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All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

TU-H-CAMPUS-IeP1-02: Validation of a CT Monte Carlo Software

Purpose: To validate the in-house developed CT Monte Carlo calculation tool GMctdospp against reference simulation sets provided by the AAPM in the new report 195. Methods: Deposited energy was calculated in four segments (test 1) and two 10 cm long cylinders (test 2) inside a CTDI phantom (following case #4 of the AAPM report 195). The x-ray point source of a given 120 kVp spectrum was collimated to a fan beam with two thicknesses (10 mm, 80 mm) for a static and a rotational setup. In addition, a given chest geometry was used to calculate deposited energy in several organs for a 0° static and a rotational beam (following case #5 of the AAPM report 195). The results of GMctdospp were compared against the particular mean value of the four quoted Monte Carlo codes (EGSnrc, Geant 4, MCNP and Penelope). Results: Calculated values showed no outliers in any of the cases. Differences between GMctdospp and the particular mean Results: Calculated values showed no outliers in any of the cases. Differences between GMctdospp and the particular mean value were always at similar magnitude compared to the quoted codes. For case #4 (CTDI phantom) the relative differences were within 1.5 %, on average 0.4 % and for case #5 (chest phantom) within 2.5 % and on average 0.85 %. Conclusion: The results confirmed an overall uncertainty of the Monte-Carlo calculation chain in GMctdospp being <2.5 %, for most cases even better. This can be considered small compared to other sources of uncertainties, e.g. virtual source and patient models. The photon transport implemented in GMctdospp inside a voxel-based patient geometry was successfully verified.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2022

Zero power experiment for molten salt reactor - code verification and validation request

Molten Salt Reactors (MSRs) are one of the six perspective Generation IV nuclear technologies that should fulfil sustainability, economics, safety, and proliferation resistance goals. In recent years these concepts have attracted increased interest in research as well as in industry. In the context of developing new, innovative technology, a zero-power experiment would be an ideal option for improving the understanding of MSR systems under low-risk and low-cost conditions. The given research was focused on cross-verification of the nuclear codes for MSR modelling and simulations as required for the development of the first design of a potential facility and a draft of an experimental programme. The analysis included the evaluation of lattice and Monte Carlo codes. The study showed that all considered nuclear tools were in good agreement with each other. Although this does not mean the results will match the real experimental data, which is essential for the regulatory approval of a novel technology. Currently, experimental data is almost absent for MSR technology. Hence, a zero power experiment is needed to assure we can rely on modelling and simulation studies for further development of MSR technology and deliver real-world experiments to support the licensing of future MSRs. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Preparatory Verification of a New Monte Carlo Code cosRMC based on VERA Core Physics Benchmark

To promote nuclear power plant system design and simulation software self-reliance development, DOE of China approved a R and D key project to develop an integrated engine for pressurized water reactor core and system design and analysis, referred as COSINE hereafter. The scope of this code package includes reactor physics, thermal hydraulics, safety analysis, severe accidents, probability safety assessment and etc. Many universities, research institutes and nuclear related corporations in China are involved in the code development, among which the State Nuclear Power Software Development Center (SNPSDC) plays the leading role. With the requirement of accurate three-dimensional modeling of the advanced, special or new complex core in reactor neutronics and shielding analysis, Monte Carlo method is becoming a more powerful tool for reactor analysis. Thus, a new Monte Carlo transport code named cosRMC (abbreviation for COSINE Reactor Monte Carlo) has been developed jointly by SNPSDC and Department of Engineering Physics, Tsinghua University, the main developer of RMC code. The cosRMC now has many functions as criticality calculation, fixed-source calculation, burnup calculation, and etc. Some techniques for geometry treatment, new burnup algorithm, source convergence acceleration, massive tally and parallel calculation have been implemented in cosRMC to improve the efficiency. In order to ensure its reliability, several types of V and V work have been and are being carried out for cosRMC. In this paper, some preparatory verification results based on VERA Core Physics Benchmark are presented. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

AETHER: A simulation platform for inductively coupled plasma

An in-house code is developed to simulate the inductively coupled plasma (ICP). The model comprises the fluid, electromagnetic and transformer submodels. Fluid equations are solved to evaluate the plasma flow parameters, including the plasma and neutral densities, ion and neutral velocities, electron flux, electron temperature, and electric potential. The model relies on the ambipolar approximation and offers the evaluation of plasma parameters without solving the sheath region. The electromagnetic model handles the calculation of the electric and magnetic fields using the magnetic vector potential. The transformer model captures the effect of the matching circuit utilized in laboratory experiments for RF power deposition. The continuity and momentum equations are solved using finite volume method. The energy, electric potential, and magnetic vector potential equations are solved using finite difference method. The resulting linear systems of equations are solved with iterative solvers including Jacobi and GMRES. The code is written using the C++ programming language, it works in parallel and has graphical user interface. The model is applied to study ICP characteristics of a plasma confined within a cylindrical chamber with dielectric walls for two different power deposition cases. The results obtained from the developed model are verified using the plasma module of COMSOL Multiphysics. The model is also applied to a plasma source configuration, and it is demonstrated that there is an overall increase in the plasma potential when current is extracted from ICP with a biased wall electrode.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2015

A cubic map chaos criterion theorem with applications in generalized synchronization based pseudorandom number generator and image encryption

This paper sets up a chaos criterion theorem on a kind of cubic polynomial discrete maps. Using this theorem, Zhou-Song's chaos criterion theorem on quadratic polynomial discrete maps and generalized synchronization (GS) theorem construct an eight-dimensional chaotic GS system. Numerical simulations have been carried out to verify the effectiveness of theoretical results. The chaotic GS system is used to design a chaos-based pseudorandom number generator (CPRNG). Using FIPS 140-2 test suit/Generalized FIPS 140-2, test suit tests the randomness of two 1000 key streams consisting of 20 000 bits generated by the CPRNG, respectively. The results show that there are 99.9%/98.5% key streams to have passed the FIPS 140-2 test suit/Generalized FIPS 140-2 test. Numerical simulations show that the different keystreams have an average 50.001% same codes. The key space of the CPRNG is larger than 2{sup 1345}. As an application of the CPRNG, this study gives an image encryption example. Experimental results show that the linear coefficients between the plaintext and the ciphertext and the decrypted ciphertexts via the 100 key streams with perturbed keys are less than 0.00428. The result suggests that the decrypted texts via the keystreams generated via perturbed keys of the CPRNG are almost completely independent on the original image text, and brute attacks are needed to break the cryptographic system.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2015

CRISTAL V2.0: A New-generation criticality package - 14597

A new release of the CRISTAL French criticality calculation package, named CRISTAL V2.0, has been developed and intensively validated by IRSN and CEA, in collaboration with AREVA. This package was designed to cover the needs of criticality practitioners in the fields of nuclear fuel cycle facilities and transportation of fissile materials. Based on an optimized modular architecture in an updated software development environment, the design of CRISTAL package V2.0 integrates new generation of cross section libraries, new calculation schemes consistent with new applications, and up-to-date calculation codes capabilities. Four calculation routes are currently available allowing multi-group and continuous energy calculations. The validation database is composed of more than 3000 critical experiments, in order to cover a wide range of configurations. The CRISTAL V2.0 package also comprises customer services covering a range of areas of activity like documentation, hotline, maintenance and training courses. A wide promotion and an increased recognition referred both nationally and internationally are foreseen, including the provision of the CRISTAL V2.0 package to the OECD / NEA Data Bank. (authors)

42 ENGINEERING↗
DOE OSTI2013

Development and application of a hybrid transport methodology for active interrogation systems

A hybrid Monte Carlo and deterministic methodology has been developed for application to active interrogation systems. The methodology consists of four steps: i) neutron flux distribution due to neutron source transport and subcritical multiplication; ii) generation of gamma source distribution from (n, 7) interactions; iii) determination of gamma current at a detector window; iv) detection of gammas by the detector. This paper discusses the theory and results of the first three steps for the case of a cargo container with a sphere of HEU in third-density water cargo. To complete the first step, a response-function formulation has been developed to calculate the subcritical multiplication and neutron flux distribution. Response coefficients are pre-calculated using the MCNP5 Monte Carlo code. The second step uses the calculated neutron flux distribution and Bugle-96 (n, 7) cross sections to find the resulting gamma source distribution. In the third step the gamma source distribution is coupled with a pre-calculated adjoint function to determine the gamma current at a detector window. The AIMS (Active Interrogation for Monitoring Special-Nuclear-Materials) software has been written to output the gamma current for a source-detector assembly scanning across a cargo container using the pre-calculated values and taking significantly less time than a reference MCNP5 calculation. (authors)

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2015

Evaluation of the concrete shield compositions from the 2010 criticality accident alarm system benchmark experiments at the CEA Valduc SILENE facility

In October 2010, a series of benchmark experiments were conducted at the French Commissariat a l'Energie Atomique et aux Energies Alternatives (CEA) Valduc SILENE facility. These experiments were a joint effort between the United States Department of Energy Nuclear Criticality Safety Program and the CEA. The purpose of these experiments was to create three benchmarks for the verification and validation of radiation transport codes and evaluated nuclear data used in the analysis of criticality accident alarm systems. This series of experiments consisted of three single-pulsed experiments with the SILENE reactor. For the first experiment, the reactor was bare (unshielded), whereas in the second and third experiments, it was shielded by lead and polyethylene, respectively. The polyethylene shield of the third experiment had a cadmium liner on its internal and external surfaces, which vertically was located near the fuel region of SILENE. During each experiment, several neutron activation foils and thermoluminescent dosimeters (TLDs) were placed around the reactor. Nearly half of the foils and TLDs had additional high-density magnetite concrete, high-density barite concrete, standard concrete, and/or BoroBond shields. CEA Saclay provided all the concrete, and the US Y-12 National Security Complex provided the BoroBond. Measurement data from the experiments were published at the 2011 International Conference on Nuclear Criticality (ICNC 2011) and the 2013 Nuclear Criticality Safety Division (NCSD 2013) topical meeting. Preliminary computational results for the first experiment were presented in the ICNC 2011 paper, which showed poor agreement between the computational results and the measured values of the foils shielded by concrete. Recently the hydrogen content, boron content, and density of these concrete shields were further investigated within the constraints of the previously available data. New computational results for the first experiment are now available that show much better agreement with the measured values.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

SU-F-T-139: Meeting the Challenges of Quality Control in the TOPAS Monte Carlo Simulation Toolkit for Proton Therapy

Purpose: Monte Carlo particle transport simulation (MC) codes have become important tools in proton therapy and biology, both for research and practice. TOPAS is an MC toolkit serving users worldwide (213 licensed users at 95 institutions in 21 countries). It provides unprecedented ease in 4D placement of geometry components, beam sources and scoring through its user-friendly and reproducible parameter file interface. Quality control (QC) of stochastic simulation software is inherently difficult, and the versatility of TOPAS introduces additional challenges. But QC is vital as the TOPAS development team implements new features, addresses user feedback and reacts to upgrades of underlying software (i.e. Geant4). Methods: Whenever code is committed to our repository, over 50 separate module tests are automatically triggered via a continuous integration service. They check that the various module options execute successfully and that their results are statistically consistent with previous reference values. Prior to each software release, longer end-to-end tests automatically validate TOPAS against experimental data and a TOPAS benchmark. These include simulating multiple properties of spread-out Bragg peaks, validating nuclear models, and searching for differences in patient simulations. Results: Continuous integration has proven effective in catching regressions at the time they are introduced, particularly when implementing new features that involve refactoring code (e.g. multithreading and ntuple output). Code coverage statistics highlight untested portions of code and guide development of new tests. The various end-to-end tests demonstrate that TOPAS accurately describes the physics of proton therapy within clinical tolerances. Conclusion: The TOPAS QC strategy of frequent short tests and pre-release long tests has led to a more reliable tool. However, the versatility of TOPAS makes it difficult to predict how users might combine different modules, and so QC ultimately remains a partnership between the developer and the user. This work was funded by National Cancer Institute grant R01 CA 140735.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2017

Employing Whisper in selecting benchmarks for traditional validation methods

The Nuclear Criticality Safety Program at Sandia National Laboratories (SNL) uses a collection of benchmarks developed from the International Handbook of Evaluated Criticality Safety Benchmark Experiments for validation of MCNP. In early 2016, an effort began to expand the benchmark collection to cover a broader range of materials and neutron spectra. The program undertook the task of updating current benchmarks to reflect best practices from Los Alamos National Laboratory (LANL) for MCNP calculations. Spring of 2016, LANL shared the Whisper code with SNL in a collaborated effort to investigate application of the code to activities at SNL. Whisper is a statistical package developed to support nuclear criticality safety validation efforts. It measures how well benchmarks correlate to a system or process and calculates a USL. Whisper has begun to develop into an asset for the SNL Nuclear Criticality Safety team to determine and reassure benchmarks selected for validation truly correlate process parameters (i.e., fall within an acceptable Area of Applicability). SNL's use of traditional statistical methods, with the abilities of Whisper to identify correlating benchmarks, has facilitated in ensuring confidence in validation and USL determination. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Active learning for accelerated design of layered materials

Hetero-structures made from vertically stacked monolayers of transition metal dichalcogenides hold great potential for optoelectronic and thermoelectric devices. Discovery of the optimal layered material for specific applications necessitates the estimation of key material properties, such as electronic band structure and thermal transport coefficients. However, screening of material properties via brute force ab initio calculations of the entire material structure space exceeds the limits of current computing resources. Moreover, the functional dependence of material properties on the structures is often complicated, making simplistic statistical procedures for prediction difficult to employ without large amounts of data collection. Here, we present a Gaussian process regression model, which predicts material properties of an input hetero-structure, as well as an active learning model based on Bayesian optimization, which can efficiently discover the optimal hetero-structure using a minimal number of ab initio calculations. The electronic band gap, conduction/valence band dispersions, and thermoelectric performance are used as representative material properties for prediction and optimization. The Materials Project platform is used for electronic structure computation, while the BoltzTraP code is used to compute thermoelectric properties. Bayesian optimization is shown to significantly reduce the computational cost of discovering the optimal structure when compared with finding an optimal structure by building a regression model to predict material properties. The models can be used for predictions with respect to any material property and our software, including data preparation code based on the Python Materials Genomics (PyMatGen) library as well as python-based machine learning code, is available open source.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Efficiency Calculation for an In-situ Beta Nuclide Monitoring System Based on a Plastic Scintillator - 18193

A plastic scintillator performance for an in-situ measurement system was analyzed using both simulation and experimental methods. The measurement results of four major pure beta-emitting radionuclides - namely H-3. C-14, P-32 and Sr-90/Y-90 - were compared with those from a Monte Carlo N-particle (MCNP) code simulation. Good agreement between the simulation and experimental results was observed for P-32 and Sr-90/Y-90. However, due to the short range of their beta particles in water, the efficiency for both H-3 and C-14 was rather low, and these radionuclides could not be detected. The optimal thickness for the plastic scintillator was identified, as was its suggested diameter. The effect of water depth on system performance was also considered. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Design of the radiation shielding for the time of flight enhanced diagnostics neutron spectrometer at Experimental Advanced Superconducting Tokamak

A radiation shielding has been designed to reduce scattered neutrons and background gamma-rays for the new double-ring Time Of Flight Enhanced Diagnostics (TOFED). The shielding was designed based on simulation with the Monte Carlo code MCNP5. Dedicated model of the EAST tokamak has been developed together with the emission neutron source profile and spectrum; the latter were simulated with the Nubeam and GENESIS codes. Significant reduction of background radiation at the detector can be achieved and this satisfies the requirement of TOFED. The intensities of the scattered and direct neutrons in the line of sight of the TOFED neutron spectrometer at EAST are studied for future data interpretation.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2013

Expansion effects on solar wind hybrid simulations

Ion kinetic simulations of the solar wind using hybrid codes can model local wave input, heating and instabilities, but generally do not include long term evolution effects in the expanding solar wind. We further develop the expanding box model used in earlier studies to include the mirror force effects and study their role in the evolution of the proton distribution functions in the outer corona and inner heliosphere. The mirror force, significant in the acceleration region of the solar wind, is required for consistency with the conservation of magnetic moment of particles in the expanding wind. We present preliminary results from the modified 1D expanding box hybrid (EBHM) simulations.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2016

Initial Validation of the Neutron Physics Codes in COSINE Using the MIT BEAVRS Benchmark

A software package for nuclear power plant design and analysis, known as COSINE, has been developed by State Nuclear Power Software Development Center (SNPSDC). A detailed benchmark called BEAVRS which includes measured reactor data is presented by MIT. The BEAVRS benchmark is based on a commercial power plant and provides a highly-detailed Pressurized Water Reactor (PWR) specification with measured operational data that has been used to validate the neutron physics codes in COSINE. In this paper, a COSINE core model is prepared to calculate the BEAVRS benchmark. The first stage validation is focused on the depletion of two cycles and initial physics test of cycle 1. The calculation results agree well with the measured data. It could be confirmed that the COSINE codes package has the ability to design and analyze a commercial PWR core, and the precision of the calculation results can meet the requirements of industrial design and analysis. A calculation of MIT BEAVRS benchmark using COSINE has been performed in this paper. A lot of measurements are given in the benchmark specification which can be used for code validation. The calculation includes a depletion calculation of both two cycles and a calculation of the initial physics test of cycle 1. COSINE neutron physics codes use the mature industrial calculation method which can achieve reasonable results with acceptable accuracy and speed. Through the calculation, it can be seen that the difference of critical boron concentration calculated from cos Nu is less than 25 pcm compared with measured data. And the relative difference of rod worth is less than 10%. The isothermal temperature coefficient difference between measurement and calculation is less than 1 pcm/deg. F. All results meet the criteria of ANS. The validation for measurements after initial physics test following a more complex power history and the validation for in-core detector signals are underway. (authors)

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2017

CONV-3D Code Application for CFD-Modeling of Lead-Bismuth Coolant

The results of testing the three-dimensional CFD-code on problems very close to real processes for lead–bismuth eutectic are presented: flow in a 19-fuel element assembly with three spacing lattices and convection in the ‘tumbler’ model. The results of convergence on a sequence of grids permit drawing the conclusion that for a 19-fuel element assembly the error in the calculation of the friction coefficient on a more finely divided grid does not exceed 10% with two-fold smaller grid step in each direction. Similar arguments are valid for the calculation of the hydraulic resistance coefficient.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Developing a GUM Compliant Uncertainty Approach for NDA Gamma Analysis Using FRAM

The Fixed-Energy Response-Function Analysis with Multiple Efficiency (FRAM) code was developed at Los Alamos National Laboratory (LANL) to analyze gamma-ray spectra for the isotopic composition of plutonium or uranium. This software provides means to easily process gamma-ray spectra to obtain isotopic information and is used globally for international safeguards applications. In a recent review of the FRAM code for the use in nuclear forensics applications, it was determined that a more complete understanding and definition of the various sources of uncertainty that contribute to the overall uncertainty of the FRAM measurement process is required. As one of the ISO 17025 accredited Bulk Special Nuclear Materials Analysis Program (BSAP) laboratories, LANL is required to perform an accurate and transparent uncertainty calculation on every analysis it performs. This paper describes a new approach to calculate a Guide to the Expression of Uncertainty in Measurement (GUM) compliant FRAM uncertainty to support nuclear forensics applications.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2017

New estimates of extensive-air-shower energies on the basis of signals in scintillation detectors

New formulas for estimating the energy of inclined extensive air showers (EASs) on the basis of signals in detectors by means of an original method and detailed tables of signals induced in scintillation detectors by photons, electrons, positrons, and muons and calculated with the aid of the GEANT4 code package were proposed in terms of the QGSJETII-04, EPOS LHC, and GHEISHA models. The parameters appearing in the proposed formulas were calculated by employing the CORSIKA code package. It is shown that, for showers of zenith angles in the range of 20◦–45◦, the standard constant-intensity-cut method, which is used to interpret data from the Yakutsk EAS array, overestimates the shower energy by a factor of 1.2 to 1.5. It is proposed to employ the calculated VEM (Vertical Equivalent Muon) signal units of 10.8 and 11.4 MeV for, respectively, ground-based and underground scintillation detectors and to take into account the dependence of signals on the azimuthal angle of the detector position and fluctuations in the development of showers.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗