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All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Godiva IV Thermal Neutron Dosimetry Modeling and Variance Reduction

The transfer of the Godiva IV experiment from the Los Alamos Critical Experiments Facility (LACEF) to the National Critical Experiments Research Center (NCERC) introduced a vastly different experiment room return to the neutron flux. The contribution of the background to the burst neutron energy spectrum is significant in the thermal and epithermal neutron energies. Target materials may be placed in various locations in the Godiva room, or outside of the room, for thermal neutron activation. Modeling of this dosimetry problem in Monte Carlo N-Particle (MCNP) presented a novel challenge compared to previous Godiva IV glory hole irradiation simulations. An advanced dosimetry modeling framework for high efficiency calculations in locations far from the Godiva IV fission source was desired. The mesh-based weight windows and point detector advanced variance reduction techniques in MCNP were implemented and tested using adaptations of the critical experiment benchmark model of the Godiva IV problem. The models were validated against measured activations of Nickel, Indium, Scandium, and Cobalt foils at locations 2 meters from the Godiva IV core. Dosimetry measurements were performed in collaboration with Sandia National Laboratory. The weight windows and point detector variance reduction coupled method resulted in the highest problem efficiency.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2012

2012 SARA Students Technical Report

The Service Academy Research Associates (SARA) program provides an opportunity for Midshipmen and Cadets from US Service Academies to participate in research at Los Alamos National Laboratory (LANL), Lawrence Livermore National Laboratory (LLNL), and Sandia National Laboratory for several weeks during the summer as part of their summer training assignments. During the summer of 2012, three Midshipmen were assigned to work with the XCP Division at LANL for approximately 5-6 weeks. As one of the nation's top national security science laboratories, LANL stretches across 36 square miles, has over 2,100 facilities, and employs over 9,000 individuals including a significant number of students and postdocs. LANL's mission is to 'apply science and technology to: ensure the safety, security, and reliability of the US nuclear deterrent, reduce global threats, and solve other emerging national security challenges.' While LANL officially operates under the US Department of Energy (DoE), fulfilling this mission requires mutual cooperation with the US Department of Defense (DoD) as well. LANL's high concentration of knowledge and experience provides interns a chance to perform research in many disciplines, and its connection with the DoD in both operation and personnel gives SARA students insight to career possibilities both during and after military service. SARA students have plenty of opportunity to enjoy hiking, camping, the Los Alamos YMCA, and many other outdoor activities in New Mexico while staying at the Buffalo Thunder Resort, located 20 miles east of the lab. XCP Division is the Computational Physics division of LANL's Weapons Department. Working with XCP Division requires individuals to be Q cleared by the DoE. This means it is significantly more convenient for SARA students to be assigned to XCP Division than their civilian counterparts as the DoD CNWDI clearance held by SARA students is easily transferred to the lab prior to the students arriving at the start of the summer. SARA students working with XCP Division were given a comprehensive introduction into nuclear engineering and physics, nuclear weapons, and radiation transport and detection via texts and lectures at various classification levels. Students also attended tours of several prominent facilities at LANL including TA-41 Ice House, TA-55 PF-4 plutonium facility, the Nicholas C. Metropolis Center for Modeling and Simulation, also known as the Secure Computing Center (SCC), and the Dual-Axis Radiological Hydro Test (DARHT) facility; in addition, SARA students accompanied by LANL staff traveled to Minot AFB in North Dakota for tours of the 5th Bomb Wing and 91st Missile Wing facilities. Students participated in a week long class on the Monte Carlo N Particle (MCNP) code to supplement their understanding of radiation transport simulations. SARA students were then tasked with using this knowledge to model radiation detectors and use MCNP to compare their models to experimental data and previously accepted models.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗
DOE OSTI2022

Computational Modeling and Simulation for Nonproliferation: The History (and Present) of Monte Carlo and the MCNP(R) Code at Los Alamos [Slides]

The emergence of the Monte Carlo method as a research tool springs from work done at Los Alamos in the 1940s.Monte Carlo and the MCNP Code have been and continue to be developed at Los Alamos for many decades. From basic science in support of understanding nuclear interaction physics to global security applications, application uses of the code are extensive. Recent R&D projects and code modernization efforts make the MCNP code a great tool for nuclear nonproliferation applications. In collaboration with nuclear safeguards experts, new training has just recently been developed to help new practitioners learn how to use the code for nuclear safeguards applications.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2017

The Nature of Scatter at the DARHT Facility and Suggestions for Improved Modeling of DARHT Facility

The U.S. Stockpile Stewardship Program [1] is designed to sustain and evaluate the nuclear weapons stockpile while foregoing underground nuclear tests. The maintenance of a smaller, aging U.S. nuclear weapons stockpile without underground testing requires complex computer calculations [14]. These calculations in turn need to be verified and benchmarked [14]. A wide range of research facilities have been used to test and evaluate nuclear weapons while respecting the Comprehensive Nuclear Test-Ban Treaty (CTBT) [2]. Some of these facilities include the National Ignition Facility (NIF) at Lawrence Livermore National Laboratory, the Z machine at Sandia National Laboratories, and the Dual Axis Radiographic Hydrodynamic Test (DARHT) facility at Los Alamos National Laboratory. This research will focus largely on DARHT (although some information from Cygnus and the Los Alamos Microtron may be used in this research) by modeling it and comparing to experimental data. DARHT is an electron accelerator that employs high-energy flash x-ray sources for imaging hydro-tests. This research proposes to address some of the issues crucial to understanding DARHT Axis II and the analysis of the radiographic images produced. Primarily, the nature of scatter at DARHT will be modeled and verified with experimental data. It will then be shown that certain design decisions can be made to optimize the scatter field for hydrotest experiments. Spectral effects will be briefly explored to determine if there is any considerable effect on the density reconstruction caused by changes in the energy spectrum caused by target changes. Finally, a generalized scatter model will be made using results from MCNP that can be convolved with the direct transmission of an object to simulate the scatter of that object at the detector plane. The region in which with this scatter model is appropriate will be explored.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗
DOE OSTI2017

A Pulsed Sphere Tutorial

Here I attempt to explain what physically happens when we pulse an object with neutrons, specifically what we expect the time dependent behavior of the neutron population to look like. Emphasis is on the time dependent emission of both prompt and delayed neutrons. I also describe how the TART Monte Carlo transport code models this situation; see the appendix for a complete description of the model used by TART. I will also show that, as we expect, MCNP and MERCURY, produce similar results using the same delayed neutron model (again, see the appendix).

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Cross section of neutrons from the H 2 ( n , 2 n ) reaction at E n = 15 MeV

In this work, the double-differential cross section of the deuteron breakup reaction 2 H(n, 2n) has been studied experimentally with a neutron beam energy of 15 MeV. Special attention has been devoted to estimation of background condition and multiple scattering effect in the scattering sample. Experimental data have been compared with models based on phase-space approximation used in the ENDF/B-VIII.0 data library and in the MCNP neutron transport code, as well as with rigorous model based on Faddeev equations used for cross section evaluations in JENDL data library. It was found that experimental data are better reproduced by Faddeev model, however, the model overestimates data in the low-energy region of the neutron spectrum (<4 MeV)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Evaluation of Erosion of the Dummy “EE” Plate 19 in YA Type ATR Fuel Element During Reactor PALM Cycles

Advanced Test Reactor (ATR) Cycle 153B-1 was a 14-day, high-power, powered axial locator mechanism (PALM) operating cycle that completed on April 12, 2013. Cycle 153B-1 was a typical operating cycle for the ATR, and did not result in any unusual plant transients. ATR was started up and shut down as scheduled. The PALM drive physically moves the selected experiments into and out of the core to simulate reactor startup and heat up, and shutdown and cooldown transients, while the reactor remains in steady-state conditions. However, after the cycle was over, when the fuel elements were removed from the core and inspected, several thousand flow-assisted erosion pits and “horseshoeing” defects were readily observed on the surface of the several YA-type fuel elements (these are aluminum “dummy” plates that contain no fuel). In order to understand these erosion phenomena, a thermal-hydraulic model of coolant channel 20 on a YA-M fuel element was generated. The boundaries of the model were the aluminum EE plate of a YA-M fuel element and a beryllium reflector block with 13 horizontal saw cuts which represented regions of zero flow. The heat generated in fuel plates 1 through 18 was modeled to be passing through the aluminum EE plate. The coolant channel 20 width was set at 0.058 in. (58 mils). It was established that the horizontal saw cuts had a significant effect on the temperature of the coolant. The flow, which was expected to vary linearly with gradual heating of the coolant as it passed through the channel, was extremely turbulent. The temperature rise, which was expected to be a smooth “S” curve, was represented by a series temperature rise “humps,” which occurred at each horizontal saw cut in the beryllium reflector block. Each of the 13 saw cuts had a chamfered edge which resulted in the coolant flow being re-directed as a jet across the coolant channel into the surface of the EE plate, which explained the temperature rise and the observed scalloping and pitting degradation on the YA-M fuel elements. In the case of scalloping (horseshoeing) a surprising similarity of that defect to those appearing on aluminum plate rolled in over-lubrication conditions, were established. In turn, this made us think that the principal feature responsible for the appearance of these defects, was horizontal cuts in the beryllium reflector block created to arrest the propagation of large vertical crack(s) in Be in PALM cycles with higher overall fluence. This assumption was fully confirmed by the results of thermo-hydraulic simulations. The neutronics data for these modeling experiments were provided using advanced irradiation simulations (MCNP, HELIOS). In the case of pitting erosion the following corrective measures were proposed based upon the results of JMatPro v.8.2 modeling (TTT- and CCT-diagrams): change the fabrication process by adding blister anneal before program anneal, immediately after cold rolling of AA6061plate. This step will allow achieving complete recrystallization, eliminating of strengthening due to metastable precipitates, and reduce the possibility of forming sharp microstructural features upon the surface.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Neutron-Neutron Correlations in the Photofission of 238U

Background: In the fission of actinides, the nearly back-to-back motion of the fission fragments has a strong effect on the kinematics of fission neutrons. This leads to a favoring of opening angles near 0? and 180? in the neutron-neutron (n-n) opening angle distributions of correlated neutron pairs from the same fission event. Purpose: To measure the n-n opening angle and energy correlations in the photofission of 238U. As of this writing, measurements of correlated n-n opening angle distributions have been reported only for the spontaneous and neutron-induced fission of actinides. This work is the first to report such a measurement using photofission and will provide useful experimental input for photofission models used in codes such as MCNP and FREYA. Method: Fission is induced using bremsstrahlung photons produced via a low duty factor, pulsed, linear electron accelerator. The bremsstrahlung photon beam impinges upon a 238U target that is surrounded by a large neutron scintillation detection system capable of measuring particle position and time of flight, from which n-n opening angle and energy are measured. Neutron-neutron angular correlations are determined by taking the ratio between a correlated neutron distribution and an uncorrelated neutron distribution formed by the pairing of neutrons produced during different beam pulses. This analysis technique greatly diminishes effects due to detector efficiencies, acceptance, and experimental drifts. Results: The angular correlation of neutrons from the photofission of 238U shows a high dependence on neutron energy as well as a dependence on the angle of the emitted neutrons with respect to the incoming photon beam. Angular correlations were also measured using neutrons from the spontaneous fission of 252Cf, showing good agreement with past measurements. Conclusions: The measured angular correlations reflect the underlying back-to-back nature of the fission fragments. An anomalous decline in n-n yield was observed for opening angles near 180? for 238U.

Burggraf, Jeff↗
DOE OSTI2019

Critical Experiment Benchmark Results using MCNP6.2 Unstructured Mesh

The purpose of this report is to document the results of five critical experiment benchmarks modeled using the unstructured mesh (UM) geometry capability in MCNP®1 code, version 6.2. LANL currently has a vast critical experiment benchmark library with over 1100 cases, modeled using constructive solid geometry (CSG) and packaged with Whisper-1.1. Validation efforts for the purposes of nuclear criticality safety typically focus on benchmark libraries built using CSG models. Results of UM geometry models are compared with CSG results and the critical experiment k-effective from the ICSBEP handbook. All computed k-effective results in this study are well within 1% of the experimental result.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Nuclear Data: What, Why How, Who? [Slides]

Nuclear data is data that describes physics in forms and formats that computer code can read to perform simulations of nuclear processes. The data is put into ”evaluations” following the combination of experiment and theory. Various physics are applied to these evaluations by the NJOY nuclear data processing code to produce application files, which simulation codes (e.g., MCNP, Partisn, etc.) can use to model various scenarios. The Nuclear Data Team at Los Alamos National Laboratory—in partnership with a variety of national and international organizations—provides nuclear data for use at LANL and throughout the world. This data is verified and validated to ensure that it performs as expected and accurately represents Mother Nature.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Alternative Source Verification for the NB5 Beamline

This report details the calculation and comparison of secondary neutron sources for the purpose of performing NB5 velocity selector shielding calculations without relying on the primary HB4 source and the use of a modified MCNP code to run the full model.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2018

SISTec: Mathematical Calibration of Large Clearance Monitors - 18265

One major path for clearance of large amount of waste arising from the decommission of nuclear power plants is based on measurement with large clearance monitors (also known as free release measurement facilities). These devices utilize multi-detector plastic scintillator arrays (large clearance monitors can consist of up to 24 plastic scintillators) to offer both, extremely low minimum detectable activities (MDA) and short measurement times for waste packages weighing up to one metric ton. However, a major problem with these devices is to obtain nuclide specific efficiencies, especially for geometrically complex large waste packages. Normal procedure for estimating the efficiency is to device dummy waste packages and attach point sources with well-known activities to unfavourable positions in these packages. In accordance with this procedure, calibration factors can be derived in a radiological conservative way. Devising such test dummies is not always possible and the procedure is prone to generating large measurement uncertainties. In this publication, we propose a new method to derive calibration factors by means of the Los Alamos National Laboratory (LANL) simulation tool Monte Carlo N-Particle (MCNP) and a statistical-phenomenological model for the detector characterization. In a multi-step approach, first the detector response for a distribution of activity in a waste package is calculated and subsequently the detectors are mathematically characterized. Applying state of the art statistical and optimizing procedures, parameters for the detector response are derived in a self-consistent way. This also allows to rigorously propagate the uncertainties to the final calculated calibration factors according to the Guide to the Expression of Uncertainty in Measurement (GUM supplement 1). To validate the derived parameters for the detector characterization a minimally invasive experimental setup was devised and installed into the large clearance monitor. It consisted of an aluminum swivel arm which could position point sources to the relevant coordinates. The so collected measurement data not only allowed to investigate the precision of the calculated calibration factors, but also to probe the location of concentrated activity sources in the waste package. Furthermore, data on measurements with dummy waste packages were collected to test the reproducibility of well-known calibration factors. We present the software tool SIStec, which will provide an intuitive user-interface to generate and simulate waste package models and execute the above-mentioned method. Using SIStec, different scenarios for the distribution of activity in the waste package can be probed. This will also allow to optimize the clearance process on various levels. For example, measurement time for pre-measurements can be reduced in exchange for unfavourable calibration factors, which would still allow to meet the regulator limits for clearance. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Computational Modeling, Simulation and Analysis of the RPI Walthousen Reactor Critical Facility (RCF)

The Walthousen Reactor Critical Facility (RCF), located in Schenectady, New York, is currently owned and operated by Rensselaer Polytechnic Institute (RPI). The facility was originally constructed by the American Locomotive Company (ALCO) for the research and development of small, portable reactors. Originally designed for highly-enriched uranium fuel, the prototype reactor first came online in August 26, 1956, and was transferred to RPI ownership in 1963. Several changes to the reactor's design and operation have been made, the most notable of which was the conversion from high-enriched uranium to low-enriched uranium fuel. The RCF is an open-pool critical assembly that operates at a low power with a license limit of less than 100 watts. This low operating power permits a much greater flexibility in configuration and operation as compared to other research reactors. Because of this, experiments may be designed that would be unfeasible for research reactors that operate at much higher powers. The RCF's functional capabilities provide opportunities to perform several reactor experiments and measurements, such as: core criticality experiments, differential and integral rod worth measurements, moderator temperature coefficient of reactivity calculations, critical rod height and critical water height measurements, etc.. These experiments are regularly performed in conjunction with RPI course-work education and Senior Reactor Operator (SRO) training. The unique capabilities and experiments that may be performed in the RCF are particularly useful for validating certain aspects of computational analysis codes. A computational model of the RCF reactor has been previously generated and validated using the Monte Carlo N-Particle (MCNP) code. For the purpose of future experiments, this model is also being adapted for use with the SHARP Toolset, which includes the PROTEUS neutronics code, the Nek5000 fluid dynamics code, and the Diablo structural mechanics code. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Methodological Aspects of Shielding and Activation Analyses for ITER Diagnostics Systems - Paper 125

It is expected that the diagnostics systems of the ITER tokamak will provide valuable measurement data about the first in its kind D-T fusion plasma with emission of 14 MeV neutrons. These systems must withstand the neutron and secondary photon radiation along the ITER operation. Under the neutron irradiation the ITER materials are getting radioactive and therefore an adequate shielding must be provided along all the time of ITER operation and shut-down for the reliable machine operation and for the safe personnel access to its components, such as the diagnostics ports. These are the prerequisites for the shielding and activation analysis of the ITER diagnostics systems discussed in this paper. This paper provides insight into the current tasks of neutronics support for the development of ITER diagnostics systems. The tasks themselves range from relatively straightforward ones such as Monte Carlo radiation transport calculations for shielding analyses of the diagnostics systems, to more advanced tasks of Shut-Down Dose Rate (SDDR) analyses for which the dedicated Rigorous 2-Step mesh-based (R2Smesh) and Direct 1-Step (D1S) methodologies were applied. The accuracy and precision of the ITER neutronics analysis is dependent on the methodological approaches and the implying modeling assumptions. In addition to the SDDR challenges, another distinguishing feature of fusion neutronics is a wide use of CAD-based neutronics modeling to design components of fusion devices such as ITER and DEMO. Basically, there are two possibilities of CAD geometry settings in Monte Carlo (MC) radiation transport models, either by the geometry format conversion of CAD models into the format of MC models by means of the interface programs such as McCad and MCAM or by the direct use of CAD geometry engine in MC calculations such as it was realized in the DAGMC code approach. Several conversion interfaces and the DAGMC software library have been tested on the 3D ITER model developed for the CAD-MCNP benchmark. It was concluded that the considered approaches have reached the maturity level to allow their application to real ITER design analyses.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2022

Experimental validation of a high fidelity Monte Carlo neutron transport model of the MIT graphite exponential pile

High-fidelity modeling and simulation were performed for the MIT graphite exponential pile (MGEP) using Monte Carlo neutron transport codes OpenMC and MCNP, and the results were validated by experimental data. The MGEP is being used as the test bed for the design of an autonomous control system for the pile's neutron flux distribution. The main contribution of this work is to generate the training data sets of neutron flux distributions with different locations of control rods that perturb the neutron flux profiles. First, code -to-code cross verification between OpenMC and MCNP was performed to ensure consistency of the numerical modeling within statistical uncertainties. To validate the accuracy of this high-fidelity model, a series of neutron flux measurements were conducted using a Helium-3 (He-3) neutron detector on a mobile platform that is placed inside the pile. Second, the neutron flux profiles were measured in four vertical layers of interest, and compared to the corresponding simulation results. The comparison results shows that the root mean square error is less than 2.5% in the two upper layers, and less than 4.5% in all four measured layers. Here the results validated the accuracy of the modeling and simulation. Finally, the relative change of the neutron flux profiles from moving control rods was analyzed, which identified the layer that has the best sensitivity regarding the control rods movements. Thus, this work identified and provided training data sets of both simulated and experimental neutron flux profiles in the most sensitive layer, paving the path forward to the real-time experimental demonstration of the autonomous control system.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

A Transient Reactor Physics Experiment with High-Fidelity, 3-D Flux Measurements for Validation and Verification

This project led to the design and construction of a system of novel micro-pocket fission detectors (MPFDs) and their deployment as part of a series of steady-state and transient experiments at the University of Wisconsin Nuclear Reactor (UWNR). The data acquired during these experiments is intended for use in validation of reactor-physics models, including those developed as part of the NEAMS Toolkit (e.g., PROTEUS) and others. Part of the work led to a completely new computational model of UWNR based on Python-automated input generation for both MCNP and Serpent, and preliminary efforts have been made to develop a mesh for models in PROTEUS and other deterministic tools.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Characterization of Uranium Foil Irradiations at the WSU TRIGA Reactor Using a New Reactor Model in Scale

A new reactor model of the Washington State University TRIGA was developed in the SCALE neutron transport code, and its fidelity was verified by comparison to MCNP and available data for several reactor parameters. The model was used to characterize irradiations designed to produce the short-lived actinides 237 U and 239 U, two key isotopes for nuclear forensics. These short-lived actinides, their decay daughters 237Np and 239Pu, and total fissions (via 99 Mo) were measured in irradiated foils at Los Alamos National Laboratory and other labs with good agreement among parent/daughter pairs and among labs. The laboratory-measured isotope ratios were used as a benchmark for the model determination of reaction rates. The flux distribution in the foils was also determined. Finally, using the continuous energy TSUNAMI-3D module of SCALE, a sensitivity/uncertainty analysis was performed to determine the effects on foil reactions caused by other system-wide reactions and by uncertainties in the known cross-sections.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗