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Results for “PULSED REACTORS”

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 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 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↗
DOE OSTI2017

Godiva II—An Unmoderated Pulse-Irradiation Reactor

Design features of Godiva II, the improved pulsed reactor successor to Lady Godiva, are discussed together with characteristics of power excursions, and performance is compared with that of the original Godiva. Measurements of the wait time between stepwise reactivity insertion and the occurrence of a burst are presented and compared with theory based on a statistical model of fission chains. Analytical and numerical solutions of the reactor equations are developed to reproduce experimental data and extrapolate to higher energy release. Here, consideration is also given to perturbations arising from room-returned neutrons. Two different modes of operation are discussed and some design problems of Godiva-type pulsed reactors are briefly mentioned. Typical bursts are illustrated with peak powers up to 13,000 Mw and widths at half-maximum down to 35 μsec.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Steady state versus pulsed tokamak reactors

Abstract We have carried out a detailed analysis that compares steady state versus pulsed tokamak reactors. The motivations are as follows. Steady state current drive has turned out to be more difficult than expected—it takes too many watts to drive an ampere, which has a negative effect on power balance and economics. This is partially compensated by the recent development of high temperature REBCO superconductors, which offers the promise of more compact, lower cost tokamak reactors, both steady state and pulsed. Of renewed interest is the reduction in size of pulsed reactors because of the possibility of higher field Ohmic transformers for a given required pulse length. Our main conclusion is that pulsed reactors may indeed be competitive with steady state reactors and this issue should be re-examined with more detailed engineering level studies.

Physics↗
DOE OSTI2018

Neptunium-Based High-Flux Pulsed Research Reactor

The design principles and parameters of a promising neutron source, based on the fission reaction of the isotope {sup 237}Np, for use in research in extracted beams are presented. Such a source can operate as a pulsing reactor or super-booster or as a super-booster multiplying the neutrons from the target of a proton accelerator by using a reactivity modulator. Owing to the threshold character of {sup 237}Np fission such a neutron source will possess significantly higher characteristics than a similar setup with plutonium, and in terms of utilization efficiency it will be the leading source in the world for research employing neutron spectroscopy.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2016

Measurement of prompt neutron generation time at the VIR-2M pulsed nuclear reactor

The prompt neutron generation time is measured in the core of the VIR-2M research nuclear reactor. The measurements are performed using the Babala method while the reactor is in the subcritical state. The VIR-2M reactor and the relevant experimental equipment are briefly described, and the experimental procedure and data processing technique are presented. It is shown that the prompt neutron generation time with empty experimental channels is 35 ± 1 μs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Monte Carlo Direct Simulation Method for Initiation Probability Problems

During the early period of a supercritical system, the growth of fission chains shows a significant stochastic feature. In multiplying medium, it is a probability issue of fail-pass type for a single neutron to sponsor a divergent fission chain, and the probability of success is named as single neutron initiation probability. Meanwhile, in the presence of a weak neutron source, the time at which fission burst occurs also exhibits a stochastic nature, which can be observed on experiments of pulsed reactors such as GODIVA I and GODIVA II. The initiation probability and burst waiting time reflect the dynamic feature of a supercritical system, which are very important for nuclear criticality safety concerns. It is desirable that a supercritical system has an initiation probability very close to zero and releases limited energy in case of criticality accidents. The stochastic process of neutron population evolution in early age in a multiplying system is governed by Bell's equation in the form of an adjoint neutron transport equation with a non-linear fission term. In order to solve this stochastic neutron transport equation, the deterministic method discretizing the phase space was proposed long ago and has been implemented in several SN codes such as Partisn, Ardra and PANDA. However, the deterministic method is of limited usage, due to its poor ability to precisely describe real-world 3D complex geometries and continuous energy neutron spectrum. Recently, Monte Carlo methods have gradually gained attention for solving Bell's equation. Greenman from LLNL proposed an analog Monte Carlo forward-transport method for calculating initiation probability in 2007. He implemented this method by introducing the maximum age parameter, the time sub-interval and the neutron progenitor tag into Mercury in time-dependent fixed-source simulation mode. This method is only suitable for initiation probability calculations, however not for analyzing the randomness of burst waiting time. Later, GANG in 2011 developed the GSMP (Generalized Semi-Markov Process method) Monte Carlo method according to Bell's equation. While the GSMP method, which can deal with both initiation probability and burst waiting time, is flexible and powerful for zero-dimensional problems, it is not favorable when extended to 3D and continuous energy problems due to severe lack of computational efficiency. In 2015, a direct Monte Carlo simulation method for stochastic systems problems was implemented in MCATK, yet this method has not been applied specifically to the burst waiting time. In this work, a new and efficient 3D Monte Carlo direct simulation method for calculating both the initiation probability and the burst waiting time is developed. By taking random factors during evolvement of fission chains into account in dynamic Monte Carlo simulation, this new method has the potential of simulating the whole process from source neutron injection to exponential growth of the neutron population, and finally to extinction of the neutron pulse. This capability is useful in criticality accidents evaluation and pulsed reactor experiments explanation. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

ARCS Replacement at TREAT for TRTR 2020 - Abstract

The Transient REActor Test (TREAT) facility was restarted in 2017 to perform reactor fuels testing under transient conditions, following shutdown in 1994 and 23 years in non-operational standby. TREAT is a transient reactor that, like most pulsing reactors, can create natural shaped power pulses. However, TREAT provides the ability to shape the power pulse and control the energy deposition rate on the test specimens by automatically controlling the amount of reactivity insertion over time. The original automatic reactor control system (ARCS) was created in the 1980s using INTEL 8086 CPU boards and software written in assembly code. The need to maintain the system for the next 40+ years of TREAT life led to the decision to replace ARCS with modern hardware while maintaining the original functionality. After a prototyping effort, the replacement was successfully completed in the first quarter of 2020. The process and results of that effort are reported here.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗