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At least 1,783 records · Page 99
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Distinguishing Research Reactor Fuels Using Neutron Multiplicity - 25615

Neutron measurements are one of the reliable means of interrogation of fissile material. Neutron measurements are used for total counting (pure metal fissile samples) or counting coincidence/multiplicity (for impure fissile samples). Various flavors of neutron measurement are used routinely by the IAEA to determine the fissile mass of an unknown sample. The objective of this work is to design an experiment to distinguish the two different fuel types using neutron multiplicity method. The design of experiment is performed using Monte Carlo N- Particle (MCNP) transport code. General Electric (GE) designed 1-MWt pool-type research reactor at UMass Lowell (UMLRR) uses Materials Testing Reactor (MTR) plate-type fuel consisting of 19 full and 2 partial fuel assemblies, controlled by four boron control blades and one regulating control rod. Two different fuel types are available at the UMLRR facility. In 2008, UMass Lowell obtained slightly burnt fuel assemblies from the decommissioned WPI research reactor. The simulation results discussed in this paper show that within the limits of statistical error, the single and double moments for the WPI and UMLRR fuels can be measured. The two fuels have very similar composition and enrichments resulting in very similar neutron multiplicity moments. This implies that neutron multiplicity measurements alone would not be sufficient to distinguish between the two fuel types at low burnups. The change in the observed single-double ratio between each fuel type is less than 6 % which falls below the uncertainty in the efficiency of the JCC71 slab detectors. The neutron signature from low burnup and low enriched uranium fuel are not strong, however, these can be measured under low background and long measurement times. This would be an important consideration in designing the experiments and parallel gamma measurements will also be planned. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Magnetic Resonance Imaging-Based Radiation-Absorbed Dose Estimation of {sup 166}Ho Microspheres in Liver Radioembolization

Purpose: To investigate the potential of magnetic resonance imaging (MRI) for accurate assessment of the three-dimensional {sup 166}Ho activity distribution to estimate radiation-absorbed dose distributions in {sup 166}Ho-loaded poly (L-lactic acid) microsphere ({sup 166}Ho-PLLA-MS) liver radioembolization. Methods and Materials: MRI, computed tomography (CT), and single photon emission CT (SPECT) experiments were conducted on an anthropomorphic gel phantom with tumor-simulating gel samples and on an excised human tumor-bearing liver, both containing known amounts of {sup 166}Ho-PLLA-MS. Three-dimensional radiation-absorbed dose distributions were estimated at the voxel level by convolving the {sup 166}Ho activity distribution, derived from quantitative MRI data, with a {sup 166}Ho dose point-kernel generated by MCNP (Monte Carlo N-Particle transport code) and from Medical Internal Radiation Dose Pamphlet 17. MRI-based radiation-absorbed dose distributions were qualitatively compared with CT and autoradiography images and quantitatively compared with SPECT-based dose distributions. Both MRI- and SPECT-based activity estimations were validated against dose calibrator measurements. Results: Evaluation on an anthropomorphic phantom showed that MRI enables accurate assessment of local {sup 166}Ho-PLLA-MS mass and activity distributions, as supported by a regression coefficient of 1.05 and a correlation coefficient of 0.99, relating local MRI-based mass and activity calculations to reference values obtained with a dose calibrator. Estimated MRI-based radiation-absorbed dose distributions of {sup 166}Ho-PLLA-MS in an ex vivo human liver visually showed high correspondence to SPECT-based radiation-absorbed dose distributions. Quantitative analysis revealed that the differences in local and total amounts of {sup 166}Ho-PLLA-MS estimated by MRI, SPECT, and the dose calibrator were within 10%. Excellent agreement was observed between MRI- and SPECT-based dose-volume histograms. Conclusions: Quantitative MRI was demonstrated to provide accurate three-dimensional {sup 166}Ho-PLLA-MS activity distributions, enabling localized intrahepatic radiation-absorbed dose estimation by convolution with a {sup 166}Ho dose point-kernel for liver radioembolization treatment optimization and evaluation.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2012

Dependences of mucosal dose on photon beams in head-and-neck intensity-modulated radiation therapy: a Monte Carlo study

Dependences of mucosal dose in the oral or nasal cavity on the beam energy, beam angle, multibeam configuration, and mucosal thickness were studied for small photon fields using Monte Carlo simulations (EGSnrc-based code), which were validated by measurements. Cylindrical mucosa phantoms (mucosal thickness = 1, 2, and 3 mm) with and without the bone and air inhomogeneities were irradiated by the 6- and 18-MV photon beams (field size = 1 Multiplication-Sign 1 cm{sup 2}) with gantry angles equal to 0 Degree-Sign , 90 Degree-Sign , and 180 Degree-Sign , and multibeam configurations using 2, 4, and 8 photon beams in different orientations around the phantom. Doses along the central beam axis in the mucosal tissue were calculated. The mucosal surface doses were found to decrease slightly (1% for the 6-MV photon beam and 3% for the 18-MV beam) with an increase of mucosal thickness from 1-3 mm, when the beam angle is 0 Degree-Sign . The variation of mucosal surface dose with its thickness became insignificant when the beam angle was changed to 180 Degree-Sign , but the dose at the bone-mucosa interface was found to increase (28% for the 6-MV photon beam and 20% for the 18-MV beam) with the mucosal thickness. For different multibeam configurations, the dependence of mucosal dose on its thickness became insignificant when the number of photon beams around the mucosal tissue was increased. The mucosal dose with bone was varied with the beam energy, beam angle, multibeam configuration and mucosal thickness for a small segmental photon field. These dosimetric variations are important to consider improving the treatment strategy, so the mucosal complications in head-and-neck intensity-modulated radiation therapy can be minimized.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2015

SU-E-T-608: Perturbation Corrections for Alanine Dosimeters in Different Phantom Materials in High-Energy Photon Beams

Purpose: Alanine dosimeters are often used for in-vivo dosimetry purposes in radiation therapy. In a Monte Carlo study the influence of 20 different surrounding/phantom materials for alanine dosimeters was investigated. The investigations were performed in high-energy photon beams, covering the whole range from {sup 60}Co up to 25 MV-X. The aim of the study is the introduction of a perturbation correction k{sub env} for alanine dosimeters accounting for the environmental material. Methods: The influence of different surrounding materials on the response of alanine dosimeters was investigated with Monte Carlo simulations using the EGSnrc code. The photon source was adapted with BEAMnrc to a {sup 60}Co unit and an Elekta (E{sub nom}=6, 10, 25 MV-X) linear accelerator. Different tissue-equivalent materials ranging from cortical bone to lung were investigated. In addition to available phantom materials, some material compositions were taken and scaled to different electron densities. The depth of the alanine detectors within the different phantom materials corresponds to 5 cm depth in water, i.e. the depth is scaled according to the electron density (n{sub e}/n{sub e,w}) of the corresponding phantom material. The dose was scored within the detector volume once for an alanine/paraffin mixture and once for a liquid water voxel. The relative response, the ratio of the absorbed dose to alanine to the absorbed dose to water, was calculated and compared to the corresponding ratio under reference conditions. Results: For each beam quality the relative response r and the correction factor for the environment kenv was calculated. k{sub env}=0.9991+0.0049 *((n{sub e}/n{sub e,w})−0.7659){sup 3} Conclusion: A perturbation correction factor k{sub env} accounting for the phantom environment has been introduced. The response of the alanine dosimeter can be considered independent of the surrounding material for relative electron densities (n{sub e}/n{sub e,w}) between 1 and 1.4. For denser materials such as bone or much less dense surroundings such as lung, a small correction would be appropriate.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2011

Characterization of Neutron and Gamma Dose in the Irradiation Cell of Texas A and M University Research Reactor

The Monte Carlo N-Particle (MCNP) code was used to develop a three dimensional computational model of the Texas A and M University Nuclear Science Center Reactor (NSCR) operating against the irradiation (dry cell) at steady state thermal power of 1 MW. The geometry of the NSCR core and the dry cell were modeled in detail. NSCR is used for a wide variety of experiments that utilizes the dry cell for neutron as well as gamma irradiation of samples. Information on the neutron and gamma radiation environment inside the dry cell is required to facilitate irradiation of samples. This paper presents the computed neutron flux, neutron and gamma dose rate, and foil reaction rates in the dry cell, obtained through MCNP5 simulations of the NSCR core. The neutron flux was measured using foil activation method and the reaction rates obtained from {sup 197}Au(n,{gamma}){sup 198}Au and {sup 54}Fe(n,p){sup 54}Mn were compared with the model and they showed agreement within {approx} 20%. The gamma dose rate at selected locations inside the dry cell was measured using radiochromic films and the results indicate slightly higher dose rates than predicted from the model. This is because the model calculated only prompt gamma dose rates during reactor operation while the radiochromic films measured gammas from activation products and fission product decayed gammas. The model was also used to calculate the neutron energy spectra for the energy range from 0.001 eV- 20 MeV. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

Design of a backscatter 14-MeV neutron time-of-flight spectrometer for experiments at ITER

Neutron energy spectrometry diagnostics play an important role in present-day experiments related to fusion energy research. Measurements and thorough analysis of the neutron emission from the fusion plasma give information on a number of basic fusion performance quantities, on the condition of the neutron source and plasma behavior. Here we discuss the backscatter Time-of-Flight (bTOF) spectrometer concept as a possible instrument for performing high resolution measurements of 14 MeV neutrons. The instrument is based on two sets of scintillators, a first scatterer exposed to a collimated neutron beam and a second detector set placed in the backward direction. The scintillators of the first set are enriched in deuterium to achieve neutron backscattering. The energy resolution and efficiency of a bTOF instrument have been determined for various geometrical configurations. A preliminary design of optimal geometry for the two scintillator sets has been obtained by Monte Carlo simulations based on the MCNPX code.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2012

Intrinsic Radiation Source Generation with the ISC Package: Data Comparisons and Benchmarking

The characterization of radioactive emissions from unstable isotopes (intrinsic radiation) is necessary for shielding and radiological-dose calculations from radioactive materials. While most radiation transport codes, e.g., MCNP [X-5 Monte Carlo Team, 2003], provide the capability to input user prescribed source definitions, such as radioactive emissions, they do not provide the capability to calculate the correct radioactive-source definition given the material compositions. Special modifications to MCNP have been developed in the past to allow the user to specify an intrinsic source, but these modification have not been implemented into the primary source base [Estes et al., 1988]. To facilitate the description of the intrinsic radiation source from a material with a specific composition, the Intrinsic Source Constructor library (LIBISC) and MCNP Intrinsic Source Constructor (MISC) utility have been written. The combination of LIBISC and MISC will be herein referred to as the ISC package. LIBISC is a statically linkable C++ library that provides the necessary functionality to construct the intrinsic-radiation source generated by a material. Furthermore, LIBISC provides the ability use different particle-emission databases, radioactive-decay databases, and natural-abundance databases allowing the user flexibility in the specification of the source, if one database is preferred over others. LIBISC also provides functionality for aging materials and producing a thick-target bremsstrahlung photon source approximation from the electron emissions. The MISC utility links to LIBISC and facilitates the description of intrinsic-radiation sources into a format directly usable with the MCNP transport code. Through a series of input keywords and arguments the MISC user can specify the material, age the material if desired, and produce a source description of the radioactive emissions from the material in an MCNP readable format. Further details of using the MISC utility can be obtained from the user guide [Solomon, 2012]. The remainder of this report presents a discussion of the databases available to LIBISC and MISC, a discussion of the models employed by LIBISC, a comparison of the thick-target bremsstrahlung model employed, a benchmark comparison to plutonium and depleted-uranium spheres, and a comparison of the available particle-emission databases.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Radiation dose and cancer risk to out-of-field and partially in-field organs from radiotherapy for symptomatic vertebral hemangiomas

Purpose: Vertebral hemangiomas (VHs) are the most common benign tumors of the spine that may cause bone resorption. Megavoltage irradiation is usually the treatment of choice for the management of symptomatic VHs. The current study was conducted to estimate the risk for carcinogenesis from radiotherapy of this benign disease on the basis of the calculated radiation doses to healthy organs. Methods: The Monte Carlo N-particle transport code was employed to simulate the irradiation with 6 MV x-rays of a VH presented in the cervical, upper thoracic, lower thoracic, and lumbar spine. The average radiation dose (D{sub av}) received by each critical organ located outside the primarily irradiated area was calculated. Three-dimensional treatment plans were also generated for the VHs occurring at the four different sites of the spinal cord based on patients’ computed tomography data. The organ equivalent dose (OED) to each radiosensitive structure, which was partly encompassed by the applied treatment fields, was calculated with the aid of differential dose–volume histograms. The D{sub av} and the OED values were combined with a linear-no-threshold model and a nonlinear mechanistic model, respectively, to estimate the organ-, age-, and gender-specific lifetime attributable risks (LARs) for cancer development. The estimated risks were compared with the respective nominal lifetime intrinsic risks (LIRs) for the unexposed population. Results: For a standard target dose of 34 Gy, the OED varied from 0.39–5.15 Gy by the organ of interest and the irradiation site. The D{sub av} range for the out-of-field organs was 4.9 × 10{sup −4} to 0.56 Gy. The LAR for the appearance of malignancies in the partially in-field organs after radiotherapy of male and female patients was (0.08%–1.8%) and (0.09%–1.9%), respectively. These risk values were 1.5–15.5 times lower when compared to the respective LIRs. The lifetime probability for out-of-field cancer induction in irradiated males and females was (2.5 × 10{sup −4} to 7.7 × 10{sup −2})% and (1.4 × 10{sup −4} to 2.6 × 10{sup −1})%, respectively. The above risks were one to four orders of magnitude lower than the LIRs. Conclusions: The probability for the development of out-of-field malignancies due to radiotherapy for VHs is trivial with respect to the nominal risk for unexposed population. The respective cancer risks to partially in-field organs are smaller than the nominal probabilities but they should not be considered as inconsiderable. These risks may be taken into account during the follow-up of patients treated for a symptomatic VH.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2022

Computational models of direct and indirect X‐ray breast imaging detectors for in silico trials

Abstract Background To facilitate in silico studies that investigate digital mammography (DM) and breast tomosynthesis (DBT), models replicating the variety in imaging performance of the DM and DBT systems, observed across manufacturers are needed. Purpose The main purpose of this work is to develop generic physics models for direct and indirect detector technology used in commercially available systems, with the goal of making them available open source to manufacturers to further tweak and develop the exact in silico replicas of their systems. Methods We recently reported on an in silico version of the SIEMENS Mammomat Inspiration DM/DBT system using an open‐source GPU‐accelerated Monte Carlo x‐ray imaging simulation code (MC‐GPU). We build on the previous version of the MC‐GPU codes to mimic the imaging performances of two other Food and Drug Administration (FDA)‐approved DM/DBT systems, such as Hologic Selenia Dimensions (HSD) and the General Electric Senographe Pristina (GSP) systems. In this work, we developed a hybrid technique to model the optical spread and signal crosstalk observed in the GSP and HSD systems. MC simulations are used to track each x‐ray photon till its first interaction within the x‐ray detector. On the other hand, the signal spread in the x‐ray detectors is modeled using previously developed analytical equations. This approach allows us to preserve the modeling accuracy offered by MC methods in the patient body, while speeding up secondary carrier transport (either electron–hole pairs or optical photons) using analytical equations in the detector. The analytical optical spread model for the indirect detector includes the depth‐dependent spread and collection of optical photons and relies on a pre‐computed set of point response functions that describe the optical spread as a function of depth. To understand the capabilities of the computational x‐ray detector models, we compared image quality metrics like modulation transfer function (MTF), normalized noise power spectrum (NNPS), and detective quantum efficiency (DQE), simulated with our models against measured data. Please note that the purpose of these comparisons with measured data would be to gauge if the model developed as part of this work could replicate commercially used direct and indirect technology in general and not to achieve perfect fits with measured data. Results We found that the simulated image quality metrics such as MTF, NNPS, and DQE were in reasonable agreement with experimental data. To demonstrate the imaging performance of the three DM/DBT systems, we integrated the detector models with the VICTRE pipeline and simulated DM images of a fatty breast model containing a spiculated mass and a calcium oxalate cluster. In general, we found that the images generated using the indirect model appeared more blurred with a different noise texture and contrast as compared to the systems with direct detectors. Conclusions We have presented computational models of three commercially available FDA‐approved DM/DBT systems, which implement both direct and indirect detector technology. The updated versions of the MC‐GPU codes that can be used to replicate three systems are available in open source format through GitHub.

Sengupta, Aunnasha↗
DOE OSTI2015

Top-pair production and decay at NLO matched with parton showers

We present a next-to-leading order (NLO) calculation of tt¯ production in hadronic collisions interfaced to shower generators according to the POWHEG method. We start from an NLO result from previous work, obtained in the zero width limit, where radiative corrections to both production and decays are included. The POWHEG interface required an extension of the POWHEG BOX framework, in order to deal with radiation from the decay of resonances. This extension is fully general (i.e. it can be applied in principle to any process considered in the zero width limit), and is here applied for the first time. In order to perform a realistic simulation, we introduce finite width effects using different approximations, that we validated by comparing with published exact NLO results. We have interfaced our POWHEG code to the PYTHIA8 shower Monte Carlo generator. At this stage, we dealt with novel issues related to the treatment of resonances, especially with regard to the initial scale for the shower that needs to be set appropriately. This procedure affects, for example, the fragmentation function of the b quark, that we have studied with particular attention. We believe that the tool presented here improves over previous generators for all aspects that have to do with top decays, and especially for the study of issues related to top mass measurements that involve B hadrons or b jets. As a result, the work presented here also constitutes a first step towards a fully consistent matching of NLO calculations involving intermediate resonances decaying into coloured particles, with parton showers.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2012

Nuclear spectroscopy for in situ soil elemental analysis: Monte Carlo simulations

We developed a model to simulate a novel inelastic neutron scattering (INS) system for in situ non-destructive analysis of soil using standard Monte Carlo Neutron Photon (MCNP5a) transport code. The volumes from which 90%, 95%, and 99% of the total signal are detected were estimated to be 0.23 m{sup 3}, 0.37 m{sup 3}, and 0.79 m{sup 3}, respectively. Similarly, we assessed the instrument's sampling footprint and depths. In addition we discuss the impact of the carbon's depth distribution on sampled depth.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2017

Spectroscopic and neutron detection properties of rare earth and titanium doped LiAlO 2 single crystals

In this study, LiAlO2 crystals doped with rare-earth elements and Ti were produced by the CZ method and spectroscopic and neutron detection properties were investigated. Photoluminescence revealed no clear luminescent activation of LiAlO2 by the rare-earth dopants though some interesting luminescence was observed from secondary phases within the crystal. Gamma-ray pulse height spectra collected using a 137Cs source exhibited only a Compton edge for the crystals. Neutron modeling using Monte Carlo N-Particle Transport Code revealed most neutrons used in the detection setup are thermalized, and while using natural lithium in the crystal growth, which contains 7.6 % 6Li, a 10 mm Ø by 10 mm sample of LiAlO2 has a 70.7 % intrinsic thermal neutron capture efficiency. Furthermore, the pulse height spectra collected using a 241Am-Be neutron source demonstrated a distinct neutron peak.

Dickens, Peter T.↗
DOE OSTI2021

Comparison of MCNP thermal scattering laws to inelastic neutron scattering data

Neutron transport computations employ thermal scattering laws (TSLs) to account for the influence of material structure and collective excitations at low energy (≲ 10 eV). Recent work has been undertaken to improve TSL accuracy and expand the number of available materials in the Evaluated Nuclear Data File (ENDF). However, comparison of TSL to measured data is often sparse. In this work, we present a method to evaluate TSL accuracy in the Monte Carlo N-Particle (MCNP) code by simulating inelastic neutron scattering (INS) measurement and comparing to data. The method is briefly described with example application to the recently released ENDF-VIII TSL library for polyethylene as well as a cryogenic PE TSL we are separately developing.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Monte Carlo N-Particle forward modeling for density reconstruction of double shell capsule radiographs

In the Double Shell Inertial Confinement Fusion concept, characterizing the shape asymmetry of imploding metal shells is vital for understanding energy-efficient compression and radiative losses of the thermonuclear fuel. The Monte Carlo N-Particle MCNP® code forward models radiography of Double Shell capsule implosions using the Advanced Radiographic Capability at the National Ignition Facility. A procedure is developed for using MCNP to reconstruct density profiles from the radiograph image intensity. For a given Double Shell imploding target geometry, MCNP radiographs predict image contrast, which can help guide experimental design. Finally, in future work, the calculated MCNP synthetic radiographs will be compared with experimental radiographs to determine the radial and azimuthal density profiles of the Double Shell capsules.

47 OTHER INSTRUMENTATION↗
DOE OSTI2023

Activation Analysis in Preparation for a Tungsten Irradiation Experiment at LANSCE

To organize the safe handling of activated material, knowing the residual dose rates is crucial. Here, in this work, we present the pre-experiment activation analysis for an experiment in which tungsten blocks are irradiated by 800-MeV protons. In this analysis, we use the Monte Carlo N-Particle (MCNP) code for radiation transport, Attila4MC for unstructured mesh generation, and Activation in Accelerator Radiation Environments (AARE), including CINDER2008, for activation analysis. If the tungsten blocks must be removed within a day after the experiment, then exposure to personnel entering the room must be reduced. One exposure-reduction strategy is to add carbon steel shielding around the tungsten blocks, efficiently reducing the dose from the activated tungsten. However, the shielding becomes activated itself during irradiation: 56Mn is the dominant contributor for short decay times. The actual schedule at the time of the experiment allowed sufficient cool-off time for the tungsten in the room so that additional shielding was not necessary. A less rigorous comparison of the calculated values with the post-experiment measurements showed reasonable agreement.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Impact of High-Reactivity Advanced Test Reactor Experiments on Photon Heating in Nearby Experiment Locations

The Advanced Test Reactor’s (ATR’s) distinctive ability to provide a wide range of irradiation conditions is attractive for programs pursuing fuel qualification experiments. These potentially high-fuel-load experiments are a relatively new development and produce unexplored effects on nearby experiments. Here, this paper explores how photon heating of such an experiment may affect other nearby experiment programs, ultimately serving to better inform decisions regarding experiment design and risks to programmatic goals. The MC21 (Monte Carlo for the 21st Century) code is used to model and study how gamma heat generation rates and axial effects impact different ATR positions. The results reveal that the proximity of a given experiment’s position to the high-fuel-load one can significantly alter that experiment’s expected axial profile.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Multi-Objective Optimization of Uranium Target Assembly–3: A Comparison of Genetic and Traditional Methods

Commonly produced as a byproduct of uranium fission, 99 Mo is a key medical isotope that is in high demand in the United States. An international goal is to switch from medical isotope production technologies that require highly enriched uranium to medical isotope production technologies that require only low-enriched uranium. Niowave Inc. is contributing to this goal by developing an accelerator-driven subcritical assembly called the Uranium Target Assembly (UTA). This work compares the performance of Dakota’s Multi-Objective Genetic Algorithm (MOGA) against traditional sensitivity analysis in the neutronic optimization of the UTA-3 system. The design objectives are k-eigenvalue (k eff ) and natural uranium fission power, which are directly correlated with the amount of 99 Mo produced. Dakota:MOGA did not perform as well as human engineering ingenuity in optimization studies with high numbers of input parameters, such as fuel rod type selection and fuel rod placement. However, Dakota:MOGA did outperform traditional sensitivity analysis in optimization studies with fewer than 20 parameters and revealed the degree to which each parameter influences the optimal design space for k eff and natural uranium fission power (to a lesser extent). As the design model became more complex in the final stage of design, the computational resources required to calculate the design objective values in the Monte Carlo N-Particle transport code from selected input parameter combinations limited Dakota:MOGA’s performance, and, unfortunately, human intervention was required to discern the optimal design space. In conclusion, future work will attempt to reduce computational resource constraints by incorporating areduced-order neutronics model into the optimization cycle.

Accelerator-driven systems↗
DOE OSTI2026

CADIS and FW-CADIS Variance Reduction in Gamma Transport for Predicting Prompt Forensics Signatures

The goal of prompt nuclear forensics is to determine the characteristics of a nuclear detonation based on the signatures available almost immediately after the explosion. An important characteristic is the reaction time history (RTH), a measure of the device’s rate of neutron multiplication. The RTH can be estimated by observation of the gamma radiation emitted from the detonation, which can be detected directly or observed indirectly as Teller light. Gamma transport simulations used to predict these radiation fields are often modeled stochastically using the Monte Carlo N-Particle (MCNP) code, which can be a computationally demanding task due to the number of particle histories needed to achieve statistical convergence. In an attempt to improve the efficiency of these calculations, we evaluate two variance reduction techniques: Consistent Adjoint-Driven Importance Sampling (CADIS) and Forward-Weighted Consistent Adjoint-Driven Importance Sampling (FW-CADIS). These methods use a deterministically calculated adjoint flux to create weight windows and source biasing that guide MCNP sampling. We study the utility of CADIS and FW-CADIS for their use in MCNP gamma transport for nuclear forensics prediction simulations. Furthermore, the results demonstrate that both CADIS and FW-CADIS improve the accuracy for forensics-focused simulations, with CADIS being most beneficial in direct detection and FW-CADIS being ideal for computing a global Teller light source.

CADIS↗