Nuclear Criticality Research Articles

COG 11.3 New Features

COG is a high fidelity, multi-particle, Monte Carlo radiation transport code in development at LLNL since 1980 in support of multiple applications including nuclear criticality safety, radiation shielding, radiography, and subcritical multiplicity analysis. COG 11.3 is the latest version scheduled for public release in 2024 representing the culmination of ten years of development since previous releases. This paper describes the major new features available in COG 11.3 including: (a) updated nuclear data libraries; (b) updated activation data; (c) advanced LLNL Fission Reaction Event Yield Algorithm; (d) new time-tagged list-mode detector feature; (e) new spontaneous fission source feature; (f) new alpha and deuteron particle transport; (g) updated electron transport (implementing EGS5 with internal electron library generation using PEGS5); (h) three new estimators for the effective delayed neutron fraction; (i) new input pre-processor options for user-friendly generation of three-dimensional rectangular and triangular lattice geometries; (j) new parallel-processing capabilities for the inverse reactor period and CritDetVR features using MPI; (k) a new imaging detector feature; and (l) details of the modernized COG website at https://cog.llnl.gov. Like previously released versions, COG11.3 will be available to users through RSICC at ORNL, the Nuclear Energy Agency Data Bank at OECD, and by special arrangement through LLNL.

Availability of Critical Benchmark Experiments for the Pebble Tanker Transportation Model for Nuclear Criticality Safety Validation of TRISO Pebbles

This study addresses the need for comprehensive investigations into TRi-structural ISOtropic (TRISO) fuel pebble transportation validation. In this work, an exploratory model, the pebble tanker(PT), was developed with the aim of facilitating the validation of nuclear criticality safety calculations in the context of industrial-scale transportation of TRISO fuel. The PT model was designed to investigate the availability and applicability of critical benchmark experiments crucial for assessing the transportation of these pebbles. This work incorporated sensitivity/uncertainty (S/U) similarity studies to quantify the applicability of critical benchmark experiments and to address nuclear data uncertainties in the context of TRISO transportation. Two container models were investigated: one for the Hermes-type pebble and one for the Pebble Bed Modular Reactor (PBMR)–type pebble. The models were simplified, considering fuel, containment, and either water or air, to enable a focus on the underlying physics of applications involving TRISO fuel pebbles using the PT model. A crucial aspect under consideration was the capacity of the transport package to hold pebbles while ensuring subcriticality in the flooded state. An approach in the criticality validation process involves assessing the similarity between systems through an integral index parameter evaluation. This involves calculating a correlation coefficient (referred to as ck) based on shared nuclear data–induced uncertainty between a benchmark experiment and the application of the PT model. To facilitate this analysis, the SCALE tools, particularly the CSAS6-Shift, TSUNAMI-3D-Shift, and TSUNAMI-IP sequences, were employed for comprehensive studies in neutronics and S/U analysis. Our findings showed that there are sufficient critical experimental benchmarks to perform this validation of the PT model in the most reactive state, i.e. when the tanker is flooded. This paper provides valuable insights into validating a transport package for Generation IV TRISO fuel pebbles.

Application of the Krylov-Schur method in three-dimensional nuclear reactor discrete ordinates criticality calculations

The accurate and efficient solution of the neutron transport equation is essential in nuclear reactor physics for understanding reactor kinetics, stability, and safety. This study investigates the application of the Krylov-Schur method to three-dimensional neutron transport criticality calculations within the Marvin code, comparing its performance to the traditional Power Iteration (PI) method. Using the Takeda benchmark problems, the Krylov-Schur solver demonstrated high accuracy in eigenvalue calculations and neutron flux distributions, closely matching reference Monte Carlo (MC) results. Additionally, the parallel efficiency of the Krylov-Schur method was evaluated, showing significant speed-up and better scalability compared to the PI method, particularly in large-scale computations. However, the method requires a larger memory footprint due to the need to store multiple Krylov subspace vectors and Schur decompositions. Despite this, the findings highlight the Krylov-Schur method's robustness and computational efficiency, making it a promising tool for neutron transport simulations in complex reactor configurations. Future work will focus on investigating the subtraction of high-order eigenvalues and eigenvectors using the Krylov-Schur method to further enhance neutron transport simulations.

Technical, Regulatory, and Policy Feasibility Assessments of Rod Consolidation in Spent Nuclear Fuel Disposal

Various approaches have been proposed to address the challenges associated with storage and disposal of spent nuclear fuel. Rod consolidation, which involves extracting only the fuel rods from fuel assemblies and combining them for storage with reduced spacing, gained attention until the 1990s for its benefits in storage efficiency. In this case study, experts from diverse backgrounds evaluate the technical and political feasibility of rod consolidation for enhancing disposal practices. This study conducts a technical and regulatory analysis of rod consolidation under disposal conditions and evaluates its feasibility from a disposal perspective. From a technical viewpoint of 2:1 and 1.5:1 rod consolidations, the rod consolidation methods demonstrate positive effects on nuclear criticality compared to the reference, non-rod consolidation case. For the advanced repository design with elevated temperature conditions of 150 °C or higher, the applicability of rod consolidation becomes more convincing, thereby achieving a reduction in the number of required disposal canisters. In addition, regulatory considerations primarily focus on the requirements for applying rod consolidation technology under the Korea Nuclear Act. Moreover, this research conducts environmental, economic, and acceptance analyses to ascertain the viability of rod consolidation. The findings provide compelling evidence supporting the feasibility of rod consolidation technology for disposal purposes, highlighting its technical, regulatory, economic, environmental, and societal acceptance advantages. However, challenges remain, particularly concerning its applicability to high-burnup nuclear fuel and unresolved regulatory and technical issues.

A Consistent Comparison of Upper Subcritical Limit Methods

This study conducts a consistent and comprehensive comparison of several methods for estimating upper subcritical limits (USLs) for nuclear criticality safety analysis. To eliminate inconsistency caused by the discrepancy between the estimated covariance data and the true degree of uncertainty present in nuclear data, the experimental system eigenvalues k exp were assumed to equal the calculated eigenvalue for the systems using nominal ENDF/B-VII.1 cross sections, and the estimated calculated eigenvalue k calc was assumed to equal the calculated eigenvalue for the systems using nuclear data from one perturbed cross-section library. USLs are estimated for a variety of validation application cases using the Whisper, TSURFER, and USLSTATS tools and are compared to a reference 95/95 limit. The TSURFER approach produced the strongest agreement with the reference USLs; the Whisper produced USLs that were reliably conservative compared to the reference USL; and the USLSTATS approach experienced difficulty consistently producing accurate USL estimates. Last, this study observed a noteworthy discrepancy when using Cholesky decomposition to randomly sample neutron cross-section covariance data that are small in magnitude.

A review of criticality dosimetry at the Y-12 National Security Complex and practical importance of dose accuracy in emergency response

A nuclear criticality results in the emission of both neutron and gamma radiation and can produce doses to personnel near the event that exceed 0.1 Gy (10 rad). The primary purpose of nuclear accident dosimetry is to rapidly identify affected personnel in need of prompt medical treatment and to reassure personnel who have been only minimally exposed. While accurate dosimetry is desired, it must be recognized that dose determinations made from whole-body dosimeters or simple triage methods are very rough estimates and contain significant uncertainties. Even when accounting for factors like varying neutron energy spectra, mean photon energies, body orientation within the radiation field, and transient effects on dosimeter response, etc., the end value is a dosimetric quantity defined for very specific radiological conditions and determined within a simple phantom usually at a single depth. Of more importance is the biological response to the radiation, which will vary by person and can be affected by the individual's radiation sensitivity, age, gender, mass, and underlying health conditions. The overall biological, person-specific response to a given dose cannot be precisely determined except by patient symptom observation and individual biological dosimetry (e.g. chromosome analysis, lymphocyte ratios, etc.). This work describes and discusses the criticality accident dosimetry program at the Y-12 National Security Complex, a United States Department of Energy National Nuclear Security Administration facility. The primary goals of the Y-12 accident dosimetry program are, among others, the rapid identification of significantly exposed persons, prompt routing of exposed workers for medical evaluation and treatment, and the ultimate processing of dosimeters to assign doses to personnel.

A three-dimensional criticality safety analysis code for spent fuel solution system

The nuclear criticality safety analysis is crucial for the nuclear safety of spent fuel reprocessing plants. Thus, a set of numerical tools with high accuracy and efficiency to predict the criticality and simulate the hypothetical accidents in the reprocessing procedure is of great importance. Among them, this paper focus on the analysis of the storage tank which is used for dissolution and storage of the spent fuel. According to the characteristics of spent fuel solution system, a parallel 3D critical safety analysis tool for fissile solution system, Hydra-TD, is developed. Generally, it contains the cross-section generation model, the three-dimensional space-time neutron kinetics model, and the R-Z two-dimensional heat conduction and radiolysis gas simulation model. The verifications based on the experiments of the SILENE and TRACY facility are conducted. The results show good accuracy of the prediction of the first fission power peak, multiplication time and total fission energy, indicating the reliability of these numerical models.

United States Navy nuclear accident dosimetry program: History and Current Status

Federal Regulations of the United States require that installations possessing sufficient quantities of fissile material to potentially constitute a critical mass, such that the excessive exposure of individuals to radiation from a nuclear accident is possible, shall provide appropriate nuclear accident dosimetry. The American National Standard ANSl/HPS N13.3–2013 Dosimetry for Criticality Accidents provides technical, quality assurance, and performance requirements for nuclear accident dosimeters (NAD). In 2023 the U.S. Navy operated 82 nuclear-powered ships, with the fleet being composed of 11 aircraft carriers, 68 submarines, having a total number of 98 reactors. Since 1968 the U.S. Navy has used fixed nuclear accident dosimeters (FNAD) mounted to the bulkheads surrounding naval nuclear propulsion reactors. Since 1968 the US Navy has used two nuclear accident criticality dosimeters. The first Navy accident dosimeter DT-518/PD was introduced in 1968. It was developed by the Naval Radiological Defense Laboratory in San Francisco, California under leadership of Eugene Tochilin. This dosimeter contains two indium foils for quick dose assessments using shipboard gamma instruments available on nuclear powered vessels and two sulfur pellets/LiF TLD-700 powder for final dose determination at the Naval Dosimetry Center. The newest Navy NAD is the DT-723/PD, which contains indium foil, gold foil, cadmium shielded gold foil, sulfur pellet and a LiF TLD-700 chip. This paper provides a brief description of the measurement procedures, results of the testing of both NADs and comparison of their performance.

Startup and Multiphysics Analysis of a Compact kW-Class Thermal Spectrum Microreactor

Self-regulating compact nuclear microreactor concepts are being developed for use in space and at terrestrial remote sites. The current emphasis is on the development of high-assay low-enriched uranium–fueled reactors that rely on metal hydrides to achieve size, specific weight, and power parity with the legacy highly enriched uranium systems. In the case of terrestrial applications, metal hydride–moderated systems can also improve economic feasibility. The neutronic design of metal hydride–moderated high-temperature reactor cores is complicated by the need to optimize and synchronize delayed spatial and temporal reactivity feedback from outer reflectors. Zebra is a thermal spectrum proof-of-principle core designed for demonstrating the reactor dynamics of hydride-moderated spectrum reactors, which is studied in this work. The reactor, including the fuel, yttrium hydride moderator, heat pipes, central control rod, structural supports, and the beryllium reflector, weighs ~425 kg. The design shares several features common with nuclear criticality test geometries routinely used in National Criticality Experiments Research Center (NCERC) experiments such that a prototype of the reactor can be readily assembled and tested. Additional features, such as heat pipes and hydrogen barrier clads, were added to minimize the potential for large thermal gradients that in turn could induce hydrogen loss or structural deformation during extended periods of operation at ~1000 K. In this work, normal and off-normal state performance of the Zebra rector core was analyzed using MCNP and Abaqus-based Reactor Multiphysics (MARM) software. Up-to-date nuclear, thermal, and mechanical performance data were used to characterize the performance of the yttrium hydride moderator. Steady-state analyses established that at a postulated power between 20 and 50 kW(thermal), the reactor core is nearly isothermal irrespective of quality of conduction coupling between heat pipes and fuel plates. Additionally, heat pipe failure modes, including simultaneous failure of all heat pipes in a quadrant of a reactor, were examined to bound the maximum credible temperature spike in the reactor core for extreme off-normal operating conditions. Finally, we detail the startup design challenges for the hydride-moderated thermal core, and analyze load-following cases to achieve “self-regulation” using the Dynamic Analysis of Reactor Transients module of MARM. The reactor is self-regulating with a reactivity temperature coefficient of ~−1 pcm/K at the operating point based solely on nuclear cross-section feedback. It can be further strengthened using additional spectral shift neutron absorbers and incorporating design features that enhance core expansion. This work captures that hydride-moderated systems are feasible for various self-regulating applications once systematic checks are verified in order to achieve a well-engineered core design.

Optimal fissile distribution in multiplying systems: Illustrative examples with Monte Carlo simulation and Pontryagin’s maximum principle

In multiplying systems, such as nuclear reactors and criticality experiments, it is desirable to place the fissile material in the optimal or ‘best’ way to reduce the critical mass of the system as well as to achieve uniform fuel burnup. This paper considers two methods, namely Pontryagin’s maximum principle (PMP) and Monte Carlo (MC) perturbation for estimating a minimum critical mass configuration. These methods are applied to an elementary multizone model of a pressurized water reactor (PWR) and a criticality experiment to estimate the minimum critical mass. It is found that while two-group diffusion theory with PMP predicts a minimum critical mass, more detailed MC simulations with MCNP5 show a consistent reduction in critical mass when fissile fuel is placed in inner zones. Such a distribution reduces the fissile material requirement but is undesirable due to the higher power peaking. MC simulations show that for a three-zone model of the KORI 1 PWR, a uniform fissile distribution gives criticality for 1.09 atomic percent (at.%) enrichment, whereas non-uniform fissile distribution (0.6, 1.6, 0.6 at.%) reduces the critical mass by 14%. The changes found from MC simulations were subsequently predicted from first- and second-order derivative sampling. It was found that substantial computational savings can be achieved for large-scale optimization problems. In the case of a criticality experiment, MC derivative sampling was also used to estimate optimal fissile distribution for minimizing the critical mass.

Demonstrating a Quantitative and Systematic Approach to Reducing Excess Conservativism in Nuclear Criticality Safety Analyses

Although reducing conservatism would alleviate unnecessary constraints in processing, storage, transportation, and disposal of nuclear materials, excessively conservative approaches are still utilized in many safety analyses. Criticality safety limits are put in place to reduce the likelihood of having a nuclear criticality accident to a value that is deemed incredible but often utilize parameters that are conservative to the point of becoming incredible themselves. The analyses that determine criticality limits are supposed to be based on credible instead of incredible events and circumstance, highlighting the need to be able to distinguish between what is in the realm of possibility and what is not. This paper provides a quantitative approach for reducing unrealistically conservative parameters by recalculating limiting factors in a state that deviates from the worst-case scenario and assigning probability distributions to these systematic deviations. This provides a technical basis for replacing excessively conservative values with something that is both objective and reasonably bounding, which may be systematically utilized in any criticality safety analyses. The assumption of “perfect sphericity” in the TRUPACT-II package’s fissile contents model was used as an example case to demonstrate the proposed approach for replacing qualitative reductions to conservatism with quantitative reductions. Through a series of Monte Carlo calculations and statistical analyses, it was shown that conservative deviations from sphericity will provide lower keff values, where the magnitude and impact of this deviation is system specific. The statistical significance from applying probabilistic conservatism will be dependent on the chosen κ value and integration limit for the exponential distribution, as it varies the degree of conservatism applied to any parameter of interest. This approach is not limited to geometric assumptions and may be applied to a variety of conservative parameters. In an effort to move toward a standard method for reducing conservatism, this objective approach may be used in lieu of or in conjunction with subjective methods for relaxing constraints.

Mathematical and computational models for simulating transient nuclear criticality excursions within plutonium nitrate systems

This paper presents a novel methodology for the analysis of transient nuclear criticality in aqueous solutions of plutonium nitrate. The transient methodology includes one-dimensional thermal hydraulics and a detailed description of radiolytic gas formation (including aqueous concentrations of radiolysis species and bubble population, nucleation radius and advection velocity). The positive temperature feedback effect in dilute plutonium solutions, due to a low lying resonance (at ≈ 0.3 eV) in the microscopic fission cross section, is investigated. The solutions’ nuclear properties, such as intrinsic neutron source and mean neutron generation time; thermophysical properties and heat transfer coefficients; linear energy transfer and dynamics of radiolytic gas nucleation; and subcritical power, are also investigated. The effect of step and ramp reactivity insertions on transient nuclear criticality behaviour is examined, alongside the effect of fuel solution ingress and egress from the system. For a solution containing 16.0 g L−1 plutonium with an acidity of 0.2 mol L−1, increasing the width of the prescribed step reactivity insertion from 2.25 to 2.50 s increased the magnitude of the initial power peak, and decreased the time to boil, by an order of magnitude. It was observed that the characteristics of the radiolytic gas were dependent on the magnitude of the initial power peak. For example, the transients with the highest initial power peak produced an initially large quantity of radiolytic gas, with defined oscillations in their quantity (and other properties) as the transient progressed. For transients with lower power peaks, the oscillations became less defined, with a reduced period and magnitude, due to the lower quantity of radiolytic gas produced due to the initial power peak. Many of the presented simulations of nuclear criticality transients were terminated when the solution started to boil, a phenomenon which is not included in the presented methodology. Boiling occurred when the initial reactivity input induced a power spike that raised the solution temperature enough for the positive reactivity feedback effect to sustain the transient even when the reactivity input was removed.

Innovative dead-time correction and background subtraction for neutron multiplicity measurements using neural networks

The number of neutrons emitted from a nuclear reaction plays a crucial role in various fields, including nuclear theory, nuclear nonproliferation, nuclear energy and nuclear criticality safety. Accurate determination of neutron multiplicities requires the application of several corrections, with dead-time correction and background subtraction being particularly significant. These corrections become more challenging for neutron detectors with time-dependent neutron capture. In this work, we perform a comprehensive study of three existing methods used for dead-time correction and background subtraction in neutron detectors with time-dependent neutron capture. The methods were tested for dead-times in the range from 0 to 1 μs using a Monte Carlo model simulating the dead-time and background effects in the standard neutron multiplicity probability distribution of 252\\documentclass[12pt]{minimal} \\usepackage{amsmath} \\usepackage{wasysym} \\usepackage{amsfonts} \\usepackage{amssymb} \\usepackage{amsbsy} \\usepackage{mathrsfs} \\usepackage{upgreek} \\setlength{\\oddsidemargin}{-69pt} \\begin{document}$$^{252}$$\\end{document}Cf. The previous methods showed larger than desired uncertainty or systematic trade off. Those uncertainties prompted the development of a novel approach using neural networks trained with data from Monte Carlo simulations. The Neural Network method enabled the correction of neutron multiplicity probabilities more accurately than the other methods with fractional errors smaller than 3% for multiplicities around the peak of 252\\documentclass[12pt]{minimal} \\usepackage{amsmath} \\usepackage{wasysym} \\usepackage{amsfonts} \\usepackage{amssymb} \\usepackage{amsbsy} \\usepackage{mathrsfs} \\usepackage{upgreek} \\setlength{\\oddsidemargin}{-69pt} \\begin{document}$$^{252}$$\\end{document}Cf. A similar approach using neural networks could be applied to problems where the system being studied can be accurately simulated without having an accurate analytical description available. The neural network method presented in this paper can be easily expanded if multiplicities greater than 10 are expected.

Steady-state analysis of the 1970 Windscale nuclear criticality incident

This paper presents several steady-state neutron transport models which are used to determine the behaviour of the nuclear criticality incident that occurred at the Windscale Works in 1970. The transfer vessel involved in this incident contained two immiscible fissile liquids (an aqueous phase and an organic phase) which had been disturbed such that an emulsion layer formed between them. Infinite domains of the aqueous, organic, and emulsion phases were modelled to enable the understanding of the effects of nuclear data library, neutron transport code and material on the calculated kinetics data, effective multiplication factor and macroscopic neutron cross sections. Six possible configurations of the 1970 Windscale criticality incident were simulated using three-dimensional models in MCNP, two-dimensional models in academic research code EVENT, and one-dimensional models in a collision probability (CP) code. Quantities of interest, such as reactivity, scalar neutron flux, neutron absorption and production rates, and kinetics data are presented. System reactivity, mean neutron generation time and neutron absorption and production rates are shown to increase with increasing emulsion thickness, whereas the total scalar neutron flux is shown to decrease. It is determined that the volume of organic phase present during the transient was likely to be around 39.0 L, as the emulsion thickness required to reach criticality (6.5 cm) was within the range cited in the literature. EVENT and the CP code showed deficiencies when calculating the scalar neutron flux through the plenum gas at the top of the transfer vessel, with the CP code overestimating the total scalar neutron flux in the system by as much as 19 %. The intrinsic neutron source, subcritical multiplication and subcritical power of the system were computed using Monte Carlo simulations. It was shown, by use of the Hansen criteria, that the subcritical system was likely to be in a strong neutron source regime (such that stochastic variations in neutron population are negligible when regarding nuclear criticality transients). In addition, it was computed that a vessel containing 39.0 L of organic solvent had a subcritical power of 10.81 ±0.03μW.

Exploration of genetic algorithms to build a balanced neutron spectra dataset useful to train unfolding techniques based on artificial neural networks

Diverse domains need neutrons unfolding technics to assess the incident neutron energy spectrum. Examples are radiation protection, nuclear reactor physics or criticality safety. Traditionally, methods based on the Bayesian approach requires an initial guess of the solution which may significantly impact the unfolding result. This work proposes a novel method for neutron spectrum reconstruction using machine learning (ML) techniques trained on a large dataset. To ensure the ML algorithm to perform on a large domain of application particular attention has been paid to the dataset creation. We propose a comparison of two methods of building large dataset where the most adequate solution is obtained using a dynamic genetic algorithm (GA). This GA targets optimal combinations of 48 parameters to generate a variety of neutron spectra. The resulting dataset is then used to train a new convolutional neural network architecture for unfolding neutron spectra. Obtained performance metrics of the tested architecture show high efficiency and emphasize the added value of the built dataset.

An introduction to Spent Nuclear Fuel decay heat for Light Water Reactors: a review from the NEA WPNCS

This paper summarized the efforts performed to understand decay heat estimation from existing spent nuclear fuel (SNF), under the auspices of the Working Party on Nuclear Criticality Safety (WPNCS) of the OECD Nuclear Energy Agency. Needs for precise estimations are related to safety, cost, and optimization of SNF handling, storage, and repository. The physical origins of decay heat (a more correct denomination would be decay power) are then introduced, to identify its main contributors (fission products and actinides) and time-dependent evolution. Due to limited absolute prediction capabilities, experimental information is crucial; measurement facilities and methods are then presented, highlighting both their relevance and our need for maintaining the unique current full-scale facility and developing new ones. The third part of this report is dedicated to the computational aspect of the decay heat estimation: calculation methods, codes, and validation. Different approaches and implementations currently exist for these three aspects, directly impacting our capabilities to predict decay heat and to inform decision-makers. Finally, recommendations from the expert community are proposed, potentially guiding future experimental and computational developments. One of the most important outcomes of this work is the consensus among participants on the need to reduce biases and uncertainties for the estimated SNF decay heat. If it is agreed that uncertainties (being one standard deviation) are on average small (less than a few percent), they still substantially impact various applications when one needs to consider up to three standard deviations, thus covering more than 95% of cases. The second main finding is the need of new decay heat measurements and validation for cases corresponding to more modern fuel characteristics: higher initial enrichment, higher average burnup, as well as shorter and longer cooling time. Similar needs exist for fuel types without public experimental data, such as MOX, VVER, or CANDU fuels. A third outcome is related to SNF assemblies for which no direct validation can be performed, representing the vast majority of cases (due to the large number of SNF assemblies currently stored, or too short or too long cooling periods of interest). A few solutions are possible, depending on the application. For the final repository, systematic measurements of quantities related to decay heat can be performed, such as neutron or gamma emission. This would provide indications of the SNF decay heat at the time of encapsulation. For other applications (short- or long-term cooling), the community would benefit from applying consistent and accepted recommendations on calculation methods, for both decay heat and uncertainties. This would improve the understanding of the results and make comparisons easier.

Godiva-IV dosimetry exercise 2022 preliminary results

Integral Experiment Request (IER) 538 is part of a series of dose characterization and nuclear accident dosimetry (NAD) exercises performed under the Department of Energy (DOE) Nuclear Criticality Safety Program (NCSP). This is the second NAD exercise using the Godiva-IV critical assembly and the third NAD exercise overall. The participating laboratories provided their own dosimeters that were mounted on the Lawrence Livermore National Laboratory (LLNL) BOttle Manikin ABsorption (BOMAB) phantoms and aluminum plates. The BOMABs and plates were placed at two, three, and four meters away from the center of Godiva. Alongside the NADs, there was a LLNL Passive Neutron Spectrometer (PNS), Atomic Weapons Establishment (AWE) PNS, and Y-12 Sphere present to measure the neutron dose from Godiva-IV. Two irradiations were conducted to test the NAD performance from each laboratory and assesses their performance to the DOE-STD-1098-2017 part 515 criteria. Neutron and gamma doses were measured prior to this exercise. This work presents a model for the neutron and gamma dose respectively to serve as the reference value. A code written in C/C++/ROOT was used to fit the measured neutron and gamma dose with the new models. It was assumed that the neutron and gamma doses are proportional to the change in temperature of Godiva after a burst irradiation. Uncertainties for the reference values were calculated using error propagation of the model’s parameters. Preliminary results (within twenty-four hours) and final results were compared for each laboratory. On average of all the participating laboratories, 32% of neutron doses and 78% of gamma doses were outside the DOE standards. One laboratory did not report their dose readings and were not included in this average. There is a bias for a lower neutron dose and a higher gamma dose based on the distribution of results. In comparison with the past Godiva-IV NAD exercise, there is an improvement in neutron dose readings by 20%.

Utilization of ACE nuclear data file toolkit ACEtk to calculate relative sensitivity coefficients of point-kinetics parameters

Sensitivity and uncertainty methods are quintessential for nuclear criticality safety and experiment design. This type of analysis relies on calculations of sensitivity coefficients; sensitivity coefficients of the effective neutron multiplication factor with respect to some nuclear data are predominantly calculated and used. As a part of the Laboratory Directed Research & Development project EUCLID (Experiments Underpinned by Computational Learning for Improvements in nuclear Data) at Los Alamos National Laboratory, sensitivity coefficients of many radiation detector measurement responses with respect to nuclear data were investigated. Specifically, this paper outlines a method to calculate point-kinetics parameters relative sensitivity coefficients with respect to nuclear data. Point-kinetics parameters such as the prompt neutron decay constant, effective delayed neutron fraction, and neutron generation time are especially important to experimenters and reactor operators designing systems with dynamic neutron populations. This method couples capabilities of the ACE (A Compact ENDF) nuclear data file toolkit, ACEtk, with the ability to load cross sections into the radiation transport code Monte Carlo N-Particle (MCNP). Key aspects of optimizing this method for a particular application and sensitivity profiles of the Jezebel criticality experiment are examined and discussed.

Comparative Analysis of Standard and Advanced USL Methodologies for Nuclear Criticality Safety

The American National Standards Institute/American Nuclear Society national standards 8.1 and 8.24 provide guidance on the requirements and recommendations for establishing confidence in the results of the computerized models used to support operation with fissionable materials. By design, the guidance is not prescriptive, leaving freedom to the analysts to determine how the various sources of uncertainties are to be statistically aggregated. Due to the involved use of statistics entangled with heuristic recipes, the resulting safety margins are often difficult to interpret. Also, these technical margins are augmented by additional administrative margins, which are required to ensure compliance with safety standards or regulations, eliminating the incentive to understand their differences. With the new resurgent wave of advanced nuclear systems, e.g., advanced reactors, fuel cycles, and fuel concepts, focused on economizing operation, there is a strong need to develop a clear understanding of the uncertainties and their consolidation methods to reduce them in manners that can be scientifically defended. In response, the current studies compare the analyses behind four notable methodologies for upper subcriticality limit estimation that have been documented in the nuclear criticality safety literature: the parametric, nonparametric, Whisper, and TSURFER methodologies. Specifically, the work offers a deep dive into the various assumptions of the noted methodologies, their adequacies, and their limitations to provide guidance on developing confidence for the emergent nuclear systems that are expected to be challenged by the scarcity of experimental data. To limit the scope, the current work focuses on the application of these methodologies to criticality safety experiments, where the goal is to calculate a bias, a bias uncertainty, and a tolerance limit for keff in support of determining an upper subcriticality limit for nuclear criticality safety.

Current Status of TRIPOLI-4® Monte Carlo Radiation Transport Code on Adult and Pediatric Computational Phantoms for Radiation Dosimetry Study

TRIPOLI-4® is a general-purpose Monte Carlo radiation transport code developed by the Service d’Études des Réacteurs et de Mathématiques Appliquées at CEA-Saclay. It uses continuous-energy nuclear data to simulate neutron, photon, electron, and positron transport in fields like radiation shielding, reactor physics, and nuclear criticality safety. To study radiation protection dosimetry in human tissues and organs, male and female adult computational phantoms from the Medical Internal Radiation Dose–Oak Ridge National Laboratory phantoms family and the International Commission on Radiological Protection (ICRP) publication 110 were recently modeled and calculated using the geometry options of TRIPOLI-4. To easily use the ICRP 110 voxel-based phantoms in different exposure scenarios, a newly developed phantom option is available in TRIPOLI-4 and its display tool T4G. This new phantom option is helpful for modeling one or more phantoms and for improving calculation performance in real irradiation environments. The 2020 published pediatric computational reference phantoms are accessible from ICRP publication 143. Male and female pediatric phantoms are also verified with the new T4G tool and TRIPOLI-4 code. This paper reports on recent works using TRIPOLI-4 on adult and pediatric computational phantoms. The modeling methods of stylized and voxel-based phantoms, the graphic displays of modeled phantoms with T4G, and the verification procedures for single-phantom and two-phantom application cases are presented. Validation for external and internal dosimetry calculations has been performed. Calculation results on organ dose S values for nuclear medicine applications are presented for single female and single male voxel phantoms using 131I and 177Lu radiation sources. Effective dose calculations for two-phantom cases using 99mTc and 18F sources are compared with traditional H*(10) calculations from nuclear medicine patient to patient caregiver.