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All sources NASA NTRS DOE OSTI

DOE OSTI2016

The Effect of Localized Pin Flux on Short-Term Gamma Intensities of Spent Fuel Assemblies

Nuclear reactor operation for electricity generation involves irradiation of a fuel assembly for an extended period of time until a certain burnup is achieved. To harness the full nuclear energy potential and flatten the flux around the core, fuel assemblies are shuffled to one or more different locations within the core. Post-irradiation practice typically involves storing the spent fuel assembly in a spent fuel pool. While in the spent fuel pool, aspects of the fuel assembly continue to evolve, particularly the isotopic concentration as nuclides decay. Numerous nondestructive assay (NDA) safeguards techniques exist and have been proposed for performing verification on spent fuel assemblies to ensure they have not been tampered with. Adigun et al. recently proposed measuring both the neutron and gamma signatures from spent fuel from discharge to final disposal, however, the research in only investigated the neutron part of this approach. In the conclusion, the authors caveated the need for the gamma results to make more certain conclusions about a spent fuel assembly, as a combination of the neutron and gamma NDA measurements will limit the range of possible initial enrichment and burnup combinations. This paper focuses on the production of the gamma source. It is important to understand the driving mechanisms behind the emissions signatures at shutdown and after long cooling time periods, to better determine the correlation between the measured passive gamma emission spectra and the gamma emissions per second of the assembly. The objective is to demonstrate how the effect of localized flux within a reactor affects the gamma emissions per second (or gamma intensity) present in an assembly at shutdown. This study examined the photons emitted from the nuclides present within the fuel assembly, and not the photons escaping from the fuel assembly. The latter point of transporting the photons present within the spent fuel assemblies to an appropriate detector will be addressed in a future paper. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Delayed-Gamma Energy Biasing with Exact Energy Sampling in MCNP 6.2.0

Characterization of delayed-particle signatures is of the utmost importance for detecting, identifying, and quantifying special nuclear material (SNM). Various non-destructive assay (NDA) techniques used in nuclear material safeguards rely on characteristic gamma-ray or neutron emissions to infer properties of a material. The all-particle, all-energy Monte-Carlo radiation transport code MCNP6 is capable of simulating delayed-particle signatures from SNM in addition to modeling the transport of emitted particles to a detector or other instrument. In some cases, NDA techniques require the measurement of low probability emissions from a given material. In physical scenarios, this problem can be bypassed by longer count-times or may be simply irrelevant due to the mass, and therefore the total activity, of the material being measured. However, when applying the Monte-Carlo method to such scenarios, sampling of low probability emissions requires a large number of histories to converge to a statistically acceptable solution. To address this issue, the delayed-neutron and gamma energy biasing feature was introduced in MCNP6 by H. Armstrong et al. in 2013 which allows the user to set up energy windows in which delayed-particle emission sampling can be adjusted. For delayed-gamma sampling, this biasing method was based on the bin-wise structure of gamma emission data for both the multi-bin gamma and line-emission sampling modes. Recent improvements have been made to the line-emission sampling algorithm which upgrade the previous mini-bin sampling treatment to an exact energy-line treatment. This upgrade was not compatible with the biasing method implemented by H. Armstrong et al. due to the sampled data no longer having a bin-wise structure in line-emission mode. For this reason the delayed-gamma energy biasing feature has been rewritten to support the exact line-sampling treatment. This paper provides a description of the delayed-gamma energy biasing (DGEB) feature in MCNP6 as well as results and discussion for a gamma-ray spectroscopy example used to demonstrate the biasing utility. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2015

Nuclear criticality safety in the United Kingdom - Work In progress

Full text of publication follows: This commentary presents a brief overview of the recent and future challenges and opportunities presented by the UK nuclear industry. To give just a pure overview would lead perhaps to a rather dry and superficial review, so I have mixed in with this my own thoughts and coloured it with some of my opinions and observations from involvement in the field of Nuclear Criticality Safety (NCS) over the last 26 years. A brief overview of industry status and developments in the UK will, however, be highlighted, in order to give a background to what I hope would be an interesting personal commentary on the subject. The creation of the Nuclear Decommissioning Authority (NDA) in 2004 have (on behalf of the UK governments Department for Energy and Climate Change (DECC)) pursued the mission to 'deliver safe and sustainable solutions to the challenge of nuclear clean-up and waste management'. NDA have responsibilities for 17 of the civil nuclear facilities in the UK. This has meant for the UK nuclear industry, and in turn UK Criticality Safety, the significant focus toward safe, efficient and effective clean-up. With this mission, integrated with the ongoing commercial ventures at many of the UK licensed sites, a great variety of challenges are presented. For non-commercial activities, criticality safety has and will continue to have a significant future part to play. The UK government's commitment to a continued nuclear defense capacity ensures a major investment in the future and will continue to demand significant capability from criticality safety to support the existing commitments and future needs. Meeting these challenges demands a holistic approach to criticality safety if it is to succeed in supporting the continued safety of the industry. Technical solutions to the challenges are only part of the solution. Maintaining a technical competency, working as part of an integrated team with other safety disciplines, along with the design teams and operations management etc. are other important aspects. Working collaboratively within these teams and beyond, both nationally and internationally, is a necessary part of achieving success. Nuclear criticality safety as a discipline in the UK has always been conscious of isolation and stagnation in development given the relatively small number of personnel in the criticality field. With this concern, it benefits immensely from national and international collaboration, and this must be encouraged to continue. (author)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2022

Analysis of cell-based diffusion acceleration for the slice balance approach

In this work, we perform analysis on the use of cell-based diffusion acceleration methodologies to accelerate the convergence of transport solutions discretized with the slice balance approach (SBA) on unstructured polygonal grids.We investigated both linear diffusion synthetic acceleration (DSA) and non linear diffusion acceleration (NDA), including its partial-current variant (pNDA). DSA and NDA were both shown to diverge for intermediate ranges of mesh optical thicknesses. However, pNDA and Krylov methods like GMRES and Broyden stabilized the acceleration schemes, including problems with degenerate cells formed by mesh refinement. (author)

42 ENGINEERING↗

DOE OSTI2026

Agentic Diagrammatica: Towards Autonomous Symbolic Computation in High Energy Physics

We present Diagrammatica, a symbolic computation extension to the HEPTAPOD agentic framework, which enables LLM agents to plan and execute multi-step theoretical calculations. Symbolic computation poses a distinctive reliability challenge for LLM agents, as correctness is governed by implicit mathematical conventions that are not encoded in a form that can be easily checked in the computational backend. We identify two complementary remedies, tool-constrained computation and targeted knowledge grounding, and pursue the first as the primary architecture. Concretely, we concentrate the agent's action distribution onto tool calls with convention-fixing semantics, in which the agent specifies a compact, human-auditable diagram specification and a trusted backend performs the symbolic or numerical manipulations exactly. The toolkit provides two complementary calculation paths consuming a shared diagram specification: Naive Dimensional Analysis (NDA) for order-of-magnitude rate estimates and Exact Diagrammatic Analysis (EDA) for tree-level symbolic calculations via automatic FeynCalc code generation, both supplemented by automatic Feynman diagram enumeration and a navigable theory knowledge base. The architecture is validated on two benchmarks: (1) an exhaustive catalog of all tree-level, single-vertex $1\to 2$ partial decay widths across scalar, fermion, and vector parents, with complete massless and threshold limits and Standard Model validation; and (2) an NDA sensitivity study of the muon decay multiplicity $μ^+ \to ν_μ\barν_e + n(e^+e^-) + e^-$, determining the maximum observable $n$ at current and planned muon experiments.

Menzo, Tony [Alabama U.; Fermilab] (ORCID:00000002↗

DOE OSTI2015

A High-Order Nonlinear Diffusion Acceleration for the SN Equations Discretized with the Discontinuous FEM II: Fourier Analysis

In this paper, a Fourier Analysis (FA) of a new high order Nonlinear Diffusion Acceleration Method (NDA) for the $S_N$ equations discretized with the Discontinuous Finite Element Method (DFEM) is presented. In contrast to coarse mesh finite difference methods (CMFD), the associated diffusion system is discretized using DFEM with a function space identical to the $S_N$ equations. The utilized $S_N$ solver is the latest addition to the Rattlesnake\cite{Wang2013} radiation transport code that is based on the MOOSE\cite{Gaston2009} framework.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2024

Small instanton-induced flavor invariants and the axion potential

Small instantons which increase the axion mass due to an appropriate modification of QCD at a UV scale ΛSI, can also enhance the effect of CP-violating operators to shift the axion potential minimum by an amount, θ ind, proportional to the flavorful couplings in the SMEFT. Since physical observables must be flavor basis independent, we construct a basis of determinant-like flavor invariants that arise from instanton calculations containing the effects of dimension-six CP-odd operators at the scale Λ CP . This new basis provides a more reliable estimate of the shift θ ind, that is severely constrained by neutron electric dipole moment experiments. In particular, for the case of four-quark, semi-leptonic and gluon dipole operators, these invariants are then used to provide improved limits on the ratio of scales Λ SI /Λ CP for different flavor scenarios. The CP-odd flavor invariants also provide a classification of the leading effects from Wilson coefficients, and as an example, we show that a semi-leptonic four-fermion operator is subdominant compared to the four-quark operators. More generally, the flavor invariants, together with an instanton NDA, can be used to more accurately estimate small instanton effects in the axion potential that arise from any SMEFT operator.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗

DOE OSTI2015

The Impact of Gate Width Setting and Gate Utilization Factors on Plutonium Assay in Passive Correlated Neutron Counting

In the field of nuclear safeguards, passive neutron multiplicity counting (PNMC) is a method typically employed in non-destructive assay (NDA) of special nuclear material (SNM) for nonproliferation, verification and accountability purposes. PNMC is generally performed using a well-type thermal neutron counter and relies on the detection of correlated pairs or higher order multiplets of neutrons emitted by an assayed item. To assay SNM, a set of parameters for a given well-counter is required to link the measured multiplicity rates to the assayed item properties. Detection efficiency, die-away time, gate utilization factors (tightly connected to die-away time) as well as optimum gate width setting are among the key parameters. These parameters along with the underlying model assumptions directly affect the accuracy of the SNM assay. In this paper we examine the role of gate utilization factors and the single exponential die-away time assumption and their impact on the measurements for a range of plutonium materials. In addition, we examine the importance of item-optimized coincidence gate width setting as opposed to using a universal gate width value. Finally, the traditional PNMC based on multiplicity shift register electronics is extended to Feynman-type analysis and application of this approach to Pu mass assay is demonstrated.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2017

Naphthalene Bis(4,8-diamino-1,5-dicarboxyl)amide Building Block for Semiconducting Polymers

We report a new naphthalene bis(4,8-diamino-1,5-dicarboxyl)amide (NBA) building block for polymeric semiconductors. Computational modeling suggests that regio-connectivity at the 2,6- or 3,7-NBA positions strongly modulates polymer backbone torsion and, therefore, intramolecular π-conjugation and aggregation. Optical, electrochemical, and X-ray diffraction characterization of 3,7- and 2,6-dithienyl-substituted NBA molecules and corresponding isomeric NBA-bithiophene copolymers P1 and P2, respectively, reveals the key role of regio-connectivity. Charge transport measurements demonstrate that while the twisted 3,7-NDA-based P1 is a poor semiconductor, the planar 2,6-functionalized NBA polymers (P2–P4) exhibit ambipolarity, with μe and μh of up to 0.39 and 0.32 cm2/(V·s), respectively.

Eckstein, Brian J.↗

DOE OSTI2016

Systematic Uncertainties in High-Rate Germanium Data

For many nuclear material safeguards inspections, spectroscopic gamma detectors are required which can achieve high event rates (in excess of 10^6 s^-1) while maintaining very good energy resolution for discrimination of neighboring gamma signatures in complex backgrounds. Such spectra can be useful for non-destructive assay (NDA) of spent nuclear fuel with long cooling times, which contains many potentially useful low-rate gamma lines, e.g., Cs-134, in the presence of a few dominating gamma lines, such as Cs-137. Detectors in use typically sacrifice energy resolution for count rate, e.g., LaBr3, or visa versa, e.g., CdZnTe. In contrast, we anticipate that beginning with a detector with high energy resolution, e.g., high-purity germanium (HPGe), and adapting the data acquisition for high throughput will be able to achieve the goals of the ideal detector. In this work, we present quantification of Cs-134 and Cs-137 activities, useful for fuel burn-up quantification, in fuel that has been cooling for 22.3 years. A segmented, planar HPGe detector is used for this inspection, which has been adapted for a high-rate throughput in excess of 500k counts/s. Using a very-high-statistic spectrum of 2.4*10^11 counts, isotope activities can be determined with very low statistical uncertainty. However, it is determined that systematic uncertainties dominate in such a data set, e.g., the uncertainty in the pulse line shape. This spectrum offers a unique opportunity to quantify this uncertainty and subsequently determine required counting times for given precision on values of interest.

burn-up↗

DOE OSTI2026

Data-Driven Performance Optimization of Gamma Spectrometers With Many Channels

In gamma spectrometers with variable spectroscopic performance across many channels (e.g., many pixels or voxels), a tradeoff exists between including data from successively worse-performing readout channels and increasing efficiency. Brute-force calculation of the optimal set of included channels is exponentially infeasible as the number of channels grows, and approximate methods are required. In this work, we present a data-driven framework for attempting to find near-optimal sets of included detector channels. The framework leverages non-negative matrix factorization (NMF) to learn the behavior of gamma spectra across the detector and clusters similarly-performing detector channels together. Performance comparisons are then made between spectra with channel clusters removed, which is more feasible than brute force. The framework is general and can be applied to arbitrary, user-defined performance metrics depending on the application. We apply this framework to optimizing gamma spectra measured by H3D M400 CdZnTe (CZT) spectrometers, which exhibit variable performance across their crystal volumes. In particular, we show several examples optimizing various performance metrics for uranium and plutonium gamma spectra in non-destructive assay (NDA) for nuclear safeguards, and explore trends in performance versus parameters such as clustering algorithm type. We also compare the NMF + clustering pipeline to several non-machine-learning (ML) algorithms, including several greedy algorithms. Although, we find that the NMF + clustering pipeline tends to find the best-performing set of detector voxels, significantly improving over the unoptimized spectra, but that a greedy accumulation of spectra segmented by detector depth can, in some cases, give similar performance improvements in much less computation time.

Energy resolution↗

DOE OSTI2024

Radiation Monitoring for Volatilized Zinc Contamination Using Gamma-Ray Imaging and Spectroscopy

Gamma-ray imaging is a tool that has grown in importance in the applications of non-destructive assay (NDA) for radioactive survey and analysis of nuclear facilities. Imaging techniques have shown great promise in providing valuable information involving radioactive waste management and contamination prevention. For the application studied in this work, 65 Zn has been identified as a radioactive contaminant during tritium extraction. Due to the volatile nature of 65 Zn under the pressure and temperature changes during extraction operations, 65 Zn can easily travel through components of the extraction system as vapor, making it difficult to trap. Previous research involving the development of a filtration system showed that the 65 Zn can be trapped, mitigating product contamination. However, during the extraction process, direct analysis of the equipment to confirm that zinc contamination is trapped in the filter and has not spread to other components is impractical. Here, in this situation, the need to assay the location of the contamination with little-to-no interference with operations is vital. In this work, we demonstrate the use of a commercialized 3D position-sensitive CdZnTe (CZT) gamma-ray imaging spectrometer to provide analysis of the 65 Zn contamination. Onsite measurements during an extraction process are studied to assess the location and migration of the 65 Zn. The results obtained from real-time glovebox monitoring demonstrate the feasibility of gamma-ray imaging for localizing the contamination and providing a preliminary qualitative assessment that is intended to be used in future work quantifying the contamination build-up and activity over time.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2013

An integrated approach to geological disposal of UK wastes containing carbon-14

Carbon-14 is a key radionuclide in the assessment of the safety of a geological disposal facility for radioactive waste because of the calculated assessment of the radiological consequences of gaseous carbon-14 bearing species [i]. It may be that such calculations are based on overly conservative assumptions and that better understanding could lead to considerably reduced assessment of the radiological consequences from these wastes. Alternatively, it may be possible to mitigate the impact of these wastes through alternative treatment, packaging or design options. The Radioactive Waste Management Directorate of the UK's Nuclear Decommissioning Authority (NDA RWMD) has established an integrated project team in which the partners are working together to develop a holistic approach to carbon-14 management in the disposal system [ii]. For a waste stream containing carbon-14 to be an issue: There must be a significant inventory of carbon-14 in the waste stream; and That waste stream has to generate carbon-14 bearing gas; and a bulk gas phase has to entrain the carbon-14 bearing gas: and these gases must migrate through the engineered barriers in significant quantities; and these gases must migrate through the overlying geological environment (either as a distinct gas phase or as dissolved gas); and these gases must interact with materials in the biosphere (i.e. plants) in a manner that leads to significant doses and risks to exposed groups or potentially exposed groups. The project team has developed and used this 'and' approach to structure and prioritise the technical work and break the problem down in a manageable way. We have also used it to develop our approach to considering alternative treatment, packaging and design options. For example, it may be possible to pre-treat some wastes to remove some of the inventory or to segregate other wastes so that they are removed from any bulk gas phase which might facilitate migration through the geosphere. Initially, the project team has undertaken a six month programme of work to examine the current understanding of these aspects and has captured this in the Phase 1 report [ii], in a modelling basis spreadsheet and in scoping assessments, which help us better understand the potential significance of carbon-14. Using the current modelling basis, but ignoring any potential benefits from the geosphere in retarding or preventing gas from reaching the surface, the calculated release of carbon- 14 is dominated by: corrosion of irradiated reactive metals (in the operational and early post-closure time frame); corrosion of irradiated stainless steel and leaching of irradiated graphite (in the longer term). The Phase 1 work has shown that there is considerable scope for reducing the calculated radiological consequence for these wastes and a roadmap has been developed for a second Phase of work. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2023

Spectral Peak Enhancement by Combining Trusted Response Elements via Machine Learning (SPECTRE-ML) v0.8.0

SPECTRE-ML (Spectral Peak Enhancement by Combining Trusted Response Elements via Machine Learning) is a machine learning based program for finding optimal clusters of radiation detector segments (i.e., pixels or voxels) in order to improve spectral performance. It provides facilities for pre-processing and analyzing training datasets, running ML algorithms, and evaluating and visualizing outputs. SPECTRE-ML outperforms simpler ad-hoc segmentation methods such as uniform depth clusters, learning detector performance trends such as dead layers, edge effects, and gain shifts. Although extensible to arbitrary highly-segmented spectroscopic radiation detectors, SPECTRE-ML currently focuses on improving spectral performance in highly-segmented CdZnTe (CZT) detectors for International Atomic Energy Agency (IAEA) non-destructive assay (NDA) safeguards tasks.

Vavrek, Jayson↗

DOE OSTI2019

Study of serious adverse drug reactions using FDA-approved drug labeling and MedDRA

Background Adverse Drug Reactions (ADRs) are of great public health concern. FDA-approved drug labeling summarizes ADRs of a drug product mainly in three sections, i.e., Boxed Warning (BW), Warnings and Precautions (WP), and Adverse Reactions (AR), where the severity of ADRs are intended to decrease in the order of BW > WP > AR. Several reported studies have extracted ADRs from labeling documents, but most, if not all, did not discriminate the severity of the ADRs by the different labeling sections. Such a practice could overstate or underestimate the impact of certain ADRs to the public health. In this study, we applied the Medical Dictionary for Regulatory Activities (MedDRA) to drug labeling and systematically analyzed and compared the ADRs from the three labeling sections with a specific emphasis on analyzing serious ADRs presented in BW, which is of most drug safety concern. Results This study investigated New Drug Application (NDA) labeling documents for 1164 single-ingredient drugs using Oracle Text search to extract MedDRA terms. We found that only a small portion of MedDRA Preferred Terms (PTs), 3819 out of 21,920 or 17.42%, were observed in a whole set of documents. In detail, 466/3819 (12.0%) PTs were in BW, 2023/3819 (53.0%) were in WP, and 2961/3819 (77.5%) were in AR sections. We also found a higher overlap of top 20 occurring BW PTs with WP sections compared to AR sections. Within the MedDRA System Organ Class levels, serious ADRs (sADRs) from BW were prevalent in Nervous System disorders and Vascular disorders. A Hierarchical Cluster Analysis (HCA) revealed that drugs within the same therapeutic category shared the same ADR patterns in BW (e.g., nervous system drug class is highly associated with drug abuse terms such as dependence, substance abuse, and respiratory depression). Conclusions This study demonstrated that combining MedDRA standard terminologies with data mining techniques facilitated computer-aided ADR analysis of drug labeling. We also highlighted the importance of labeling sections that differ in seriousness and application in drug safety. Using sADRs primarily related to BW sections, we illustrated a prototype approach for computer-aided ADR monitoring and studies which can be applied to other public health documents.

59 BASIC BIOLOGICAL SCIENCES↗

DOE OSTI2022

Demonstration of MOOSE-based Griffin reactor physics, code for heterogeneous lead-cooled fast reactor analysis

The MOOSE-based reactor physics code Griffin was assessed on a heterogeneous pin-resolved model of a prototype lead-cooled fast reactor assembly. This model was developed in preparation for future use in MOOSE-based multiphysics calculations for computing hot channel factors. Heterogeneous multigroup cross sections were prepared using the fast reactor multi-group cross section processing code MC{sup 2}-3 using a two-step method. Griffin simulations were performed using the DFEM-SN solver on 576 cores on Argonne's LCRC cluster. Diffusion-based acceleration methods were applied (NDA and CMFD). Reference solutions were generated with continuous energy MCNP and the hybrid MOC/finite element solver PROTEUS-MOC for code-to-code comparison. Space-angle convergence studies were conducted to observe convergence in k-eigenvalue and axial pin power distributions. The fully resolved Griffin calculation was within 68 pcm of the MCNP eigenvalue and exhibited max 1.2% relative error in the axial pin power distribution. Griffin produced nearly identical results to PROTEUS-MOC when using the same 9-group multigroup cross-section set. Griffin demonstrated favorable scaling in wall-clock time and memory usage when using diffusion-based acceleration methods. Griffin is capable of simulating the pin-resolved heterogeneous LFR assembly with good accuracy and performance, and is suitable for future use in coupled high-fidelity hot channel factor simulations. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2026

EV Watts Public Database

With the rapid increase in vehicle electrification, there is a need for up-to-date, publicly available national data to understand end user charging and driving patterns, as well as vehicle and infrastructure performance, to inform research planning. Energetics worked with various partners to collect and analyze plug-in electric vehicle (PEV) and electric vehicle supply equipment (EVSE) data from 2019 to 2022. All sensitive attributes have been removed from this publicly available dataset. Researchers from one of the partner national labs under non-disclosure agreement (NDA) can request access to additional attributes by reaching out to evwattsdata@energetics.com.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗

DOE OSTI2013

Evaluation and Testing of the ADVANTG Code on SNM Detection

Pacific Northwest National Laboratory (PNNL) has been tasked with evaluating the effectiveness of ORNL’s new hybrid transport code, ADVANTG, on scenarios of interest to our NA-22 sponsor, specifically of detection of diversion of special nuclear material (SNM). PNNL staff have determined that acquisition and installation of ADVANTG was relatively straightforward for a code in its phase of development, but probably not yet sufficient for mass distribution to the general user. PNNL staff also determined that with little effort, ADVANTG generated weight windows that typically worked for the problems and generated results consistent with MCNP. With slightly greater effort of choosing a finer mesh around detectors or sample reaction tally regions, the figure of merit (FOM) could be further improved in most cases. This does take some limited knowledge of deterministic transport methods. The FOM could also be increased by limiting the energy range for a tally to the energy region of greatest interest. It was then found that an MCNP run with the full energy range for the tally showed improved statistics in the region used for the ADVANTG run. The specific case of interest chosen by the sponsor is the CIPN project from Las Alamos National Laboratory (LANL), which is an active interrogation, non-destructive assay (NDA) technique to quantify the fissile content in a spent fuel assembly and is also sensitive to cases of material diversion. Unfortunately, weight windows for the CIPN problem cannot currently be properly generated with ADVANTG due to inadequate accommodations for source definition. ADVANTG requires that a fixed neutron source be defined within the problem and cannot account for neutron multiplication. As such, it is rendered useless in active interrogation scenarios. It is also interesting to note that this is a difficult problem to solve and that the automated weight windows generator in MCNP actually slowed down the problem. Therefore, PNNL had determined that there is not an effective tool available for speeding up MCNP for problems such as the CIPN scenario. With regard to the Benchmark scenarios, ADVANTG performed very well for most of the difficult, long-running, standard radiation detection scenarios. Specifically, run time speedups were observed for spatially large scenarios, or those having significant shielding or scattering geometries. ADVANTG performed on par with existing codes for moderate sized scenarios, or those with little to moderate shielding, or multiple paths to the detectors. ADVANTG ran slower than MCNP for very simply, spatially small cases with little to no shielding that run very quickly anyway. Lastly, ADVANTG could not solve problems that did not consist of fixed source to detector geometries. For example, it could not solve scenarios with multiple detectors or secondary particles, such as active interrogation, neutron induced gamma, or fission neutrons.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗