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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

Delayed Neutron Precursor Group Parameter and Spectra Generation from Fast Fission of 235U in SCALE

Delayed neutron precursor (DNP) group data are important for modeling reactor dynamics. Although the data for individual DNPs have been developed over time, the DNP group data present in the Evaluated Nuclear Data Files (ENDF) have not been updated in the past 20 years. In this work, we use SCALE to recreate the Godiva experiment that was used to generate the original DNP group structure for fast fission of 235U. However, each DNP is modeled using up-to-date data, and the results are then converted into a newly updated group structure. This conversion uses an iterative linear least squares solver to minimize chi-squared. The approaches used in this work also enable energy spectrum generation and uncertainty tracking. The method used in this paper for fast 235U fission DNP group structure updating can be applied to different energies and fissile nuclides. Demonstration of the uncertainty tracking in reactor kinetics and dynamics simulations is shown using point reactor kinetics simulations. Results show that there are data discrepancies between the International Atomic Energy Agency database and data used in ORIGEN, which are currently being fixed. Results also show that the proposed method for group spectra generation performs well.

Seifert, Luke [University of Illinois]↗
DOE OSTI2018

OPTIMIZATION OF FAST FISSION GAS RELEASE MODEL PARAMETERS USING MACHINE LEARNING ACCELERATED EVOLUTIONARY ALGORITHMS

Fission gas release models used in FRAPCON and FAST are based on equations that describe the physical transport phenomena that control release from the fuel pellet to the rod void volume of fission gas produced during irradiation. Although these models are physically based, many of the parameters that control the release such as the impact of burnup on diffusion and bubble formation and saturation are critical to the accurate prediction of fission gas release, but cannot be measured in any quantitative way. To overcome this problem, these parameters have been empirically derived to provide a best fit to the available data that includes rod puncture data and electron probe microanalysis (EMPA) and X-ray fluorescence (XRF) data of radial location of residual gas within the pellet. In general, a heuristic approach for deriving these parameters via practical implementation of expert analyses on just a few dependent variables has been shown to perform quite well in most situations. In this paper, we present an approach for applying machine learning to expedite model development. This approach is based on developing a deep artificial neural network which describes FRAPCON’s fission gas release models, and optimizes parameters by a differential evaluation algorithm. This approach allows us to quickly and accurately tune physical models based on expert judgment, and works as a human-in-the-loop approach, to assist the modeler in identifying and addressing regions of high uncertainty in a multi-parameter space. Results of the updated fission gas release model will be shown for all the FAST assessment data.

Johns, Jesse M.↗
DOE OSTI2019

Optimal arrangements of 1,3-diphenylisobenzofuran molecule pairs for fast singlet fission

A simplified version of the frontier orbital model has been applied to pairs of C 2 , C 2v , C s , and C 1 symmetry 1,3-diphenylisobenzofuran rotamers to determine their best packing for fast singlet fission (SF). For each rotamer the square of the electronic matrix element for SF was calculated at 2.2 × 10 9 pair geometries and a few thousand most significant physically accessible local maxima were identified in the six-dimensional space of mutual arrangements. At these pair geometries, SF energy balance was evaluated, relative SF rate constants were approximated using Marcus theory, and the SF rate constant k SF was maximized by further optimization of the geometry of the molecular pair. The process resulted in 142, 67, 214, and 291 unique geometries for the C 2 , C 2v , C s , and C 1 symmetry molecular pairs, respectively, predicted to be superior to the C 2 symmetrized known crystal pair structure. Furthermore, these optimized pair geometries and their triplet biexciton binding energies are reported as targets for crystal engineering and/or covalent dimer synthesis, and as possible starting points for high-level pair geometry optimizations.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
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↗
DOE OSTI2016

Detection of special nuclear material from delayed neutron emission induced by a dual-particle monoenergetic source

Detection of unique signatures of special nuclear materials is critical for their interdiction in a variety of nuclear security and nonproliferation scenarios. We report on the observation of delayed neutrons from fission of uranium induced in dual-particle active interrogation based on the {sup 11}B(d,n γ){sup 12}C nuclear reaction. Majority of the fissions are attributed to fast fission induced by the incident quasi-monoenergetic neutrons. A Li-doped glass–polymer composite scintillation neutron detector, which displays excellent neutron/γ discrimination at low energies, was used in the measurements, along with a recoil-based liquid scintillation detector. Time-dependent buildup and decay of delayed neutron emission from {sup 238}U were measured between the interrogating beam pulses and after the interrogating beam was turned off, respectively. Characteristic buildup and decay time profiles were compared to the common parametrization into six delayed neutron groups, finding a good agreement between the measurement and nuclear data. This method is promising for detecting fissile and fissionable materials in cargo scanning applications and can be readily integrated with transmission radiography using low-energy nuclear reaction sources.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2010

FUEL CYCLE ISOTOPE EVOLUTION BY TRANSMUTATION DYNAMICS OVER MULTIPLE RECYCLES

Because all actinides have the ability to fission appreciably in a fast neutron spectrum, these types of reactor systems are usually not associated with the buildup of higher mass actinides: curium, berkelium and californium. These higher actinides have high specific decay heat power, gamma and neutron source strengths, and are usually considered as a complication to the fuel manufacturing and transportation of fresh recycled transuranic fuel. This buildup issue has been studied widely for thermal reactor fuels. However, recent studies have shown that the transmutation physics associated with "gateway isotopes" dictates Cm-Bk-Cf buildup, even in fast burner reactors. Assuming a symbiotic fuel relationship with light water reactors (LWR), Pu-242 and Am-243 are formed in the LWRs and then are externally fed to the fast reactor as part of its overall transuranic fuel supply. These isotopes are created much more readily in a thermal than in fast spectrum systems due to the differences in the fast fission (i.e., above the fission threshold for non-fissile actinides) contribution. In a strictly breeding fast reactor this dependency on LWR transuranics would not exist, and thus avoids the introduction of LWR derived gateway isotopes into the fast reactor system. However in a transuranic burning fast reactor, the external supply of these gateway isotopes behaves as an external driving force towards the creation and build-up of Cm-Bk-Cf in the fuel cycle. It was found that though the Cm-Bk-Cf concentration in the equilibrium fuel cycle is dictated by the fast neutron spectrum, the time required to reach that equilibrium concentration is dictated by recycle, transmutation and decay storage dynamics.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Differential Die-Away Instrument: Report on Initial Simulations of Spent Fuel Experiment

New Monte Carlo simulations of the differential die-away (DDA) instrument response to the assay of spent and fresh fuel helped to redefine the signal-to-Background ratio and the effects of source neutron tailoring on the system performance. Previously, burst neutrons from the neutron generator together with all neutrons from a fission chain started by a fast fission of 238 U were considered to contribute to active background counts. However, through additional simulations, the magnitude of the 238 U first fission contribution was found to not affect the DDA performance in reconstructing 239 Pu eff . As a result, the newly adopted DDA active background definition considers now any neutrons within a branch of the fission chain that does not include at least one fission event induced by a thermal neutron, before being detected, to be the active background. The active background, consisting thus of neutrons from a fission chain or its individual branches composed entirely of sequence of fast fissions on any fissile or fissionable nuclei, is not expected to change significantly with different fuel assemblies. Additionally, while source tailoring materials surrounding the neutron generator were found to influence and possibly improve the instrument performance, the effect was not substantial.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2010

Studies of heavy ion reactions around Coulomb barrier

The angular distributions of fission fragments for the {sup 32}S+{sup 184}W reaction near Coulomb barrier energies were measured. The experimental fission excitation function is obtained. The measured fission cross sections are decomposed into fusion-fission, quasifission and fast fission contributions by the dinuclear system model. The hindrance to complete fusion both at small and large collision energies is explained. The fusion excitation functions of {sup 32}S+{sup 90,96}Zr in an energy range from above to below Coulomb barrier were measured and analyzed. The obvious effect of positive Q-value neutron transfers has been observed in the {sup 32}S+{sup 96}Zr system. In addition, the excitation functions of quasielastic scattering at a backward angle have been measured for the systems of {sup 16}O+{sup 208}Pb, {sup 196}Pt, {sup 184}W, and {sup 154,152}Sm at energies well below the Coulomb barrier. Considering the coupling effects, the extracted diffuseness parameters are close to the values extracted from the systematic analysis of elastic and inelastic scattering data.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2012

ADRIANA project: Identification of research infrastructures for the SFR, within the frame of European industrial initiative for sustainable nuclear fission

Fast neutron reactors have a large potential as sustainable energy source. In particular, Sodium Fast Reactors (SFR) with a closed fuel cycle and potential for minor actinide burning may allow minimization of volume and heat load of high level waste and provide improved use of natural resources (as compared to only 1% energy recovery in the current once-through fuel cycle, with Thermal Reactors, such as EPR). The coordinating action ADRIANA (Advanced Reactor Initiative And Network Arrangement) has been initiated to set up a network dedicated to the construction and operation of research infrastructures in support of developments for the European Industrial Initiative for sustainable nuclear fission. The Project sets these objectives for the following reactor systems and related technologies: Sodium Fast Reactor (SFR), Lead Fast Reactor (LFR), Gas Fast Reactor (GFR, including very high temperature technologies), Instrumentation, diagnostics and experimental devices, Irradiation facilities and hot laboratories, Zero power reactors. Among the fast reactor systems, the sodium cooled reactor has the most comprehensive technological basis as result of the experience gained from worldwide operation of several experimental, prototype and commercial size reactors, since the forties (see Appendix I). This concept is currently considered as the reference, within the European strategy. Innovations are needed to further enhance safety, reduce capital cost and improve efficiency reliability and operability, making the Generation IV SFR an attractive option for electricity production. Currently, in France, a moderate (500 to 600 MWe) power demonstrator named ASTRID (Advanced Sodium Test Reactor for Industrial Demonstration) has been proposed and endorsed by EU. Presently, the reference configuration is a pool concept. General R and D needs have been identified and experimental facilities required to satisfy these needs have been listed for the following domains: material and coolant, thermal-hydraulics and safety, severe accidents, systems and components validation, instrumentation and Education and Training. Then, the contributors have detailed the different types of experimental facilities, existing as well as lacking facilities. A ranking procedure of the facilities has been carried out, then a list of priorities and a road-map for investment has been established. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Advantages and applications of megawatt-sized heat-pipe reactors

Recently, worldwide interest in nuclear energy has focused on small reactors (10 to 300 MWe) to address emerging energy needs in remote locations. These designs are new to varying degrees but share similar approaches and common weaknesses with regard to primary heat rejection that differ little from reactor designs of the late 1950's. Here, an innovative concept, heat-pipe reactors, is discussed. The concept is unique in its simplicity and potential for safe, affordable, and reliable energy. Given the potential for reactors to meet worldwide energy needs and the pivotal role of heat rejection in overall reactor safety, the potential societal impact of this type of innovation is substantial. Heat-pipe-cooled, fast-spectrum reactors have been proposed for government applications requiring a robust, reliable, remotely controlled system with capacity much less than 1 MWe; however, they have not been designed for power ranges greater than 1 MWe. Los Alamos National Laboratory has initiated a study to design heat-pipe-cooled, fast-fission reactors and to generate a point design of a > 10-MWe-class machine suitable for next-generation compact reactors at remote locations. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

The 14 MeV Neutron Irradiation Facility in MARIA Reactor

The MARIA reactor with thermal neutron flux density up to 3x10{sup 14} cm{sup -2} s{sup -1} and a number of vertical channels is well suited to material testing by thermal neutron treatment. Beside of that some fast neutron irradiation facilities are operated in MARIA reactor as well. One of them is thermal to 14 MeV neutron converter launched in 2014. It is especially devoted to fusion devices material testing irradiation. The ITER and DEMO research thermonuclear facilities are to be run using the deuterium - tritium fusion reaction. Fast neutrons (of energy approximately 14 MeV) resulting from the reaction are essential to carry away the released thermonuclear energy and to breed tritium. However, constructional materials of which thermonuclear reactors are to be built must be specially selected to survive intense fluxes of fast neutrons. Strong sources of 14 MeV neutrons are needed if research on resistance of candidate materials to such fluxes is to be carried out effectively. Nuclear reactor-based converter capable to convert thermal neutrons into 14 MeV fast neutrons may be used to that purpose. The converter based on two stage nuclear reaction on lithium-6 and deuterium compounds leading to 14 MeV neutron production. The reaction chain is begun by thermal neutron capture by lithium-6 nucleus resulted in triton release. The neutron and triton transport calculations have been therefore carried-out to estimate the thermal to 14 MeV neutron conversion efficiency and optimize converter construction. The usable irradiation space of ca. 60 cm{sup 3} has been obtained. The released energy have been calculated. Heat transport has been asses to ensure proper device cooling. A set of thermocouples has been installed in converter to monitor its temperature distribution on-line. Influence of converter on reactor operation has been studied. Safety analyses of steady states and transients have been done. Performed calculations and analyses allow designing the converter and formulate its operation limits and conditions. During first tested operation of the converter the 14 MeV neutron flux density was estimated to 10{sup 9} cm{sup -2} s{sup -1}, whereas fast fission neutrons inside converter achieved 10{sup 12} cm{sup -2} s{sup -1}, and thermal neutrons were reduced down to 109 cm-2 s-1. Taking into account the feasibility of almost incessant converter operation for a number of months, its arisen as one of the most powerful (in terms of fluence), currently available 14 MeV neutron source. Such a converter currently under operation in the MARIA reactor core will be presented. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Optimal Arrangements of Tetracene Molecule Pairs for Fast Singlet Fission

In search for a qualitative understanding of the effects of molecular packing on singlet fission (SF) rate, a simplified version of the frontier orbital model is described and illustrated on a pair of tetracene molecules. To identify all favorable physically accessible pair geometries, all significant local maxima of the square of the electronic matrix element for SF have been located within the six-dimensional space of possible arrangements of two rigid bodies, using a grid of over 4.7 × 10 8 pair geometries. Those at which the molecules interpenetrate were excluded using a hard-sphere model. The effects of intermolecular interaction on the SF energy balance and thus its rate constant k SF were approximated using Marcus theory at each of the maxima using the same simplified version of the frontier orbital model. Starting at these local maxima, the pair geometries were optimized for maximum k SF and the 21 best are reported along with their computed Davydov splitting and triplet biexciton binding energies. The optimal pair structures at the resulting maxima follow qualitative rules published previously and further elaborated here.

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

On the combination of delayed neutron and delayed gamma techniques for fission rate measurement in nuclear fuel

Novel techniques to measure newly induced fissions in spent fuel after re-irradiation at low power have been developed and tested at the Proteus zero-power research reactor. The two techniques are based on the detection of high energy gamma-rays emitted by short-lived fission products and delayed neutrons. The two techniques relate the measured signals to the total fission rate, the isotopic composition of the fuel, and nuclear data. They can be combined to derive better estimates on each of these parameters. This has potential for improvement in many areas. Spent fuel characterisation and safeguard applications can benefit from these techniques for non-destructive assay of plutonium content. Another application of choice is the reduction of uncertainties on nuclear data. As a first application of the combination of the delayed neutron and gamma measurement techniques, this paper shows how to reduce the uncertainties on the relative abundances of the longest delayed neutron group for thermal fissions in {sup 235}U, {sup 239}Pu and fast fissions in {sup 238}U. The proposed experiments are easily achievable in zero-power research reactors using fresh UO{sub 2} and MOX fuel and do not require fast extraction systems. The relative uncertainties (1{sigma}) on the relative abundances are expected to be reduced from 13% to 4%, 16% to 5%, and 38% to 12% for {sup 235}U, {sup 238}U and {sup 239}Pu, respectively. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Short-lived fission product measurements from >0.1 MeV neutron-induced fission using boron carbide.

A boron carbide shield was designed, custom fabricated, and used to create a fast fission energy neutron spectrum. The fissionable isotopes 233, 235, 238U, 237Np, and 239Pu were separately placed inside of this shield and irradiated under pulsed conditions at the Washington State University 1 MW TRIGA reactor. A unique set of fission product gamma spectra were collected at short times (4 minutes to 1 week) post-fission. Gamma spectra were collected on single-crystal high purity germanium detectors and on Pacific Northwest National Laboratory's (PNNL's) Direct Simultaneous Measurement (DSM) system composed of HPGe detectors connected in coincidence. This work defines the experimental methods used to produce and collect the gamma data, and demonstrates the validity of the measurements. It is important to fully document this information so the data can be used with high confidence for the advancement of nuclear science and non-proliferation applications. The gamma spectra collected in these and other experiments will be made publicly available at https://spcollab.pnl.gov/sites/gammadata or via the link at http://rdnsgroup.pnl.gov. A revised version of this publication will be posted with the data to make the experimental details available to those using the data.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Effects of fission yield data in the calculation of antineutrino spectra for U 235 ( n , fission ) at thermal and fast neutron energies

Fission yields form an integral part of the prediction of antineutrino spectra generated by nuclear reactors, but little attention has been paid to the quality and reliability of the data used in current calculations. Following a critical review of the thermal and fast ENDF/B-VII.1 235 U fission yields, deficiencies are identified and improved yields are obtained, based on corrections of erroneous yields, consistency between decay and fission yield data, and updated isomeric ratios. These corrected yields are used to calculate antineutrino spectra using the summation method. An anomalous value for the thermal fission yield of 86 Ge generates an excess of antineutrinos at 5–7 MeV, a feature which is no longer present when the corrected yields are used. Thermal spectra calculated with two distinct fission yield libraries (corrected ENDF/B and JEFF) differ by up to 6% in the 0–7 MeV energy window, allowing for a basic estimate of the uncertainty involved in the fission yield component of summation calculations. Lastly, the fast neutron antineutrino spectrum is calculated, which at the moment can only be obtained with the summation method and may be relevant for short baseline reactor experiments using highly enriched uranium fuel.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

Final design for thermal epithermal experiments (TEX) with ZPPR plutonium/aluminium plates with polyethylene and tantalum - 13592

Lawrence Livermore National Laboratory has completed the final design for the first phase of the Thermal Epithermal experiments (TEX), which focused on critical configurations composed of plutonium/aluminum Zero Power Physics Reactor (ZPPR) plates moderated by polyethylene with and without a tantalum diluent. {sup 239}Pu and {sup 240}Pu were identified as being the number one and two nuclear data need by the nuclear criticality safety and nuclear data communities, with special emphasis placed on performance issues of the Pu cross sections in the intermediate energy region. Preliminary TEX design recommended tantalum as the first diluent to pursue for final design, as it represented a high priority data need, showed the highest sensitivity and largest contribution to uncertainty with the Pu ZPPR system, and a large number of high quality Ta plates from the ZPPR inventory in Idaho was readily available. The goals of the new critical experiments are 1) to create assemblies that span a wide range of fission energy spectra, from thermal (below 0.625 eV), through the intermediate (0.65 eV to 100 keV), and to fast energies (above 100 keV), and 2) to easily modify the assemblies to include high priority materials like tantalum. A total of ten critical configurations were designed as part of CED-2. Five TEX experiments (Experiments 1-5) were designed to establish baseline configurations with the Pu ZPPR plates covering the thermal, intermediate, and fast fission energy regimes. For all five experiments, the ZPPR plates will be arranged in layers of 24 plates (6 plates by 4 plates), resulting in approximately a 12 inch by 12 inch footprint. Multiple layers will be stacked together with varying thicknesses of interspersed polyethylene placed between the layers to fine tune the neutron spectrum of the assembly. The other five experiments (Experiments 6-10) will be similar to the five baseline experiments, except that a tantalum layer will be placed next to each plutonium plate layer, thus diluting the stack and allowing for tests of the neutron cross section for tantalum. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2011

Advanced Monte Carlo modeling of prompt fission neutrons for thermal and fast neutron-induced fission reactions on {sup 239}Pu

Prompt fission neutrons following the thermal and 0.5 MeV neutron-induced fission reaction of {sup 239}Pu are calculated using a Monte Carlo approach to the evaporation of the excited fission fragments. Exclusive data such as the multiplicity distribution P({nu}), the average multiplicity as a function of fragment mass {nu}-bar(A), and many others are inferred in addition to the most used average prompt fission neutron spectrum {chi}(E{sub in},E{sub out}), as well as average neutron multiplicity {nu}-bar. Experimental information on these more exclusive data help constrain the Monte Carlo model parameters. The calculated average total neutron multiplicity is {nu}-bar{sub c}=2.871 in very close agreement with the evaluated value {nu}-bar{sub e}=2.8725 present in the ENDF/B-VII.0 library. The neutron multiplicity distribution P({nu}) is in very good agreement with the evaluation by Holden and Zucker. The calculated average spectrum differs in shape from the ENDF/B-VII.0 spectrum, evaluated with the Madland-Nix model. In particular, we predict more neutrons in the low-energy tail of the spectrum (below about 300 keV) than the Madland-Nix calculations, casting some doubts on how much scission neutrons contribute to the shape of the low-energy tail of the spectrum. The spectrum high-energy tail is very sensitive to the total kinetic energy distribution of the fragments as well as to the total excitation energy sharing at scission. Present experimental uncertainties on measured spectra above 6 MeV are too large to distinguish between various theoretical hypotheses. Finally, comparisons of the Monte Carlo results with experimental data on {nu}-bar(A) indicate that more neutrons are emitted from the light fragments than the heavy ones, in agreement with previous works.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗