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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Kinetics calculation of fast periodic pulsed reactors using MCNP6

Fast periodic pulsed reactor is a type of reactor in which the fission bursts are formed entirely with external reactivity modulation with a specified time periodicity. This type of reactors could generate much larger intensity of neutron beams for experimental use, compared with the steady state reactors. In the design of fast periodic pulsed reactors, the time dependent simulation of the power pulse is majorly based on a point kinetic model, which is known to have limitations. A more accurate calculation method is desired for the design analyses of fast periodic pulsed reactors. Monte Carlo computer code MCNP6 is used for this task due to its three dimensional transport capability with a continuous energy library. Some new routines were added to simulate the rotation of the movable reflector parts in the time dependent calculation. Fast periodic pulsed reactor IBR-2M was utilized to validate the new routines. This reactor is periodically in prompt supercritical state, which lasts for similar to 400 mu s, during the equilibrium state. This generates long neutron fission chains, which requires tremendously large amount of computation time during Monte Carlo simulations. Russian Roulette was applied for these very long neutron chains in MCNP6 calculation, combined with other approaches to improve the efficiency of the simulations. In the power pulse of the IBR-2M at equilibrium state, there is some discrepancy between the experimental measurements and the calculated results using the point kinetics model. MCNP6 results matches better the experimental measurements, which shows the merit of using MCNP6 calculation relative to the point kinetics model.

Zhong, Z.↗
DOE OSTI2017

Parameters of a Pulsed Reactor for Pumping a Large-Size NPL

The design of a fast pulsed reactor with a large axial cavity that can contain a multichannel laser module to generate high-power pulses of optical radiation is proposed and substantiated by calculations. The results of calculations of the reactor parameters, as well as those of displacements and tensions in its fuel rods, are provided. Preliminary conclusions concerning the procedure of generation of fission pulses in the proposed reactor are formulated.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2011

Reactor pulse repeatability studies at the annular core research reactor

The Annular Core Research Reactor (ACRR) at Sandia National Laboratories is a water-moderated pool-type reactor designed for testing many types of objects in the pulse and steady-state mode of operations. Personnel at Sandia began working to improve the repeatability of pulse operations for experimenters in the facility. The ACRR has a unique UO{sub 2}-BeO fuel that makes the task of producing repeatable pulses difficult with the current operating procedure. The ACRR produces a significant quantity of photoneutrons through the {sup 9}Be({gamma}, n){sup 8}Be reaction in the fuel elements. The photoneutrons are the result of the gammas produced during fission and in fission product decay, so their production is very much dependent on the reactor power history and changes throughout the day/week of experiments in the facility. Because the photoneutrons interfere with the delayed-critical measurements required for accurate pulse reactivity prediction, a new operating procedure was created. The photoneutron effects at delayed critical are minimized when using the modified procedure. In addition, the pulse element removal time is standardized for all pulse operations with the modified procedure, and this produces less variation in reactivity removal times. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2011

Development of an active detector for the characterization of the late-time radiation environment from a reactor pulse

Document available in abstract form only, full text of document follows: This paper discusses the use of a commercially available {sup 235}U fission chamber, with a matching compensating ion chamber, originally sold as a single-ended detector with the signal conducted over the shield of a coaxial cable. The authors designed an aluminum housing that isolates the two detectors and converts the signals to full differential mode as a noise-reduction technique. The signals are processed using the switched resistor technique to extend the signal range to longer times from the peak of the pulse [Luker, S. M., Griffin, P. J., King, D. B., and Suo-Anttila, A. J., 'Improved Diagnostics for Analysis of a Reactor Pulse Radiation Environment,' 13. International Symposium on Reactor Dosimetry, Akersloot, Netherlands, May 25, 2008, pp. 4-6.]. The newly configured fission chamber assembly has been used at the annular core research reactor at Sandia National Laboratories to provide a high-fidelity characterization of the neutron time profile from a pulsed operation. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

Integral cross section measurement of the 235 U(n,n') 235m U reaction in a pulsed reactor

The integral measurement of the neutron inelastic cross section leading to the 26-minute half-life 235m U isomer in a fission-like neutron spectrum is presented. The experiment has been performed at a pulsed reactor, where the internal conversion decay of the isomer was measured using a dedicated electron detector after activation. The sample preparation, efficiency measurement, irradiation, radiochemistry purification, and isomer decay measurement will be presented. We determined the integral cross section for the 235 U(n,n') 235m U reaction to be 1.00 ± 0.13b. Furthermore, this result supports an evaluation performed with talys-1.4 code with respect to the isomer excitation as well as the total neutron inelastic scattering cross section.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2018

Annular Core Research Reactor Pulse History Report: 2003-2017

This report documents and describes the tabulation and analysis of historical pulse operation data from the Annular Core Research Reactor (ACRR) at Sandia National Laboratories (SNL). The pulse data was obtained from a combination of pulse log files generated at the control console and pulse diagnostics system data. The pulses presented were performed within the time period of April 2003 and December 2017. A brief analysis of the data is included to characterize the aggregate behavior of ACRR pulses with respect to theoretical treatments based on the point reactor kinetics model. It is expected that the data presented will provide an organized and consolidated resource to reference historical pulse data at the ACRR for use in analyses, verification and validation, and general understanding of the machine. A comprehensive set of data is presented to the reader in the appendices.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2011

New Measurements and Calculations to Characterize the Caliban Pulsed Reactor Cavity Neutron Spectrum by the Foil Activation Method

Caliban is a cylindrical metallic core reactor mainly composed of uranium 235. It is operated by the Criticality and Neutron Science Research Laboratory located at the French Atomic Energy Commission research center in Valduc. As with other fast burst reactors, Caliban is used extensively for determining the responses of electronic parts or other objects and materials to neutron-induced displacements. Therefore, Caliban's irradiation characteristics, and especially its central cavity neutron spectrum, have to be very accurately evaluated. The foil activation method has been used in the past by the Criticality and Neutron Science Research Laboratory to evaluate the neutron spectrum of the different facilities it operated, and in particular to characterize the Caliban cavity spectrum. In order to strengthen and to improve our knowledge of the Caliban cavity neutron spectrum and to reduce the uncertainties associated with the available evaluations, new measurements have been performed on the reactor and interpreted by the foil activation method. A sensor set has been selected to sample adequately the studied spectrum. Experimental measured reaction rates have been compared to the results from UMG spectrum unfolding software and to values obtained with the activation code Fispact. Experimental and simulation results are overall in good agreement, although gaps exist for some sensors. UMG software has also been used to rebuild the Caliban cavity neutron spectrum from activation measurements. For this purpose, a default spectrum is needed, and one has been calculated with the Monte-Carlo transport code Tripoli 4 using the benchmarked Caliban description. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Preliminary Analysis of Damage to MPFDs Caused by Reactor Pulses

The need for next-generation, accident-tolerant fuels has generated significant interest in the restart of domestic fuels testing at the Transient Reactor Test (TREAT) facility. The TREAT reactor is designed to produce large, transient fluxes that replicate the effects of a wide variety of accident conditions on an irradiated fuel sample. Peak powers generated during short pulses approach 20 GW with a neutron flux of approximately 10{sup 17} n cm{sup -2} s{sup -1}. As part of the restart effort, new vehicle designs are being explored for sample irradiation, including the Static Environment Rodlet Transient Test Apparatus (SERTTA) and variants. As part of the SERTTA designs, micro-pocket fission detectors (MPFDs) are being explored for localized flux measurements before, during, and after transient operations. These detectors are also being considered as part of future, core-wide instrumentation for the TREAT reactor. However, the high fluxes associated with the pulses in the TREAT core present certain thermodynamic concerns given their intensity and the amount of energy they will release within the MPFDs within such a short amount of time. Assuming each fission that occurs in a MPFD results in a fission fragment depositing 85 MeV of energy in the substrate on average and each MPFD has 0.5 μg of natural uranium deposited on it, a 100 ms long pulse in the TREAT reactor at 1017 n cm{sup -2} s{sup -1} flux will produce approximately 4.6 x 10{sup 6} fissions in an MPFD. This pulse deposits 6.3 x 10{sup -5} J in the substrate over a 100 ms duration. That amount of energy is sufficient to melt roughly 33% of the Pt under the UO{sub 2} deposition or a circular spot with a diameter of approximately 860 μm. Of course, the above estimate makes a number of simplifying assumptions and ignores heat transfer, but does serve to illustrate the point that this rapid heating effect could be detrimental to the integrity and life of these in-core MPFDs. To study how these layers are affected by large transient neutron fluxes, five samples were prepared, pre-analyzed, and then subjected to a pulse in the central thimble of the Kansas State University TRIGA Mark II reactor to a peak power of 1.58 GW. After being allowed to decay to safer radiation levels, the sample substrates and coatings were inspected using scanning electron microscopy (SEM) to identify any changes to their surface morphology. Indeed, certain changes were found and are discussed in greater detail in this paper.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Annular Core Research Reactor (ACRR) Pulse Curve Characterization

Reactor pulse characterization at the Annular Core Research Reactor (ACRR) at Sandia Technical Area V (TA-V) is commonly done through photo conductive detection (PCD) and calorimeter detectors. Each of these offer a mode of analyzing a digital signal with different advantages/methods for determination of integrated dose or temporal based metrology. This report outlines a method and code that takes the millions of data points from such detectors and delivers a characteristic pulse trendline through two main methods: digital signal filtration and machine learning, in particular, Support Vector Machines (SVMs). Each method's endpoint is to deliver a characteristic curve for the many bucket environments of ACRR while considering other points of interest including delayed gamma removal for prompt dose metrics. This work draws and adds on previous work detailing the delayed gamma fraction contributions from CINDER simulations of the ACRR. Results from this project show a method to determine characteristic curves in a way that has previously been limited by data set size.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗