Sutton T.M.,Knolls Atomic Power Laboratory
Nuclear Science and Engineering | Year: 2017
In the study of Monte Carlo statistical uncertainties for iterated-fission-source calculations, an important distinction is made between the real and the apparent variances. The former is the actual variance of a Monte Carlo calculation result, while the latter is an estimate of the former obtained using the results of the fission generations in the formula for uncorrelated random variates. For years it has been known that the apparent variance is a biased estimate of the real variance, and the reason for the bias has been understood. More recently, several authors have noted various interesting phenomena regarding the apparent and the real variances and the relationships among them. Some of these are an increase in the apparent variance near surfaces with reflecting boundary conditions, a nonuniform spatial distribution of the ratio of the apparent-to-real variance, the dependence of this ratio on the size of the region over which the result is tallied, and a rate of convergence of the real variance that is less than the inverse of the number of neutron histories run. This paper discusses a theoretical description of the Monte Carlo process using a discretized phase-space and then uses it to explain the causes of these phenomena. © American Nuclear Society.
Ramsey S.,Fred Hutchinson Cancer Research Center |
Blough D.,Knolls Atomic Power Laboratory |
Kirchhoff A.,University of Utah |
Kreizenbeck K.,Hutchinson Institute for Cancer Outcomes Research |
And 5 more authors.
Health Affairs | Year: 2013
Much has been written about the relationship between high medical expenses and the likelihood of filing for bankruptcy, but the relationship between receiving a cancer diagnosis and filing for bankruptcy is less well understood. We estimated the incidence and relative risk of bankruptcy for people age twenty-one or older diagnosed with cancer compared to people the same age without cancer by conducting a retrospective cohort analysis that used a variety of medical, personal, legal, and bankruptcy sources covering the Western District of Washington State in US Bankruptcy Court for the period 1995-2009. We found that cancer patients were 2.65 times more likely to go bankrupt than people without cancer. Younger cancer patients had 2-5 times higher rates of bankruptcy than cancer patients age sixty-five or older, which indicates that Medicare and Social Security may mitigate bankruptcy risk for the older group. The findings suggest that employers and governments may have a policy role to play in creating programs and incentives that could help people cover expenses in the first year following a cancer diagnosis. © 2013 Project HOPE-The People-to-People Health Foundation, Inc.
News Article | October 23, 2015
After tinkering for months with thousands of lines of computer code, Benjamin L. Magolan believes he is finally getting somewhere: “I’ve had a breakthrough with my implementation and now everything is coming together,” he says. “I’m capturing the appropriate turbulence trends in my model.” A master’s student in the Department of Nuclear Science and Engineering (NSE) with a focus on computational fluid dynamics (CFD), Magolan has been devising and testing algorithms to help model the flow of coolant water inside the core of a nuclear reactor. This work, which offers insights into many reactor thermal-hydraulic phenomena, recently won him a Rickover Fellowship in Nuclear Engineering. His project is part of a $122 million U.S. Department of Energy initiative, the Consortium for Advanced Simulation of Light Water Reactors, which aims to create a virtual nuclear reactor. An ensemble of state-of-the-art codes, including the one Magolan is collaborating on, is intended to both improve operating and safety performance of the current generation of commercial reactors and help shape the future of nuclear power generation. “It’s exciting and cool that files I create might someday be essential in the design of the next generation of light water reactors,” Magolan says. In early spring, Magolan was finishing up a paper describing his contributions to a CFD code, called Hydra-TH, created at Los Alamos National Laboratory. Magolan’s work involved the code development, implementation, and validation of an advanced turbulence model in Hydra-TH. This model, once fully integrated into the code, will characterize and predict the highly complex behavior of coolant water as it rushes up through assemblies of fuel rods in a reactor core. Under the direction of NSE Assistant Professor Emilio Baglietto, Magolan tests his contributions on a personalized version of the Hydra-TH code, “playing with it to get it to work,” he says. This means he compares measurements of coolant water behavior in actual, small-scale reactor assembly experiments against his CFD simulations. His work will then be incorporated into the full version of Hydra-TH. Until recently, researchers have not been able to capture with high-grain precision the twisting paths, velocity, and temperature profiles of water as it shoots through fuel-rod assemblies. “The whole purpose of my work is to predict the turbulence-driven secondary flows that cause the coolant to spiral through reactor fuel assemblies,” says Magolan. “Capturing this behavior leads to better flow predictions.” To Magolan’s satisfaction, the rigorously defined mathematical equations he has been integrating into the Hydra-TH code are starting to match data capturing the idiosyncratic geometries and characteristics of coolant water in real single-phase nuclear assemblies for pressurized water reactors — reactors where water heats up, but doesn’t boil. He is only half-way through his validation study, but when his parts of the code are perfected and combined with comparably sophisticated software models of other reactor behaviors (such as materials design and reactor physics), the result will be a fully functioning simulation of a nuclear reactor. This model, unlike an operating nuclear plant, can serve as an experimental subject “to test and examine reactor scenarios that are too difficult or expensive to study by experimental means,” says Magolan. This virtual plant will model novel methods for optimizing reactor operation while sustaining vital safety margins. Implementing and validating code for the Hydra-TH program will satisfy the requirements for Magolan’s master’s degree, but it is just the first step in his graduate research. As a doctoral student next fall, he will be creating his own multiphase turbulence model. “I will be trying to develop a new model that will better predict how both liquid and vapor interact to influence turbulence structures as they move through a reactor,” he says. “It’s going to take a lot of work to move this forward.” These next few years of research will be supported by the Rickover Fellowship, sponsored by the Naval Reactors Division of the Department of Energy. As part of the fellowship agreement, once Magolan earns his PhD he will work at the Knolls Atomic Power Laboratory near Albany, New York, designing and modeling nuclear reactors for Navy submarines. While Magolan’s immediate future revolves around nuclear engineering, he notes that his research in computational fluid dynamics can be applied to optimize many systems. “It’s a tool you can use to model flow for a myriad of engineering applications, including submarines, rockets, airplanes, or cars,” Magolan says. “If I learn it well, I could take many routes in my career, and do different things. It’s the kind of flexibility I really like.”
Tucker J.D.,Oregon State University |
Miller M.K.,Oak Ridge National Laboratory |
Young G.A.,Knolls Atomic Power Laboratory
Acta Materialia | Year: 2015
Duplex stainless steels are desirable for use in power generation systems because of their attractive combination of strength, corrosion resistance and cost. However, thermal embrittlement at intermediate homologous temperatures of ∼475 °C and below, limits upper service temperatures for many applications. New lean grade duplex alloys have improved thermal stability over standard grades and potentially increase the upper service temperature or the lifetime at a given temperature for this class of material. The present work compares the thermal stability of lean grade, alloy 2003, to standard grade, alloy 2205, through a series of isothermal agings between 260 °C and 482 °C for times between 1 and 10,000 h. Aged samples were characterized by changes in microhardness and impact toughness. Additionally, atom probe tomography was performed to illustrate the evolution of the α-α′ phase separation in both alloys at select conditions. Atom probe tomography confirmed that phase separation occurs via spinodal decomposition for both alloys, and identified the presence of Ni-Cu-Si-Mn-P clusters in alloy 2205, which may contribute to the embrittlement of this alloy. The impact toughness model predictions for the upper service temperature show that alloy 2003 may be viable for use in 288 °C applications for 80-year service lifetimes based on a Charpy V-notch criteria of 47 J at room temperature. In comparison, alloy 2205 should be limited to 260 °C applications for the same room temperature toughness of 47 J. © 2014 Acta Materialia Inc. Published by Elsevier Ltd. All rights reserved.
Smith K.V.,Knolls Atomic Power Laboratory
Engineering Fracture Mechanics | Year: 2011
This paper presents the results of numerical simulations of fatigue crack growth performed using three-dimensional elastic-plastic finite element analysis. A simple node release scheme is used to simulate crack advancement. The crack front is assumed to be straight. Crack growth following a tensile overload is simulated. The total energy dissipated per cycle is calculated directly from the finite element analysis and used to predict fatigue crack growth. For comparison, fatigue crack growth rate experiments were performed on Type 304 stainless steel C(T) specimens to determine the effect of a single tensile overload. The dissipated energy per cycle is found to correlate well with the measured fatigue crack growth rate following an overload. © 2011 Elsevier Ltd.
Caro E.,Knolls Atomic Power Laboratory
Annals of Nuclear Energy | Year: 2016
MC21 is a continuous-energy Monte Carlo radiation transport code for the calculation of the steady-state spatial distributions of reaction rates in three-dimensional models. The code supports neutron and photon transport in fixed source problems, as well as iterated-fission source (eigenvalue) neutron problems. The capability to simulate the production of photon-induced neutrons has been added to the code. In this paper, relativistic two-body kinematics are used to derive exact expressions for the secondary energy and angle distribution of photoneutrons. This treatment, implemented in MC21, is important in cases where the evaluated photonuclear data does not give an explicit energy distribution. Comparisons of the relativistic relations were made to approximations in MCNP5 and TRIPOLI-4, highlighting the magnitude of the error of those approximations. © 2016 Elsevier Ltd
Trumbull T.H.,Knolls Atomic Power Laboratory |
Fieno T.E.,Knolls Atomic Power Laboratory
Annals of Nuclear Energy | Year: 2013
The classical treatment for determining the scattering neutron energy and angle given epithermal scattering from heavy nuclides is to sample the target nuclei velocity using the Maxwellian ideal gas law while assuming a constant scattering cross section. This treatment has been shown to be inadequate to capture the upscattering effects from certain heavy nuclides with strong epithermal scattering resonances, e.g., 238U. A method to correctly account for the resonance scattering effects has been implemented in the MC21 Monte Carlo code. The method is based on the Doppler Broadening Rejection Correction (DBRC) method and also allows for the use of weight adjustment in lieu of rejection sampling. Whereas previous work has focused on applying the resonance scattering correction to 238U only, the MC21 implementation applies the rejection correction more broadly, including all uranium and plutonium isotopes in the models analyzed. This work demonstrates the effect of applying the DBRC to both LEU and MOX pin cell depletions on eigenvalue and nuclide inventories as a function of burn-up. As compared to a reference case lacking any resonance scattering correction, the effect on uranium and plutonium inventories is significant when applying the DBRC. In the case of the MOX fuel system, significant differences in the plutonium concentrations are also seen when DBRC is limited to just the 238U, suggesting that the DBRC method should be applied more broadly. © 2013 Elsevier Ltd. All rights reserved.
Romano P.K.,Knolls Atomic Power Laboratory
Computer Physics Communications | Year: 2014
An algorithm for generating random variates from the Madland-Nix fission energy spectrum assuming a constant compound nucleus cross section is given based on physics considerations. A program was written to generate variates using the algorithm developed, and it was shown that the generated variates match the probability density function. This algorithm can be used by Monte Carlo particle transport codes to sample secondary energies for neutrons born from fission when the underlying data is given as parameters to a Madland-Nix energy spectrum. © 2014 Elsevier B.V.
Trumbull T.H.,Knolls Atomic Power Laboratory
Physics of Reactors 2016, PHYSOR 2016: Unifying Theory and Experiments in the 21st Century | Year: 2016
At thermal neutron energies, it is essential to properly model neutron interactions with bound atom moderators to achieve the correct post collision energy and angle of the scattered neutrons. This is complicated by the fact that at thermal energies, the neutrons are at energies comparable to the characteristic rotational and vibrational energies of atoms in molecules and crystals, and the de Broglie wavelength is comparable to the interatomic distances. For the mainstream nuclear data libraries, moderator scattering data is provided in ENDF-formatted files. In general, these data need to be further processed by downstream codes into a format that is usable by neutron transport codes. This paper provides details of the techniques used by the NDEX processing code to process the ENDF thermal scattering data for the MC21 Monte Carlo code. Results of the integrated scattering cross section and secondary angle and energy distribution functions for several moderator materials are provided. Some comparisons of the NDEX computed quantities to other processing codes that perform similar functions are also provided.
Sutton T.M.,Knolls Atomic Power Laboratory
Mathematics and Computations, Supercomputing in Nuclear Applications and Monte Carlo International Conference, M and C+SNA+MC 2015 | Year: 2015
In the study of Monte Carlo statistical uncertainties for iterated-fission-source calculations, an important distinction is made between the 'real' and 'apparent' variances. The former is the actual variance of a Monte Carlo calculation result, while the latter is an estimate of the former obtained using the results of the fission generations in the formula for uncorrelated random variates. That the apparent variance is a biased estimate of the real variance has been known and the reason for the bias understood for years. More recently, several authors have noted various interesting phenomena regarding the apparent and real variances and the relationship between them. Some of these are: an increase in the apparent variance near surfaces with reflecting boundary conditions, a non-uniform spatial distribution of the ratio of the apparent-to-real variance, the dependence of this ratio on the size of the region over which the result is tallied, and a rate of convergence of the real variance that is less than the inverse of the number of neutron histories run. This paper discusses a theoretical description of the Monte Carlo process using a discretized phase space, and then uses it to explain the causes of these phenomena.