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At least 199 records · Page 11
All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

Testing the Delayed Gamma Capability in MCNP6 - Paper 86

The mission of the Domestic Nuclear Detection Office (DNDO) is to quickly and reliably detect unauthorized attempts to import or transport special nuclear material (SNM) for use against the United States. Developing detection equipment to meet this objective requires accurate simulation of both the detectable signature and detection mechanism. MCNP6 is capable of transporting the various types of radiation encountered in these scenarios; however, the accuracy of MCNP6 must be quantified such that we can be confident in the results of the simulations. Delayed gamma transport is one important component of these detection scenarios. Delayed gamma transport is defined as the combination of gamma emission from radioactive and subsequent transport through matter. MCNP always possessed the ability to emit delayed neutrons from fission; however, these delayed neutron spectra and intensity were functional fits and not directly sampled from the created fission products. Furthermore, there existed no method to emit the decay photons from radioactive fission products (or any radioactive nucleus) A delayed particle capability was initially added to MCNPX 2.6.A in 2005 to sample the radioactive fission product parents and emit decay particles resulting from the decay chain. At the point of capture, a residual nucleus is created and decayed using the CINDER90 depletion code. MCNP only sends CINDER90 a residual radioactive nucleus and not a flux (i.e. capture rate). As a result, the Bateman equations for the solution are coupled first order differential equations with constant coefficients. Theoretically, these equations could be solved analytically; however, CINDER90 employs a numerical Markov linear chain scheme to solve these equations. This paper addresses a more simplified approach to benchmarking in order to best assess: (1) whether or not deficiencies exist in the current methodology for certain types of decay chains; and (2) paths forward for improving the delayed particle emission capability. We utilize closed form analytic solutions to the granddaughter equations for particular sets of isotope systems to compare the accuracy of the delayed gamma capability in MCNP6. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2014

ADVANTG Results for a Proposed NDA System for Detecting Pin Diversion from SNF - Paper 120

This summary describes the results of applying the ADVANTG variance reduction parameter generator to accelerate continuous-energy Monte Carlo simulations of the {sup 252}Cf Interrogation with Prompt Neutron (CIPN) detection system proposed by Hu et al. for detecting the diversion of fuel elements from spent nuclear fuel (SNF) assemblies. In previous work, ADVANTG was applied to a suite of nine radiation detection, special nuclear material (SNM) movement detection, and safeguards test cases. That effort demonstrated that ADVANTG was able to provide large increases in tally figures of merit (FOMs) in nearly all of the cases without introducing bias into the Monte Carlo results. The objective of the work described here is to further evaluate ADVANTG, and also to identify gaps in capability with regard to nuclear safeguards applications, by applying the code to a challenging problem that is of a different nature from the ones previously studied. The CIPN nondestructive assay (NDA) system consists of three fission chambers embedded in a C-shaped block of polyethylene and a polyethylene-moderated {sup 252}Cf source that are placed on opposite sides of a subject SNF assembly. The idea of the detection system is to measure the difference between the background neutron count rate, due to spontaneous fission in the fuel material, and the count rate observed once the source is introduced, which is dominated by multiplication within the assembly. Based on the difference between these count rates, the fissile content of the assembly can be inferred, and this information can be used to determine if some fraction of the original fuel pins have been replaced with depleted uranium substitutes. The authors report that the system can detect the replacement of at least eight pins (3% of total mass). The ADVANTG code generates space- and energy-dependent weight-window bounds and biased source distributions based on approximate transport solutions that are efficiently generated by the Denovo 3-D discrete ordinates package. ADVANTG implements the Consistent Adjoint Driven Importance Sampling (CADIS) method for accelerating individual tallies and the Forward-Weighted CADIS (FW-CADIS) method for obtaining relatively uniform uncertainties across tallies over multiple regions and/or energy bins. The variance reduction parameters are generated in a format directly usable by standard versions of MCNP5 and MCNPX. In this work, the ADVANTG variance reduction parameter generator was applied to accelerate continuous energy MCNPX simulations of the {sup 252}Cf Interrogation with Prompt Neutron detection system for detecting the diversion of fuel elements from SNF assemblies. Variance reduction parameters generated with the existing version of ADVANTG actually slowed down tally convergence in all cases. The cause of the poor performance was found to be the lack of an induced fission source treatment in the fixed-source discrete ordinates calculations employed by ADVANTG. Because fission was treated as capture and not as a source, the importance maps generated by ADVANTG significantly underestimated the importance of thermal neutrons in and around the fuel elements. In order to obtain improved results, we developed an experimental implementation of a fission source outer iteration for the adjoint calculations. We used this prototype capability to generate new variance reduction parameters and found that the mean fission rates agreed well with those predicted by the conventional MCNPX simulations, as was expected. With these new parameters, speedup factors of 59 to 80 were obtained for the case with the background source, and factors of 17 to 24 were obtained for the cases with the interrogating source. These results demonstrate that a fission source treatment is needed to improve the effectiveness of ADVANTG in problems where the effects of induced fission are important. This includes many types of NDA problems that are relevant to safeguards applications. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

Design of Nuclear Criticality Safety Framework for Hands-on Construction of Fast Systems Over Ranges of Multiplication

Since 1945, Los Alamos National Laboratory (LANL) has performed critical experiments, primarily at the Los Alamos Critical Experiments Facility (LACEF) at Technical Area 18 (TA-18). Since 2011, these experiments have been conducted by the National Criticality Experiments Research Center (NCERC), operated by LANL, at the Device Assembly Facility (DAF) in the Nevada National Security Site (NNSS). These experiments utilize various types of Special Nuclear Material (SNM). Some of these experiments utilize what is referred to as the Rocky Flat Shells, which are called such as they came from the Rocky Flats Plant. These concentric hemi-shells are made of Highly Enriched Uranium (HEU), which is 93 w/o 235U. Fig. 1. Subset of the Rocky Flat Shells LANL is designing a new subcritical hands-on experiment with the goal of achieving a neutron multiplication in the range of 50 to 200, which correlates to Keff values of 0.98 to 0.995. ANSI/ANS-1 is the standard for Conduct of Critical Experiments which governs critical operations at NCERC. Section 3.9 of ANSI/ANS-1 states that when manipulating a critical assembly by hand, the predicted k eff of a known configuration should not exceed 0.95 (a neutron multiplication of 20). This presents a challenge, as this system would have a higher multiplication than 20 and the assembly would have to remain subcritical in normal and credible accident scenarios. To ensure the safety of such an assembly, the different normal and abnormal conditions that could alter the criticality 1 MCNP® and Monte Carlo N-Particle® are registered trademarks owned by Triad National Security, LLC, manager and operator of Los Alamos National Laboratory. Any third party use of such registered marks should be properly of the system must be considered and analyzed. If such a condition is deemed to be credible, it will be further investigated using the Monte Carlo N-Particle (MCNP) transport code. If the results of these simulations show that a k eff larger than one could be possible, modifications to the assembly will be made until such an event is deemed no longer possible.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Field Testing of Safeguards Technologies in the Hot Fuel Examination Facility

Recent developments in nuclear fuel reprocessing techniques have yielded more efficient processes and fuel cycle options that strengthen the nuclear industry and production of clean energy. One such area of interest is pyroprocessing of used oxide fuel. However, with these advances in the back end of the nuclear fuel cycle, advances in safeguards instrumentation, measurements, and approaches are needed to ensure special nuclear material (SNM) is accounted for according to regulatory requirements. As a high-level overview of a nominal pyroprocessing approach, used oxide fuel from commercial light water reactors (LWR) is mechanically removed from the metallic cladding. Then the fuel is crushed and randomized representative samples are taken and sent to an analytical lab for analysis. The analytical results of the feed material are used for input accountancy into the rest of the process. The crushed oxide fuel is then moved to the oxide reduction (OR) furnace where it is reduced from an oxide to metallic form. The OR product is distilled to remove salt and then moved to an electrorefiner (ER), where it is immersed in a eutectic mixture of lithium chloride potassium chloride (LiCl-KCl) that typically ranges between 450-550 ?. Within the ER, the usable uranium is electrochemically transported through the molten salt from the anode to the cathode, and then subsequently removed as a relatively pure U product.. A simplified model of pyroprocessing techniques with added emphasis on the safeguards can be seen below in Fig. 1

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

FRAM v5.2 estimation of plutonium and uranium isotopics using digitized 3-D position-sensitive CdZnTe detectors

Gamma-ray spectroscopy is an important tool used in international safeguards to measure plutonium and uranium isotopics. Recently developed, digital, 3-D position-sensitive CdZnTe gamma-ray spectrometers are promising for safeguards, offering improved performance over commonly available CdZnTe systems, with energy resolutions approaching that of high-purity germanium. CdZnTe gamma-ray spectra from plutonium and uranium samples were processed using the commercially available software FRAM v5.2, a tool that estimates plutonium and uranium isotopics from high-purity germanium spectra, across a wide range of materials. Furthermore using digital CdZnTe systems 239 Pu/ 240 Pu isotopic composition and effective 240 Pu percentage are evaluated to within less than 3 sigma statistical measurement uncertainty across reactor to weapons-grade samples, while low and highly-enriched uranium are easily distinguished.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2021

Impact of Spontaneous Fission Neutron Emission Energy on Neutron Detector Response

A parametric study was simulated in MCNP to investigate the impact of spontaneous fission neutron energy on neutron detector response. A total of 880 configurations were simulated by varying 3 He pressure and polyethylene reflector thickness for 8 different isotopes. Results showed that deviations between the simulated isotopes become more prevalent at higher energies with 238 U showing the greatest deviation. It was also observed that configurations comparable to most detection setups, 10 atm and 6 - 8 cm polyethylene, were closer to the upper bound of deviations from a 252 Cf. 240 Pu, a common spontaneous fission source of interest, ranged from 2% - 3% within this space. This suggests that further investigation is required to determine exact correction factors when calibrating neutron detectors with a 252 Cf. Although demonstrated here for a 4π-detection system, future plans include simulating a more detailed model of the MC-15 detection setup to determine exact correction factors for future experiments.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2019

Adding ($\mathcal{α}$,n) Emission to an Existing Subcritical Benchmark Statistical Uncertainty Method

Subcritical measurements often use radiation detectors to infer parameters related to the physics of a measured system. One type of subcritical measurements typically performed are called neutron noise measurements. These types of measurements utilize the fact that prompt neutrons are produced from fission immediately after a fission event and are therefore correlated in time. Subcritical benchmarks require evaluation of the statistical and systematic uncertainties of a measurement. Previous work established methods for determining statistical and systematic uncertainties in subcritical benchmarks. The data were analyzed using the Hage-Cifarelli formalism of the Feynman Variance-to-Mean method. The previous work made an assumption that the ($\mathcal{α}$,n) emission rate was negligible. This assumption was used because the previous work focused on systems with metal Plutonium. This work extends the existing statistical uncertainty method to include ($\mathcal{α}$,n) emissions.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

FRAM, Version 6.1 User Manual

This manual describes the use of version 6.1 of the FRAM gamma-ray isotopic analysis software developed in the Safeguards Science and Technology Group, NEN-1, Nuclear Engineering & Nonproliferation Division, Los Alamos National Laboratory. The software analyzes the gamma ray spectrum from plutonium-bearing items and determines the isotopic distribution of the plutonium, 241Am content, and concentration of other isotopes in the item. The software can also determine the isotopic distribution of uranium isotopes in items containing only uranium. The body of this manual describes the generic version of the code.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2020

Characteristics of Uranium-Plutonium Mixed Oxide Nondestructive Assay Standards

A set of six well-characterized, homogeneous, mixed HEU-WGPu oxide standards has been manufactured for and are proposed for qualification as NDA working reference standards with SAFE-NMCA. These standards, if approved, will provide the basis for qualifying the FRAM isotopic systems for use in accountability measurements of 235U in the presence of Pu for unknown mixed U-Pu items by nondestructive assay. The U to Pu elemental mass ratios of the six analytical standards are nominally in the proportions 0.1, 0.25, 1.0, 2.5, 10, and 20; these values were chosen to span the U/Pu mass ratios of inventory items that ultimately will be measured by nondestructive assay and which currently reside in long term storage. The measured characteristics, including plutonium and uranium elemental masses and isotopic distribution, of each standard and relevant source documentation are reported here.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

FRAM In-Depth [Slides]

FRAM is an isotopic analysis code nominally designed for plutonium and uranium. It can also analyze other materials such as spent fuel (off all ages), medical isotopes.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2024

Independent Review of Non-Destructive Assay for the K-25/K-27 D&D Project: November 30-December 3, 2004, Oak Ridge, TN

The DOE with concurrence from the D&D contractor has chartered an assessment of the methods and procedures used to obtain the original NDA results of holdup deposits at the K-25 and K-27 uranium gaseous diffusion facilities in Oak Ridge. The assessment has been performed by an external review team with the participation of the D&D contractor, DOE, and DOE/OR. This document is the final report for the assessment. The DOE provided the charter for this review. The review team spent a week in Oak Ridge attending meetings and participating in discussions with expert contractor staff. Substantial additional input for the review came from numerous written materials, unpublished and published. This final report of observations, findings, and recommendations in areas defined by the charter is based on information from the meetings, discussions, and documents. The assessment of the Review Team is that the DOE/OR and BJC approach in using historical NDA data for the D&D of K-25 and K-27 is appropriate and generally acceptable. Resolving issues, addressing findings, and implementing the recommendations documented in this report will reconcile specific technical concerns.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

FRAM Version 7.1’s Bias

The Fixed-Energy Response-Function Analysis with Multiple Efficiency (FRAM) code was developed at Los Alamos National Laboratory to measure the gamma-ray spectrometry of the isotopic composition of plutonium, uranium, and other actinides. For FRAM versions 4 and earlier, the reported uncertainties of the results come from the propagation of the statistics in the peak areas only. No systematic error components are included in the reported uncertainties. For FRAM versions 5 and 6, we examined the FRAM analytical results of both the archival plutonium data and the data specifically acquired for the isotopic uncertainty analysis project and found the relationship between the bias and other parameters. We worked out the equations representing the biases of the measured isotopes from each measurement using internal spectral parameters, such as peak resolution and shape, region of analysis, and burnup (for plutonium) or enrichment (for uranium). The resulting biases were included in the reported uncertainties of FRAM v.5 and v.6.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2019

Characterization of Mechanical and Thermal Factors Related to Glovebox Container Water Ingress (Final Report)

This report summarizes tests conducted to quantify water ingress into nuclear material containers when damaged. The damage inflicted is meant to replicate realistic worst-case seismic/fire scenario at the Los Alamos National Laboratory (LANL). Four types (SAVY, Hagan, K-flange, and ROCK) of three-quart containers were subject to drop and then furnace testing. At each stage (as-delivered, post-drop, and post-furnace), water ingress potential was tested by subjecting the containers to a static water column. Key results of these experiments include: • LANL provided two water ingress testing criteria. The first, and most stringent, was a 24 inAq column for a period of one week. These parameters were used for baseline testing on as-delivered containers. The second was a 6 inAq column for a period of two hours. These parameters were used to test containers after impact and furnace testing. • Under the more severe criteria for as-delivered containers, the measured water ingress was: SAVY 7–14 g; Hagan 168–493 g; K-flange 4–5 g; and ROCK 24–280 g of water ingress. • Under the less severe criteria and after drop testing, all containers permitted less than 50 g of water ingress with one exception a ROCK container that had one of three latches release during impact. • Finally, after furnace testing (500°C, 45 minutes), which destroyed the o-ring/gasket materials, and under the same 6 inAq, 2 hr column test, the measured water ingress exceeded 50 g for all but one K-flange and one ROCK container.

42 ENGINEERING↗
DOE OSTI2015

Muon Tracking to Detect Special Nuclear Materials

Previous experiments have proven that high-Z materials can be located in shielded configurations and that nuclear assemblies can be imaged and identified inside of shipping containers using vertical trajectory cosmic-ray muons with two-sided imaging. These experiments have further demonstrated that nuclear assemblies can be identified by detecting fission products in coincidence with tracked muons. By developing these technologies, advanced sensors can be designed for a variety of warhead monitoring and detection applications. The focus of this project is to develop tomographic-mode imaging using near-horizontal trajectory muons for weapons and spent fuel applications. This will allow passive imaging in-situ without the need to relocate the objects.

cosmic ray muons, muon tracking, SNM, spent fuel↗