new fissile materials, in this case U-233. Consequently a certain amount of fissile material must be present at the start-up of the reactor in order to guarantee its operation. The thorium fuel can be used in both once-through and recycle options, and in both fast and thermal spectrum systems. The present study has been aimed by the necessity of investigating the option of using reprocessed plutonium/TRU, from a once-through reference LEU scenario (50 GWd/ tIHM), mixed with natural thorium and the need of collect data (mass fractions, cross-sections etc.) for this particular fuel cycle scenario. As previously pointed out, the fissile plutonium is needed to guarantee the operation of the reactor.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2012

Safeguard Requirements for Fusion Power Plants

Nuclear proliferation risks from magnetic fusion energy associated with access to fissile materials can be divided into three main categories: 1) clandestine production of fissile material in an undeclared facility, 2) covert production and diversion of such material in a declared and safeguarded facility, and 3) use of a declared facility in a breakout scenario, in which a state openly produces fissile material in violation of international agreements. The degree of risk in each of these categories is assessed, taking into account both state and non-state actors, and it is found that safeguards are required for fusion energy to be highly attractive from a non-proliferation standpoint. Specific safeguard requirements and R&D needs are outlined for each category of risk, and the technical capability of the ITER experiment, under construction, to contribute to this R&D is noted. A preliminary analysis indicates a potential legal pathway for fusion power systems to be brought under the Treaty for the Non-Proliferation of Nuclear Weapons. "Vertical" proliferation risks associated with tritium and with the knowledge that can be gained from inertial fusion energy R&D are outlined.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2013

LWR First Recycle of TRU with Thorium Oxide for Transmutation and Cross Sections

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2013

Investigation of the Performance of D 2 O-Cooled High-Conversion Reactors for Fuel Cycle Calculations

This report presents FY13 activities for the analysis of D 2 O cooled tight-pitch High-Conversion PWRs (HCPWRs) with U-Pu and Th-U fueled cores aiming at break-even or near breeder conditions while retaining the negative void reactivity. The analyses are carried out from several aspects which could not be covered in FY12 activities. SCALE 6.1 code system is utilized, and a series of simple 3D fuel pin-cell models are developed in order to perform Monte Carlo based criticality and burnup calculations. The performance of U-Pu fueled cores with axial and internal blankets is analyzed in terms of their impact on the relative fissile Pu mass balance, initial Pu enrichment, and void coefficient. In FY12, Pu conversion performances of D 2 O-cooled HCPWRs fueled with MOX were evaluated with small sized axial/internal DU blankets (approximately 4cm of axial length) in order to ensure the negative void reactivity, which evidently limits the conversion performance of HCPWRs. In this fiscal year report, the axial sizes of DU blankets are extended up to 30 cm in order to evaluate the amount of DU necessary to reach break-even and/or breeding conditions. Several attempts are made in order to attain the milestone of the HCPWR designs (i.e., break-even condition and negative void reactivity) by modeling of HCPWRs under different conditions such as boiling of D 2 O coolant, MOX with different 235 U enrichment, and different target burnups. A similar set of analyses are performed for Th-U fueled cores. Several promising characteristics of 233 U over other fissile like 239 Pu and 235 U, most notably its higher fission neutrons per absorption in thermal and epithermal ranges combined with lower ___ in the fast range than 239 Pu allows Th-U cores to be taller than MOX ones. Such an advantage results in 4% higher relative fissile mass balance than that of U-Pu fueled cores while retaining the negative void reactivity until the target burnup of 51 GWd/t. Several other distinctions between U-Pu and Th-U fueled cores are identified by evaluating the sensitivity coefficients of keff, mass balance, and void coefficient. The effect of advanced iron alloy cladding (i.e., FeCrAl) on the performance of Pu conversion in MOX fueled cores is studied instead of using standard stainless-steel cladding. Variations in clad thickness and coolant-to-fuel volume ratio are also exercised. The use of FeCrAl instead of SS as a cladding alloy reduces the required Pu enrichment and improves the Pu conversion rate primarily due to the absence of nickel in the cladding alloy that results in the reduction of the neutron absorption. Also the difference in void coefficients between SS and FeCrAl alloys is nearly 500 pcm over the entire burnup range. The report also shows sensitivity and uncertainty analyses in order to characterize D 2 O cooled HCPWRs from different aspects. The uncertainties of integral parameters (keff and void coefficient) for selected reactor cores are evaluated at different burnup points in order to find similarities and trends respect to D 2 O-HCPWR.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

IER 203 CED-2 Report: LLNL Final Design for BERP Ball With a Composite Reflector of Thin Polyethylene Backed by Nickel

This report documents the results of final design (CED-2) for IER 203, BERP Ball Composite Reflection, and focuses on critical configurations with a 4.5 kg α-phase plutonium sphere reflected by a combination of thin high-density polyethylene (HDPE) backed by a thick nickel reflector. The Lawrence Livermore National Laboratory’s (LLNL’s) Nuclear Criticality Safety Division, in support of fissile material operations, calculated surprisingly reactive configurations when a fissile core was surrounded by a thin, moderating reflector backed by a thick metal reflector. These composite reflector configurations were much more reactive than either of the single reflector materials separately. The calculated findings have resulted in a stricter-than-anticipated criticality control set, impacting programmatic work. IER 203 was requested in response to these seemingly anomalous calculations to see if the composite reflection effect could be shown experimentally. This report focuses on the Beryllium Reflected Plutonium (BERP) ball as a fissile material core reflected by polyethylene and nickel. A total of four critical configurations were designed as part of CED-2. Fabrication costs are estimated to be $98,500, largely due to the cost of the large nickel reflectors. The IER 203 experiments could reasonably be expected to begin in early FY2017.

42 ENGINEERING↗

DOE OSTI2017

Hybrid Gama Emission Tomography (HGET): FY16 Annual Report

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2017

Scattering Angular Distributions in the ENDF/B Nuclear Data Library

Nuclear reactors function by controlling the chain reaction of fission events in a fissile material. Understanding the gain and loss of neutrons is crucial for controlling this chain reaction. The role of fission in liberating neutrons from nuclei is well known. The fact that neutrons can scatter out of a reactor and therefore be lost is less well appreciated. Scattering can happen within fissile fuel or off of any number of the non-fissile structural components holding a reactor together. The scattering angular distributions for reactor material are the quantity that give the probability for a particle (usually a neutron) to scatter off a nucleus into a given angle. The Evaluated Nuclear Data File/B (ENDF/B) nuclear data library is the most complete and authoritative reference for neutron scattering data, containing reaction cross sections and outgoing particle distributions (including scattering angular distributions). In this project, we will compare the scattering angular distributions given in these ENDF/B data sets with those that can be derived using the Blatt-Biedenharn formalism and the resolved resonance data also in the ENDF/B files. More comparisons between the scattering angular distributions of relevant isotopes will help identify and correct any errors in the ENDF/B files. The results of this study will be an important component of the next major release of the ENDF/B library.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2019

Simulations of Pipe Overpack Container Compaction at the Waste Isolation Pilot Plant

The Waste Isolation Pilot Plant (WIPP) is an operating geologic repository in southeastern New Mexico for transuranic (TRU) waste from nuclear defense activities. Past nuclear criticality concerns have generally been low at the WIPP due to the low initial concentration of fissile material and the natural tendency of fissile solute to disperse during fluid transport in porous media (Rechard et al. 2000). On the other hand, the list of acceptable WIPP waste types has expanded over the years to include Criticality Control Overpack (CCO) containers and Pipe Overpack (POP) containers. Containers bound for WIPP are bundled together in hexagon shaped 7-packs (six containers surround one container in the center). Two 7-packs are often combined into a TRUPACT-II package for a total of 14 containers. Most TRUPACT-II packages are restricted to a maximum fissile mass equivalent to plutonium (FMEP) between 0.1 and 0.38 kg, but a CCO TRUPACT-II package and a POP TRUPACT-II package are respectively permitted to have 5.32 kg and 2.80 kg FMEP (see Section 3 of US DOE (2013)). Consequently, CCO container criticality after emplacement at the WIPP was evaluated in Saylor and Scaglione (2018), and Oak Ridge National Laboratories is currently at work on POP container criticality analyses.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

NASA NTRS2011

Low Cost Nuclear Thermal Rocket Cermet Fuel Element Environment Testing

Deep space missions with large payloads require high specific impulse and relatively high thrust to achieve mission goals in reasonable time frames.1,2 Conventional storable propellants produce average specific impulse. Nuclear thermal rockets capable of producing high specific impulse are proposed. Nuclear thermal rockets employ heat produced by fission reaction to heat and therefore accelerate hydrogen, which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000 K), and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited.3 The primary concern is the mechanical failure of fuel elements that employ high-melting-point metals, ceramics, or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. The purpose of the testing is to obtain data to assess the properties of the non-nuclear support materials, as-fabricated, and determine their ability to survive and maintain thermal performance in a prototypical NTR reactor environment of exposure to hydrogen at very high temperatures. The fission process of the planned fissile material and the resulting heating performance is well known and does not therefore require that active fissile material be integrated in this testing. A small-scale test bed designed to heat fuel element samples via non-contact radio frequency heating and expose samples to hydrogen is being developed to assist in optimal material and manufacturing process selection without employing fissile material. This paper details the test bed design and results of testing conducted to date.

Bradley, D. E.↗

DOE OSTI2013

Update on Establishing the Feasibility of Lead Slowing Down Spectroscopy for Direct Measurement of Plutonium in Used Fuel

Developing a method for the accurate, direct, and independent assay of the fissile isotopes in bulk materials (such as used fuel) of next-generation domestic nuclear fuel cycles is a goal of the Office of Nuclear Energy, Fuel Cycle R&D, Material Protection and Control Technology (MPACT) Campaign. To meet this goal, MPACT supports a multi-institutional collaboration to address the feasibility of Lead Slowing Down Spectroscopy (LSDS) as an active, nondestructive assay method. LSDS has the potential to provide independent, direct measurement of Pu and U isotopic masses in used fuel with an uncertainty considerably lower than today’s confirmatory assay methods, for which typical uncertainties are approximately 10%. LSDS techniques are sensitive to the fission resonances in the energy range of ~0.1-1000 eV, enabling their use to determine the mass content of the fissile isotopes in used fuel. This paper will present an update with regard to applying LSDS for used fuel assay and the development of algorithms to extract fissile isotopic masses from the used fuel.

Lead Slowing Down Spectroscopy, Used Fuel, Spent F↗

DOE OSTI2019

Apparatus and method for generating medical isotopes

An apparatus for generating medical isotopes provides an annular fissile solution vessel surrounding a neutron generator. The annular fissile solution vessel provides for good capture of the emitted neutrons and a geometry that provides enhanced stability in an aqueous reactor. A neutron multiplier and/or a neutron moderator may be used to improve the efficiency and control the criticality of the reaction in the annular fissile solution vessel.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2017

Hybrid Gamma Emission Tomography for the Verification of Unirradiated Fuel: A Viability Study

Current International Atomic Energy Agency (IAEA) methodologies for the verification of fresh low-enriched uranium (LEU) and mixed oxide (MOX) fuel assemblies are volume-averaging methods that lack sensitivity to individual pins. Further, as unirradiated fuel assemblies become more and more complex (e.g., heavy gadolinium loading, high degrees of axial and radial variation in fissile concentration), the accuracy of current IAEA instruments degrades and measurement time increases. Particularly in light of the fact that no special tooling is required to remove individual pins from modern fuel assemblies, new capabilities for the verification of unirradiated (i.e., fresh LEU and MOX) assemblies are needed to ensure that fissile material has not been diverted. Passive gamma emission tomography has demonstrated potential to provide pin-level verification of spent fuel, but gamma-ray emission rates from unirradiated fuel emissions are significantly lower, precluding purely passive tomography methods. The work presented here introduces the concept of Hybrid Gamma Emission Tomography (HGET) for verification of unirradiated fuels, in which a neutron source is used to actively interrogate the fuel assembly and the resulting gamma-ray emissions are imaged using tomographic methods to provide pin-level verification of fissile material concentration. This paper describes the status of a viability study on the HGET concept, including: envisioned use-case scenarios and corresponding definitions of fuel assemblies; modeling framework based on Monte Carlo and deterministic transport methods, and its validation; quantitative assessment of candidate HGET signatures with a focus on prompt fission gamma rays and delayed fission gamma rays; a nominal HGETv1 instrument design; candidate HGET-specific tomographic reconstruction methods that fully incorporate declared information; and examples of simulation-based predictions of HGET performance.

Miller, Erin A.↗

DOE OSTI2022

Apparatus and method for generating medical isotopes

Piefer, Gregory Richard↗

DOE OSTI2022

Coated fuel pellets with enhanced water and steam oxidation resistance

Disclosed herein is a method comprising coating a fissile, uranium-containing ceramic material with a water-resistant layer, the layer being non-reactive with the fissile, uranium-containing ceramic material. The coating is applied to a surface of the fissile, uranium-containing ceramic material. Also disclosed is a fuel for use in a nuclear reactor.

Lahoda, Edward J.↗

DOE OSTI2017

Challenges and Recent Developments of Neutron Non-Destructive Assay with UO{sub 2}F{sub 2}-Bearing Materials - 17336

In the enrichment process of an operating Gaseous Diffusion Plant (GDP), such as the Portsmouth and Paducah facilities, UF6 gas will immediately hydrolyze forming the chemical compound UO{sub 2}F{sub 2} (Uranyl Fluoride). This mostly happens in Alumina (Al{sub 2}O{sub 3}) traps where the UF6 adsorbs to moistened Alumina pellets forming a UO{sub 2}F{sub 2} layer around the pellet. Similarly, NaF.UF{sub 6} absorption traps produce the complex Na{sub 2}UF{sub 8} which also hydrolyzes when exposed to moist air. UO{sub 2}F{sub 2} can also be formed in many other processes where UF6 gas comes into contact with moisture within structures and objects such as pipes, converters compressors and coolers. When the structures are refurbished, or designated for D and D, UO{sub 2}F{sub 2}-bearing objects are generated both in situ and ex situ requiring fissile (U-235) characterization. In addition, ex situ waste streams are generated when UO{sub 2}F{sub 2}-bearing materials are placed in drummed or boxed waste containers. Gamma spectroscopy Non-Destructive Assay (NDA) is commonly utilized to characterize UO{sub 2}F{sub 2}-bearing waste but generally has high associated Total Measurement Uncertainties (TMU's) and, in some cases, Data Quality Objectives (DQO's) cannot be met. In addition, high UO{sub 2}F{sub 2} mass loadings, such as hold-up materials, require gamma modeling when infinitely thick to the Uranium energy lines. A common neutron NDA technique, or method, is to exploit the neutrons generated from the F-19(a, n)Na-22 reaction which has an appreciable reaction cross-section. The isotopes of Uranium are a-emitters, and, in particular, U-234 has a significantly higher a-decay reaction cross-section leading to a significant relative neutron Specific Activity (nSA). Neutron NDA is appealing since, in some cases, it has potentially lower TMU's than gamma NDA and neutron detector slabs facilitate surrounding large, complex-shaped, objects. Characterizing UO{sub 2}F{sub 2}-bearing waste for fissile content using neutrons generated from U-234, requires the U-235/U-234 ratio as well as the enrichment as well as the corresponding nSA's. It is also noted that U-238 has a considerable, relative, spontaneous fission nSA and this is a higher concern for low-enrichments where the mass of U-238 dominates. Additional neutron interfering elements include cosmic-ray induced spallation within the objects and background from other, in situ, UO{sub 2}F{sub 2}-bearing objects. In this paper we present the challenges and recently developed hardware and algorithmic strategies for quantifying fissile mass both in situ and ex situ using neutron NDA that works in conjunction to gamma spectroscopy NDA, or process knowledge, for objects and drummed/boxed waste containing UO{sub 2}F{sub 2}-bearing materials leading to minimal TMU's and achievable DQO's. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2015

Increased fissionable material loading of vitrified borosilicate glass in support of efficient disposition of nuclear materials at Savannah River - 15299

The US DOE is vitrifying liquid HLW, converting it into a glass form suitable for long term storage and disposal. The waste includes fissionable isotopes and, as a consequence, may possess an inadvertent criticality hazard. The technical basis for determining when an inadvertent criticality hazard exists is unclear. ANS criticality safety standards do not provide relevant subcritical safety limits and little experimental data exist to support development of such a limit. As a result, the DOE has implemented a conservative limit of 897 grams of fissile nuclides per cubic meter of glass. However, this limit has the potential to impede efficient disposal of various fissionable materials at the Savannah River Site. Work is being conducted to establish an alternate fissionable material limit which would be less restrictive, but still ensure criticality safety. A series of computer models are constructed and analyzed using a general purpose Monte Carlo N-Particle neutron transport code. First, a baseline model is established and analyzed. Next, numerous variations to the baseline model (i.e., perturbations) are analyzed with the code to understand the impact of the change. Based on the results, it is evident the atomic ratio of B-10 to equivalent Pu-239 would be an effective means to ensure subcriticality of HLW vitrified in a borosilicate glass matrix. The proposed limit, its technical basis, and all necessary application constraints are discussed. This proposed limit is expected to permit significantly higher quantities of fissile nuclides in the vitrified waste form. It will also account for several non-fissile, fissionable nuclides which are expected to add reactivity to the glass waste form. The proposed limit embodies large conservatisms intended to provide assurance of safety in spite of the significant uncertainties associated with its development. The single most important uncertainty arises from the inability to validate the calculations which form its technical basis. While they are done with the latest available nuclear data and well established analytical methods, they are, nonetheless, not validated as required for subcritical safety limits. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

Heavy Water Moderated, Molten Uranium Cooled Reactor MCNP Model

Two of the major challenges facing the nuclear industry are the large amounts of nuclear waste produced by the current light water reactors and the risk of nuclear proliferation. Most proposed solutions for the waste problem involve fast breeder reactors, but that doesn't help with the nuclear proliferation problem. It is generally understood that a uranium fueled (U-238 + U-235) thermal reactor cannot be a breeder because there aren't enough neutrons. That analysis does not include the effects of fast fission of U-238. This is not surprising because the percentage of fast fissions in existing thermal reactor designs is low, around 3% (in Candu reactors) or less. Table I shows the neutrons involved in 1,000 fissions in a thermal reactor for a conversion ratio of 1.000 when the fuel is mostly U-238 and the fissile content is low enough so that most fast fissions are of U-238. The top section of the table is for U-235 in U-238, the bottom section is for Pu-239 in U-238. The bottom line in each section shows the number of neutrons available for all other losses (escape from the reactor, absorption in non fuel components, and absorption in xenon, samarium, and other fission products). If the percentage of fast fissions of U-238 is 10 or 15 % a thermal breeder may be feasible. This is largely independent of the percentage of fissile content of the fuel. If the reactor starting fissile content is mostly U-235, as the U-235 is replaced with Pu-239 the number of available neutrons increases even if the conversion ratio is only 1.000. The percentage of fast fissions in a reactor is a question of reactor geometry, as the minimum dimension of the fuel elements increases the average distance a fast neutron travels in the fuel before exiting increases and with it the probability of fast fission of U-238.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗