Estimation of relative biological effectiveness for boron neutron capture therapy using the PHITS code coupled with a microdosimetric kinetic model

PubMed Central

Horiguchi, Hironori; Sato, Tatsuhiko; Kumada, Hiroaki; Yamamoto, Tetsuya; Sakae, Takeji

2015-01-01

Abstract The absorbed doses deposited by boron neutron capture therapy (BNCT) can be categorized into four components: α and 7Li particles from the 10B(n, α)7Li reaction, 0.54-MeV protons from the 14N(n, p)14C reaction, the recoiled protons from the 1H(n, n) 1H reaction, and photons from the neutron beam and 1H(n, γ)2H reaction. For evaluating the irradiation effect in tumors and the surrounding normal tissues in BNCT, it is of great importance to estimate the relative biological effectiveness (RBE) for each dose component in the same framework. We have, therefore, established a new method for estimating the RBE of all BNCT dose components on the basis of the microdosimetric kinetic model. This method employs the probability density of lineal energy, y, in a subcellular structure as the index for expressing RBE, which can be calculated using the microdosimetric function implemented in the particle transport simulation code (PHITS). The accuracy of this method was tested by comparing the calculated RBE values with corresponding measured data in a water phantom irradiated with an epithermal neutron beam. The calculation technique developed in this study will be useful for biological dose estimation in treatment planning for BNCT. PMID:25428243

Kinetics of silver release from microfuel with taking into account the limited-solubility effect

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ivanov, A. S., E-mail: asi.kiae@gmail.com; Rusinkevich, A. A., E-mail: rusinkevich_andr@mail.ru

2014-12-15

The effect of a limited solubility of silver in silicon carbide on silver release from a microfuel with a TRISO coating is studied. It is shown that a limited solubility affects substantially both concentration profiles and silver release from a microfuel over a broad range of temperatures. A procedure is developed for obtaining fission-product concentration profiles in a microfuel and graphs representing the flow and integrated release of fission products on the basis of data from neutron-physics calculations and results obtained by calculating thermodynamics with the aid of the Ivtanthermo code and kinetics with the aid of the FP-Kinetics code.more » This procedure takes into account a limited solubility of fission products in protective coatings of microfuel.« less

Application of ATHLET/DYN3D coupled codes system for fast liquid metal cooled reactor steady state simulation

NASA Astrophysics Data System (ADS)

Ivanov, V.; Samokhin, A.; Danicheva, I.; Khrennikov, N.; Bouscuet, J.; Velkov, K.; Pasichnyk, I.

2017-01-01

In this paper the approaches used for developing of the BN-800 reactor test model and for validation of coupled neutron-physic and thermohydraulic calculations are described. Coupled codes ATHLET 3.0 (code for thermohydraulic calculations of reactor transients) and DYN3D (3-dimensional code of neutron kinetics) are used for calculations. The main calculation results of reactor steady state condition are provided. 3-D model used for neutron calculations was developed for start reactor BN-800 load. The homogeneous approach is used for description of reactor assemblies. Along with main simplifications, the main reactor BN-800 core zones are described (LEZ, MEZ, HEZ, MOX, blankets). The 3D neutron physics calculations were provided with 28-group library, which is based on estimated nuclear data ENDF/B-7.0. Neutron SCALE code was used for preparation of group constants. Nodalization hydraulic model has boundary conditions by coolant mass-flow rate for core inlet part, by pressure and enthalpy for core outlet part, which can be chosen depending on reactor state. Core inlet and outlet temperatures were chosen according to reactor nominal state. The coolant mass flow rate profiling through the core is based on reactor power distribution. The test thermohydraulic calculations made with using of developed model showed acceptable results in coolant mass flow rate distribution through the reactor core and in axial temperature and pressure distribution. The developed model will be upgraded in future for different transient analysis in metal-cooled fast reactors of BN type including reactivity transients (control rods withdrawal, stop of the main circulation pump, etc.).

Sensitivity Analysis of Cf-252 (sf) Neutron and Gamma Observables in CGMF

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carter, Austin Lewis; Talou, Patrick; Stetcu, Ionel

CGMF is a Monte Carlo code that simulates the decay of primary fission fragments by emission of neutrons and gamma rays, according to the Hauser-Feshbach equations. As the CGMF code was recently integrated into the MCNP6.2 transport code, great emphasis has been placed on providing optimal parameters to CGMF such that many different observables are accurately represented. Of these observables, the prompt neutron spectrum, prompt neutron multiplicity, prompt gamma spectrum, and prompt gamma multiplicity are crucial for accurate transport simulations of criticality and nonproliferation applications. This contribution to the ongoing efforts to improve CGMF presents a study of the sensitivitymore » of various neutron and gamma observables to several input parameters for Californium-252 spontaneous fission. Among the most influential parameters are those that affect the input yield distributions in fragment mass and total kinetic energy (TKE). A new scheme for representing Y(A,TKE) was implemented in CGMF using three fission modes, S1, S2 and SL. The sensitivity profiles were calculated for 17 total parameters, which show that the neutron multiplicity distribution is strongly affected by the TKE distribution of the fragments. The total excitation energy (TXE) of the fragments is shared according to a parameter RT, which is defined as the ratio of the light to heavy initial temperatures. The sensitivity profile of the neutron multiplicity shows a second order effect of RT on the mean neutron multiplicity. A final sensitivity profile was produced for the parameter alpha, which affects the spin of the fragments. Higher values of alpha lead to higher fragment spins, which inhibit the emission of neutrons. Understanding the sensitivity of the prompt neutron and gamma observables to the many CGMF input parameters provides a platform for the optimization of these parameters.« less

Multi-scale modeling of irradiation effects in spallation neutron source materials

NASA Astrophysics Data System (ADS)

Yoshiie, T.; Ito, T.; Iwase, H.; Kaneko, Y.; Kawai, M.; Kishida, I.; Kunieda, S.; Sato, K.; Shimakawa, S.; Shimizu, F.; Hashimoto, S.; Hashimoto, N.; Fukahori, T.; Watanabe, Y.; Xu, Q.; Ishino, S.

2011-07-01

Changes in mechanical property of Ni under irradiation by 3 GeV protons were estimated by multi-scale modeling. The code consisted of four parts. The first part was based on the Particle and Heavy-Ion Transport code System (PHITS) code for nuclear reactions, and modeled the interactions between high energy protons and nuclei in the target. The second part covered atomic collisions by particles without nuclear reactions. Because the energy of the particles was high, subcascade analysis was employed. The direct formation of clusters and the number of mobile defects were estimated using molecular dynamics (MD) and kinetic Monte-Carlo (kMC) methods in each subcascade. The third part considered damage structural evolutions estimated by reaction kinetic analysis. The fourth part involved the estimation of mechanical property change using three-dimensional discrete dislocation dynamics (DDD). Using the above four part code, stress-strain curves for high energy proton irradiated Ni were obtained.

Analysis of fluid fuel flow to the neutron kinetics on molten salt reactor FUJI-12

DOE Office of Scientific and Technical Information (OSTI.GOV)

Aji, Indarta Kuncoro, E-mail: indartaaji@s.itb.ac.id; Waris, Abdul, E-mail: awaris@fi.itb.ac.id; Permana, Sidik

Molten Salt Reactor is a reactor are operating with molten salt fuel flowing. This condition interpret that the neutron kinetics of this reactor is affected by the flow rate of the fuel. This research analyze effect by the alteration velocity of the fuel by MSR type Fuji-12, with fuel composition LiF-BeF{sub 2}-ThF{sub 4}-{sup 233}UF{sub 4} respectively 71.78%-16%-11.86%-0.36%. Calculation process in this study is performed numerically by SOR and finite difference method use C programming language. Data of reactivity, neutron flux, and the macroscopic fission cross section for calculation process obtain from SRAC-CITATION (Standard thermal Reactor Analysis Code) and JENDL-4.0 datamore » library. SRAC system designed and developed by JAEA (Japan Atomic Energy Agency). This study aims to observe the effect of the velocity of fuel salt to the power generated from neutron precursors at fourth year of reactor operate (last critical condition) with number of multiplication effective; 1.0155.« less

Estimation of relative biological effectiveness for boron neutron capture therapy using the PHITS code coupled with a microdosimetric kinetic model.

PubMed

Horiguchi, Hironori; Sato, Tatsuhiko; Kumada, Hiroaki; Yamamoto, Tetsuya; Sakae, Takeji

2015-03-01

The absorbed doses deposited by boron neutron capture therapy (BNCT) can be categorized into four components: α and (7)Li particles from the (10)B(n, α)(7)Li reaction, 0.54-MeV protons from the (14)N(n, p)(14)C reaction, the recoiled protons from the (1)H(n, n) (1)H reaction, and photons from the neutron beam and (1)H(n, γ)(2)H reaction. For evaluating the irradiation effect in tumors and the surrounding normal tissues in BNCT, it is of great importance to estimate the relative biological effectiveness (RBE) for each dose component in the same framework. We have, therefore, established a new method for estimating the RBE of all BNCT dose components on the basis of the microdosimetric kinetic model. This method employs the probability density of lineal energy, y, in a subcellular structure as the index for expressing RBE, which can be calculated using the microdosimetric function implemented in the particle transport simulation code (PHITS). The accuracy of this method was tested by comparing the calculated RBE values with corresponding measured data in a water phantom irradiated with an epithermal neutron beam. The calculation technique developed in this study will be useful for biological dose estimation in treatment planning for BNCT. © The Author 2014. Published by Oxford University Press on behalf of The Japan Radiation Research Society and Japanese Society for Radiation Oncology.

Particle-in-cell modeling for MJ scale dense plasma focus with varied anode shape

DOE Office of Scientific and Technical Information (OSTI.GOV)

Link, A., E-mail: link6@llnl.gov; Halvorson, C., E-mail: link6@llnl.gov; Schmidt, A.

2014-12-15

Megajoule scale dense plasma focus (DPF) Z-pinches with deuterium gas fill are compact devices capable of producing 10{sup 12} neutrons per shot but past predictive models of large-scale DPF have not included kinetic effects such as ion beam formation or anomalous resistivity. We report on progress of developing a predictive DPF model by extending our 2D axisymmetric collisional kinetic particle-in-cell (PIC) simulations from the 4 kJ, 200 kA LLNL DPF to 1 MJ, 2 MA Gemini DPF using the PIC code LSP. These new simulations incorporate electrodes, an external pulsed-power driver circuit, and model the plasma from insulator lift-off throughmore » the pinch phase. To accommodate the vast range of relevant spatial and temporal scales involved in the Gemini DPF within the available computational resources, the simulations were performed using a new hybrid fluid-to-kinetic model. This new approach allows single simulations to begin in an electron/ion fluid mode from insulator lift-off through the 5-6 μs run-down of the 50+ cm anode, then transition to a fully kinetic PIC description during the run-in phase, when the current sheath is 2-3 mm from the central axis of the anode. Simulations are advanced through the final pinch phase using an adaptive variable time-step to capture the fs and sub-mm scales of the kinetic instabilities involved in the ion beam formation and neutron production. Validation assessments are being performed using a variety of different anode shapes, comparing against experimental measurements of neutron yield, neutron anisotropy and ion beam production.« less

Plasma kinetic effects on atomistic mix in one dimension and at structured interfaces (II)

NASA Astrophysics Data System (ADS)

Albright, Brian; Yin, Lin; Cooley, James; Haack, Jeffrey; Douglas, Melissa

2017-10-01

The Marble campaign seeks to develop a platform for studying mix evolution in turbulent, inhomogeneous, high-energy-density plasmas at the NIF. Marble capsules contain engineered CD foams, the pores of which are filled with hydrogen and tritium. During implosion, hydrodynamic stirring and plasma diffusivity mix tritium fuel into the surrounding CD plasma, leading to both DD and DT fusion neutron production. In this presentation, building upon prior work, kinetic particle-in-cell simulations using the VPIC code are used to examine kinetic effects on thermonuclear burn in Marble-like settings. Departures from Maxwellian distributions are observed near the interface and TN burn rates and inferred temperatures from synthetic neutron time of flight diagnostics are compared with those from treating the background species as Maxwellian. Work performed under the auspices of the U.S. DOE by the Los Alamos National Security, LLC Los Alamos National Laboratory and supported by the ASC and Science programs.

Keno-Nr a Monte Carlo Code Simulating the Californium -252-SOURCE-DRIVEN Noise Analysis Experimental Method for Determining Subcriticality

NASA Astrophysics Data System (ADS)

Ficaro, Edward Patrick

The ^{252}Cf -source-driven noise analysis (CSDNA) requires the measurement of the cross power spectral density (CPSD) G_ {23}(omega), between a pair of neutron detectors (subscripts 2 and 3) located in or near the fissile assembly, and the CPSDs, G_{12}( omega) and G_{13}( omega), between the neutron detectors and an ionization chamber 1 containing ^{252}Cf also located in or near the fissile assembly. The key advantage of this method is that the subcriticality of the assembly can be obtained from the ratio of spectral densities,{G _sp{12}{*}(omega)G_ {13}(omega)over G_{11 }(omega)G_{23}(omega) },using a point kinetic model formulation which is independent of the detector's properties and a reference measurement. The multigroup, Monte Carlo code, KENO-NR, was developed to eliminate the dependence of the measurement on the point kinetic formulation. This code utilizes time dependent, analog neutron tracking to simulate the experimental method, in addition to the underlying nuclear physics, as closely as possible. From a direct comparison of simulated and measured data, the calculational model and cross sections are validated for the calculation, and KENO-NR can then be rerun to provide a distributed source k_ {eff} calculation. Depending on the fissile assembly, a few hours to a couple of days of computation time are needed for a typical simulation executed on a desktop workstation. In this work, KENO-NR demonstrated the ability to accurately estimate the measured ratio of spectral densities from experiments using capture detectors performed on uranium metal cylinders, a cylindrical tank filled with aqueous uranyl nitrate, and arrays of safe storage bottles filled with uranyl nitrate. Good agreement was also seen between simulated and measured values of the prompt neutron decay constant from the fitted CPSDs. Poor agreement was seen between simulated and measured results using composite ^6Li-glass-plastic scintillators at large subcriticalities for the tank of uranyl nitrate. It is believed that the response of these detectors is not well known and is incorrectly modeled in KENO-NR. In addition to these tests, several benchmark calculations were also performed to provide insight into the properties of the point kinetic formulation.

Equilibrium and Stability Properties of Low Aspect Ratio Mirror Systems: from Neutron Source Design to the Parker Spiral

NASA Astrophysics Data System (ADS)

Peterson, Ethan; Anderson, Jay; Clark, Mike; Egedal, Jan; Endrizzi, Douglass; Flanagan, Ken; Harvey, Robert; Lynn, Jacob; Milhone, Jason; Wallace, John; Waleffe, Roger; Mirnov, Vladimir; Forest, Cary

2017-10-01

Equilibrium reconstructions of rotating magnetospheres in the lab are computed using a user-friendly extended Grad-Shafranov solver written in Python and various magnetic and kinetic measurements. The stability of these equilibria are investigated using the NIMROD code with two goals: understand the onset of the classic ``wobble'' in the heliospheric current sheet and demonstrating proof-of-principle for a laboratory source of high- β turbulence. Using the same extended Grad-Shafranov solver, equilibria for an axisymmetric, non-paraxial magnetic mirror are used as a design foundation for a high-field magnetic mirror neutron source. These equilibria are numerically shown to be stable to the m=1 flute instability, with higher modes likely stabilized by FLR effects; this provides stability to gross MHD modes in an axisymmetric configuration. Numerical results of RF heating and neutral beam injection (NBI) from the GENRAY/CQL3D code suite show neutron fluxes promising for medical radioisotope production as well as materials testing. Synergistic effects between NBI and high-harmonic fast wave heating show large increases in neutron yield for a modest increase in RF power. work funded by DOE, NSF, NASA.

Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

NASA Astrophysics Data System (ADS)

Mohammed, Abdul Aziz; Pauzi, Anas Muhamad; Rahman, Shaik Mohmmed Haikhal Abdul; Zin, Muhamad Rawi Muhammad; Jamro, Rafhayudi; Idris, Faridah Mohamad

2016-01-01

In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 (233U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintaining the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.

Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mohammed, Abdul Aziz, E-mail: azizM@uniten.edu.my; Rahman, Shaik Mohmmed Haikhal Abdul; Pauzi, Anas Muhamad, E-mail: anas@uniten.edu.my

2016-01-22

In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 ({sup 233}U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintainingmore » the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.« less

Temperature dependence of the symmetry energy and neutron skins in Ni, Sn, and Pb isotopic chains

NASA Astrophysics Data System (ADS)

Antonov, A. N.; Kadrev, D. N.; Gaidarov, M. K.; Sarriguren, P.; de Guerra, E. Moya

2017-02-01

The temperature dependence of the symmetry energy for isotopic chains of even-even Ni, Sn, and Pb nuclei is investigated in the framework of the local density approximation (LDA). The Skyrme energy density functional with two Skyrme-class effective interactions, SkM* and SLy4, is used in the calculations. The temperature-dependent proton and neutron densities are calculated through the hfbtho code that solves the nuclear Skyrme-Hartree-Fock-Bogoliubov problem by using the cylindrical transformed deformed harmonic-oscillator basis. In addition, two other density distributions of 208Pb, namely the Fermi-type density determined within the extended Thomas-Fermi (TF) method and symmetrized-Fermi local density obtained within the rigorous density functional approach, are used. The kinetic energy densities are calculated either by the hfbtho code or, for a comparison, by the extended TF method up to second order in temperature (with T2 term). Alternative ways to calculate the symmetry energy coefficient within the LDA are proposed. The results for the thermal evolution of the symmetry energy coefficient in the interval T =0 -4 MeV show that its values decrease with temperature. The temperature dependence of the neutron and proton root-mean-square radii and corresponding neutron skin thickness is also investigated, showing that the effect of temperature leads mainly to a substantial increase of the neutron radii and skins, especially in the more neutron-rich nuclei, a feature that may have consequences on astrophysical processes and neutron stars.

Optimizing Dense Plasma Focus Neutron Yields with Fast Gas Jets

NASA Astrophysics Data System (ADS)

McMahon, Matthew; Kueny, Christopher; Stein, Elizabeth; Link, Anthony; Schmidt, Andrea

2016-10-01

We report a study using the particle-in-cell code LSP to perform fully kinetic simulations modeling dense plasma focus (DPF) devices with high density gas jets on axis. The high density jet models fast gas puffs which allow for more mass on axis while maintaining the optimal pressure for the DPF. As the density of the jet compared to the background fill increases we find the neutron yield increases, as does the variability in the neutron yield. Introducing perturbations in the jet density allow for consistent seeding of the m =0 instability leading to more consistent ion acceleration and higher neutron yields with less variability. Jets with higher on axis density are found to have the greatest yield. The optimal jet configuration is explored. This work performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344.

Initial Neutronics Analyses for HEU to LEU Fuel Conversion of the Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kontogeorgakos, D.; Derstine, K.; Wright, A.

2013-06-01

The purpose of the TREAT reactor is to generate large transient neutron pulses in test samples without over-heating the core to simulate fuel assembly accident conditions. The power transients in the present HEU core are inherently self-limiting such that the core prevents itself from overheating even in the event of a reactivity insertion accident. The objective of this study was to support the assessment of the feasibility of the TREAT core conversion based on the present reactor performance metrics and the technical specifications of the HEU core. The LEU fuel assembly studied had the same overall design, materials (UO 2more » particles finely dispersed in graphite) and impurities content as the HEU fuel assembly. The Monte Carlo N–Particle code (MCNP) and the point kinetics code TREKIN were used in the analyses.« less

A Neutronic Program for Critical and Nonequilibrium Study of Mobile Fuel Reactors: The Cinsf1D Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lecarpentier, David; Carpentier, Vincent

2003-01-15

Molten salt reactors (MSRs) have the distinction of having a liquid fuel that is also the coolant. The transport of delayed-neutron precursors by the fuel modifies the precursors' equation. As a consequence, it is necessary to adapt the methods currently used for solid fuel reactors to achieve critical or kinetics calculations for an MSR. A program is presented for which this adaptation has been carried out within the framework of the two-energy-group diffusion theory with one dimension of space. This program has been called Cinsf1D (Cinetique pour reacteur a sels fondus 1D)

Optimizing Dense Plasma Focus Neutron Yields With Fast Gas Jets

NASA Astrophysics Data System (ADS)

McMahon, Matthew; Stein, Elizabeth; Higginson, Drew; Kueny, Christopher; Link, Anthony; Schmidt, Andrea

2017-10-01

We report a study using the particle-in-cell code LSP to perform fully kinetic simulations modeling dense plasma focus (DPF) devices with high density gas jets on axis. The high-density jets are modeled in the large-eddy Navier-Stokes code CharlesX, which is suitable for modeling both sub-sonic and supersonic gas flow. The gas pattern, which is essentially static on z-pinch time scales, is imported from CharlesX to LSP for neutron yield predictions. Fast gas puffs allow for more mass on axis while maintaining the optimal pressure for the DPF. As the density of a subsonic jet increases relative to the background fill, we find the neutron yield increases, as does the variability in the neutron yield. Introducing perturbations in the jet density via super-sonic flow (also known as Mach diamonds) allow for consistent seeding of the m =0 instability leading to more consistent ion acceleration and higher neutron yields with less variability. Jets with higher on axis density are found to have the greatest yield. The optimal jet configuration and the necessary jet conditions for increasing neutron yield and reducing yield variability are explored. Simulations of realistic jet profiles are performed and compared to the ideal scenario. This work performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344 and supported by the Laboratory Directed Research and Development Program (15-ERD-034) at LLNL.

Correlated Production and Analog Transport of Fission Neutrons and Photons using Fission Models FREYA, FIFRELIN and the Monte Carlo Code TRIPOLI-4® .

NASA Astrophysics Data System (ADS)

Verbeke, Jérôme M.; Petit, Odile; Chebboubi, Abdelhazize; Litaize, Olivier

2018-01-01

Fission modeling in general-purpose Monte Carlo transport codes often relies on average nuclear data provided by international evaluation libraries. As such, only average fission multiplicities are available and correlations between fission neutrons and photons are missing. Whereas uncorrelated fission physics is usually sufficient for standard reactor core and radiation shielding calculations, correlated fission secondaries are required for specialized nuclear instrumentation and detector modeling. For coincidence counting detector optimization for instance, precise simulation of fission neutrons and photons that remain correlated in time from birth to detection is essential. New developments were recently integrated into the Monte Carlo transport code TRIPOLI-4 to model fission physics more precisely, the purpose being to access event-by-event fission events from two different fission models: FREYA and FIFRELIN. TRIPOLI-4 simulations can now be performed, either by connecting via an API to the LLNL fission library including FREYA, or by reading external fission event data files produced by FIFRELIN beforehand. These new capabilities enable us to easily compare results from Monte Carlo transport calculations using the two fission models in a nuclear instrumentation application. In the first part of this paper, broad underlying principles of the two fission models are recalled. We then present experimental measurements of neutron angular correlations for 252Cf(sf) and 240Pu(sf). The correlations were measured for several neutron kinetic energy thresholds. In the latter part of the paper, simulation results are compared to experimental data. Spontaneous fissions in 252Cf and 240Pu are modeled by FREYA or FIFRELIN. Emitted neutrons and photons are subsequently transported to an array of scintillators by TRIPOLI-4 in analog mode to preserve their correlations. Angular correlations between fission neutrons obtained independently from these TRIPOLI-4 simulations, using either FREYA or FIFRELIN, are compared to experimental results. For 240Pu(sf), the measured correlations were used to tune the model parameters.

Computational Transport Modeling of High-Energy Neutrons Found in the Space Environment

NASA Technical Reports Server (NTRS)

Cox, Brad; Theriot, Corey A.; Rohde, Larry H.; Wu, Honglu

2012-01-01

The high charge and high energy (HZE) particle radiation environment in space interacts with spacecraft materials and the human body to create a population of neutrons encompassing a broad kinetic energy spectrum. As an HZE ion penetrates matter, there is an increasing chance of fragmentation as penetration depth increases. When an ion fragments, secondary neutrons are released with velocities up to that of the primary ion, giving some neutrons very long penetration ranges. These secondary neutrons have a high relative biological effectiveness, are difficult to effectively shield, and can cause more biological damage than the primary ions in some scenarios. Ground-based irradiation experiments that simulate the space radiation environment must account for this spectrum of neutrons. Using the Particle and Heavy Ion Transport Code System (PHITS), it is possible to simulate a neutron environment that is characteristic of that found in spaceflight. Considering neutron dosimetry, the focus lies on the broad spectrum of recoil protons that are produced in biological targets. In a biological target, dose at a certain penetration depth is primarily dependent upon recoil proton tracks. The PHITS code can be used to simulate a broad-energy neutron spectrum traversing biological targets, and it account for the recoil particle population. This project focuses on modeling a neutron beamline irradiation scenario for determining dose at increasing depth in water targets. Energy-deposition events and particle fluence can be simulated by establishing cross-sectional scoring routines at different depths in a target. This type of model is useful for correlating theoretical data with actual beamline radiobiology experiments. Other work exposed human fibroblast cells to a high-energy neutron source to study micronuclei induction in cells at increasing depth behind water shielding. Those findings provide supporting data describing dose vs. depth across a water-equivalent medium. This poster presents PHITS data suggesting an increase in dose, up to roughly 10 cm depth, followed by a continual decrease as neutrons come to a stop in the target.

Benchmark of Atucha-2 PHWR RELAP5-3D control rod model by Monte Carlo MCNP5 core calculation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pecchia, M.; D'Auria, F.; Mazzantini, O.

2012-07-01

Atucha-2 is a Siemens-designed PHWR reactor under construction in the Republic of Argentina. Its geometrical complexity and peculiarities require the adoption of advanced Monte Carlo codes for performing realistic neutronic simulations. Therefore core models of Atucha-2 PHWR were developed using MCNP5. In this work a methodology was set up to collect the flux in the hexagonal mesh by which the Atucha-2 core is represented. The scope of this activity is to evaluate the effect of obliquely inserted control rod on neutron flux in order to validate the RELAP5-3D{sup C}/NESTLE three dimensional neutron kinetic coupled thermal-hydraulic model, applied by GRNSPG/UNIPI formore » performing selected transients of Chapter 15 FSAR of Atucha-2. (authors)« less

Research on stellarator-mirror fission-fusion hybrid

NASA Astrophysics Data System (ADS)

Moiseenko, V. E.; Kotenko, V. G.; Chernitskiy, S. V.; Nemov, V. V.; Ågren, O.; Noack, K.; Kalyuzhnyi, V. N.; Hagnestål, A.; Källne, J.; Voitsenya, V. S.; Garkusha, I. E.

2014-09-01

The development of a stellarator-mirror fission-fusion hybrid concept is reviewed. The hybrid comprises of a fusion neutron source and a powerful sub-critical fast fission reactor core. The aim is the transmutation of spent nuclear fuel and safe fission energy production. In its fusion part, neutrons are generated in deuterium-tritium (D-T) plasma, confined magnetically in a stellarator-type system with an embedded magnetic mirror. Based on kinetic calculations, the energy balance for such a system is analyzed. Neutron calculations have been performed with the MCNPX code, and the principal design of the reactor part is developed. Neutron outflux at different outer parts of the reactor is calculated. Numerical simulations have been performed on the structure of a magnetic field in a model of the stellarator-mirror device, and that is achieved by switching off one or two coils of toroidal field in the Uragan-2M torsatron. The calculations predict the existence of closed magnetic surfaces under certain conditions. The confinement of fast particles in such a magnetic trap is analyzed.

Particle-In-Cell Modeling For MJ Dense Plasma Focus with Varied Anode Shape

NASA Astrophysics Data System (ADS)

Link, A.; Halvorson, C.; Schmidt, A.; Hagen, E. C.; Rose, D.; Welch, D.

2014-10-01

Megajoule scale dense plasma focus (DPF) Z-pinches with deuterium gas fill are compact devices capable of producing 1012 neutrons per shot but past predictive models of large-scale DPF have not included kinetic effects such as ion beam formation or anomalous resistivity. We report on progress of developing a predictive DPF model by extending our 2D axisymmetric collisional kinetic particle-in-cell (PIC) simulations to the 1 MJ, 2 MA Gemini DPF using the PIC code LSP. These new simulations incorporate electrodes, an external pulsed-power driver circuit, and model the plasma from insulator lift-off through the pinch phase. The simulations were performed using a new hybrid fluid-to-kinetic model transitioning from a fluid description to a fully kinetic PIC description during the run-in phase. Simulations are advanced through the final pinch phase using an adaptive variable time-step to capture the fs and sub-mm scales of the kinetic instabilities involved in the ion beam formation and neutron production. Results will be present on the predicted effects of different anode configurations. This work performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory (LLNL) under Contract DE-AC52-07NA27344 and supported by the Laboratory Directed Research and Development Program (11-ERD-063) and the Computing Grand Challenge program at LLNL. This work supported by Office of Defense Nuclear Nonproliferation Research and Development within U.S. Department of Energy's National Nuclear Security Administration.

SONTRAC: A solar neutron track chamber detector

NASA Technical Reports Server (NTRS)

Frye, G. M., Jr.; Jenkins, T. L.; Owens, A.

1985-01-01

The recent detection on the solar maximum mission (SMM) satellite of high energy neutrons emitted during large solar flares has provided renewed incentive to design a neutron detector which has the sensitivity, energy resolution, and time resolution to measure the neutron time and energy spectra with sufficient precision to improve our understanding of the basic flare processes. Over the past two decades a variety of neutron detectors has been flown to measure the atmospheric neutron intensity above 10 MeV and to search for solar neutrons. The SONTRAC (Solar Neutron Track Chamber) detector, a new type of neutron detector which utilizes n-p scattering and has a sensitivity 1-3 orders of magnitude greater than previous instruments in the 20-200 MeV range is described. The energy resolution is 1% for neutron kinetic energy, T sub n 50 MeV. When used with a coded aperture mask at 50 m (as would be possible on the space station) an angular resolution of approx. 4 arc sec could be achieved, thereby locating the sites of high energy nuclear interactions with an angular precision comparable to the existing x-ray experiments on SMM. The scintillation chamber is investigated as a track chamber for high energy physics, either by using arrays of scintillating optical fibers or by optical imaging of particle trajectories in a block of scintillator.

Fully kinetic simulations of dense plasma focus Z-pinch devices.

PubMed

Schmidt, A; Tang, V; Welch, D

2012-11-16

Dense plasma focus Z-pinch devices are sources of copious high energy electrons and ions, x rays, and neutrons. The mechanisms through which these physically simple devices generate such high-energy beams in a relatively short distance are not fully understood. We now have, for the first time, demonstrated a capability to model these plasmas fully kinetically, allowing us to simulate the pinch process at the particle scale. We present here the results of the initial kinetic simulations, which reproduce experimental neutron yields (~10(7)) and high-energy (MeV) beams for the first time. We compare our fluid, hybrid (kinetic ions and fluid electrons), and fully kinetic simulations. Fluid simulations predict no neutrons and do not allow for nonthermal ions, while hybrid simulations underpredict neutron yield by ~100x and exhibit an ion tail that does not exceed 200 keV. Only fully kinetic simulations predict MeV-energy ions and experimental neutron yields. A frequency analysis in a fully kinetic simulation shows plasma fluctuations near the lower hybrid frequency, possibly implicating lower hybrid drift instability as a contributor to anomalous resistivity in the plasma.

Interface requirements to couple thermal-hydraulic codes to 3D neutronic codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Langenbuch, S.; Austregesilo, H.; Velkov, K.

1997-07-01

The present situation of thermalhydraulics codes and 3D neutronics codes is briefly described and general considerations for coupling of these codes are discussed. Two different basic approaches of coupling are identified and their relative advantages and disadvantages are discussed. The implementation of the coupling for 3D neutronics codes in the system ATHLET is presented. Meanwhile, this interface is used for coupling three different 3D neutronics codes.

Implementing and testing theoretical fission fragment yields in a Hauser-Feshbach statistical decay framework

NASA Astrophysics Data System (ADS)

Jaffke, Patrick; Möller, Peter; Stetcu, Ionel; Talou, Patrick; Schmitt, Christelle

2018-03-01

We implement fission fragment yields, calculated using Brownian shape-motion on a macroscopic-microscopic potential energy surface in six dimensions, into the Hauser-Feshbach statistical decay code CGMF. This combination allows us to test the impact of utilizing theoretically-calculated fission fragment yields on the subsequent prompt neutron and γ-ray emission. We draw connections between the fragment yields and the total kinetic energy TKE of the fission fragments and demonstrate that the use of calculated yields can introduce a difference in the 〈TKE〉 and, thus, the prompt neutron multiplicity v, as compared with experimental fragment yields. We deduce the uncertainty on the 〈TKE〉 and v from this procedure and identify possible applications.

Correlated prompt fission data in transport simulations

DOE PAGES

Talou, P.; Vogt, R.; Randrup, J.; ...

2018-01-24

Detailed information on the fission process can be inferred from the observation, modeling and theoretical understanding of prompt fission neutron and γ-ray observables. Beyond simple average quantities, the study of distributions and correlations in prompt data, e.g., multiplicity-dependent neutron and γ-ray spectra, angular distributions of the emitted particles, n -n, n - γ, and γ - γ correlations, can place stringent constraints on fission models and parameters that would otherwise be free to be tuned separately to represent individual fission observables. The FREYA and CGMF codes have been developed to follow the sequential emissions of prompt neutrons and γ raysmore » from the initial excited fission fragments produced right after scission. Both codes implement Monte Carlo techniques to sample initial fission fragment configurations in mass, charge and kinetic energy and sample probabilities of neutron and γ emission at each stage of the decay. This approach naturally leads to using simple but powerful statistical techniques to infer distributions and correlations among many observables and model parameters. The comparison of model calculations with experimental data provides a rich arena for testing various nuclear physics models such as those related to the nuclear structure and level densities of neutron-rich nuclei, the γ-ray strength functions of dipole and quadrupole transitions, the mechanism for dividing the excitation energy between the two nascent fragments near scission, and the mechanisms behind the production of angular momentum in the fragments, etc. Beyond the obvious interest from a fundamental physics point of view, such studies are also important for addressing data needs in various nuclear applications. The inclusion of the FREYA and CGMF codes into the MCNP6.2 and MCNPX - PoliMi transport codes, for instance, provides a new and powerful tool to simulate correlated fission events in neutron transport calculations important in nonproliferation, safeguards, nuclear energy, and defense programs. Here, this review provides an overview of the topic, starting from theoretical considerations of the fission process, with a focus on correlated signatures. It then explores the status of experimental correlated fission data and current efforts to address some of the known shortcomings. Numerical simulations employing the FREYA and CGMF codes are compared to experimental data for a wide range of correlated fission quantities. The inclusion of those codes into the MCNP6.2 and MCNPX - PoliMi transport codes is described and discussed in the context of relevant applications. The accuracy of the model predictions and their sensitivity to model assumptions and input parameters are discussed. Lastly, a series of important experimental and theoretical questions that remain unanswered are presented, suggesting a renewed effort to address these shortcomings.« less

Correlated prompt fission data in transport simulations

NASA Astrophysics Data System (ADS)

Talou, P.; Vogt, R.; Randrup, J.; Rising, M. E.; Pozzi, S. A.; Verbeke, J.; Andrews, M. T.; Clarke, S. D.; Jaffke, P.; Jandel, M.; Kawano, T.; Marcath, M. J.; Meierbachtol, K.; Nakae, L.; Rusev, G.; Sood, A.; Stetcu, I.; Walker, C.

2018-01-01

Detailed information on the fission process can be inferred from the observation, modeling and theoretical understanding of prompt fission neutron and γ-ray observables. Beyond simple average quantities, the study of distributions and correlations in prompt data, e.g., multiplicity-dependent neutron and γ-ray spectra, angular distributions of the emitted particles, n - n, n - γ, and γ - γ correlations, can place stringent constraints on fission models and parameters that would otherwise be free to be tuned separately to represent individual fission observables. The FREYA and CGMF codes have been developed to follow the sequential emissions of prompt neutrons and γ rays from the initial excited fission fragments produced right after scission. Both codes implement Monte Carlo techniques to sample initial fission fragment configurations in mass, charge and kinetic energy and sample probabilities of neutron and γ emission at each stage of the decay. This approach naturally leads to using simple but powerful statistical techniques to infer distributions and correlations among many observables and model parameters. The comparison of model calculations with experimental data provides a rich arena for testing various nuclear physics models such as those related to the nuclear structure and level densities of neutron-rich nuclei, the γ-ray strength functions of dipole and quadrupole transitions, the mechanism for dividing the excitation energy between the two nascent fragments near scission, and the mechanisms behind the production of angular momentum in the fragments, etc. Beyond the obvious interest from a fundamental physics point of view, such studies are also important for addressing data needs in various nuclear applications. The inclusion of the FREYA and CGMF codes into the MCNP6.2 and MCNPX - PoliMi transport codes, for instance, provides a new and powerful tool to simulate correlated fission events in neutron transport calculations important in nonproliferation, safeguards, nuclear energy, and defense programs. This review provides an overview of the topic, starting from theoretical considerations of the fission process, with a focus on correlated signatures. It then explores the status of experimental correlated fission data and current efforts to address some of the known shortcomings. Numerical simulations employing the FREYA and CGMF codes are compared to experimental data for a wide range of correlated fission quantities. The inclusion of those codes into the MCNP6.2 and MCNPX - PoliMi transport codes is described and discussed in the context of relevant applications. The accuracy of the model predictions and their sensitivity to model assumptions and input parameters are discussed. Finally, a series of important experimental and theoretical questions that remain unanswered are presented, suggesting a renewed effort to address these shortcomings.

Correlated prompt fission data in transport simulations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Talou, P.; Vogt, R.; Randrup, J.

Detailed information on the fission process can be inferred from the observation, modeling and theoretical understanding of prompt fission neutron and γ-ray observables. Beyond simple average quantities, the study of distributions and correlations in prompt data, e.g., multiplicity-dependent neutron and γ-ray spectra, angular distributions of the emitted particles, n -n, n - γ, and γ - γ correlations, can place stringent constraints on fission models and parameters that would otherwise be free to be tuned separately to represent individual fission observables. The FREYA and CGMF codes have been developed to follow the sequential emissions of prompt neutrons and γ raysmore » from the initial excited fission fragments produced right after scission. Both codes implement Monte Carlo techniques to sample initial fission fragment configurations in mass, charge and kinetic energy and sample probabilities of neutron and γ emission at each stage of the decay. This approach naturally leads to using simple but powerful statistical techniques to infer distributions and correlations among many observables and model parameters. The comparison of model calculations with experimental data provides a rich arena for testing various nuclear physics models such as those related to the nuclear structure and level densities of neutron-rich nuclei, the γ-ray strength functions of dipole and quadrupole transitions, the mechanism for dividing the excitation energy between the two nascent fragments near scission, and the mechanisms behind the production of angular momentum in the fragments, etc. Beyond the obvious interest from a fundamental physics point of view, such studies are also important for addressing data needs in various nuclear applications. The inclusion of the FREYA and CGMF codes into the MCNP6.2 and MCNPX - PoliMi transport codes, for instance, provides a new and powerful tool to simulate correlated fission events in neutron transport calculations important in nonproliferation, safeguards, nuclear energy, and defense programs. Here, this review provides an overview of the topic, starting from theoretical considerations of the fission process, with a focus on correlated signatures. It then explores the status of experimental correlated fission data and current efforts to address some of the known shortcomings. Numerical simulations employing the FREYA and CGMF codes are compared to experimental data for a wide range of correlated fission quantities. The inclusion of those codes into the MCNP6.2 and MCNPX - PoliMi transport codes is described and discussed in the context of relevant applications. The accuracy of the model predictions and their sensitivity to model assumptions and input parameters are discussed. Lastly, a series of important experimental and theoretical questions that remain unanswered are presented, suggesting a renewed effort to address these shortcomings.« less

Modeling of displacement damage in silicon carbide detectors resulting from neutron irradiation

NASA Astrophysics Data System (ADS)

Khorsandi, Behrooz

There is considerable interest in developing a power monitor system for Generation IV reactors (for instance GT-MHR). A new type of semiconductor radiation detector is under development based on silicon carbide (SiC) technology for these reactors. SiC has been selected as the semiconductor material due to its superior thermal-electrical-neutronic properties. Compared to Si, SiC is a radiation hard material; however, like Si, the properties of SiC are changed by irradiation by a large fluence of energetic neutrons, as a consequence of displacement damage, and that irradiation decreases the life-time of detectors. Predictions of displacement damage and the concomitant radiation effects are important for deciding where the SiC detectors should be placed. The purpose of this dissertation is to develop computer simulation methods to estimate the number of various defects created in SiC detectors, because of neutron irradiation, and predict at what positions of a reactor, SiC detectors could monitor the neutron flux with high reliability. The simulation modeling includes several well-known---and commercial---codes (MCNP5, TRIM, MARLOWE and VASP), and two kinetic Monte Carlo codes written by the author (MCASIC and DCRSIC). My dissertation will highlight the displacement damage that may happen in SiC detectors located in available positions in the OSURR, GT-MHR and IRIS. As extra modeling output data, the count rates of SiC for the specified locations are calculated. A conclusion of this thesis is SiC detectors that are placed in the thermal neutron region of a graphite moderator-reflector reactor have a chance to survive at least one reactor refueling cycle, while their count rates are acceptably high.

3D neutronic codes coupled with thermal-hydraulic system codes for PWR, and BWR and VVER reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Langenbuch, S.; Velkov, K.; Lizorkin, M.

1997-07-01

This paper describes the objectives of code development for coupling 3D neutronics codes with thermal-hydraulic system codes. The present status of coupling ATHLET with three 3D neutronics codes for VVER- and LWR-reactors is presented. After describing the basic features of the 3D neutronic codes BIPR-8 from Kurchatov-Institute, DYN3D from Research Center Rossendorf and QUABOX/CUBBOX from GRS, first applications of coupled codes for different transient and accident scenarios are presented. The need of further investigations is discussed.

The total kinetic energy release in the fast neutron-induced fission of 232Th

DOE Office of Scientific and Technical Information (OSTI.GOV)

King, Jonathan; Yanez, Ricardo; Loveland, Walter

Here, the post-emission total kinetic energy release (TKE) in the neutron-induced fission of 232Th was measured (using white spectrum neutrons from LANSCE) for neutron energies from E n=3 to 91MeV. In this energy range the average post-neutron total kinetic energy release decreases from 162.3±0.3 at E n=3 MeV to 154.9±0.3 MeV at E n=91 MeV. Analysis of the fission mass distributions indicates that the decrease in TKE with increasing neutron energy is a combination of increasing yields of symmetric fission (which has a lower associated TKE) and a decrease in the TKE release in asymmetric fission.

The total kinetic energy release in the fast neutron-induced fission of 232Th

DOE PAGES

King, Jonathan; Yanez, Ricardo; Loveland, Walter; ...

2017-12-15

Here, the post-emission total kinetic energy release (TKE) in the neutron-induced fission of 232Th was measured (using white spectrum neutrons from LANSCE) for neutron energies from E n=3 to 91MeV. In this energy range the average post-neutron total kinetic energy release decreases from 162.3±0.3 at E n=3 MeV to 154.9±0.3 MeV at E n=91 MeV. Analysis of the fission mass distributions indicates that the decrease in TKE with increasing neutron energy is a combination of increasing yields of symmetric fission (which has a lower associated TKE) and a decrease in the TKE release in asymmetric fission.

Neutronic calculation of fast reactors by the EUCLID/V1 integrated code

NASA Astrophysics Data System (ADS)

Koltashev, D. A.; Stakhanova, A. A.

2017-01-01

This article considers neutronic calculation of a fast-neutron lead-cooled reactor BREST-OD-300 by the EUCLID/V1 integrated code. The main goal of development and application of integrated codes is a nuclear power plant safety justification. EUCLID/V1 is integrated code designed for coupled neutronics, thermomechanical and thermohydraulic fast reactor calculations under normal and abnormal operating conditions. EUCLID/V1 code is being developed in the Nuclear Safety Institute of the Russian Academy of Sciences. The integrated code has a modular structure and consists of three main modules: thermohydraulic module HYDRA-IBRAE/LM/V1, thermomechanical module BERKUT and neutronic module DN3D. In addition, the integrated code includes databases with fuel, coolant and structural materials properties. Neutronic module DN3D provides full-scale simulation of neutronic processes in fast reactors. Heat sources distribution, control rods movement, reactivity level changes and other processes can be simulated. Neutron transport equation in multigroup diffusion approximation is solved. This paper contains some calculations implemented as a part of EUCLID/V1 code validation. A fast-neutron lead-cooled reactor BREST-OD-300 transient simulation (fuel assembly floating, decompression of passive feedback system channel) and cross-validation with MCU-FR code results are presented in this paper. The calculations demonstrate EUCLID/V1 code application for BREST-OD-300 simulating and safety justification.

Real Time Computation of Kinetic Constraints to Support Equilibrium Reconstruction

NASA Astrophysics Data System (ADS)

Eggert, W. J.; Kolemen, E.; Eldon, D.

2016-10-01

A new method for quickly and automatically applying kinetic constraints to EFIT equilibrium reconstructions using readily available data is presented. The ultimate goal is to produce kinetic equilibrium reconstructions in real time and use them to constrain the DCON stability code as part of a disruption avoidance scheme. A first effort presented here replaces CPU-time expensive modules, such as the fast ion pressure profile calculation, with a simplified model. We show with a DIII-D database analysis that we can achieve reasonable predictions for selected applications by modeling the fast ion pressure profile and determining the fit parameters as functions of easily measured quantities including neutron rate and electron temperature on axis. Secondly, we present a strategy for treating Thomson scattering and Charge Exchange Recombination data to automatically form constraints for a kinetic equilibrium reconstruction, a process that historically was performed by hand. Work supported by US DOE DE-AC02-09CH11466 and DE-FC02-04ER54698.

Development of high-fidelity multiphysics system for light water reactor analysis

NASA Astrophysics Data System (ADS)

Magedanz, Jeffrey W.

There has been a tendency in recent years toward greater heterogeneity in reactor cores, due to the use of mixed-oxide (MOX) fuel, burnable absorbers, and longer cycles with consequently higher fuel burnup. The resulting asymmetry of the neutron flux and energy spectrum between regions with different compositions causes a need to account for the directional dependence of the neutron flux, instead of the traditional diffusion approximation. Furthermore, the presence of both MOX and high-burnup fuel in the core increases the complexity of the heat conduction. The heat transfer properties of the fuel pellet change with irradiation, and the thermal and mechanical expansion of the pellet and cladding strongly affect the size of the gap between them, and its consequent thermal resistance. These operational tendencies require higher fidelity multi-physics modeling capabilities, and this need is addressed by the developments performed within this PhD research. The dissertation describes the development of a High-Fidelity Multi-Physics System for Light Water Reactor Analysis. It consists of three coupled codes -- CTF for Thermal Hydraulics, TORT-TD for Neutron Kinetics, and FRAPTRAN for Fuel Performance. It is meant to address these modeling challenges in three ways: (1) by resolving the state of the system at the level of each fuel pin, rather than homogenizing entire fuel assemblies, (2) by using the multi-group Discrete Ordinates method to account for the directional dependence of the neutron flux, and (3) by using a fuel-performance code, rather than a Thermal Hydraulics code's simplified fuel model, to account for the material behavior of the fuel and its feedback to the hydraulic and neutronic behavior of the system. While the first two are improvements, the third, the use of a fuel-performance code for feedback, constitutes an innovation in this PhD project. Also important to this work is the manner in which such coupling is written. While coupling involves combining codes into a single executable, they are usually still developed and maintained separately. It should thus be a design objective to minimize the changes to those codes, and keep the changes to each code free of dependence on the details of the other codes. This will ease the incorporation of new versions of the code into the coupling, as well as re-use of parts of the coupling to couple with different codes. In order to fulfill this objective, an interface for each code was created in the form of an object-oriented abstract data type. Object-oriented programming is an effective method for enforcing a separation between different parts of a program, and clarifying the communication between them. The interfaces enable the main program to control the codes in terms of high-level functionality. This differs from the established practice of a master/slave relationship, in which the slave code is incorporated into the master code as a set of subroutines. While this PhD research continues previous work with a coupling between CTF and TORT-TD, it makes two major original contributions: (1) using a fuel-performance code, instead of a thermal-hydraulics code's simplified built-in models, to model the feedback from the fuel rods, and (2) the design of an object-oriented interface as an innovative method to interact with a coupled code in a high-level, easily-understandable manner. The resulting code system will serve as a tool to study the question of under what conditions, and to what extent, these higher-fidelity methods will provide benefits to reactor core analysis. (Abstract shortened by UMI.)

An information theoretic approach to use high-fidelity codes to calibrate low-fidelity codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lewis, Allison, E-mail: lewis.allison10@gmail.com; Smith, Ralph; Williams, Brian

For many simulation models, it can be prohibitively expensive or physically infeasible to obtain a complete set of experimental data to calibrate model parameters. In such cases, one can alternatively employ validated higher-fidelity codes to generate simulated data, which can be used to calibrate the lower-fidelity code. In this paper, we employ an information-theoretic framework to determine the reduction in parameter uncertainty that is obtained by evaluating the high-fidelity code at a specific set of design conditions. These conditions are chosen sequentially, based on the amount of information that they contribute to the low-fidelity model parameters. The goal is tomore » employ Bayesian experimental design techniques to minimize the number of high-fidelity code evaluations required to accurately calibrate the low-fidelity model. We illustrate the performance of this framework using heat and diffusion examples, a 1-D kinetic neutron diffusion equation, and a particle transport model, and include initial results from the integration of the high-fidelity thermal-hydraulics code Hydra-TH with a low-fidelity exponential model for the friction correlation factor.« less

Evaluating the performance of two neutron spectrum unfolding codes based on iterative procedures and artificial neural networks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ortiz-Rodriguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.

In this work the performance of two neutron spectrum unfolding codes based on iterative procedures and artificial neural networks is evaluated. The first one code based on traditional iterative procedures and called Neutron spectrometry and dosimetry from the Universidad Autonoma de Zacatecas (NSDUAZ) use the SPUNIT iterative algorithm and was designed to unfold neutron spectrum and calculate 15 dosimetric quantities and 7 IAEA survey meters. The main feature of this code is the automated selection of the initial guess spectrum trough a compendium of neutron spectrum compiled by the IAEA. The second one code known as Neutron spectrometry and dosimetrymore » with artificial neural networks (NDSann) is a code designed using neural nets technology. The artificial intelligence approach of neural net does not solve mathematical equations. By using the knowledge stored at synaptic weights on a neural net properly trained, the code is capable to unfold neutron spectrum and to simultaneously calculate 15 dosimetric quantities, needing as entrance data, only the rate counts measured with a Bonner spheres system. Similarities of both NSDUAZ and NSDann codes are: they follow the same easy and intuitive user's philosophy and were designed in a graphical interface under the LabVIEW programming environment. Both codes unfold the neutron spectrum expressed in 60 energy bins, calculate 15 dosimetric quantities and generate a full report in HTML format. Differences of these codes are: NSDUAZ code was designed using classical iterative approaches and needs an initial guess spectrum in order to initiate the iterative procedure. In NSDUAZ, a programming routine was designed to calculate 7 IAEA instrument survey meters using the fluence-dose conversion coefficients. NSDann code use artificial neural networks for solving the ill-conditioned equation system of neutron spectrometry problem through synaptic weights of a properly trained neural network. Contrary to iterative procedures, in neural net approach it is possible to reduce the rate counts used to unfold the neutron spectrum. To evaluate these codes a computer tool called Neutron Spectrometry and dosimetry computer tool was designed. The results obtained with this package are showed. The codes here mentioned are freely available upon request to the authors.« less

Evaluating the performance of two neutron spectrum unfolding codes based on iterative procedures and artificial neural networks

NASA Astrophysics Data System (ADS)

Ortiz-Rodríguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Solís Sánches, L. O.; Miranda, R. Castañeda; Cervantes Viramontes, J. M.; Vega-Carrillo, H. R.

2013-07-01

In this work the performance of two neutron spectrum unfolding codes based on iterative procedures and artificial neural networks is evaluated. The first one code based on traditional iterative procedures and called Neutron spectrometry and dosimetry from the Universidad Autonoma de Zacatecas (NSDUAZ) use the SPUNIT iterative algorithm and was designed to unfold neutron spectrum and calculate 15 dosimetric quantities and 7 IAEA survey meters. The main feature of this code is the automated selection of the initial guess spectrum trough a compendium of neutron spectrum compiled by the IAEA. The second one code known as Neutron spectrometry and dosimetry with artificial neural networks (NDSann) is a code designed using neural nets technology. The artificial intelligence approach of neural net does not solve mathematical equations. By using the knowledge stored at synaptic weights on a neural net properly trained, the code is capable to unfold neutron spectrum and to simultaneously calculate 15 dosimetric quantities, needing as entrance data, only the rate counts measured with a Bonner spheres system. Similarities of both NSDUAZ and NSDann codes are: they follow the same easy and intuitive user's philosophy and were designed in a graphical interface under the LabVIEW programming environment. Both codes unfold the neutron spectrum expressed in 60 energy bins, calculate 15 dosimetric quantities and generate a full report in HTML format. Differences of these codes are: NSDUAZ code was designed using classical iterative approaches and needs an initial guess spectrum in order to initiate the iterative procedure. In NSDUAZ, a programming routine was designed to calculate 7 IAEA instrument survey meters using the fluence-dose conversion coefficients. NSDann code use artificial neural networks for solving the ill-conditioned equation system of neutron spectrometry problem through synaptic weights of a properly trained neural network. Contrary to iterative procedures, in neural net approach it is possible to reduce the rate counts used to unfold the neutron spectrum. To evaluate these codes a computer tool called Neutron Spectrometry and dosimetry computer tool was designed. The results obtained with this package are showed. The codes here mentioned are freely available upon request to the authors.

Production mechanism of new neutron-rich heavy nuclei in the 136Xe +198Pt reaction

NASA Astrophysics Data System (ADS)

Li, Cheng; Wen, Peiwei; Li, Jingjing; Zhang, Gen; Li, Bing; Xu, Xinxin; Liu, Zhong; Zhu, Shaofei; Zhang, Feng-Shou

2018-01-01

The multinucleon transfer reaction of 136Xe +198Pt at Elab = 7.98 MeV/nucleon is investigated by using the improved quantum molecular dynamics model. The quasielastic, deep-inelastic, and quasifission collision mechanisms are studied via analyzing the angular distributions of fragments and the energy dissipation processes during the collisions. The measured isotope production cross sections of projectile-like fragments are reasonably well reproduced by the calculation of the ImQMD model together with the GEMINI code. The isotope production cross sections for the target-like fragments and double differential cross sections of 199Pt, 203Pt, and 208Pt are calculated. It is shown that about 50 new neutron-rich heavy nuclei can be produced via deep-inelastic collision mechanism, where the production cross sections are from 10-3 to 10-6 mb. The corresponding emission angle and the kinetic energy for these new neutron-rich nuclei locate at 40∘-60∘ and 100-200 MeV, respectively.

Neutrons Flux Distributions of the Pu-Be Source and its Simulation by the MCNP-4B Code

NASA Astrophysics Data System (ADS)

Faghihi, F.; Mehdizadeh, S.; Hadad, K.

Neutron Fluence rate of a low intense Pu-Be source is measured by Neutron Activation Analysis (NAA) of 197Au foils. Also, the neutron fluence rate distribution versus energy is calculated using the MCNP-4B code based on ENDF/B-V library. Theoretical simulation as well as our experimental performance are a new experience for Iranians to make reliability with the code for further researches. In our theoretical investigation, an isotropic Pu-Be source with cylindrical volume distribution is simulated and relative neutron fluence rate versus energy is calculated using MCNP-4B code. Variation of the fast and also thermal neutrons fluence rate, which are measured by NAA method and MCNP code, are compared.

Overview and Current Status of Analyses of Potential LEU Design Concepts for TREAT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Connaway, H. M.; Kontogeorgakos, D. C.; Papadias, D. D.

2015-10-01

Neutronic and thermal-hydraulic analyses have been performed to evaluate the performance of different low-enriched uranium (LEU) fuel design concepts for the conversion of the Transient Reactor Test Facility (TREAT) from its current high-enriched uranium (HEU) fuel. TREAT is an experimental reactor developed to generate high neutron flux transients for the testing of nuclear fuels. The goal of this work was to identify an LEU design which can maintain the performance of the existing HEU core while continuing to operate safely. A wide variety of design options were considered, with a focus on minimizing peak fuel temperatures and optimizing the powermore » coupling between the TREAT core and test samples. Designs were also evaluated to ensure that they provide sufficient reactivity and shutdown margin for each control rod bank. Analyses were performed using the core loading and experiment configuration of historic M8 Power Calibration experiments (M8CAL). The Monte Carlo code MCNP was utilized for steady-state analyses, and transient calculations were performed with the point kinetics code TREKIN. Thermal analyses were performed with the COMSOL multi-physics code. Using the results of this study, a new LEU Baseline design concept is being established, which will be evaluated in detail in a future report.« less

Neutron detection using the superconducting Nb-based current-biased kinetic inductance detector

NASA Astrophysics Data System (ADS)

Shishido, Hiroaki; Yamaguchi, Hiroyuki; Miki, Yuya; Miyajima, Shigeyuki; Oikawa, Kenichi; Harada, Masahide; Hidaka, Mutsuo; Oku, Takayuki; Arai, Masatoshi; Fujimaki, Akira; Ishida, Takekazu

2017-09-01

We demonstrate neutron detection using a solid-state 3He-free superconducting current-biased kinetic inductance detector (CB-KID), which consists of a superconducting Nb meander line and 10B neutron absorption layer. The CB-KID is based on the transient process of kinetic inductance of Cooper pairs induced by the nuclear reaction between 10B and neutrons. Therefore, the CB-KID can be operated in a wide superconducting region in the bias current-temperature diagram, as demonstrated in this paper. The transient change of the kinetic inductance induces the electromagnetic wave pulse under a DC bias current. The signal propagates along the meander line toward both sides with opposite polarity, where the signal polarity is dominated by the bias current direction. The full width at half maximum of the signals remains on the order of a few tens of ns, which confirms the high-speed operation of our detectors. We determine the neutron incident position within 1.3 mm accuracy in one dimension using the multichannel CB-KIDs.

Neutron lifetimes behavior analysis considering the two-region kinetic model in the IPEN/MB-01 reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gonnelli, Eduardo; Diniz, Ricardo

2014-11-11

This is a complementary work about the behavior analysis of the neutron lifetimes that was developed in the IPEN/MB-01 nuclear reactor facility. The macroscopic neutron noise technique was experimentally employed using pulse mode detectors for two stages of control rods insertion, where a total of twenty levels of subcriticality have been carried out. It was also considered that the neutron reflector density was treated as an additional group of delayed neutrons, being a sophisticated approach in the two-region kinetic theoretical model.

Gravitational wave content and stability of uniformly, rotating, triaxial neutron stars in general relativity.

PubMed

Tsokaros, Antonios; Ruiz, Milton; Paschalidis, Vasileios; Shapiro, Stuart L; Baiotti, Luca; Uryū, Kōji

2017-06-15

Targets for ground-based gravitational wave interferometers include continuous, quasiperiodic sources of gravitational radiation, such as isolated, spinning neutron stars. In this work, we perform evolution simulations of uniformly rotating, triaxially deformed stars, the compressible analogs in general relativity of incompressible, Newtonian Jacobi ellipsoids. We investigate their stability and gravitational wave emission. We employ five models, both normal and supramassive, and track their evolution with different grid setups and resolutions, as well as with two different evolution codes. We find that all models are dynamically stable and produce a strain that is approximately one-tenth the average value of a merging binary system. We track their secular evolution and find that all our stars evolve toward axisymmetry, maintaining their uniform rotation, rotational kinetic energy, and angular momentum profiles while losing their triaxiality.

TEMPEST II--A NEUTRON THERMALIZATION CODE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shudde, R.H.; Dyer, J.

The TEMPEST II neutron thermalization code in Fortran for IBM 709 or 7090 calculates thermal neutron flux spectra based upon the Wigner-Wilkins equation, the Wilkins equation, or the Maxwellian distribution. When a neutron spectrum is obtained, TEMPEST II provides microscopic and macroscopic cross section averages over that spectrum. Equations used by the code and sample input and output data are given. (auth)

TORT/MCNP coupling method for the calculation of neutron flux around a core of BWR.

PubMed

Kurosawa, Masahiko

2005-01-01

For the analysis of BWR neutronics performance, accurate data are required for neutron flux distribution over the In-Reactor Pressure Vessel equipments taking into account the detailed geometrical arrangement. The TORT code can calculate neutron flux around a core of BWR in a three-dimensional geometry model, but has difficulties in fine geometrical modelling and lacks huge computer resource. On the other hand, the MCNP code enables the calculation of the neutron flux with a detailed geometry model, but requires very long sampling time to give enough number of particles. Therefore, a TORT/MCNP coupling method has been developed to eliminate the two problems mentioned above in each code. In this method, the TORT code calculates angular flux distribution on the core surface and the MCNP code calculates neutron spectrum at the points of interest using the flux distribution. The coupling method will be used as the DOT-DOMINO-MORSE code system. This TORT/MCNP coupling method was applied to calculate the neutron flux at points where induced radioactivity data were measured for 54Mn and 60Co and the radioactivity calculations based on the neutron flux obtained from the above method were compared with the measured data.

Delayed Neutrons Effect on Power Reactor with Variation of Fluid Fuel Velocity at MSR Fuji-12

NASA Astrophysics Data System (ADS)

Kuncoro Aji, Indarta; Pramuditya, Syeilendra; Novitrian; Irwanto, Dwi; Waris, Abdul

2017-01-01

As the nuclear reactor operate with liquid fuel, controlling velocity of the fuel flow on the Molten salt reactor very influence on the neutron kinetics in that reactor system. The effect of the pace fuel changes to the populations number of neutrons and power density on vertical direction (1 dimension) from the first until fifth year reactor operating had been analyzed on this research. This research had been conducted on MSR Fuji-12 with a two meters core high, and LiF-BeF2-ThF4-233UF4 as fuel composition respectively 71.78%-16%-11.86%-0.36%. Data of reactivity, neutron flux, and the macroscopic fission cross section obtained from ouput of SRAC (neutronic calculation code has been developed by JAEA, with JENDL-4.0 as data library on the SRAC calculation) was being used for the calculation process of this research. The calculation process of this research had been performed numerically by SOR (successive over relaxation) and finite difference methode, as well as using C programing language. From the calculation, regarding to the value of power density resulting from delayed neutrons, concluded that 20 m/s is the optimum fuel flow velocity in all the years reactor had operated. Where the increases number of power are inversely proportional with the fuel flow speed.

A comparison of the COG and MCNP codes in computational neutron capture therapy modeling, Part I: boron neutron capture therapy models.

PubMed

Culbertson, C N; Wangerin, K; Ghandourah, E; Jevremovic, T

2005-08-01

The goal of this study was to evaluate the COG Monte Carlo radiation transport code, developed and tested by Lawrence Livermore National Laboratory, for neutron capture therapy related modeling. A boron neutron capture therapy model was analyzed comparing COG calculational results to results from the widely used MCNP4B (Monte Carlo N-Particle) transport code. The approach for computing neutron fluence rate and each dose component relevant in boron neutron capture therapy is described, and calculated values are shown in detail. The differences between the COG and MCNP predictions are qualified and quantified. The differences are generally small and suggest that the COG code can be applied for BNCT research related problems.

PROTEUS-SN User Manual

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shemon, Emily R.; Smith, Micheal A.; Lee, Changho

2016-02-16

PROTEUS-SN is a three-dimensional, highly scalable, high-fidelity neutron transport code developed at Argonne National Laboratory. The code is applicable to all spectrum reactor transport calculations, particularly those in which a high degree of fidelity is needed either to represent spatial detail or to resolve solution gradients. PROTEUS-SN solves the second order formulation of the transport equation using the continuous Galerkin finite element method in space, the discrete ordinates approximation in angle, and the multigroup approximation in energy. PROTEUS-SN’s parallel methodology permits the efficient decomposition of the problem by both space and angle, permitting large problems to run efficiently on hundredsmore » of thousands of cores. PROTEUS-SN can also be used in serial or on smaller compute clusters (10’s to 100’s of cores) for smaller homogenized problems, although it is generally more computationally expensive than traditional homogenized methodology codes. PROTEUS-SN has been used to model partially homogenized systems, where regions of interest are represented explicitly and other regions are homogenized to reduce the problem size and required computational resources. PROTEUS-SN solves forward and adjoint eigenvalue problems and permits both neutron upscattering and downscattering. An adiabatic kinetics option has recently been included for performing simple time-dependent calculations in addition to standard steady state calculations. PROTEUS-SN handles void and reflective boundary conditions. Multigroup cross sections can be generated externally using the MC2-3 fast reactor multigroup cross section generation code or internally using the cross section application programming interface (API) which can treat the subgroup or resonance table libraries. PROTEUS-SN is written in Fortran 90 and also includes C preprocessor definitions. The code links against the PETSc, METIS, HDF5, and MPICH libraries. It optionally links against the MOAB library and is a part of the SHARP multi-physics suite for coupled multi-physics analysis of nuclear reactors. This user manual describes how to set up a neutron transport simulation with the PROTEUS-SN code. A companion methodology manual describes the theory and algorithms within PROTEUS-SN.« less

Kinetics of Thermal Neutrons in a Time-of-Flight Spectrometer. I. Probability of Transmission of Neutrons through a Revolving Slit; CENETICA DEI NEUTRONI LENTI IN UNO SPETTROMETRO A TEMPO DI VOLO. I-PROBABILITA DI TRANSMISSIONE DEI NEUTRONI ATTRAVERSO UNA FENDITURA RUOTANTE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marseguerra, M.; Pauli, G.

1958-07-01

The kinetic behavior of thermal neutrons in a time-offlight spectrometer is examined. An analytical method for obtaining the expressions for the probability for slow neutron transmission through a revolving slit (the general case of a curved slit is considered) is presented and discussed in detail. (auth)

Validating the performance of correlated fission multiplicity implementation in radiation transport codes with subcritical neutron multiplication benchmark experiments

DOE PAGES

Arthur, Jennifer; Bahran, Rian; Hutchinson, Jesson; ...

2018-06-14

Historically, radiation transport codes have uncorrelated fission emissions. In reality, the particles emitted by both spontaneous and induced fissions are correlated in time, energy, angle, and multiplicity. This work validates the performance of various current Monte Carlo codes that take into account the underlying correlated physics of fission neutrons, specifically neutron multiplicity distributions. The performance of 4 Monte Carlo codes - MCNP®6.2, MCNP®6.2/FREYA, MCNP®6.2/CGMF, and PoliMi - was assessed using neutron multiplicity benchmark experiments. In addition, MCNP®6.2 simulations were run using JEFF-3.2 and JENDL-4.0, rather than ENDF/B-VII.1, data for 239Pu and 240Pu. The sensitive benchmark parameters that in this workmore » represent the performance of each correlated fission multiplicity Monte Carlo code include the singles rate, the doubles rate, leakage multiplication, and Feynman histograms. Although it is difficult to determine which radiation transport code shows the best overall performance in simulating subcritical neutron multiplication inference benchmark measurements, it is clear that correlations exist between the underlying nuclear data utilized by (or generated by) the various codes, and the correlated neutron observables of interest. This could prove useful in nuclear data validation and evaluation applications, in which a particular moment of the neutron multiplicity distribution is of more interest than the other moments. It is also quite clear that, because transport is handled by MCNP®6.2 in 3 of the 4 codes, with the 4th code (PoliMi) being based on an older version of MCNP®, the differences in correlated neutron observables of interest are most likely due to the treatment of fission event generation in each of the different codes, as opposed to the radiation transport.« less

Validating the performance of correlated fission multiplicity implementation in radiation transport codes with subcritical neutron multiplication benchmark experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Arthur, Jennifer; Bahran, Rian; Hutchinson, Jesson

Historically, radiation transport codes have uncorrelated fission emissions. In reality, the particles emitted by both spontaneous and induced fissions are correlated in time, energy, angle, and multiplicity. This work validates the performance of various current Monte Carlo codes that take into account the underlying correlated physics of fission neutrons, specifically neutron multiplicity distributions. The performance of 4 Monte Carlo codes - MCNP®6.2, MCNP®6.2/FREYA, MCNP®6.2/CGMF, and PoliMi - was assessed using neutron multiplicity benchmark experiments. In addition, MCNP®6.2 simulations were run using JEFF-3.2 and JENDL-4.0, rather than ENDF/B-VII.1, data for 239Pu and 240Pu. The sensitive benchmark parameters that in this workmore » represent the performance of each correlated fission multiplicity Monte Carlo code include the singles rate, the doubles rate, leakage multiplication, and Feynman histograms. Although it is difficult to determine which radiation transport code shows the best overall performance in simulating subcritical neutron multiplication inference benchmark measurements, it is clear that correlations exist between the underlying nuclear data utilized by (or generated by) the various codes, and the correlated neutron observables of interest. This could prove useful in nuclear data validation and evaluation applications, in which a particular moment of the neutron multiplicity distribution is of more interest than the other moments. It is also quite clear that, because transport is handled by MCNP®6.2 in 3 of the 4 codes, with the 4th code (PoliMi) being based on an older version of MCNP®, the differences in correlated neutron observables of interest are most likely due to the treatment of fission event generation in each of the different codes, as opposed to the radiation transport.« less

FY17 Status Report on NEAMS Neutronics Activities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, C. H.; Jung, Y. S.; Smith, M. A.

2017-09-30

Under the U.S. DOE NEAMS program, the high-fidelity neutronics code system has been developed to support the multiphysics modeling and simulation capability named SHARP. The neutronics code system includes the high-fidelity neutronics code PROTEUS, the cross section library and preprocessing tools, the multigroup cross section generation code MC2-3, the in-house meshing generation tool, the perturbation and sensitivity analysis code PERSENT, and post-processing tools. The main objectives of the NEAMS neutronics activities in FY17 are to continue development of an advanced nodal solver in PROTEUS for use in nuclear reactor design and analysis projects, implement a simplified sub-channel based thermal-hydraulic (T/H)more » capability into PROTEUS to efficiently compute the thermal feedback, improve the performance of PROTEUS-MOCEX using numerical acceleration and code optimization, improve the cross section generation tools including MC2-3, and continue to perform verification and validation tests for PROTEUS.« less

Comprehensive approach to fast ion measurements in the beam-driven FRC

NASA Astrophysics Data System (ADS)

Magee, Richard; Smirnov, Artem; Onofri, Marco; Dettrick, Sean; Korepanov, Sergey; Knapp, Kurt; the TAE Team

2015-11-01

The C-2U experiment combines tangential neutral beam injection, edge biasing, and advanced recycling control to explore the sustainment of field-reversed configuration (FRC) plasmas. To study fast ion confinement in such advanced, beam-driven FRCs, a synergetic technique was developed that relies on the measurements of the DD fusion reaction products and the hybrid code Q2D, which treats the plasma as a fluid and the fast ions kinetically. Data from calibrated neutron and proton detectors are used in a complementary fashion to constrain the simulations: neutron detectors measure the volume integrated fusion rate to constrain the total number of fast ions, while proton detectors with multiple lines of sight through the plasma constrain the axial profile of fast ions. One application of this technique is the diagnosis of fast ion energy transfer and pitch angle scattering. A parametric numerical study was conducted, in which additional ad hoc loss and scattering terms of varying strengths were introduced in the code and constrained with measurement. Initial results indicate that the energy transfer is predominantly classical, while, in some cases, non-classical pitch angle scattering can be observed.

GUINEVERE experiment: Kinetic analysis of some reactivity measurement methods by deterministic and Monte Carlo codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bianchini, G.; Burgio, N.; Carta, M.

The GUINEVERE experiment (Generation of Uninterrupted Intense Neutrons at the lead Venus Reactor) is an experimental program in support of the ADS technology presently carried out at SCK-CEN in Mol (Belgium). In the experiment a modified lay-out of the original thermal VENUS critical facility is coupled to an accelerator, built by the French body CNRS in Grenoble, working in both continuous and pulsed mode and delivering 14 MeV neutrons by bombardment of deuterons on a tritium-target. The modified lay-out of the facility consists of a fast subcritical core made of 30% U-235 enriched metallic Uranium in a lead matrix. Severalmore » off-line and on-line reactivity measurement techniques will be investigated during the experimental campaign. This report is focused on the simulation by deterministic (ERANOS French code) and Monte Carlo (MCNPX US code) calculations of three reactivity measurement techniques, Slope ({alpha}-fitting), Area-ratio and Source-jerk, applied to a GUINEVERE subcritical configuration (namely SC1). The inferred reactivity, in dollar units, by the Area-ratio method shows an overall agreement between the two deterministic and Monte Carlo computational approaches, whereas the MCNPX Source-jerk results are affected by large uncertainties and allow only partial conclusions about the comparison. Finally, no particular spatial dependence of the results is observed in the case of the GUINEVERE SC1 subcritical configuration. (authors)« less

A neutron spectrum unfolding computer code based on artificial neural networks

NASA Astrophysics Data System (ADS)

Ortiz-Rodríguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Cervantes Viramontes, J. M.; Vega-Carrillo, H. R.

2014-02-01

The Bonner Spheres Spectrometer consists of a thermal neutron sensor placed at the center of a number of moderating polyethylene spheres of different diameters. From the measured readings, information can be derived about the spectrum of the neutron field where measurements were made. Disadvantages of the Bonner system are the weight associated with each sphere and the need to sequentially irradiate the spheres, requiring long exposure periods. Provided a well-established response matrix and adequate irradiation conditions, the most delicate part of neutron spectrometry, is the unfolding process. The derivation of the spectral information is not simple because the unknown is not given directly as a result of the measurements. The drawbacks associated with traditional unfolding procedures have motivated the need of complementary approaches. Novel methods based on Artificial Intelligence, mainly Artificial Neural Networks, have been widely investigated. In this work, a neutron spectrum unfolding code based on neural nets technology is presented. This code is called Neutron Spectrometry and Dosimetry with Artificial Neural networks unfolding code that was designed in a graphical interface. The core of the code is an embedded neural network architecture previously optimized using the robust design of artificial neural networks methodology. The main features of the code are: easy to use, friendly and intuitive to the user. This code was designed for a Bonner Sphere System based on a 6LiI(Eu) neutron detector and a response matrix expressed in 60 energy bins taken from an International Atomic Energy Agency compilation. The main feature of the code is that as entrance data, for unfolding the neutron spectrum, only seven rate counts measured with seven Bonner spheres are required; simultaneously the code calculates 15 dosimetric quantities as well as the total flux for radiation protection purposes. This code generates a full report with all information of the unfolding in the HTML format. NSDann unfolding code is freely available, upon request to the authors.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gorelenkov, N. N.; Heidbrink, W. W.; Kramer, G. J.

The redistribution and potential loss of energetic particles due to MHD modes can limit the performance of fusion plasmas by reducing the plasma heating rate. In this work, we present validation studies of the 1.5D critical gradient model (CGM) for Alfvén eigenmode (AE) induced EP transport in NSTX and DIII-D neutral beam heated plasmas. In previous comparisons with a single DIII-D L-mode case, the CGM model was found to be responsible for 75% of measured AE induced neutron deficit [1]. A fully kinetic HINST is used to compute mode stability for the non-perturbative version of CGM (or nCGM). We have found that AEs show strong local instability drive up tomore » $$\\gamma /\\omega \\sim 20\\%$$ violating assumptions of perturbative approaches used in NOVA-K code. Lastly, we demonstrate that both models agree with each other and both underestimate the neutron deficit measured in DIII-D shot by approximately a factor of 2.« less

Preliminary study on new configuration with LEU fuel assemblies for the Dalat nuclear research reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Van Lam Pham; Vinh Vinh Le; Ton Nghiem Huynh

2008-07-15

The fuel conversion of the Dalat Nuclear Research Reactor (DNRR) is being realized. The DNRR is a pool type research reactor which was reconstructed from the 250 kW TRIGA- MARK II reactor. The reconstructed reactor attained its nominal power of 500 kW in February 1984. According to the results of design and safety analyses performed by the joint study between RERTR Program at Argonne National Laboratory (ANL) and Vietnam Atomic Energy Commission (VAEC) the mixed core of irradiated HEU and new LEU WWR-M2 fuel assemblies will be created soon. This paper presents the results of preliminary study on new configurationmore » with only LEU fuel assemblies for the DNRR. The codes MCNP, REBUS and VARI3D are used to calculate neutron flux performance in irradiation positions and kinetics parameters. The idea of change of Beryllium rod reloading enables to get working configuration assured shutdown margin, thermal-hydraulic safety and increase in thermal neutron flux in neutron trap at the center of DNRR active core. (author)« less

Load Designs For MJ Dense Plasma Foci

NASA Astrophysics Data System (ADS)

Link, A.; Povlius, A.; Anaya, R.; Anderson, M. G.; Angus, J. R.; Cooper, C. M.; Falabella, S.; Goerz, D.; Higginson, D.; Holod, I.; McMahon, M.; Mitrani, J.; Koh, E. S.; Pearson, A.; Podpaly, Y. A.; Prasad, R.; van Lue, D.; Watson, J.; Schmidt, A. E.

2017-10-01

Dense plasma focus (DPF) Z-pinches are compact pulse power driven devices with coaxial electrodes. The discharge of DPF consists of three distinct phases: first generation of a plasma sheath, plasma rail gun phase where the sheath is accelerated down the electrodes and finally an implosion phase where the plasma stagnates into a z-pinch geometry. During the z-pinch phase, DPFs can produce MeV ion beams, x-rays and neutrons. Megaampere class DPFs with deuterium fills have demonstrated neutron yields in the 1012 neutrons/shot range with pulse durations of 10-100 ns. Kinetic simulations using the code Chicago are being used to evaluate various load configurations from initial sheath formation to the final z-pinch phase for DPFs with up to 5 MA and 1 MJ coupled to the load. Results will be presented from the preliminary design simulations. LLNL-ABS-734785 This work performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory (LLNL) under Contract DE-AC52-07NA27344 and with support from the Computing Grand Challenge program at LLNL.

Verification and Validation of Monte Carlo n-Particle Code 6 (MCNP6) with Neutron Protection Factor Measurements of an Iron Box

DTIC Science & Technology

2014-03-27

VERIFICATION AND VALIDATION OF MONTE CARLO N- PARTICLE CODE 6 (MCNP6) WITH NEUTRON PROTECTION FACTOR... PARTICLE CODE 6 (MCNP6) WITH NEUTRON PROTECTION FACTOR MEASUREMENTS OF AN IRON BOX THESIS Presented to the Faculty Department of Engineering...STATEMENT A. APPROVED FOR PUBLIC RELEASE; DISTRIBUTION UNLIMITED iv AFIT-ENP-14-M-05 VERIFICATION AND VALIDATION OF MONTE CARLO N- PARTICLE CODE 6

Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core.

PubMed

Lashkari, A; Khalafi, H; Kazeminejad, H

2013-05-01

In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR_PC package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change.

Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core

PubMed Central

Lashkari, A.; Khalafi, H.; Kazeminejad, H.

2013-01-01

In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR_PC package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change. PMID:24976672

Developmental assessment of the Fort St. Vrain version of the Composite HTGR Analysis Program (CHAP-2)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stroh, K.R.

1980-01-01

The Composite HTGR Analysis Program (CHAP) consists of a model-independent systems analysis mainframe named LASAN and model-dependent linked code modules, each representing a component, subsystem, or phenomenon of an HTGR plant. The Fort St. Vrain (FSV) version (CHAP-2) includes 21 coded modules that model the neutron kinetics and thermal response of the core; the thermal-hydraulics of the reactor primary coolant system, secondary steam supply system, and balance-of-plant; the actions of the control system and plant protection system; the response of the reactor building; and the relative hazard resulting from fuel particle failure. FSV steady-state and transient plant data are beingmore » used to partially verify the component modeling and dynamic smulation techniques used to predict plant response to postulated accident sequences.« less

Development of MCNPX-ESUT computer code for simulation of neutron/gamma pulse height distribution

NASA Astrophysics Data System (ADS)

Abolfazl Hosseini, Seyed; Vosoughi, Naser; Zangian, Mehdi

2015-05-01

In this paper, the development of the MCNPX-ESUT (MCNPX-Energy Engineering of Sharif University of Technology) computer code for simulation of neutron/gamma pulse height distribution is reported. Since liquid organic scintillators like NE-213 are well suited and routinely used for spectrometry in mixed neutron/gamma fields, this type of detectors is selected for simulation in the present study. The proposed algorithm for simulation includes four main steps. The first step is the modeling of the neutron/gamma particle transport and their interactions with the materials in the environment and detector volume. In the second step, the number of scintillation photons due to charged particles such as electrons, alphas, protons and carbon nuclei in the scintillator material is calculated. In the third step, the transport of scintillation photons in the scintillator and lightguide is simulated. Finally, the resolution corresponding to the experiment is considered in the last step of the simulation. Unlike the similar computer codes like SCINFUL, NRESP7 and PHRESP, the developed computer code is applicable to both neutron and gamma sources. Hence, the discrimination of neutron and gamma in the mixed fields may be performed using the MCNPX-ESUT computer code. The main feature of MCNPX-ESUT computer code is that the neutron/gamma pulse height simulation may be performed without needing any sort of post processing. In the present study, the pulse height distributions due to a monoenergetic neutron/gamma source in NE-213 detector using MCNPX-ESUT computer code is simulated. The simulated neutron pulse height distributions are validated through comparing with experimental data (Gohil et al. Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, 664 (2012) 304-309.) and the results obtained from similar computer codes like SCINFUL, NRESP7 and Geant4. The simulated gamma pulse height distribution for a 137Cs source is also compared with the experimental data.

A Study of Neutron Leakage in Finite Objects

NASA Technical Reports Server (NTRS)

Wilson, John W.; Slaba, Tony C.; Badavi, Francis F.; Reddell, Brandon D.; Bahadori, Amir A.

2015-01-01

A computationally efficient 3DHZETRN code capable of simulating High charge (Z) and Energy (HZE) and light ions (including neutrons) under space-like boundary conditions with enhanced neutron and light ion propagation was recently developed for simple shielded objects. Monte Carlo (MC) benchmarks were used to verify the 3DHZETRN methodology in slab and spherical geometry, and it was shown that 3DHZETRN agrees with MC codes to the degree that various MC codes agree among themselves. One limitation in the verification process is that all of the codes (3DHZETRN and three MC codes) utilize different nuclear models/databases. In the present report, the new algorithm, with well-defined convergence criteria, is used to quantify the neutron leakage from simple geometries to provide means of verifying 3D effects and to provide guidance for further code development.

Simulations of neutron transport at low energy: a comparison between GEANT and MCNP.

PubMed

Colonna, N; Altieri, S

2002-06-01

The use of the simulation tool GEANT for neutron transport at energies below 20 MeV is discussed, in particular with regard to shielding and dose calculations. The reliability of the GEANT/MICAP package for neutron transport in a wide energy range has been verified by comparing the results of simulations performed with this package in a wide energy range with the prediction of MCNP-4B, a code commonly used for neutron transport at low energy. A reasonable agreement between the results of the two codes is found for the neutron flux through a slab of material (iron and ordinary concrete), as well as for the dose released in soft tissue by neutrons. These results justify the use of the GEANT/MICAP code for neutron transport in a wide range of applications, including health physics problems.

The CCONE Code System and its Application to Nuclear Data Evaluation for Fission and Other Reactions

NASA Astrophysics Data System (ADS)

Iwamoto, O.; Iwamoto, N.; Kunieda, S.; Minato, F.; Shibata, K.

2016-01-01

A computer code system, CCONE, was developed for nuclear data evaluation within the JENDL project. The CCONE code system integrates various nuclear reaction models needed to describe nucleon, light charged nuclei up to alpha-particle and photon induced reactions. The code is written in the C++ programming language using an object-oriented technology. At first, it was applied to neutron-induced reaction data on actinides, which were compiled into JENDL Actinide File 2008 and JENDL-4.0. It has been extensively used in various nuclear data evaluations for both actinide and non-actinide nuclei. The CCONE code has been upgraded to nuclear data evaluation at higher incident energies for neutron-, proton-, and photon-induced reactions. It was also used for estimating β-delayed neutron emission. This paper describes the CCONE code system indicating the concept and design of coding and inputs. Details of the formulation for modelings of the direct, pre-equilibrium and compound reactions are presented. Applications to the nuclear data evaluations such as neutron-induced reactions on actinides and medium-heavy nuclei, high-energy nucleon-induced reactions, photonuclear reaction and β-delayed neutron emission are mentioned.

The new double energy-velocity spectrometer VERDI

NASA Astrophysics Data System (ADS)

Jansson, Kaj; Frégeau, Marc Olivier; Al-Adili, Ali; Göök, Alf; Gustavsson, Cecilia; Hambsch, Franz-Josef; Oberstedt, Stephan; Pomp, Stephan

2017-09-01

VERDI (VElocity foR Direct particle Identification) is a fission-fragment spectrometer recently put into operation at JRC-Geel. It allows measuring the kinetic energy and velocity of both fission fragments simultaneously. The velocity provides information about the pre-neutron mass of each fission fragment when isotropic prompt-neutron emission from the fragments is assumed. The kinetic energy, in combination with the velocity, provides the post-neutron mass. From the difference between pre- and post-neutron masses, the number of neutrons emitted by each fragment can be determined. Multiplicity as a function of fragment mass and total kinetic energy is one important ingredient, essential for understanding the sharing of excitation energy between fission fragments at scission, and may be used to benchmark nuclear de-excitation models. The VERDI spectrometer design is a compromise between geometrical efficiency and mass resolution. The spectrometer consists of an electron detector located close to the target and two arrays of silicon detectors, each located 50 cm away from the target. In the present configuration pre-neutron and post-neutron mass distributions are in good agreement with reference data were obtained. Our latest measurements performed with spontaneously fissioning 252Cf is presented along with the developed calibration procedure to obtain pulse height defect and plasma delay time corrections.

Benchmark test of transport calculations of gold and nickel activation with implications for neutron kerma at Hiroshima.

PubMed

Hoshi, M; Hiraoka, M; Hayakawa, N; Sawada, S; Munaka, M; Kuramoto, A; Oka, T; Iwatani, K; Shizuma, K; Hasai, H

1992-11-01

A benchmark test of the Monte Carlo neutron and photon transport code system (MCNP) was performed using a 252Cf fission neutron source to validate the use of the code for the energy spectrum analyses of Hiroshima atomic bomb neutrons. Nuclear data libraries used in the Monte Carlo neutron and photon transport code calculation were ENDF/B-III, ENDF/B-IV, LASL-SUB, and ENDL-73. The neutron moderators used were granite (the main component of which is SiO2, with a small fraction of hydrogen), Newlight [polyethylene with 3.7% boron (natural)], ammonium chloride (NH4Cl), and water (H2O). Each moderator was 65 cm thick. The neutron detectors were gold and nickel foils, which were used to detect thermal and epithermal neutrons (4.9 eV) and fast neutrons (> 0.5 MeV), respectively. Measured activity data from neutron-irradiated gold and nickel foils in these moderators decreased to about 1/1,000th or 1/10,000th, which correspond to about 1,500 m ground distance from the hypocenter in Hiroshima. For both gold and nickel detectors, the measured activities and the calculated values agreed within 10%. The slopes of the depth-yield relations in each moderator, except granite, were similar for neutrons detected by the gold and nickel foils. From the results of these studies, the Monte Carlo neutron and photon transport code was verified to be accurate enough for use with the elements hydrogen, carbon, nitrogen, oxygen, silicon, chlorine, and cadmium, and for the incident 252Cf fission spectrum neutrons.

Measurement of absolute response functions and detection efficiencies of an NE213 scintillator up to 600 MeV

NASA Astrophysics Data System (ADS)

Kajimoto, Tsuyoshi; Shigyo, Nobuhiro; Sanami, Toshiya; Ishibashi, Kenji; Haight, Robert C.; Fotiades, Nikolaos

2011-02-01

Absolute neutron response functions and detection efficiencies of an NE213 liquid scintillator that was 12.7 cm in diameter and 12.7 cm in thickness were measured for neutron energies between 15 and 600 MeV at the Weapons Neutron Research facility of the Los Alamos Neutron Science Center. The experiment was performed with continuous-energy neutrons on a spallation neutron source by 800-MeV proton incidence. The incident neutron flux was measured using a 238U fission ionization chamber. Measured response functions and detection efficiencies were compared with corresponding calculations using the SCINFUL-QMD code. The calculated and experimental values were in good agreement for data below 70 MeV. However, there were discrepancies in the energy region between 70 and 150 MeV. Thus, the code was partly modified and the revised code provided better agreement with the experimental data.

Hexagonal Uniformly Redundant Arrays (HURAs) for scintillator based coded aperture neutron imaging

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gamage, K.A.A.; Zhou, Q.

2015-07-01

A series of Monte Carlo simulations have been conducted, making use of the EJ-426 neutron scintillator detector, to investigate the potential of using hexagonal uniformly redundant arrays (HURAs) for scintillator based coded aperture neutron imaging. This type of scintillator material has a low sensitivity to gamma rays, therefore, is of particular use in a system with a source that emits both neutrons and gamma rays. The simulations used an AmBe source, neutron images have been produced using different coded-aperture materials (boron- 10, cadmium-113 and gadolinium-157) and location error has also been estimated. In each case the neutron image clearly showsmore » the location of the source with a relatively small location error. Neutron images with high resolution can be easily used to identify and locate nuclear materials precisely in nuclear security and nuclear decommissioning applications. (authors)« less

Neutron flux measurements on a mock-up of a storage cask for high-level nuclear waste using 2.5 MeV neutrons.

PubMed

Suárez, H Saurí; Becker, F; Klix, A; Pang, B; Döring, T

2018-06-07

To store and dispose spent nuclear fuel, shielding casks are employed to reduce the emitted radiation. To evaluate the exposure of employees handling such casks, Monte Carlo radiation transport codes can be employed. Nevertheless, to assess the reliability of these codes and nuclear data, experimental checks are required. In this study, a neutron generator (NG) producing neutrons of 2.5 MeV was employed to simulate neutrons produced in spent nuclear fuel. Different configurations of shielding layers of steel and polyethylene were positioned between the target of the NG and a NE-213 detector. The results of the measurements of neutron and γ radiation and the corresponding simulations with the code MCNP6 are presented. Details of the experimental set-up as well as neutron and photon flux spectra are provided as reference points for such NG investigations with shielding structures.

A novel neutron energy spectrum unfolding code using particle swarm optimization

NASA Astrophysics Data System (ADS)

Shahabinejad, H.; Sohrabpour, M.

2017-07-01

A novel neutron Spectrum Deconvolution using Particle Swarm Optimization (SDPSO) code has been developed to unfold the neutron spectrum from a pulse height distribution and a response matrix. The Particle Swarm Optimization (PSO) imitates the bird flocks social behavior to solve complex optimization problems. The results of the SDPSO code have been compared with those of the standard spectra and recently published Two-steps Genetic Algorithm Spectrum Unfolding (TGASU) code. The TGASU code have been previously compared with the other codes such as MAXED, GRAVEL, FERDOR and GAMCD and shown to be more accurate than the previous codes. The results of the SDPSO code have been demonstrated to match well with those of the TGASU code for both under determined and over-determined problems. In addition the SDPSO has been shown to be nearly two times faster than the TGASU code.

Energy dynamics and current sheet structure in fluid and kinetic simulations of decaying magnetohydrodynamic turbulence

DOE Office of Scientific and Technical Information (OSTI.GOV)

Makwana, K. D., E-mail: kirit.makwana@gmx.com; Cattaneo, F.; Zhdankin, V.

Simulations of decaying magnetohydrodynamic (MHD) turbulence are performed with a fluid and a kinetic code. The initial condition is an ensemble of long-wavelength, counter-propagating, shear-Alfvén waves, which interact and rapidly generate strong MHD turbulence. The total energy is conserved and the rate of turbulent energy decay is very similar in both codes, although the fluid code has numerical dissipation, whereas the kinetic code has kinetic dissipation. The inertial range power spectrum index is similar in both the codes. The fluid code shows a perpendicular wavenumber spectral slope of k{sub ⊥}{sup −1.3}. The kinetic code shows a spectral slope of k{submore » ⊥}{sup −1.5} for smaller simulation domain, and k{sub ⊥}{sup −1.3} for larger domain. We estimate that collisionless damping mechanisms in the kinetic code can account for the dissipation of the observed nonlinear energy cascade. Current sheets are geometrically characterized. Their lengths and widths are in good agreement between the two codes. The length scales linearly with the driving scale of the turbulence. In the fluid code, their thickness is determined by the grid resolution as there is no explicit diffusivity. In the kinetic code, their thickness is very close to the skin-depth, irrespective of the grid resolution. This work shows that kinetic codes can reproduce the MHD inertial range dynamics at large scales, while at the same time capturing important kinetic physics at small scales.« less

OBJECT KINETIC MONTE CARLO SIMULATIONS OF RADIATION DAMAGE IN TUNGSTEN SUBJECTED TO NEUTRON FLUX WITH PKA SPECTRUM CORRESPONDING TO THE HFIR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nandipati, Giridhar; Setyawan, Wahyu; Heinisch, Howard L.

2015-12-31

The objective of this work is to study the damage accumulation in pure tungsten (W) subjected to neutron bombardment with a primary knock-on atom (PKA) spectrum corresponding to the High Flux Isotope Reactor (HFIR), using the object kinetic Monte Carlo (OKMC) method.

Application of the MCNPX-McStas interface for shielding calculations and guide design at ESS

NASA Astrophysics Data System (ADS)

Klinkby, E. B.; Knudsen, E. B.; Willendrup, P. K.; Lauritzen, B.; Nonbøl, E.; Bentley, P.; Filges, U.

2014-07-01

Recently, an interface between the Monte Carlo code MCNPX and the neutron ray-tracing code MCNPX was developed [1, 2]. Based on the expected neutronic performance and guide geometries relevant for the ESS, the combined MCNPX-McStas code is used to calculate dose rates along neutron beam guides. The generation and moderation of neutrons is simulated using a full scale MCNPX model of the ESS target monolith. Upon entering the neutron beam extraction region, the individual neutron states are handed to McStas via the MCNPX-McStas interface. McStas transports the neutrons through the beam guide, and by using newly developed event logging capability, the neutron state parameters corresponding to un-reflected neutrons are recorded at each scattering. This information is handed back to MCNPX where it serves as neutron source input for a second MCNPX simulation. This simulation enables calculation of dose rates in the vicinity of the guide. In addition the logging mechanism is employed to record the scatterings along the guides which is exploited to simulate the supermirror quality requirements (i.e. m-values) needed at different positions along the beam guide to transport neutrons in the same guide/source setup.

Analysis of a Neutronic Experiment on a Simulated Mercury Spallation Neutron Target Assembly Bombarded by Giga-Electron-Volt Protons

DOE Office of Scientific and Technical Information (OSTI.GOV)

Maekawa, Fujio; Meigo, Shin-ichiro; Kasugai, Yoshimi

2005-05-15

A neutronic benchmark experiment on a simulated spallation neutron target assembly was conducted by using the Alternating Gradient Synchrotron at Brookhaven National Laboratory and was analyzed to investigate the prediction capability of Monte Carlo simulation codes used in neutronic designs of spallation neutron sources. The target assembly consisting of a mercury target, a light water moderator, and a lead reflector was bombarded by 1.94-, 12-, and 24-GeV protons, and the fast neutron flux distributions around the target and the spectra of thermal neutrons leaking from the moderator were measured in the experiment. In this study, the Monte Carlo particle transportmore » simulation codes NMTC/JAM, MCNPX, and MCNP-4A with associated cross-section data in JENDL and LA-150 were verified based on benchmark analysis of the experiment. As a result, all the calculations predicted the measured quantities adequately; calculated integral fluxes of fast and thermal neutrons agreed approximately within {+-}40% with the experiments although the overall energy range encompassed more than 12 orders of magnitude. Accordingly, it was concluded that these simulation codes and cross-section data were adequate for neutronics designs of spallation neutron sources.« less

Neutron Capture Gamma-Ray Libraries for Nuclear Applications

NASA Astrophysics Data System (ADS)

Sleaford, B. W.; Firestone, R. B.; Summers, N.; Escher, J.; Hurst, A.; Krticka, M.; Basunia, S.; Molnar, G.; Belgya, T.; Revay, Z.; Choi, H. D.

2011-06-01

The neutron capture reaction is useful in identifying and analyzing the gamma-ray spectrum from an unknown assembly as it gives unambiguous information on its composition. This can be done passively or actively where an external neutron source is used to probe an unknown assembly. There are known capture gamma-ray data gaps in the ENDF libraries used by transport codes for various nuclear applications. The Evaluated Gamma-ray Activation file (EGAF) is a new thermal neutron capture database of discrete line spectra and cross sections for over 260 isotopes that was developed as part of an IAEA Coordinated Research Project. EGAF is being used to improve the capture gamma production in ENDF libraries. For medium to heavy nuclei the quasi continuum contribution to the gamma cascades is not experimentally resolved. The continuum contains up to 90% of all the decay energy and is modeled here with the statistical nuclear structure code DICEBOX. This code also provides a consistency check of the level scheme nuclear structure evaluation. The calculated continuum is of sufficient accuracy to include in the ENDF libraries. This analysis also determines new total thermal capture cross sections and provides an improved RIPL database. For higher energy neutron capture there is less experimental data available making benchmarking of the modeling codes more difficult. We are investigating the capture spectra from higher energy neutrons experimentally using surrogate reactions and modeling this with Hauser-Feshbach codes. This can then be used to benchmark CASINO, a version of DICEBOX modified for neutron capture at higher energy. This can be used to simulate spectra from neutron capture at incident neutron energies up to 20 MeV to improve the gamma-ray spectrum in neutron data libraries used for transport modeling of unknown assemblies.

Neutron streaming studies along JET shielding penetrations

NASA Astrophysics Data System (ADS)

Stamatelatos, Ion E.; Vasilopoulou, Theodora; Batistoni, Paola; Obryk, Barbara; Popovichev, Sergey; Naish, Jonathan

2017-09-01

Neutronic benchmark experiments are carried out at JET aiming to assess the neutronic codes and data used in ITER analysis. Among other activities, experiments are performed in order to validate neutron streaming simulations along long penetrations in the JET shielding configuration. In this work, neutron streaming calculations along the JET personnel entrance maze are presented. Simulations were performed using the MCNP code for Deuterium-Deuterium and Deuterium- Tritium plasma sources. The results of the simulations were compared against experimental data obtained using thermoluminescence detectors and activation foils.

Validating predictive models for fast ion profile relaxation in burning plasmas

NASA Astrophysics Data System (ADS)

Gorelenkov, N. N.; Heidbrink, W. W.; Kramer, G. J.; Lestz, J. B.; Podesta, M.; Van Zeeland, M. A.; White, R. B.

2016-11-01

The redistribution and potential loss of energetic particles due to MHD modes can limit the performance of fusion plasmas by reducing the plasma heating rate. In this work, we present validation studies of the 1.5D critical gradient model (CGM) for Alfvén eigenmode (AE) induced EP transport in NSTX and DIII-D neutral beam heated plasmas. In previous comparisons with a single DIII-D L-mode case, the CGM model was found to be responsible for 75% of measured AE induced neutron deficit [1]. A fully kinetic HINST is used to compute mode stability for the non-perturbative version of CGM (or nCGM). We have found that AEs show strong local instability drive up to γ /ω ∼ 20% violating assumptions of perturbative approaches used in NOVA-K code. We demonstrate that both models agree with each other and both underestimate the neutron deficit measured in DIII-D shot by approximately a factor of 2. On the other hand in NSTX the application of CGM shows good agreement for the measured flux deficit predictions. We attempt to understand these results with the help of the so-called kick model which is based on the guiding center code ORBIT. The kick model comparison gives important insight into the underlying velocity space dependence of the AE induced EP transport as well as it allows the estimate of the neutron deficit in the presence of the low frequency Alfvénic modes. Within the limitations of used models we infer that there are missing modes in the analysis which could improve the agreement with the experiments.

Studying fission neutrons with 2E-2v and 2E

NASA Astrophysics Data System (ADS)

Al-Adili, Ali; Jansson, Kaj; Tarrío, Diego; Hambsch, Franz-Josef; Göök, Alf; Oberstedt, Stephan; Olivier Frégeau, Marc; Gustavsson, Cecilia; Lantz, Mattias; Mattera, Andrea; Prokofiev, Alexander V.; Rakopoulos, Vasileios; Solders, Andreas; Vidali, Marzio; Österlund, Michael; Pomp, Stephan

2018-03-01

This work aims at measuring prompt-fission neutrons at different excitation energies of the nucleus. Two independent techniques, the 2E-2v and the 2E techniques, are used to map the characteristics of the mass-dependent prompt fission neutron multiplicity, v(A), when the excitation energy is increased. The VERDI 2E-2v spectrometer is being developed at JRC-GEEL. The Fission Fragment (FF) energies are measured using two arrays of 16 silicon (Si) detectors each. The FFs velocities are obtained by time-of-flight, measured between micro-channel plates (MCP) and Si detectors. With MCPs placed on both sides of the fission source, VERDI allows for independent timing measurements for both fragments. 252Cf(sf) was measured and the present results revealed particular features of the 2E-2v technique. Dedicated simulations were also performed using the GEF code to study important aspects of the 2E-2v technique. Our simulations show that prompt neutron emission has a non-negligible impact on the deduced fragment data and affects also the shape of v(A). Geometrical constraints lead to a total-kinetic energy-dependent detection efficiency. The 2E technique utilizes an ionization chamber together with two liquid scintillator detectors. Two measurements have been performed, one of 252Cf(sf) and another one of thermal-neutron induced fission in 235U(n,f). Results from 252Cf(sf) are reported here.

NSDann2BS, a neutron spectrum unfolding code based on neural networks technology and two bonner spheres

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ortiz-Rodriguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.

In this work a neutron spectrum unfolding code, based on artificial intelligence technology is presented. The code called ''Neutron Spectrometry and Dosimetry with Artificial Neural Networks and two Bonner spheres'', (NSDann2BS), was designed in a graphical user interface under the LabVIEW programming environment. The main features of this code are to use an embedded artificial neural network architecture optimized with the ''Robust design of artificial neural networks methodology'' and to use two Bonner spheres as the only piece of information. In order to build the code here presented, once the net topology was optimized and properly trained, knowledge stored atmore » synaptic weights was extracted and using a graphical framework build on the LabVIEW programming environment, the NSDann2BS code was designed. This code is friendly, intuitive and easy to use for the end user. The code is freely available upon request to authors. To demonstrate the use of the neural net embedded in the NSDann2BS code, the rate counts of {sup 252}Cf, {sup 241}AmBe and {sup 239}PuBe neutron sources measured with a Bonner spheres system.« less

NSDann2BS, a neutron spectrum unfolding code based on neural networks technology and two bonner spheres

NASA Astrophysics Data System (ADS)

Ortiz-Rodríguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Solís Sánches, L. O.; Miranda, R. Castañeda; Cervantes Viramontes, J. M.; Vega-Carrillo, H. R.

2013-07-01

In this work a neutron spectrum unfolding code, based on artificial intelligence technology is presented. The code called "Neutron Spectrometry and Dosimetry with Artificial Neural Networks and two Bonner spheres", (NSDann2BS), was designed in a graphical user interface under the LabVIEW programming environment. The main features of this code are to use an embedded artificial neural network architecture optimized with the "Robust design of artificial neural networks methodology" and to use two Bonner spheres as the only piece of information. In order to build the code here presented, once the net topology was optimized and properly trained, knowledge stored at synaptic weights was extracted and using a graphical framework build on the LabVIEW programming environment, the NSDann2BS code was designed. This code is friendly, intuitive and easy to use for the end user. The code is freely available upon request to authors. To demonstrate the use of the neural net embedded in the NSDann2BS code, the rate counts of 252Cf, 241AmBe and 239PuBe neutron sources measured with a Bonner spheres system.

Validation of DRAGON4/DONJON4 simulation methodology for a typical MNSR by calculating reactivity feedback coefficient and neutron flux

NASA Astrophysics Data System (ADS)

Al Zain, Jamal; El Hajjaji, O.; El Bardouni, T.; Boukhal, H.; Jaï, Otman

2018-06-01

The MNSR is a pool type research reactor, which is difficult to model because of the importance of neutron leakage. The aim of this study is to evaluate a 2-D transport model for the reactor compatible with the latest release of the DRAGON code and 3-D diffusion of the DONJON code. DRAGON code is then used to generate the group macroscopic cross sections needed for full core diffusion calculations. The diffusion DONJON code, is then used to compute the effective multiplication factor (keff), the feedback reactivity coefficients and neutron flux which account for variation in fuel and moderator temperatures as well as the void coefficient have been calculated using the DRAGON and DONJON codes for the MNSR research reactor. The cross sections of all the reactor components at different temperatures were generated using the DRAGON code. These group constants were used then in the DONJON code to calculate the multiplication factor and the neutron spectrum at different water and fuel temperatures using 69 energy groups. Only one parameter was changed where all other parameters were kept constant. Finally, Good agreements between the calculated and measured have been obtained for every of the feedback reactivity coefficients and neutron flux.

Calculation of response matrix of CaSO 4:Dy based neutron dosimeter using Monte Carlo code FLUKA and measurement of 241Am-Be spectra

NASA Astrophysics Data System (ADS)

Chatterjee, S.; Bakshi, A. K.; Tripathy, S. P.

2010-09-01

Response matrix for CaSO 4:Dy based neutron dosimeter was generated using Monte Carlo code FLUKA in the energy range thermal to 20 MeV for a set of eight Bonner spheres of diameter 3-12″ including the bare one. Response of the neutron dosimeter was measured for the above set of spheres for 241Am-Be neutron source covered with 2 mm lead. An analytical expression for the response function was devised as a function of sphere mass. Using Frascati Unfolding Iteration Tool (FRUIT) unfolding code, the neutron spectrum of 241Am-Be was unfolded and compared with standard IAEA spectrum for the same.

Pile noise experiment in MINERVE reactor to estimate kinetic parameters using various data processing methods

DOE Office of Scientific and Technical Information (OSTI.GOV)

Geslot, Benoit; Gruel, Adrien; Pepino, Alexandra

2015-07-01

MINERVE is a two-zone pool type zero power reactor operated by CEA (Cadarache, France). Kinetic parameters of the core (prompt neutron decay constant, delayed neutron fraction, generation time) have been recently measured using various pile noise experimental techniques, namely Feynman-α, Rossi-α and Cohn-α. Results are discussed and compared to each other's. The measurement campaign has been conducted in the framework of a tri-partite collaboration between CEA, SCK.CEN and PSI. Results presented in this paper were obtained thanks to a time-stamping acquisition system developed by CEA. PSI performed simultaneous measurements which are presented in a companion paper. Signals come from twomore » high efficiency fission chambers located in the graphite reflector next to the core driver zone. Experiments were conducted at critical state with a reactor power of 0.2 W. The core integral fission rate is obtained from a calibrated miniature fission chamber located at the center of the core. Other results obtained in two sub-critical configurations will be presented elsewhere. Best estimate delayed neutron fraction comes from the Cohn-α method: 747 ± 15 pcm (1σ). In this case, the prompt decay constant is 79 ± 0.5 s{sup -1} and the generation time is 94.5 ± 0.7 μs. Other methods give consistent results within the confidence intervals. Experimental results are compared to calculated values obtained from a full 3D core modeling with the CEA-developed Monte Carlo code TRIPOLI4.9 associated with its continuous energy JEFF3.1.1-based library. A very good agreement is observed for the calculated delayed neutron fraction (748.7 ± 0.4 pcm at 1σ), that is a difference of -0.3% with the experiment. On the contrary, a 10% discrepancy is observed for the calculated generation time (104.4 ± 0.1 μs at 1σ). (authors)« less

The specific purpose Monte Carlo code McENL for simulating the response of epithermal neutron lifetime well logging tools

NASA Astrophysics Data System (ADS)

Prettyman, T. H.; Gardner, R. P.; Verghese, K.

1993-08-01

A new specific purpose Monte Carlo code called McENL for modeling the time response of epithermal neutron lifetime tools is described. The weight windows technique, employing splitting and Russian roulette, is used with an automated importance function based on the solution of an adjoint diffusion model to improve the code efficiency. Complete composition and density correlated sampling is also included in the code, and can be used to study the effect on tool response of small variations in the formation, borehole, or logging tool composition and density. An illustration of the latter application is given for the density of a thermal neutron filter. McENL was benchmarked against test-pit data for the Mobil pulsed neutron porosity tool and was found to be very accurate. Results of the experimental validation and details of code performance are presented.

A Monte-Carlo Benchmark of TRIPOLI-4® and MCNP on ITER neutronics

NASA Astrophysics Data System (ADS)

Blanchet, David; Pénéliau, Yannick; Eschbach, Romain; Fontaine, Bruno; Cantone, Bruno; Ferlet, Marc; Gauthier, Eric; Guillon, Christophe; Letellier, Laurent; Proust, Maxime; Mota, Fernando; Palermo, Iole; Rios, Luis; Guern, Frédéric Le; Kocan, Martin; Reichle, Roger

2017-09-01

Radiation protection and shielding studies are often based on the extensive use of 3D Monte-Carlo neutron and photon transport simulations. ITER organization hence recommends the use of MCNP-5 code (version 1.60), in association with the FENDL-2.1 neutron cross section data library, specifically dedicated to fusion applications. The MCNP reference model of the ITER tokamak, the `C-lite', is being continuously developed and improved. This article proposes to develop an alternative model, equivalent to the 'C-lite', but for the Monte-Carlo code TRIPOLI-4®. A benchmark study is defined to test this new model. Since one of the most critical areas for ITER neutronics analysis concerns the assessment of radiation levels and Shutdown Dose Rates (SDDR) behind the Equatorial Port Plugs (EPP), the benchmark is conducted to compare the neutron flux through the EPP. This problem is quite challenging with regard to the complex geometry and considering the important neutron flux attenuation ranging from 1014 down to 108 n•cm-2•s-1. Such code-to-code comparison provides independent validation of the Monte-Carlo simulations, improving the confidence in neutronic results.

Neutron-fragment and Neutron-neutron Correlations in Low-energy Fission

NASA Astrophysics Data System (ADS)

Lestone, J. P.

2016-01-01

A computational method has been developed to simulate neutron emission from thermal-neutron induced fission of 235U and from spontaneous fission of 252Cf. Measured pre-emission mass-yield curves, average total kinetic energies and their variances, both as functions of mass split, are used to obtain a representation of the distribution of fragment velocities. Measured average neutron multiplicities as a function of mass split and their dependence on total kinetic energy are used. Simulations can be made to reproduce measured factorial moments of neutron-multiplicity distributions with only minor empirical adjustments to some experimental inputs. The neutron-emission spectra in the rest-frame of the fragments are highly constrained by ENDF/B-VII.1 prompt-fission neutron-spectra evaluations. The n-f correlation measurements of Vorobyev et al. (2010) are consistent with predictions where all neutrons are assumed to be evaporated isotropically from the rest frame of fully accelerated fragments. Measured n-f and n-n correlations of others are a little weaker than the predictions presented here. These weaker correlations could be used to infer a weak scission-neutron source. However, the effect of neutron scattering on the experimental results must be studied in detail before moving away from a null hypothesis that all neutrons are evaporated from the fragments.

Numerical simulation of exploding pusher targets

NASA Astrophysics Data System (ADS)

Atzeni, S.; Rosenberg, M. J.; Gatu Johnson, M.; Petrasso, R. D.

2017-10-01

Exploding pusher targets, i.e. gas-filled large aspect-ratio glass or plastic shells, driven by a strong laser-generated shock, are widely used as pulsed sources of neutrons and fast charged particles. Recent experiments on exploding pushers provided evidence for the transition from a purely fluid behavior to a kinetic one. Indeed, fluid models largely overpredict yield and temperature as the Knudsen number Kn (ratio of ion mean-free path to compressed gas radius) is comparable or larger than one. At Kn = 0.3 - 1, fluid codes reasonably estimate integral quantities as yield and neutron-averaged temperatures, but do not reproduce burn radii, burn profiles and DD/DHe3 yield ratio. This motivated a detailed simulation study of intermediate-Kn exploding pushers. We will show how simulation results depend on models for laser-interaction, electron conductivity (flux-limited local vs nonlocal), viscosity (physical vs artificial), and ion mixing. Work partially supported by Sapienza Project C26A15YTMA, Sapienza 2016 (n. 257584), and Eurofusion Project AWP17-ENR-IFE-CEA-01.

Validating predictive models for fast ion profile relaxation in burning plasmas

DOE PAGES

Gorelenkov, N. N.; Heidbrink, W. W.; Kramer, G. J.; ...

2016-07-22

The redistribution and potential loss of energetic particles due to MHD modes can limit the performance of fusion plasmas by reducing the plasma heating rate. In this work, we present validation studies of the 1.5D critical gradient model (CGM) for Alfvén eigenmode (AE) induced EP transport in NSTX and DIII-D neutral beam heated plasmas. In previous comparisons with a single DIII-D L-mode case, the CGM model was found to be responsible for 75% of measured AE induced neutron deficit [1]. A fully kinetic HINST is used to compute mode stability for the non-perturbative version of CGM (or nCGM). We have found that AEs show strong local instability drive up tomore » $$\\gamma /\\omega \\sim 20\\%$$ violating assumptions of perturbative approaches used in NOVA-K code. Lastly, we demonstrate that both models agree with each other and both underestimate the neutron deficit measured in DIII-D shot by approximately a factor of 2.« less

Binary Neutron Stars with Arbitrary Spins in Numerical Relativity

NASA Astrophysics Data System (ADS)

Pfeiffer, Harald; Tacik, Nick; Foucart, Francois; Haas, Roland; Kaplan, Jeffrey; Muhlberger, Curran; Duez, Matt; Kidder, Lawrence; Scheel, Mark; Szilagyi, Bela

2015-04-01

We present a code to construct initial data for binary neutron star where the stars are rotating. Our code, based on the formalism developed by Tichy, allows for arbitrary rotation axes of the neutron stars and is able to achieve rotation rates near rotational breakup. We demonstrate that orbital eccentricity of the binary neutron stars can be controlled to ~ 0 . 1 % . Preliminary evolutions show that spin- and orbit-precession of Neutron stars is well described by post-Newtonian approximation. The neutron stars show quasi-normal mode oscillations at an amplitude which increases with the rotation rate of the stars.

Energy dynamics and current sheet structure in fluid and kinetic simulations of decaying magnetohydrodynamic turbulence

DOE PAGES

Makwana, K. D.; Zhdankin, V.; Li, H.; ...

2015-04-10

We performed simulations of decaying magnetohydrodynamic (MHD) turbulence with a fluid and a kinetic code. The initial condition is an ensemble of long-wavelength, counter-propagating, shear-Alfvén waves, which interact and rapidly generate strong MHD turbulence. The total energy is conserved and the rate of turbulent energy decay is very similar in both codes, although the fluid code has numerical dissipation, whereas the kinetic code has kinetic dissipation. The inertial range power spectrum index is similar in both the codes. The fluid code shows a perpendicular wavenumber spectral slope of k-1.3⊥k⊥-1.3. The kinetic code shows a spectral slope of k-1.5⊥k⊥-1.5 for smallermore » simulation domain, and k-1.3⊥k⊥-1.3 for larger domain. We then estimate that collisionless damping mechanisms in the kinetic code can account for the dissipation of the observed nonlinear energy cascade. Current sheets are geometrically characterized. Their lengths and widths are in good agreement between the two codes. The length scales linearly with the driving scale of the turbulence. In the fluid code, their thickness is determined by the grid resolution as there is no explicit diffusivity. In the kinetic code, their thickness is very close to the skin-depth, irrespective of the grid resolution. Finally, this work shows that kinetic codes can reproduce the MHD inertial range dynamics at large scales, while at the same time capturing important kinetic physics at small scales.« less

Energy dynamics and current sheet structure in fluid and kinetic simulations of decaying magnetohydrodynamic turbulence

DOE Office of Scientific and Technical Information (OSTI.GOV)

Makwana, K. D.; Zhdankin, V.; Li, H.

We performed simulations of decaying magnetohydrodynamic (MHD) turbulence with a fluid and a kinetic code. The initial condition is an ensemble of long-wavelength, counter-propagating, shear-Alfvén waves, which interact and rapidly generate strong MHD turbulence. The total energy is conserved and the rate of turbulent energy decay is very similar in both codes, although the fluid code has numerical dissipation, whereas the kinetic code has kinetic dissipation. The inertial range power spectrum index is similar in both the codes. The fluid code shows a perpendicular wavenumber spectral slope of k-1.3⊥k⊥-1.3. The kinetic code shows a spectral slope of k-1.5⊥k⊥-1.5 for smallermore » simulation domain, and k-1.3⊥k⊥-1.3 for larger domain. We then estimate that collisionless damping mechanisms in the kinetic code can account for the dissipation of the observed nonlinear energy cascade. Current sheets are geometrically characterized. Their lengths and widths are in good agreement between the two codes. The length scales linearly with the driving scale of the turbulence. In the fluid code, their thickness is determined by the grid resolution as there is no explicit diffusivity. In the kinetic code, their thickness is very close to the skin-depth, irrespective of the grid resolution. Finally, this work shows that kinetic codes can reproduce the MHD inertial range dynamics at large scales, while at the same time capturing important kinetic physics at small scales.« less

Modeling the Martian neutron and gamma-ray leakage fluxes using Geant4

NASA Astrophysics Data System (ADS)

Pirard, Benoit; Desorgher, Laurent; Diez, Benedicte; Gasnault, Olivier

A new evaluation of the Martian neutron and gamma-ray (continuum and line) leakage fluxes has been performed using the Geant4 code. Even if numerous studies have recently been carried out with Monte Carlo methods to characterize planetary radiation environments, only a few however have been able to reproduce in detail the neutron and gamma-ray spectra observed in orbit. We report on the efforts performed to adapt and validate the Geant4-based PLAN- ETOCOSMICS code for use in planetary neutron and gamma-ray spectroscopy data analysis. Beside the advantage of high transparency and modularity common to Geant4 applications, the new code uses reviewed nuclear cross section data, realistic atmospheric profiles and soil layering, as well as specific effects such as gravity acceleration for low energy neutrons. Results from first simulations are presented for some Martian reference compositions and show a high consistency with corresponding neutron and gamma-ray spectra measured on board Mars Odyssey. Finally we discuss the advantages and perspectives of the improved code for precise simulation of planetary radiation environments.

Dark Kinetic Heating of Neutron Stars and an Infrared Window on WIMPs, SIMPs, and Pure Higgsinos

NASA Astrophysics Data System (ADS)

Baryakhtar, Masha; Bramante, Joseph; Li, Shirley Weishi; Linden, Tim; Raj, Nirmal

2017-09-01

We identify a largely model-independent signature of dark matter (DM) interactions with nucleons and electrons. DM in the local galactic halo, gravitationally accelerated to over half the speed of light, scatters against and deposits kinetic energy into neutron stars, heating them to infrared blackbody temperatures. The resulting radiation could potentially be detected by the James Webb Space Telescope, the Thirty Meter Telescope, or the European Extremely Large Telescope. This mechanism also produces optical emission from neutron stars in the galactic bulge, and x-ray emission near the galactic center because dark matter is denser in these regions. For GeV-PeV mass dark matter, dark kinetic heating would initially unmask any spin-independent or spin-dependent dark matter-nucleon cross sections exceeding 2 ×10-45 cm2, with improved sensitivity after more telescope exposure. For lighter-than-GeV dark matter, cross-section sensitivity scales inversely with dark matter mass because of Pauli blocking; for heavier-than-PeV dark matter, it scales linearly with mass as a result of needing multiple scatters for capture. Future observations of dark sector-warmed neutron stars could determine whether dark matter annihilates in or only kinetically heats neutron stars. Because inelastic interstate transitions of up to a few GeV would occur in relativistic scattering against nucleons, elusive inelastic dark matter like pure Higgsinos can also be discovered.

Dark Kinetic Heating of Neutron Stars and an Infrared Window on WIMPs, SIMPs, and Pure Higgsinos.

PubMed

Baryakhtar, Masha; Bramante, Joseph; Li, Shirley Weishi; Linden, Tim; Raj, Nirmal

2017-09-29

We identify a largely model-independent signature of dark matter (DM) interactions with nucleons and electrons. DM in the local galactic halo, gravitationally accelerated to over half the speed of light, scatters against and deposits kinetic energy into neutron stars, heating them to infrared blackbody temperatures. The resulting radiation could potentially be detected by the James Webb Space Telescope, the Thirty Meter Telescope, or the European Extremely Large Telescope. This mechanism also produces optical emission from neutron stars in the galactic bulge, and x-ray emission near the galactic center because dark matter is denser in these regions. For GeV-PeV mass dark matter, dark kinetic heating would initially unmask any spin-independent or spin-dependent dark matter-nucleon cross sections exceeding 2×10^{-45}  cm^{2}, with improved sensitivity after more telescope exposure. For lighter-than-GeV dark matter, cross-section sensitivity scales inversely with dark matter mass because of Pauli blocking; for heavier-than-PeV dark matter, it scales linearly with mass as a result of needing multiple scatters for capture. Future observations of dark sector-warmed neutron stars could determine whether dark matter annihilates in or only kinetically heats neutron stars. Because inelastic interstate transitions of up to a few GeV would occur in relativistic scattering against nucleons, elusive inelastic dark matter like pure Higgsinos can also be discovered.

Thermal Neutron Imaging Using A New Pad-Based Position Sensitive Neutron Detector

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dioszegi I.; Vanier P.E.; Salwen C.

2016-10-29

Thermal neutrons (with mean energy of 25 meV) have a scattering mean free path of about 20 m in air. Therefore it is feasible to find localized thermal neutron sources up to ~30 m standoff distance using thermal neutron imaging. Coded aperture thermal neutron imaging was developed in our laboratory in the nineties, using He-3 filled wire chambers. Recently a new generation of coded-aperture neutron imagers has been developed. In the new design the ionization chamber has anode and cathode planes, where the anode is composed of an array of individual pads. The charge is collected on each of themore » individual 5x5 mm2 anode pads, (48x48 in total, corresponding to 24x24 cm2 sensitive area) and read out by application specific integrated circuits (ASICs). The high sensitivity of the ASICs allows unity gain operation mode. The new design has several advantages for field deployable imaging applications, compared to the previous generation of wire-grid based neutron detectors. Among these are the rugged design, lighter weight and use of non-flammable stopping gas. For standoff localization of thermalized neutron sources a low resolution (11x11 pixel) coded aperture mask has been fabricated. Using the new larger area detector and the coarse resolution mask we performed several standoff experiments using moderated californium and plutonium sources at Idaho National Laboratory. In this paper we will report on the development and performance of the new pad-based neutron camera, and present long range coded-aperture images of various thermalized neutron sources.« less

The IAEA neutron coincidence counting (INCC) and the DEMING least-squares fitting programs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Krick, M.S.; Harker, W.C.; Rinard, P.M.

1998-12-01

Two computer programs are described: (1) the INCC (IAEA or International Neutron Coincidence Counting) program and (2) the DEMING curve-fitting program. The INCC program is an IAEA version of the Los Alamos NCC (Neutron Coincidence Counting) code. The DEMING program is an upgrade of earlier Windows{reg_sign} and DOS codes with the same name. The versions described are INCC 3.00 and DEMING 1.11. The INCC and DEMING codes provide inspectors with the software support needed to perform calibration and verification measurements with all of the neutron coincidence counting systems used in IAEA inspections for the nondestructive assay of plutonium and uranium.

Prompt neutron emission and energy balance in 235U(n,f)

NASA Astrophysics Data System (ADS)

Göök, Alf; Hambsch, Franz-Josef; Oberstedt, Stephan

2017-09-01

Investigations of prompt fission neutron (PFN) emission are of importance in understanding the fission process in general and the sharing of excitation energy among the fission fragments in particular. Experimental activities at JRC-Geel on PFN emission in response to OECD/NEA nuclear data requests is presented in this contribution. The focus lies on on-going investigations of PFN emission from the reaction 235U(n,f) in the region of the resolved resonances taking place at the GELINA facility. For this reaction strong fluctuations of fission fragment mass distributions and mean total kinetic energy have been observed as a function of incident neutron energy in the resonance region. In addition, fluctuations of prompt neutron multiplicities have also been observed. The goal of the present study is to verify the current knowledge of PFN multiplicity fluctuations and to study correlations with fission fragment properties. The experiment employs a scintillation detector array for neutron detection, while fission fragment properties are determined via the double kinetic energy technique using a position sensitive twin ionization chamber. Results on PFN multiplicity correlations with fission fragment properties from the present study show significant differences compared to earlier studies on this reaction, induced by thermal neutrons. Specifically, the total kinetic energy dependence of the neutron multiplicity per fission shows an inverse slope FX1TKE/FX2ν approximately 35% weaker than observed in earlier studies of thermal neutron induced fission on 235U. The inverse slope is related to the energy carried away per emitted neutron and is, thereby, closely connected to the energy balance of the fission reaction. The present result should have strong impact on the modeling of both prompt neutron and prompt γ-ray emission in fission of the 236U compound nucleus.

FY16 Status Report on NEAMS Neutronics Activities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, C. H.; Shemon, E. R.; Smith, M. A.

2016-09-30

The goal of the NEAMS neutronics effort is to develop a neutronics toolkit for use on sodium-cooled fast reactors (SFRs) which can be extended to other reactor types. The neutronics toolkit includes the high-fidelity deterministic neutron transport code PROTEUS and many supporting tools such as a cross section generation code MC 2-3, a cross section library generation code, alternative cross section generation tools, mesh generation and conversion utilities, and an automated regression test tool. The FY16 effort for NEAMS neutronics focused on supporting the release of the SHARP toolkit and existing and new users, continuing to develop PROTEUS functions necessarymore » for performance improvement as well as the SHARP release, verifying PROTEUS against available existing benchmark problems, and developing new benchmark problems as needed. The FY16 research effort was focused on further updates of PROTEUS-SN and PROTEUS-MOCEX and cross section generation capabilities as needed.« less

Current and anticipated uses of thermalhydraulic and neutronic codes at PSI

DOE Office of Scientific and Technical Information (OSTI.GOV)

Aksan, S.N.; Zimmermann, M.A.; Yadigaroglu, G.

1997-07-01

The thermalhydraulic and/or neutronic codes in use at PSI mainly provide the capability to perform deterministic safety analysis for Swiss NPPs and also serve as analysis tools for experimental facilities for LWR and ALWR simulations. In relation to these applications, physical model development and improvements, and assessment of the codes are also essential components of the activities. In this paper, a brief overview is provided on the thermalhydraulic and/or neutronic codes used for safety analysis of LWRs, at PSI, and also of some experiences and applications with these codes. Based on these experiences, additional assessment needs are indicated, together withmore » some model improvement needs. The future needs that could be used to specify both the development of a new code and also improvement of available codes are summarized.« less

Preliminary estimates of nucleon fluxes in a water target exposed to solar-flare protons: BRYNTRN versus Monte Carlo code

NASA Technical Reports Server (NTRS)

Shinn, Judy L.; Wilson, John W.; Lone, M. A.; Wong, P. Y.; Costen, Robert C.

1994-01-01

A baryon transport code (BRYNTRN) has previously been verified using available Monte Carlo results for a solar-flare spectrum as the reference. Excellent results were obtained, but the comparisons were limited to the available data on dose and dose equivalent for moderate penetration studies that involve minor contributions from secondary neutrons. To further verify the code, the secondary energy spectra of protons and neutrons are calculated using BRYNTRN and LAHET (Los Alamos High-Energy Transport code, which is a Monte Carlo code). These calculations are compared for three locations within a water slab exposed to the February 1956 solar-proton spectrum. Reasonable agreement was obtained when various considerations related to the calculational techniques and their limitations were taken into account. Although the Monte Carlo results are preliminary, it appears that the neutron albedo, which is not currently treated in BRYNTRN, might be a cause for the large discrepancy seen at small penetration depths. It also appears that the nonelastic neutron production cross sections in BRYNTRN may underestimate the number of neutrons produced in proton collisions with energies below 200 MeV. The notion that the poor energy resolution in BRYNTRN may cause a large truncation error in neutron elastic scattering requires further study.

New measurements on isobaric fission product yields and mean kinetic energy for 241Pu thermal neutron-induced fission

NASA Astrophysics Data System (ADS)

Julien-Laferrière, Sylvain; Kessedjian, Grégoire; Serot, Olivier; Chebboubi, Abdelaziz; Bernard, David; Blanc, Aurélien; Köster, Ulli; Litaize, Olivier; Materna, Thomas; Meplan, Olivier; Rapala, Michal; Sage, Christophe

2018-03-01

Nuclear fission yields data measurements for thermal neutron induced fission of 241Pu have been carried out at the Institut Laue Langevin (ILL) in Grenoble, using the Lohengrin mass spectrometer. Mass, isotopic and isomeric yields have been extracted for the last measurements. A focus is given in this document to the mass yield results which are obtained for almost the entire heavy peak and most of the light high yields masses, along with the covariance matrix. The mean kinetic energy as a function of the fission product mass has also been extracted from the measurements. The total mean kinetic energy pre and post neutron emission have been assessed and compared to other works showing a rather good agreement.

Radiation shielding quality assurance

NASA Astrophysics Data System (ADS)

Um, Dallsun

For the radiation shielding quality assurance, the validity and reliability of the neutron transport code MCNP, which is now one of the most widely used radiation shielding analysis codes, were checked with lot of benchmark experiments. And also as a practical example, follows were performed in this thesis. One integral neutron transport experiment to measure the effect of neutron streaming in iron and void was performed with Dog-Legged Void Assembly in Knolls Atomic Power Laboratory in 1991. Neutron flux was measured six different places with the methane detectors and a BF-3 detector. The main purpose of the measurements was to provide benchmark against which various neutron transport calculation tools could be compared. Those data were used in verification of Monte Carlo Neutron & Photon Transport Code, MCNP, with the modeling for that. Experimental results and calculation results were compared in both ways, as the total integrated value of neutron fluxes along neutron energy range from 10 KeV to 2 MeV and as the neutron spectrum along with neutron energy range. Both results are well matched with the statistical error +/-20%. MCNP results were also compared with those of TORT, a three dimensional discrete ordinates code which was developed by Oak Ridge National Laboratory. MCNP results are superior to the TORT results at all detector places except one. This means that MCNP is proved as a very powerful tool for the analysis of neutron transport through iron & air and further it could be used as a powerful tool for the radiation shielding analysis. For one application of the analysis of variance (ANOVA) to neutron and gamma transport problems, uncertainties for the calculated values of critical K were evaluated as in the ANOVA on statistical data.

Analysis of neutron and gamma-ray streaming along the maze of NRCAM thallium production target room.

PubMed

Raisali, G; Hajiloo, N; Hamidi, S; Aslani, G

2006-08-01

Study of the shield performance of a thallium-203 production target room has been investigated in this work. Neutron and gamma-ray equivalent dose rates at various points of the maze are calculated by simulating the transport of streaming neutrons, and photons using Monte Carlo method. For determination of neutron and gamma-ray source intensities and their energy spectrum, we have applied SRIM 2003 and ALICE91 computer codes to Tl target and its Cu substrate for a 145 microA of 28.5 MeV protons beam. The MCNP/4C code has been applied with neutron source term in mode n p to consider both prompt neutrons and secondary gamma-rays. Then the code is applied for the prompt gamma-rays as the source term. The neutron-flux energy spectrum and equivalent dose rates for neutron and gamma-rays in various positions in the maze have been calculated. It has been found that the deviation between calculated and measured dose values along the maze is less than 20%.

Neutron Deep Penetration Calculations in Light Water with Monte Carlo TRIPOLI-4® Variance Reduction Techniques

NASA Astrophysics Data System (ADS)

Lee, Yi-Kang

2017-09-01

Nuclear decommissioning takes place in several stages due to the radioactivity in the reactor structure materials. A good estimation of the neutron activation products distributed in the reactor structure materials impacts obviously on the decommissioning planning and the low-level radioactive waste management. Continuous energy Monte-Carlo radiation transport code TRIPOLI-4 has been applied on radiation protection and shielding analyses. To enhance the TRIPOLI-4 application in nuclear decommissioning activities, both experimental and computational benchmarks are being performed. To calculate the neutron activation of the shielding and structure materials of nuclear facilities, the knowledge of 3D neutron flux map and energy spectra must be first investigated. To perform this type of neutron deep penetration calculations with the Monte Carlo transport code, variance reduction techniques are necessary in order to reduce the uncertainty of the neutron activation estimation. In this study, variance reduction options of the TRIPOLI-4 code were used on the NAIADE 1 light water shielding benchmark. This benchmark document is available from the OECD/NEA SINBAD shielding benchmark database. From this benchmark database, a simplified NAIADE 1 water shielding model was first proposed in this work in order to make the code validation easier. Determination of the fission neutron transport was performed in light water for penetration up to 50 cm for fast neutrons and up to about 180 cm for thermal neutrons. Measurement and calculation results were benchmarked. Variance reduction options and their performance were discussed and compared.

Experimental Evidence of Kinetic Effects in Indirect-Drive Inertial Confinement Fusion Hohlraums

NASA Astrophysics Data System (ADS)

Shan, L. Q.; Cai, H. B.; Zhang, W. S.; Tang, Q.; Zhang, F.; Song, Z. F.; Bi, B.; Ge, F. J.; Chen, J. B.; Liu, D. X.; Wang, W. W.; Yang, Z. H.; Qi, W.; Tian, C.; Yuan, Z. Q.; Zhang, B.; Yang, L.; Jiao, J. L.; Cui, B.; Zhou, W. M.; Cao, L. F.; Zhou, C. T.; Gu, Y. Q.; Zhang, B. H.; Zhu, S. P.; He, X. T.

2018-05-01

We present the first experimental evidence supported by simulations of kinetic effects launched in the interpenetration layer between the laser-driven hohlraum plasma bubbles and the corona plasma of the compressed pellet at the Shenguang-III prototype laser facility. Solid plastic capsules were coated with carbon-deuterium layers; as the implosion neutron yield is quenched, DD fusion yield from the corona plasma provides a direct measure of the kinetic effects inside the hohlraum. An anomalous large energy spread of the DD neutron signal (˜282 keV ) and anomalous scaling of the neutron yield with the thickness of the carbon-deuterium layers cannot be explained by the hydrodynamic mechanisms. Instead, these results can be attributed to kinetic shocks that arise in the hohlraum-wall-ablator interpenetration region, which result in efficient acceleration of the deuterons (˜28.8 J , 0.45% of the total input laser energy). These studies provide novel insight into the interactions and dynamics of a vacuum hohlraum and near-vacuum hohlraum.

Model-Based Least Squares Reconstruction of Coded Source Neutron Radiographs: Integrating the ORNL HFIR CG1D Source Model

DOE Office of Scientific and Technical Information (OSTI.GOV)

Santos-Villalobos, Hector J; Gregor, Jens; Bingham, Philip R

2014-01-01

At the present, neutron sources cannot be fabricated small and powerful enough in order to achieve high resolution radiography while maintaining an adequate flux. One solution is to employ computational imaging techniques such as a Magnified Coded Source Imaging (CSI) system. A coded-mask is placed between the neutron source and the object. The system resolution is increased by reducing the size of the mask holes and the flux is increased by increasing the size of the coded-mask and/or the number of holes. One limitation of such system is that the resolution of current state-of-the-art scintillator-based detectors caps around 50um. Tomore » overcome this challenge, the coded-mask and object are magnified by making the distance from the coded-mask to the object much smaller than the distance from object to detector. In previous work, we have shown via synthetic experiments that our least squares method outperforms other methods in image quality and reconstruction precision because of the modeling of the CSI system components. However, the validation experiments were limited to simplistic neutron sources. In this work, we aim to model the flux distribution of a real neutron source and incorporate such a model in our least squares computational system. We provide a full description of the methodology used to characterize the neutron source and validate the method with synthetic experiments.« less

A Point Kinetics Model for Estimating Neutron Multiplication of Bare Uranium Metal in Tagged Neutron Measurements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tweardy, Matthew C.; McConchie, Seth; Hayward, Jason P.

An extension of the point kinetics model is developed in this paper to describe the neutron multiplicity response of a bare uranium object under interrogation by an associated particle imaging deuterium-tritium (D-T) measurement system. This extended model is used to estimate the total neutron multiplication of the uranium. Both MCNPX-PoliMi simulations and data from active interrogation measurements of highly enriched and depleted uranium geometries are used to evaluate the potential of this method and to identify the sources of systematic error. The detection efficiency correction for measured coincidence response is identified as a large source of systematic error. If themore » detection process is not considered, results suggest that the method can estimate total multiplication to within 13% of the simulated value. Values for multiplicity constants in the point kinetics equations are sensitive to enrichment due to (n, xn) interactions by D-T neutrons and can introduce another significant source of systematic bias. This can theoretically be corrected if isotopic composition is known a priori. Finally, the spatial dependence of multiplication is also suspected of introducing further systematic bias for high multiplication uranium objects.« less

A Point Kinetics Model for Estimating Neutron Multiplication of Bare Uranium Metal in Tagged Neutron Measurements

DOE PAGES

Tweardy, Matthew C.; McConchie, Seth; Hayward, Jason P.

2017-06-13

An extension of the point kinetics model is developed in this paper to describe the neutron multiplicity response of a bare uranium object under interrogation by an associated particle imaging deuterium-tritium (D-T) measurement system. This extended model is used to estimate the total neutron multiplication of the uranium. Both MCNPX-PoliMi simulations and data from active interrogation measurements of highly enriched and depleted uranium geometries are used to evaluate the potential of this method and to identify the sources of systematic error. The detection efficiency correction for measured coincidence response is identified as a large source of systematic error. If themore » detection process is not considered, results suggest that the method can estimate total multiplication to within 13% of the simulated value. Values for multiplicity constants in the point kinetics equations are sensitive to enrichment due to (n, xn) interactions by D-T neutrons and can introduce another significant source of systematic bias. This can theoretically be corrected if isotopic composition is known a priori. Finally, the spatial dependence of multiplication is also suspected of introducing further systematic bias for high multiplication uranium objects.« less

Experimental determination of neutron lifetimes through macroscopic neutron noise in the IPEN/MB-01 reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gonnelli, Eduardo; Diniz, Ricardo

2013-05-06

The neutron lifetimes of the core, reflector, and global were experimentally obtained through macroscopic neutron noise in the IPEN/MB-01 reactor for five levels of subcriticality. The theoretical Auto Power Spectral Densities were derived by point kinetic equations taking the reflector effect into account, and one of the approaches consider an additional group of delayed neutrons.

Fission fragment yields and total kinetic energy release in neutron-induced fission of235,238U,and239Pu

NASA Astrophysics Data System (ADS)

Tovesson, F.; Duke, D.; Geppert-Kleinrath, V.; Manning, B.; Mayorov, D.; Mosby, S.; Schmitt, K.

2018-03-01

Different aspects of the nuclear fission process have been studied at Los Alamos Neutron Science Center (LANSCE) using various instruments and experimental techniques. Properties of the fragments emitted in fission have been investigated using Frisch-grid ionization chambers, a Time Projection Chamber (TPC), and the SPIDER instrument which employs the 2v-2E method. These instruments and experimental techniques have been used to determine fission product mass yields, the energy dependent total kinetic energy (TKE) release, and anisotropy in neutron-induced fission of U-235, U-238 and Pu-239.

Fission-fragment total kinetic energy and mass yields for neutron-induced fission of 235U and 238U with En =200 keV - 30 MeV

NASA Astrophysics Data System (ADS)

Duke, D. L.; Tovesson, F.; Brys, T.; Geppert-Kleinrath, V.; Hambsch, F.-J.; Laptev, A.; Meharchand, R.; Manning, B.; Mayorov, D.; Meierbachtol, K.; Mosby, S.; Perdue, B.; Richman, D.; Shields, D.; Vidali, M.

2017-09-01

The average Total Kinetic Energy (TKE) release and fission-fragment yields in neutron-induced fission of 235U and 238U was measured using a Frisch-gridded ionization chamber. These observables are important nuclear data quantites that are relevant to applications and for informing the next generation of fission models. The measurements were performed a the Los Alamos Neutron Science Center and cover En = 200 keV - 30 MeV. The double-energy (2E) method was used to determine the fission-fragment yields and two methods of correcting for prompt-neutron emission were explored. The results of this study are correlated mass and TKE data.

Characterization of gamma rays existing in the NMIJ standard neutron field.

PubMed

Harano, H; Matsumoto, T; Ito, Y; Uritani, A; Kudo, K

2004-01-01

Our laboratory provides national standards on fast neutron fluence. Neutron fields are always accompanied by gamma rays produced in neutron sources and surroundings. We have characterised these gamma rays in the 5.0 MeV standard neutron field. Gamma ray measurement was performed using an NE213 liquid scintillator. Pulse shape discrimination was incorporated to separate the events induced by gamma rays from those by neutrons. The measured gamma ray spectra were unfolded with the HEPRO program package to obtain the spectral fluences using the response matrix prepared with the EGS4 code. Corrections were made for the gamma rays produced by neutrons in the detector assembly using the MCNP4C code. The effective dose equivalents were estimated to be of the order of 25 microSv at the neutron fluence of 10(7) neutrons cm(-2).

Kinetic Parameter Measurements in the MINERVE Reactor

NASA Astrophysics Data System (ADS)

Perret, Grégory; Geslot, Benoit; Gruel, Adrien; Blaise, Patrick; Di-Salvo, Jacques; De Izarra, Grégoire; Jammes, Christian; Hursin, Mathieu; Pautz, Andréas

2017-01-01

In the framework of an international collaboration, teams of the PSI and CEA research institutes measure the critical decay constant (α0 = β/A), delayed neutron fraction (β) and generation time (A) of the Minerve reactor using the Feynman-α, Power Spectral Density and Rossi-α neutron noise measurement techniques. These measurements contribute to the experimental database of kinetic parameters used to improve nuclear data files and validate modern methods in Monte Carlo codes. Minerve is a zero-power pool reactor composed of a central experimental test lattice surrounded by a large aluminum buffer and four high-enriched driver regions. Measurements are performed in three slightly subcritical configurations (-2 cents to -30 cents) using two high-efficiency 235U fission chambers in the driver regions. Measurement of α0 and β obtained by the two institutes and with the different techniques are consistent for the configurations envisaged. Slight increases of the β values are observed with the subcriticality level. Best estimate values are obtained with the Cross-Power Spectral Density technique at -2 cents, and are worth: β = 716.9±9.0 pcm, α0 = 79.0±0.6 s-1 and A = 90.7±1.4 μs. The kinetic parameters are predicted with MCNP5-v1.6 and TRIPOLI4.9 and the JEFF-3.1/3.1.1 and ENDF/B-VII.1 nuclear data libraries. The predictions for β and α0 overestimate the experimental results by 3-5% and 10-12%, respectively; that for A underestimate the experimental result by 6-7%. The discrepancies are suspected to come from the driven system nature of Minerve and the location of the detectors in the driver regions, which prevent accounting for the full reactor.

Comparison of the PHISICS/RELAP5-3D Ring and Block Model Results for Phase I of the OECD MHTGR-350 Benchmark

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gerhard Strydom

2014-04-01

The INL PHISICS code system consists of three modules providing improved core simulation capability: INSTANT (performing 3D nodal transport core calculations), MRTAU (depletion and decay heat generation) and a perturbation/mixer module. Coupling of the PHISICS code suite to the thermal hydraulics system code RELAP5-3D has recently been finalized, and as part of the code verification and validation program the exercises defined for Phase I of the OECD/NEA MHTGR 350 MW Benchmark were completed. This paper provides an overview of the MHTGR Benchmark, and presents selected results of the three steady state exercises 1-3 defined for Phase I. For Exercise 1,more » a stand-alone steady-state neutronics solution for an End of Equilibrium Cycle Modular High Temperature Reactor (MHTGR) was calculated with INSTANT, using the provided geometry, material descriptions, and detailed cross-section libraries. Exercise 2 required the modeling of a stand-alone thermal fluids solution. The RELAP5-3D results of four sub-cases are discussed, consisting of various combinations of coolant bypass flows and material thermophysical properties. Exercise 3 combined the first two exercises in a coupled neutronics and thermal fluids solution, and the coupled code suite PHISICS/RELAP5-3D was used to calculate the results of two sub-cases. The main focus of the paper is a comparison of the traditional RELAP5-3D “ring” model approach vs. a much more detailed model that include kinetics feedback on individual block level and thermal feedbacks on a triangular sub-mesh. The higher fidelity of the block model is illustrated with comparison results on the temperature, power density and flux distributions, and the typical under-predictions produced by the ring model approach are highlighted.« less

Neutron Angular Scatter Effects in 3DHZETRN: Quasi-Elastic

NASA Technical Reports Server (NTRS)

Wilson, John W.; Werneth, Charles M.; Slaba, Tony C.; Badavi, Francis F.; Reddell, Brandon D.; Bahadori, Amir A.

2017-01-01

The current 3DHZETRN code has a detailed three dimensional (3D) treatment of neutron transport based on a forward/isotropic assumption and has been compared to Monte Carlo (MC) simulation codes in various geometries. In most cases, it has been found that 3DHZETRN agrees with the MC codes to the extent they agree with each other. However, a recent study of neutron leakage from finite geometries revealed that further improvements to the 3DHZETRN formalism are needed. In the present report, angular scattering corrections to the neutron fluence are provided in an attempt to improve fluence estimates from a uniform sphere. It is found that further developments in the nuclear production models are required to fully evaluate the impact of transport model updates. A model for the quasi-elastic neutron production spectra is therefore developed and implemented into 3DHZETRN.

Direct detection of albedo neutron decay electrons at the inner edge of the radiation belt and experimental determination of neutron density in near-Earth space

NASA Astrophysics Data System (ADS)

Li, X.; Selesnick, R.; Schiller, Q. A.; Zhang, K.; Zhao, H.; Baker, D. N.; Temerin, M. A.

2017-12-01

The galaxy is filled with cosmic ray particles, mostly protons with kinetic energy above hundreds of mega-electron volts (MeV). Soon after the discovery of Earth's Van Allen radiation belts almost six decades ago, it was recognized that the main source of inner belt protons, with kinetic energies of tens to hundreds of MeV, is Cosmic Ray Albedo Neutron Decay (CRAND). In this process, cosmic rays reaching the upper atmosphere from throughout the galaxy interact with neutral atoms to produce albedo neutrons which, being unstable to 𝛽 decay, are a potential source of geomagnetically trapped protons and electrons. Protons retain most of the neutrons' kinetic energy while the electrons have lower energies, mostly below 1 MeV. The viability of the electron source was, however, uncertain because measurements showed that electron intensity can vary greatly while the neutron decay rate should be almost constant. Recent measurements from the Relativistic Electron and Proton Telescope integrated little experiment (REPTile) onboard the Colorado Student Space Weather Experiment (CSSWE) CubeSat now show that CRAND is the main electron source for the radiation belt near its inner edge, and also contributes to the inner belt elsewhere. Furthermore, measurement of the CRAND electron intensity provides the first experimental determination of the neutron density in near-Earth space, 2x10-9/cm3, confirming earlier theoretical estimates.

Neutron spectrometry in a mixed field of neutrons and protons with a phoswich neutron detector Part I: response functions for photons and neutrons of the phoswich neutron detector

NASA Astrophysics Data System (ADS)

Takada, M.; Taniguchi, S.; Nakamura, T.; Nakao, N.; Uwamino, Y.; Shibata, T.; Fujitaka, K.

2001-06-01

We have developed a phoswich neutron detector consisting of an NE213 liquid scintillator surrounded by an NE115 plastic scintillator to distinguish photon and neutron events in a charged-particle mixed field. To obtain the energy spectra by unfolding, the response functions to neutrons and photons were obtained by the experiment and calculation. The response functions to photons were measured with radionuclide sources, and were calculated with the EGS4-PRESTA code. The response functions to neutrons were measured with a white neutron source produced by the bombardment of 135 MeV protons onto a Be+C target using a TOF method, and were calculated with the SCINFUL code, which we revised in order to calculate neutron response functions up to 135 MeV. Based on these experimental and calculated results, response matrices for photons up to 20 MeV and neutrons up to 132 MeV could finally be obtained.

Synthetic neutron camera and spectrometer in JET based on AFSI-ASCOT simulations

NASA Astrophysics Data System (ADS)

Sirén, P.; Varje, J.; Weisen, H.; Koskela, T.; contributors, JET

2017-09-01

The ASCOT Fusion Source Integrator (AFSI) has been used to calculate neutron production rates and spectra corresponding to the JET 19-channel neutron camera (KN3) and the time-of-flight spectrometer (TOFOR) as ideal diagnostics, without detector-related effects. AFSI calculates fusion product distributions in 4D, based on Monte Carlo integration from arbitrary reactant distribution functions. The distribution functions were calculated by the ASCOT Monte Carlo particle orbit following code for thermal, NBI and ICRH particle reactions. Fusion cross-sections were defined based on the Bosch-Hale model and both DD and DT reactions have been included. Neutrons generated by AFSI-ASCOT simulations have already been applied as a neutron source of the Serpent neutron transport code in ITER studies. Additionally, AFSI has been selected to be a main tool as the fusion product generator in the complete analysis calculation chain: ASCOT - AFSI - SERPENT (neutron and gamma transport Monte Carlo code) - APROS (system and power plant modelling code), which encompasses the plasma as an energy source, heat deposition in plant structures as well as cooling and balance-of-plant in DEMO applications and other reactor relevant analyses. This conference paper presents the first results and validation of the AFSI DD fusion model for different auxiliary heating scenarios (NBI, ICRH) with very different fast particle distribution functions. Both calculated quantities (production rates and spectra) have been compared with experimental data from KN3 and synthetic spectrometer data from ControlRoom code. No unexplained differences have been observed. In future work, AFSI will be extended for synthetic gamma diagnostics and additionally, AFSI will be used as part of the neutron transport calculation chain to model real diagnostics instead of ideal synthetic diagnostics for quantitative benchmarking.

Effect of fast neutron, gamma-ray and combined radiations on the thermal decomposition of ammonium perchlorate single crystals

NASA Technical Reports Server (NTRS)

Herley, P. J.; Wang, C. S.; Varsi, G.; Levy, P. W.

1974-01-01

The thermal decomposition kinetics have been determined for ammonium perchlorate crystals subjected to a fast neutron irradiation or to a fast neutron irradiation followed by a gamma-ray irradiation. Qualitatively, the radiation induced changes are similar to those obtained in this and in previous studies, with samples exposed only to gamma rays. The induction period is shortened and the rate constants, obtained from an Avrami-Erofeyev kinetic analysis, are modified. The acceleratory period constant increases and the decay period constant decreases. When compared on an equal deposited energy basis, the fast neutron induced changes are appreciably larger than the gamma-ray induced changes. Some, or all, of the fast neutron induced effects might be attributable to the introduction of localized regions of concentrated radiation damage ('spikes') by lattice atom recoils which become thermal decomposition sites when the crystals are heated.

The infinite medium Green's function for neutron transport in plane geometry 40 years later

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ganapol, B.D.

1993-01-01

In 1953, the first of what was supposed to be two volumes on neutron transport theory was published. The monograph, entitled [open quotes]Introduction to the Theory of Neutron Diffusion[close quotes] by Case et al., appeared as a Los Alamos National Laboratory report and was to be followed by a second volume, which never appeared as intended because of the death of Placzek. Instead, Case and Zweifel collaborated on the now classic work entitled Linear Transport Theory 2 in which the underlying mathematical theory of linear transport was presented. The initial monograph, however, represented the coming of age of neutron transportmore » theory, which had its roots in radiative transfer and kinetic theory. In addition, it provided the first benchmark results along with the mathematical development for several fundamental neutron transport problems. In particular, one-dimensional infinite medium Green's functions for the monoenergetic transport equation in plane and spherical geometries were considered complete with numerical results to be used as standards to guide code development for applications. Unfortunately, because of the limited computational resources of the day, some numerical results were incorrect. Also, only conventional mathematics and numerical methods were used because the transport theorists of the day were just becoming acquainted with more modern mathematical approaches. In this paper, Green's function solution is revisited in light of modern numerical benchmarking methods with an emphasis on evaluation rather than theoretical results. The primary motivation for considering the Green's function at this time is its emerging use in solving finite and heterogeneous media transport problems.« less

KAOS/LIB-V: A library of nuclear response functions generated by KAOS-V code from ENDF/B-V and other data files

DOE Office of Scientific and Technical Information (OSTI.GOV)

Farawila, Y.; Gohar, Y.; Maynard, C.

1989-04-01

KAOS/LIB-V: A library of processed nuclear responses for neutronics analyses of nuclear systems has been generated. The library was prepared using the KAOS-V code and nuclear data from ENDF/B-V. The library includes kerma (kinetic energy released in materials) factors and other nuclear response functions for all materials presently of interest in fusion and fission applications for 43 nonfissionable and 15 fissionable isotopes and elements. The nuclear response functions include gas production and tritium-breeding functions, and all important reaction cross sections. KAOS/LIB-V employs the VITAMIN-E weighting function and energy group structure of 174 neutron groups. Auxiliary nuclear data bases, e.g., themore » Japanese evaluated nuclear data library JENDL-2 were used as a source of isotopic cross sections when these data are not provided in ENDF/B-V files for a natural element. These are needed mainly to estimate average quantities such as effective Q-values for the natural element. This analysis of local energy deposition was instrumental in detecting and understanding energy balance deficiencies and other problems in the ENDF/B-V data. Pertinent information about the library and a graphical display of the main nuclear response functions for all materials in the library are given. 35 refs.« less

Neutrino-driven Explosion of a 20 Solar-mass Star in Three Dimensions Enabled by Strange-quark Contributions to Neutrino-Nucleon Scattering

NASA Astrophysics Data System (ADS)

Melson, Tobias; Janka, Hans-Thomas; Bollig, Robert; Hanke, Florian; Marek, Andreas; Müller, Bernhard

2015-08-01

Interactions with neutrons and protons play a crucial role for the neutrino opacity of matter in the supernova core. Their current implementation in many simulation codes, however, is rather schematic and ignores not only modifications for the correlated nuclear medium of the nascent neutron star, but also free-space corrections from nucleon recoil, weak magnetism, or strange quarks, which can easily add up to changes of several 10% for neutrino energies in the spectral peak. In the Garching supernova simulations with the Prometheus-Vertex code, such sophistications have been included for a long time except for the strange-quark contributions to the nucleon spin, which affect neutral-current neutrino scattering. We demonstrate on the basis of a 20 {M}⊙ progenitor star that a moderate strangeness-dependent contribution of {g}{{a}}{{s}}=-0.2 to the axial-vector coupling constant {g}{{a}}≈ 1.26 can turn an unsuccessful three-dimensional (3D) model into a successful explosion. Such a modification is in the direction of current experimental results and reduces the neutral-current scattering opacity of neutrons, which dominate in the medium around and above the neutrinosphere. This leads to increased luminosities and mean energies of all neutrino species and strengthens the neutrino-energy deposition in the heating layer. Higher nonradial kinetic energy in the gain layer signals enhanced buoyancy activity that enables the onset of the explosion at ˜300 ms after bounce, in contrast to the model with vanishing strangeness contributions to neutrino-nucleon scattering. Our results demonstrate the close proximity to explosion of the previously published, unsuccessful 3D models of the Garching group.

TREAT Transient Analysis Benchmarking for the HEU Core

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kontogeorgakos, D. C.; Connaway, H. M.; Wright, A. E.

2014-05-01

This work was performed to support the feasibility study on the potential conversion of the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory from the use of high enriched uranium (HEU) fuel to the use of low enriched uranium (LEU) fuel. The analyses were performed by the GTRI Reactor Conversion staff at the Argonne National Laboratory (ANL). The objective of this study was to benchmark the transient calculations against temperature-limited transients performed in the final operating HEU TREAT core configuration. The MCNP code was used to evaluate steady-state neutronics behavior, and the point kinetics code TREKIN was used tomore » determine core power and energy during transients. The first part of the benchmarking process was to calculate with MCNP all the neutronic parameters required by TREKIN to simulate the transients: the transient rod-bank worth, the prompt neutron generation lifetime, the temperature reactivity feedback as a function of total core energy, and the core-average temperature and peak temperature as a functions of total core energy. The results of these calculations were compared against measurements or against reported values as documented in the available TREAT reports. The heating of the fuel was simulated as an adiabatic process. The reported values were extracted from ANL reports, intra-laboratory memos and experiment logsheets and in some cases it was not clear if the values were based on measurements, on calculations or a combination of both. Therefore, it was decided to use the term “reported” values when referring to such data. The methods and results from the HEU core transient analyses will be used for the potential LEU core configurations to predict the converted (LEU) core’s performance.« less

A comparison of the COG and MCNP codes in computational neutron capture therapy modeling, Part II: gadolinium neutron capture therapy models and therapeutic effects.

PubMed

Wangerin, K; Culbertson, C N; Jevremovic, T

2005-08-01

The goal of this study was to evaluate the COG Monte Carlo radiation transport code, developed and tested by Lawrence Livermore National Laboratory, for gadolinium neutron capture therapy (GdNCT) related modeling. The validity of COG NCT model has been established for this model, and here the calculation was extended to analyze the effect of various gadolinium concentrations on dose distribution and cell-kill effect of the GdNCT modality and to determine the optimum therapeutic conditions for treating brain cancers. The computational results were compared with the widely used MCNP code. The differences between the COG and MCNP predictions were generally small and suggest that the COG code can be applied to similar research problems in NCT. Results for this study also showed that a concentration of 100 ppm gadolinium in the tumor was most beneficial when using an epithermal neutron beam.

Measurement of neutron spectra in the AWE workplace using a Bonner sphere spectrometer.

PubMed

Danyluk, Peter

2010-12-01

A Bonner sphere spectrometer has been used to measure the neutron spectra in eight different workplace areas at AWE (Atomic Weapons Establishment). The spectra were analysed by the National Physical Laboratory using their principal unfolding code STAY'SL and the results were also analysed by AWE using a bespoke parametrised unfolding code. The bespoke code was designed specifically for the AWE workplace and is very simple to use. Both codes gave results, in good agreement. It was found that the measured fluence rate varied from 2 to 70 neutrons cm⁻² s⁻¹ (± 10%) and the ambient dose equivalent H*(10) varied from 0.5 to 57 µSv h⁻¹ (± 20%). A detailed description of the development and use of the bespoke code is presented.

GAMSOR: Gamma Source Preparation and DIF3D Flux Solution

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, M. A.; Lee, C. H.; Hill, R. N.

2016-12-15

Nuclear reactors that rely upon the fission reaction have two modes of thermal energy deposition in the reactor system: neutron absorption and gamma absorption. The gamma rays are typically generated by neutron absorption reactions or during the fission process which means the primary driver of energy production is of course the neutron interaction. In conventional reactor physics methods, the gamma heating component is ignored such that the gamma absorption is forced to occur at the gamma emission site. For experimental reactor systems like EBR-II and FFTF, the placement of structural pins and assemblies internal to the core leads to problemsmore » with power heating predictions because there is no fission power source internal to the assembly to dictate a spatial distribution of the power. As part of the EBR-II support work in the 1980s, the GAMSOR code was developed to assist analysts in calculating the gamma heating. The GAMSOR code is a modified version of DIF3D and actually functions within a sequence of DIF3D calculations. The gamma flux in a conventional fission reactor system does not perturb the neutron flux and thus the gamma flux calculation can be cast as a fixed source problem given a solution to the steady state neutron flux equation. This leads to a sequence of DIF3D calculations, called the GAMSOR sequence, which involves solving the neutron flux, then the gamma flux, then combining the results to do a summary edit. In this manuscript, we go over the GAMSOR code and detail how it is put together and functions. We also discuss how to setup the GAMSOR sequence and input for each DIF3D calculation in the GAMSOR sequence. With the GAMSOR capability, users can take any valid steady state DIF3D calculation and compute the power distribution due to neutron and gamma heating. The MC2-3 code is the preferable companion code to use for generating neutron and gamma cross section data, but the GAMSOR code can accept cross section data from other sources. To further this aspect, an additional utility code was created which demonstrates how to merge the neutron and gamma cross section data together to carry out a simultaneous solve of the two systems.« less

Comparison of the thermal neutron scattering treatment in MCNP6 and GEANT4 codes

NASA Astrophysics Data System (ADS)

Tran, H. N.; Marchix, A.; Letourneau, A.; Darpentigny, J.; Menelle, A.; Ott, F.; Schwindling, J.; Chauvin, N.

2018-06-01

To ensure the reliability of simulation tools, verification and comparison should be made regularly. This paper describes the work performed in order to compare the neutron transport treatment in MCNP6.1 and GEANT4-10.3 in the thermal energy range. This work focuses on the thermal neutron scattering processes for several potential materials which would be involved in the neutron source designs of Compact Accelerator-based Neutrons Sources (CANS), such as beryllium metal, beryllium oxide, polyethylene, graphite, para-hydrogen, light water, heavy water, aluminium and iron. Both thermal scattering law and free gas model, coming from the evaluated data library ENDF/B-VII, were considered. It was observed that the GEANT4.10.03-patch2 version was not able to account properly the coherent elastic process occurring in crystal lattice. This bug is treated in this work and it should be included in the next release of the code. Cross section sampling and integral tests have been performed for both simulation codes showing a fair agreement between the two codes for most of the materials except for iron and aluminium.

Measurements and parameterization of neutron energy spectra from targets bombarded with 120 GeV protons

NASA Astrophysics Data System (ADS)

Kajimoto, T.; Shigyo, N.; Sanami, T.; Iwamoto, Y.; Hagiwara, M.; Lee, H. S.; Soha, A.; Ramberg, E.; Coleman, R.; Jensen, D.; Leveling, A.; Mokhov, N. V.; Boehnlein, D.; Vaziri, K.; Sakamoto, Y.; Ishibashi, K.; Nakashima, H.

2014-10-01

The energy spectra of neutrons were measured by a time-of-flight method for 120 GeV protons on thick graphite, aluminum, copper, and tungsten targets with an NE213 scintillator at the Fermilab Test Beam Facility. Neutron energy spectra were obtained between 25 and 3000 MeV at emission angles of 30°, 45°, 120°, and 150°. The spectra were parameterized as neutron emissions from three moving sources and then compared with theoretical spectra calculated by PHITS and FLUKA codes. The yields of the theoretical spectra were substantially underestimated compared with the yields of measured spectra. The integrated neutron yields from 25 to 3000 MeV calculated with PHITS code were 16-36% of the experimental yields and those calculated with FLUKA code were 26-57% of the experimental yields for all targets and emission angles.

Neutron skyshine from intense 14-MeV neutron source facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nakamura, T.; Hayashi, K.; Takahashi, A.

1985-07-01

The dose distribution and the spectrum variation of neutrons due to the skyshine effect have been measured with the high-efficiency rem counter, the multisphere spectrometer, and the NE-213 scintillator in the environment surrounding an intense 14-MeV neutron source facility. The dose distribution and the energy spectra of neutrons around the facility used as a skyshine source have also been measured to enable the absolute evaluation of the skyshine effect. The skyshine effect was analyzed by two multigroup Monte Carlo codes, NIMSAC and MMCR-2, by two discrete ordinates S /sub n/ codes, ANISN and DOT3.5, and by the shield structure designmore » code for skyshine, SKYSHINE-II. The calculated results show good agreement with the measured results in absolute values. These experimental results should be useful as benchmark data for shyshine analysis and for shielding design of fusion facilities.« less

Neutron flux spectrum revealed by Nb-based current-biased kinetic inductance detector with a 10B conversion layer

NASA Astrophysics Data System (ADS)

Miyajima, Shigeyuki; Shishido, Hiroaki; Narukami, Yoshito; Yoshioka, Naohito; Fujimaki, Akira; Hidaka, Mutsuo; Oikawa, Kenichi; Harada, Masahide; Oku, Takayuki; Arai, Masatoshi; Ishida, Takekazu

2017-01-01

We successfully derived the time-dependent flux of pulsed neutrons using a superconducting Nb-based current-biased kinetic inductance detector (CB-KID) with a 10B conversion layer at Japan Proton Accelerator Research Complex. Our CB-KID is a meander line made of a 40-nm-thick Nb thin film with 1 - μm line width, which is covered with a 150-nm-thick 10B conversion layer. The detector works at a temperature below 4 K. The evaluated detection efficiency of the CB-KID in this experiment is 0.23 % at the neutron energy of 25.4 meV. The time-dependent flux spectra of pulsed neutrons thus obtained are in good agreement with the results obtained by the Monte Carlo simulations.

Neutron detection using a current biased kinetic inductance detector

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shishido, Hiroaki, E-mail: shishido@pe.osakafu-u.ac.jp; Miyajima, Shigeyuki; Ishida, Takekazu

2015-12-07

We demonstrate neutron detection using a solid state superconducting current biased kinetic inductance detector (CB-KID), which consists of a superconducting Nb meander line of 1 μm width and 40 nm thickness. {sup 10}B-enriched neutron absorber layer of 150 nm thickness is placed on top of the CB-KID. Our neutron detectors are able to operate in a wide superconducting region in the bias current–temperature diagram. This is in sharp contrast with our preceding current-biased transition edge detector, which can operate only in a narrow range just below the superconducting critical temperature. The full width at half maximum of the signals remains of the ordermore » of a few tens of ns, which confirms the high speed operation of our detectors.« less

Least-Squares Neutron Spectral Adjustment with STAYSL PNNL

NASA Astrophysics Data System (ADS)

Greenwood, L. R.; Johnson, C. D.

2016-02-01

The STAYSL PNNL computer code, a descendant of the STAY'SL code [1], performs neutron spectral adjustment of a starting neutron spectrum, applying a least squares method to determine adjustments based on saturated activation rates, neutron cross sections from evaluated nuclear data libraries, and all associated covariances. STAYSL PNNL is provided as part of a comprehensive suite of programs [2], where additional tools in the suite are used for assembling a set of nuclear data libraries and determining all required corrections to the measured data to determine saturated activation rates. Neutron cross section and covariance data are taken from the International Reactor Dosimetry File (IRDF-2002) [3], which was sponsored by the International Atomic Energy Agency (IAEA), though work is planned to update to data from the IAEA's International Reactor Dosimetry and Fusion File (IRDFF) [4]. The nuclear data and associated covariances are extracted from IRDF-2002 using the third-party NJOY99 computer code [5]. The NJpp translation code converts the extracted data into a library data array format suitable for use as input to STAYSL PNNL. The software suite also includes three utilities to calculate corrections to measured activation rates. Neutron self-shielding corrections are calculated as a function of neutron energy with the SHIELD code and are applied to the group cross sections prior to spectral adjustment, thus making the corrections independent of the neutron spectrum. The SigPhi Calculator is a Microsoft Excel spreadsheet used for calculating saturated activation rates from raw gamma activities by applying corrections for gamma self-absorption, neutron burn-up, and the irradiation history. Gamma self-absorption and neutron burn-up corrections are calculated (iteratively in the case of the burn-up) within the SigPhi Calculator spreadsheet. The irradiation history corrections are calculated using the BCF computer code and are inserted into the SigPhi Calculator workbook for use in correcting the measured activities. Output from the SigPhi Calculator is automatically produced, and consists of a portion of the STAYSL PNNL input file data that is required to run the spectral adjustment calculations. Within STAYSL PNNL, the least-squares process is performed in one step, without iteration, and provides rapid results on PC platforms. STAYSL PNNL creates multiple output files with tabulated results, data suitable for plotting, and data formatted for use in subsequent radiation damage calculations using the SPECTER computer code (which is not included in the STAYSL PNNL suite). All components of the software suite have undergone extensive testing and validation prior to release and test cases are provided with the package.

UFO: A THREE-DIMENSIONAL NEUTRON DIFFUSION CODE FOR THE IBM 704

DOE Office of Scientific and Technical Information (OSTI.GOV)

Auerbach, E.H.; Jewett, J.P.; Ketchum, M.A.

A description of UFO, a code for the solution of the fewgroup neutron diffusion equation in three-dimensional Cartesian coordinates on the IBM 704, is given. An accelerated Liebmann flux iteration scheme is used, and optimum parameters can be calculated by the code whenever they are required. The theory and operation of the program are discussed. (auth)

Implementation of a tree algorithm in MCNP code for nuclear well logging applications.

PubMed

Li, Fusheng; Han, Xiaogang

2012-07-01

The goal of this paper is to develop some modeling capabilities that are missing in the current MCNP code. Those missing capabilities can greatly help for some certain nuclear tools designs, such as a nuclear lithology/mineralogy spectroscopy tool. The new capabilities to be developed in this paper include the following: zone tally, neutron interaction tally, gamma rays index tally and enhanced pulse-height tally. The patched MCNP code also can be used to compute neutron slowing-down length and thermal neutron diffusion length. Copyright © 2011 Elsevier Ltd. All rights reserved.

Intercomparison of Monte Carlo radiation transport codes to model TEPC response in low-energy neutron and gamma-ray fields.

PubMed

Ali, F; Waker, A J; Waller, E J

2014-10-01

Tissue-equivalent proportional counters (TEPC) can potentially be used as a portable and personal dosemeter in mixed neutron and gamma-ray fields, but what hinders this use is their typically large physical size. To formulate compact TEPC designs, the use of a Monte Carlo transport code is necessary to predict the performance of compact designs in these fields. To perform this modelling, three candidate codes were assessed: MCNPX 2.7.E, FLUKA 2011.2 and PHITS 2.24. In each code, benchmark simulations were performed involving the irradiation of a 5-in. TEPC with monoenergetic neutron fields and a 4-in. wall-less TEPC with monoenergetic gamma-ray fields. The frequency and dose mean lineal energies and dose distributions calculated from each code were compared with experimentally determined data. For the neutron benchmark simulations, PHITS produces data closest to the experimental values and for the gamma-ray benchmark simulations, FLUKA yields data closest to the experimentally determined quantities. © The Author 2013. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

The NASA Neutron Star Grand Challenge: The coalescences of Neutron Star Binary System

NASA Astrophysics Data System (ADS)

Suen, Wai-Mo

1998-04-01

NASA funded a Grand Challenge Project (9/1996-1999) for the development of a multi-purpose numerical treatment for relativistic astrophysics and gravitational wave astronomy. The coalescence of binary neutron stars is chosen as the model problem for the code development. The institutes involved in it are the Argonne Lab, Livermore lab, Max-Planck Institute at Potsdam, StonyBrook, U of Illinois and Washington U. We have recently succeeded in constructing a highly optimized parallel code which is capable of solving the full Einstein equations coupled with relativistic hydrodynamics, running at over 50 GFLOPS on a T3E (the second milestone point of the project). We are presently working on the head-on collisions of two neutron stars, and the inclusion of realistic equations of state into the code. The code will be released to the relativity and astrophysics community in April of 1998. With the full dynamics of the spacetime, relativistic hydro and microphysics all combined into a unified 3D code for the first time, many interesting large scale calculations in general relativistic astrophysics can now be carried out on massively parallel computers.

Measurement of electrons from albedo neutron decay and neutron density in near-Earth space.

PubMed

Li, Xinlin; Selesnick, Richard; Schiller, Quintin; Zhang, Kun; Zhao, Hong; Baker, Daniel N; Temerin, Michael A

2017-12-21

The Galaxy is filled with cosmic-ray particles, mostly protons with kinetic energies greater than hundreds of megaelectronvolts. Around Earth, trapped energetic protons, electrons and other particles circulate at altitudes from about 500 to 40,000 kilometres in the Van Allen radiation belts. Soon after these radiation belts were discovered six decades ago, it was recognized that the main source of inner-belt protons (with kinetic energies of tens to hundreds of megaelectronvolts) is cosmic-ray albedo neutron decay (CRAND). In this process, cosmic rays that reach the upper atmosphere interact with neutral atoms to produce albedo neutrons, which, being prone to β-decay, are a possible source of geomagnetically trapped protons and electrons. These protons would retain most of the kinetic energy of the neutrons, while the electrons would have lower energies, mostly less than one megaelectronvolt. The viability of CRAND as an electron source has, however, been uncertain, because measurements have shown that the electron intensity in the inner Van Allen belt can vary greatly, while the neutron-decay rate should be almost constant. Here we report measurements of relativistic electrons near the inner edge of the inner radiation belt. We demonstrate that the main source of these electrons is indeed CRAND, and that this process also contributes to electrons in the inner belt elsewhere. Furthermore, measurement of the intensity of electrons generated by CRAND provides an experimental determination of the neutron density in near-Earth space-2 × 10 -9 per cubic centimetre-confirming theoretical estimates.

Benchmark Analysis of Pion Contribution from Galactic Cosmic Rays

NASA Technical Reports Server (NTRS)

Aghara, Sukesh K.; Blattnig, Steve R.; Norbury, John W.; Singleterry, Robert C., Jr.

2008-01-01

Shielding strategies for extended stays in space must include a comprehensive resolution of the secondary radiation environment inside the spacecraft induced by the primary, external radiation. The distribution of absorbed dose and dose equivalent is a function of the type, energy and population of these secondary products. A systematic verification and validation effort is underway for HZETRN, which is a space radiation transport code currently used by NASA. It performs neutron, proton and heavy ion transport explicitly, but it does not take into account the production and transport of mesons, photons and leptons. The question naturally arises as to what is the contribution of these particles to space radiation. The pion has a production kinetic energy threshold of about 280 MeV. The Galactic cosmic ray (GCR) spectra, coincidentally, reaches flux maxima in the hundreds of MeV range, corresponding to the pion production threshold. We present results from the Monte Carlo code MCNPX, showing the effect of lepton and meson physics when produced and transported explicitly in a GCR environment.

Analysis of the Temporal Response of Coupled Asymmetrical Zero-Power Subcritical Bare Metal Reactor Systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Klain, Kimberly L.

The behavior of symmetrical coupled-core systems has been extensively studied, yet there is a dearth of research on asymmetrical systems due to the increased complexity of the analysis of such systems. In this research, the multipoint kinetics method is applied to asymmetrical zeropower, subcritical, bare metal reactor systems. Existing research on asymmetrical reactor systems assumes symmetry in the neutronic coupling; however, it will be shown that this cannot always be assumed. Deep subcriticality adds another layer of complexity and requires modification of the multipoint kinetics equations to account for the effect of the external neutron source. A modified set ofmore » multipoint kinetics equations is derived with this in mind. Subsequently, the Rossi-alpha equations are derived for a two-region asymmetrical reactor system. The predictive capabilities of the radiation transport code MCNP6 for neutron noise experiments are shown in a comparison to the results of a series of Rossi-alpha measurements performed by J. Mihalczo utilizing a coupled set of symmetrical bare highly-enriched uranium (HEU) cylinders. The ptrac option within MCNP6 can generate time-tagged counts in a cell (list-mode data). The list-mode data can then be processed similarly to measured data to obtain values for system parameters such as the dual prompt neutron decay constants observable in a coupled system. The results from the ptrac simulations agree well with the historical measured values. A series of case studies are conducted to study the effects of geometrical asymmetry in the coupling between two bare metal HEU cylinders. While the coupling behavior of symmetrical systems has been reported on extensively, that of asymmetrical systems remains sparse. In particular, it appears that there has been no previous research in obtaining the coupling time constants for asymmetrically-coupled systems. The difficulty in observing such systems is due in part to the inability to determine the individual coupling coefficients from measurement: unlike the symmetrical cases, only the product of the values can be obtained. A method is proposed utilizing MCNP6 tally ratios to separate the coupling coefficients for such systems. This work provides insight into the behavior of asymmetrically-coupled systems as the separation distance between the two cores is changed and also as the asymmetry is increased. As the asymmetry increases, both the slower and the faster observable prompt neutron decay constants increase in magnitude. The coupling time constants are determined from the measured decay constants. As the separation distance increases, both coupling coefficients decrease as expected. Based on these findings, an effective computational method utilizing MCNP6 and the Rossialpha technique can be applied to the prediction of asymmetrical coupled system measurements.« less

Binary neutron stars with arbitrary spins in numerical relativity

NASA Astrophysics Data System (ADS)

Tacik, Nick; Foucart, Francois; Pfeiffer, Harald P.; Haas, Roland; Ossokine, Serguei; Kaplan, Jeff; Muhlberger, Curran; Duez, Matt D.; Kidder, Lawrence E.; Scheel, Mark A.; Szilágyi, Béla

2015-12-01

We present a code to construct initial data for binary neutron star systems in which the stars are rotating. Our code, based on a formalism developed by Tichy, allows for arbitrary rotation axes of the neutron stars and is able to achieve rotation rates near rotational breakup. We compute the neutron star angular momentum through quasilocal angular momentum integrals. When constructing irrotational binary neutron stars, we find a very small residual dimensionless spin of ˜2 ×10-4 . Evolutions of rotating neutron star binaries show that the magnitude of the stars' angular momentum is conserved, and that the spin and orbit precession of the stars is well described by post-Newtonian approximation. We demonstrate that orbital eccentricity of the binary neutron stars can be controlled to ˜0.1 % . The neutron stars show quasinormal mode oscillations at an amplitude which increases with the rotation rate of the stars.

Some Experimental and Monte Carlo Investigations of the Plastic Scintillators for the Current Mode Measurements at Pulsed Neutron Sources

NASA Astrophysics Data System (ADS)

Rogov, A.; Pepyolyshev, Yu.; Carta, M.; d'Angelo, A.

Scintillation detector (SD) is widely used in neutron and gamma-spectrometry in a count mode. The organic scintillators for the count mode of the detector operation are investigated rather well. Usually, they are applied for measurement of amplitude and time distributions of pulses caused by single interaction events of neutrons or gamma's with scintillator material. But in a large area of scientific research scintillation detectors can alternatively be used on a current mode by recording the average current from the detector. For example,the measurements of the neutron pulse shape at the pulsed reactors or another pulsed neutron sources. So as to get a rather large volume of experimental data at pulsed neutron sources, it is necessary to use the current mode detector for registration of fast neutrons. Many parameters of the SD are changed with a transition from an accounting mode to current one. For example, the detector efficiency is different in counting and current modes. Many effects connected with time accuracy become substantial. Besides, for the registration of solely fast neutrons, as must be in many measurements, in the mixed radiation field of the pulsed neutron sources, SD efficiency has to be determined with a gamma-radiation shield present. Here is no calculations or experimental data on SD current mode operation up to now. The response functions of the detectors can be either measured in high-precision reference fields or calculated by a computer simulation. We have used the MCNP code [1] and carried out some experiments for investigation of the plastic performances in a current mode. There are numerous programs performing simulating similar to the MCNP code. For example, for neutrons there are [2-4], for photons - [5-8]. However, all known codes to use (SCINFUL, NRESP4, SANDYL, EGS49) have more stringent restrictions on the source, geometry and detector characteristics. In MCNP code a lot of these restrictions are absent and you need only to write special additions for proton and electron recoil and transfer energy to light output. These code modifications allow taking into account all processes in organic scintillator influence the light yield.

CAFNA{reg{underscore}sign}, coded aperture fast neutron analysis for contraband detection: Preliminary results

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhang, L.; Lanza, R.C.

1999-12-01

The authors have developed a near field coded aperture imaging system for use with fast neutron techniques as a tool for the detection of contraband and hidden explosives through nuclear elemental analysis. The technique relies on the prompt gamma rays produced by fast neutron interactions with the object being examined. The position of the nuclear elements is determined by the location of the gamma emitters. For existing fast neutron techniques, in Pulsed Fast Neutron Analysis (PFNA), neutrons are used with very low efficiency; in Fast Neutron Analysis (FNS), the sensitivity for detection of the signature gamma rays is very low.more » For the Coded Aperture Fast Neutron Analysis (CAFNA{reg{underscore}sign}) the authors have developed, the efficiency for both using the probing fast neutrons and detecting the prompt gamma rays is high. For a probed volume of n{sup 3} volume elements (voxels) in a cube of n resolution elements on a side, they can compare the sensitivity with other neutron probing techniques. As compared to PFNA, the improvement for neutron utilization is n{sup 2}, where the total number of voxels in the object being examined is n{sup 3}. Compared to FNA, the improvement for gamma-ray imaging is proportional to the total open area of the coded aperture plane; a typical value is n{sup 2}/2, where n{sup 2} is the number of total detector resolution elements or the number of pixels in an object layer. It should be noted that the actual signal to noise ratio of a system depends also on the nature and distribution of background events and this comparison may reduce somewhat the effective sensitivity of CAFNA. They have performed analysis, Monte Carlo simulations, and preliminary experiments using low and high energy gamma-ray sources. The results show that a high sensitivity 3-D contraband imaging and detection system can be realized by using CAFNA.« less

Optimization of beam shaping assembly based on D-T neutron generator and dose evaluation for BNCT

NASA Astrophysics Data System (ADS)

Naeem, Hamza; Chen, Chaobin; Zheng, Huaqing; Song, Jing

2017-04-01

The feasibility of developing an epithermal neutron beam for a boron neutron capture therapy (BNCT) facility based on a high intensity D-T fusion neutron generator (HINEG) and using the Monte Carlo code SuperMC (Super Monte Carlo simulation program for nuclear and radiation process) is proposed in this study. The Monte Carlo code SuperMC is used to determine and optimize the final configuration of the beam shaping assembly (BSA). The optimal BSA design in a cylindrical geometry which consists of a natural uranium sphere (14 cm) as a neutron multiplier, AlF3 and TiF3 as moderators (20 cm each), Cd (1 mm) as a thermal neutron filter, Bi (5 cm) as a gamma shield, and Pb as a reflector and collimator to guide neutrons towards the exit window. The epithermal neutron beam flux of the proposed model is 5.73 × 109 n/cm2s, and other dosimetric parameters for the BNCT reported by IAEA-TECDOC-1223 have been verified. The phantom dose analysis shows that the designed BSA is accurate, efficient and suitable for BNCT applications. Thus, the Monte Carlo code SuperMC is concluded to be capable of simulating the BSA and the dose calculation for BNCT, and high epithermal flux can be achieved using proposed BSA.

Nuclear reactor transient analysis via a quasi-static kinetics Monte Carlo method

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jo, YuGwon; Cho, Bumhee; Cho, Nam Zin, E-mail: nzcho@kaist.ac.kr

2015-12-31

The predictor-corrector quasi-static (PCQS) method is applied to the Monte Carlo (MC) calculation for reactor transient analysis. To solve the transient fixed-source problem of the PCQS method, fission source iteration is used and a linear approximation of fission source distributions during a macro-time step is introduced to provide delayed neutron source. The conventional particle-tracking procedure is modified to solve the transient fixed-source problem via MC calculation. The PCQS method with MC calculation is compared with the direct time-dependent method of characteristics (MOC) on a TWIGL two-group problem for verification of the computer code. Then, the results on a continuous-energy problemmore » are presented.« less

Effect of driver impedance on dense plasma focus Z-pinch neutron yield

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sears, Jason, E-mail: sears8@llnl.gov, E-mail: schmidt36@llnl.gov; Link, Anthony, E-mail: sears8@llnl.gov, E-mail: schmidt36@llnl.gov; Schmidt, Andrea, E-mail: sears8@llnl.gov, E-mail: schmidt36@llnl.gov

2014-12-15

The Z-pinch phase of a dense plasma focus (DPF) heats the plasma by rapid compression and accelerates ions across its intense electric fields, producing neutrons through both thermonuclear and beam-target fusion. Driver characteristics have empirically been shown to affect performance, as measured by neutron yield per unit of stored energy. We are exploring the effect of driver characteristics on DPF performance using particle-in-cell (PIC) simulations of a kJ scale DPF. In this work, our PIC simulations are fluid for the run-down phase and transition to fully kinetic for the pinch phase, capturing kinetic instabilities, anomalous resistivity, and beam formation duringmore » the pinch. The anode-cathode boundary is driven by a circuit model of the capacitive driver, including system inductance, the load of the railgap switches, the guard resistors, and the coaxial transmission line parameters. It is known that the driver impedance plays an important role in the neutron yield: first, it sets the peak current achieved at pinch time; and second, it affects how much current continues to flow through the pinch when the pinch inductance and resistance suddenly increase. Here we show from fully kinetic simulations how total neutron yield depends on the impedance of the driver and the distributed parameters of the transmission circuit. Direct comparisons between the experiment and simulations enhance our understanding of these plasmas and provide predictive design capability for neutron source applications.« less

Accurate Ray-tracing of Realistic Neutron Star Atmospheres for Constraining Their Parameters

NASA Astrophysics Data System (ADS)

Vincent, Frederic H.; Bejger, Michał; Różańska, Agata; Straub, Odele; Paumard, Thibaut; Fortin, Morgane; Madej, Jerzy; Majczyna, Agnieszka; Gourgoulhon, Eric; Haensel, Paweł; Zdunik, Leszek; Beldycki, Bartosz

2018-03-01

Thermal-dominated X-ray spectra of neutron stars in quiescent, transient X-ray binaries and neutron stars that undergo thermonuclear bursts are sensitive to mass and radius. The mass–radius relation of neutron stars depends on the equation of state (EoS) that governs their interior. Constraining this relation accurately is therefore of fundamental importance to understand the nature of dense matter. In this context, we introduce a pipeline to calculate realistic model spectra of rotating neutron stars with hydrogen and helium atmospheres. An arbitrarily fast-rotating neutron star with a given EoS generates the spacetime in which the atmosphere emits radiation. We use the LORENE/NROTSTAR code to compute the spacetime numerically and the ATM24 code to solve the radiative transfer equations self-consistently. Emerging specific intensity spectra are then ray-traced through the neutron star’s spacetime from the atmosphere to a distant observer with the GYOTO code. Here, we present and test our fully relativistic numerical pipeline. To discuss and illustrate the importance of realistic atmosphere models, we compare our model spectra to simpler models like the commonly used isotropic color-corrected blackbody emission. We highlight the importance of considering realistic model-atmosphere spectra together with relativistic ray-tracing to obtain accurate predictions. We also insist upon the crucial impact of the star’s rotation on the observables. Finally, we close a controversy that has been ongoing in the literature in the recent years, regarding the validity of the ATM24 code.

Experimental measurements with Monte Carlo corrections and theoretical calculations of neutron inelastic scattering cross section of 115In

NASA Astrophysics Data System (ADS)

Wang, Chao; Xiao, Jun; Luo, Xiaobing

2016-10-01

The neutron inelastic scattering cross section of 115In has been measured by the activation technique at neutron energies of 2.95, 3.94, and 5.24 MeV with the neutron capture cross sections of 197Au as an internal standard. The effects of multiple scattering and flux attenuation were corrected using the Monte Carlo code GEANT4. Based on the experimental values, the 115In neutron inelastic scattering cross sections data were theoretically calculated between the 1 and 15 MeV with the TALYS software code, the theoretical results of this study are in reasonable agreement with the available experimental results.

Monte Carlo calculations of thermal neutron capture in gadolinium: a comparison of GEANT4 and MCNP with measurements.

PubMed

Enger, Shirin A; Munck af Rosenschöld, Per; Rezaei, Arash; Lundqvist, Hans

2006-02-01

GEANT4 is a Monte Carlo code originally implemented for high-energy physics applications and is well known for particle transport at high energies. The capacity of GEANT4 to simulate neutron transport in the thermal energy region is not equally well known. The aim of this article is to compare MCNP, a code commonly used in low energy neutron transport calculations and GEANT4 with experimental results and select the suitable code for gadolinium neutron capture applications. To account for the thermal neutron scattering from chemically bound atoms [S(alpha,beta)] in biological materials a comparison of thermal neutron fluence in tissue-like poly(methylmethacrylate) phantom is made with MCNP4B, GEANT4 6.0 patch1, and measurements from the neutron capture therapy (NCT) facility at the Studsvik, Sweden. The fluence measurements agreed with MCNP calculated results considering S(alpha,beta). The location of the thermal neutron peak calculated with MCNP without S(alpha,beta) and GEANT4 is shifted by about 0.5 cm towards a shallower depth and is 25%-30% lower in amplitude. Dose distribution from the gadolinium neutron capture reaction is then simulated by MCNP and compared with measured data. The simulations made by MCNP agree well with experimental results. As long as thermal neutron scattering from chemically bound atoms are not included in GEANT4 it is not suitable for NCT applications.

Investigation of some possible changes in Am-Be neutron source configuration in order to increase the thermal neutron flux using Monte Carlo code

NASA Astrophysics Data System (ADS)

Basiri, H.; Tavakoli-Anbaran, H.

2018-01-01

Am-Be neutrons source is based on (α, n) reaction and generates neutrons in the energy range of 0-11 MeV. Since the thermal neutrons are widely used in different fields, in this work, we investigate how to improve the source configuration in order to increase the thermal flux. These suggested changes include a spherical moderator instead of common cylindrical geometry, a reflector layer and an appropriate materials selection in order to achieve the maximum thermal flux. All calculations were done by using MCNP1 Monte Carlo code. Our final results indicated that a spherical paraffin moderator, a layer of beryllium as a reflector can efficiently increase the thermal neutron flux of Am-Be source.

A Monte Carlo model system for core analysis and epithermal neutron beam design at the Washington State University Radiation Center

DOE Office of Scientific and Technical Information (OSTI.GOV)

Burns, T.D. Jr.

1996-05-01

The Monte Carlo Model System (MCMS) for the Washington State University (WSU) Radiation Center provides a means through which core criticality and power distributions can be calculated, as well as providing a method for neutron and photon transport necessary for BNCT epithermal neutron beam design. The computational code used in this Model System is MCNP4A. The geometric capability of this Monte Carlo code allows the WSU system to be modeled very accurately. A working knowledge of the MCNP4A neutron transport code increases the flexibility of the Model System and is recommended, however, the eigenvalue/power density problems can be run withmore » little direct knowledge of MCNP4A. Neutron and photon particle transport require more experience with the MCNP4A code. The Model System consists of two coupled subsystems; the Core Analysis and Source Plane Generator Model (CASP), and the BeamPort Shell Particle Transport Model (BSPT). The CASP Model incorporates the S({alpha}, {beta}) thermal treatment, and is run as a criticality problem yielding, the system eigenvalue (k{sub eff}), the core power distribution, and an implicit surface source for subsequent particle transport in the BSPT Model. The BSPT Model uses the source plane generated by a CASP run to transport particles through the thermal column beamport. The user can create filter arrangements in the beamport and then calculate characteristics necessary for assessing the BNCT potential of the given filter want. Examples of the characteristics to be calculated are: neutron fluxes, neutron currents, fast neutron KERMAs and gamma KERMAs. The MCMS is a useful tool for the WSU system. Those unfamiliar with the MCNP4A code can use the MCMS transparently for core analysis, while more experienced users will find the particle transport capabilities very powerful for BNCT filter design.« less

Coupled multi-group neutron photon transport for the simulation of high-resolution gamma-ray spectroscopy applications

NASA Astrophysics Data System (ADS)

Burns, Kimberly Ann

The accurate and efficient simulation of coupled neutron-photon problems is necessary for several important radiation detection applications. Examples include the detection of nuclear threats concealed in cargo containers and prompt gamma neutron activation analysis for nondestructive determination of elemental composition of unknown samples. In these applications, high-resolution gamma-ray spectrometers are used to preserve as much information as possible about the emitted photon flux, which consists of both continuum and characteristic gamma rays with discrete energies. Monte Carlo transport is the most commonly used modeling tool for this type of problem, but computational times for many problems can be prohibitive. This work explores the use of coupled Monte Carlo-deterministic methods for the simulation of neutron-induced photons for high-resolution gamma-ray spectroscopy applications. RAdiation Detection Scenario Analysis Toolbox (RADSAT), a code which couples deterministic and Monte Carlo transport to perform radiation detection scenario analysis in three dimensions [1], was used as the building block for the methods derived in this work. RADSAT was capable of performing coupled deterministic-Monte Carlo simulations for gamma-only and neutron-only problems. The purpose of this work was to develop the methodology necessary to perform coupled neutron-photon calculations and add this capability to RADSAT. Performing coupled neutron-photon calculations requires four main steps: the deterministic neutron transport calculation, the neutron-induced photon spectrum calculation, the deterministic photon transport calculation, and the Monte Carlo detector response calculation. The necessary requirements for each of these steps were determined. A major challenge in utilizing multigroup deterministic transport methods for neutron-photon problems was maintaining the discrete neutron-induced photon signatures throughout the simulation. Existing coupled neutron-photon cross-section libraries and the methods used to produce neutron-induced photons were unsuitable for high-resolution gamma-ray spectroscopy applications. Central to this work was the development of a method for generating multigroup neutron-photon cross-sections in a way that separates the discrete and continuum photon emissions so the neutron-induced photon signatures were preserved. The RADSAT-NG cross-section library was developed as a specialized multigroup neutron-photon cross-section set for the simulation of high-resolution gamma-ray spectroscopy applications. The methodology and cross sections were tested using code-to-code comparison with MCNP5 [2] and NJOY [3]. A simple benchmark geometry was used for all cases compared with MCNP. The geometry consists of a cubical sample with a 252Cf neutron source on one side and a HPGe gamma-ray spectrometer on the opposing side. Different materials were examined in the cubical sample: polyethylene (C2H4), P, N, O, and Fe. The cross sections for each of the materials were compared to cross sections collapsed using NJOY. Comparisons of the volume-averaged neutron flux within the sample, volume-averaged photon flux within the detector, and high-purity gamma-ray spectrometer response (only for polyethylene) were completed using RADSAT and MCNP. The code-to-code comparisons show promising results for the coupled Monte Carlo-deterministic method. The RADSAT-NG cross-section production method showed good agreement with NJOY for all materials considered although some additional work is needed in the resonance region and in the first and last energy bin. Some cross section discrepancies existed in the lowest and highest energy bin, but the overall shape and magnitude of the two methods agreed. For the volume-averaged photon flux within the detector, typically the five most intense lines agree to within approximately 5% of the MCNP calculated flux for all of materials considered. The agreement in the code-to-code comparisons cases demonstrates a proof-of-concept of the method for use in RADSAT for coupled neutron-photon problems in high-resolution gamma-ray spectroscopy applications. One of the primary motivators for using the coupled method over pure Monte Carlo method is the potential for significantly lower computational times. For the code-to-code comparison cases, the run times for RADSAT were approximately 25--500 times shorter than for MCNP, as shown in Table 1. This was assuming a 40 mCi 252Cf neutron source and 600 seconds of "real-world" measurement time. The only variance reduction technique implemented in the MCNP calculation was forward biasing of the source toward the sample target. Improved MCNP runtimes could be achieved with the addition of more advanced variance reduction techniques.

Total Kinetic Energy and Fragment Mass Distribution of Neutron-Induced Fission of U-233

DOE Office of Scientific and Technical Information (OSTI.GOV)

Higgins, Daniel James; Schmitt, Kyle Thomas; Mosby, Shea Morgan

Properties of fission in U-233 were studied at the Los Alamos Neutron Science Center (LANSCE) at incident neutron energies from thermal to 40 MeV at both the Lujan Neutron Scattering Center flight path 12 and at WNR flight path 90-Left from Dec 2016 to Jan 2017. Fission fragments are observed in coincidence using a twin ionization chamber with Frisch grids. The average total kinetic energy (TKE) released from fission and fragment mass distributions are calculated from observations of energy deposited in the detector and conservation of mass and momentum. Accurate experimental measurements of these parameters are necessary to better understandmore » the fission process and obtain data necessary for calculating criticality. The average TKE released from fission has been well characterized for several isotopes at thermal neutron energy, however, few measurements have been made at fast neutron energies. This experiment expands on previous successful experiments using an ionization chamber to measure TKE and fragment mass distributions of U-235, U-238, and Pu-239. This experiment requires the full spectrum of neutron energies and can therefore only be performed at a small number of facilities in the world. The required full neutron energy spectrum is obtained by combining measurements from WNR 90L and Lujan FP12 at LANSCE.« less

Corrigendum to “Accelerated materials evaluation for nuclear applications” [J. Nucl. Mater. 488 (2017) 46–62

DOE PAGES

Griffiths, Malcolm; Walters, L.; Greenwood, L. R.; ...

2017-09-21

The original article addresses the opportunities and complexities of using materials test reactors with high neutron fluxes to perform accelerated studies of material aging in power reactors operating at lower neutron fluxes and with different neutron flux spectra. Radiation damage and gas production in different reactors have been compared using the code, SPECTER. This code provides a common standard from which to compare neutron damage data generated by different research groups using a variety of reactors. This Corrigendum identifies a few typographical errors. Tables 2 and 3 are included in revised form.

A new irradiation method with a neutron filter for silicon neutron transmutation doping at the Japan research reactor no. 3 (JRR-3).

PubMed

Komeda, Masao; Kawasaki, Kozo; Obara, Toru

2013-04-01

We studied a new silicon irradiation holder with a neutron filter designed to make the vertical neutron flux profile uniform. Since an irradiation holder has to be made of a low activation material, we applied aluminum blended with B4C as the holder material. Irradiation methods to achieve uniform flux with a filter are discussed using Monte-Carlo calculation code MVP. Validation of the use of the MVP code for the holder's analyses is also discussed via characteristic experiments. Copyright © 2013 Elsevier Ltd. All rights reserved.

CFD Code Development for Combustor Flows

NASA Technical Reports Server (NTRS)

Norris, Andrew

2003-01-01

During the lifetime of this grant, work has been performed in the areas of model development, code development, code validation and code application. For model development, this has included the PDF combustion module, chemical kinetics based on thermodynamics, neural network storage of chemical kinetics, ILDM chemical kinetics and assumed PDF work. Many of these models were then implemented in the code, and in addition many improvements were made to the code, including the addition of new chemistry integrators, property evaluation schemes, new chemistry models and turbulence-chemistry interaction methodology. Validation of all new models and code improvements were also performed, while application of the code to the ZCET program and also the NPSS GEW combustor program were also performed. Several important items remain under development, including the NOx post processing, assumed PDF model development and chemical kinetic development. It is expected that this work will continue under the new grant.

Unfolding the neutron spectrum of a NE213 scintillator using artificial neural networks.

PubMed

Sharghi Ido, A; Bonyadi, M R; Etaati, G R; Shahriari, M

2009-10-01

Artificial neural networks technology has been applied to unfold the neutron spectra from the pulse height distribution measured with NE213 liquid scintillator. Here, both the single and multi-layer perceptron neural network models have been implemented to unfold the neutron spectrum from an Am-Be neutron source. The activation function and the connectivity of the neurons have been investigated and the results have been analyzed in terms of the network's performance. The simulation results show that the neural network that utilizes the Satlins transfer function has the best performance. In addition, omitting the bias connection of the neurons improve the performance of the network. Also, the SCINFUL code is used for generating the response functions in the training phase of the process. Finally, the results of the neural network simulation have been compared with those of the FORIST unfolding code for both (241)Am-Be and (252)Cf neutron sources. The results of neural network are in good agreement with FORIST code.

MEASUREMENTS OF NEUTRON SPECTRA IN 0.8-GEV AND 1.6-GEV PROTON-IRRADIATED<2 OF 2>NA THICK TARGETS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Titarenko, Y. E.; Batyaev, V. F.; Zhivun, V. M.

2001-01-01

Measurements of neutron spectra in W, and Na targets irradiated by 0.8 GeV and 1.6 GeV protons are presented. Measurements were made by the TOF techniques using the proton beam from ITEP U-10 synchrotron. Neutrons were detected with BICRON-511 liquid scintillator-based detectors. The neutron detection efficiency was calculated via the SCINFUL and CECIL codes. The W results are compared with the similar data obtained elsewhere. The measured neutron spectra are compared with the LAHET and CEM2k code simulations results. Attempt is made to explain some observed disagreements between experiments and simulations. The presented results are of interest both in termsmore » of nuclear data buildup and as a benchmark of the up-to-date predictive power of the simulation codes used in designing the hybrid accelerator-driven system (ADS) facilities with sodium-cooled tungsten targets.« less

TRACE/PARCS analysis of the OECD/NEA Oskarshamn-2 BWR stability benchmark

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kozlowski, T.; Downar, T.; Xu, Y.

2012-07-01

On February 25, 1999, the Oskarshamn-2 NPP experienced a stability event which culminated in diverging power oscillations with a decay ratio of about 1.4. The event was successfully modeled by the TRACE/PARCS coupled code system, and further analysis of the event is described in this paper. The results show very good agreement with the plant data, capturing the entire behavior of the transient including the onset of instability, growth of the oscillations (decay ratio) and oscillation frequency. This provides confidence in the prediction of other parameters which are not available from the plant records. The event provides coupled code validationmore » for a challenging BWR stability event, which involves the accurate simulation of neutron kinetics (NK), thermal-hydraulics (TH), and TH/NK. coupling. The success of this work has demonstrated the ability of the 3-D coupled systems code TRACE/PARCS to capture the complex behavior of BWR stability events. The problem was released as an international OECD/NEA benchmark, and it is the first benchmark based on measured plant data for a stability event with a DR greater than one. Interested participants are invited to contact authors for more information. (authors)« less

Disclosure of the oscillations in kinetics of the reactor pressure vessel steel damage at fast neutron intensity decreasing

NASA Astrophysics Data System (ADS)

Krasikov, E.; Nikolaenko, V.

2017-01-01

Fast neutron intensity influence on reactor materials radiation damage is a critically important question in the problem of the correct use of the accelerated irradiation tests data for substantiation of the materials workability in real irradiation conditions that is low neutron intensity. Investigations of the fast neutron intensity (flux) influence on radiation damage and experimental data scattering reveal the existence of non-monotonous sections in kinetics of the reactor pressure vessels (RPV) steel damage. Discovery of the oscillations as indicator of the self-organization processes presence give reasons for new ways searching on reactor pressure vessel (RPV) steel radiation stability increasing and attempt of the self-restoring metal elaboration. Revealing of the wavelike process in the form of non monotonous parts of the kinetics of radiation embrittlement testifies that periodic transformation of the structure take place. This fact actualizes the problem of more precise definition of the RPV materials radiation embrittlement mechanisms and gives reasons for search of the ways to manage the radiation stability (nanostructuring and so on to stimulate the radiation defects annihilation), development of the means for creating of more stableness self recovering smart materials.

Characteristic evaluation of a Lithium-6 loaded neutron coincidence spectrometer.

PubMed

Hayashi, M; Kaku, D; Watanabe, Y; Sagara, K

2007-01-01

Characteristics of a (6)Li-loaded neutron coincidence spectrometer were investigated from both measurements and Monte Carlo simulations. The spectrometer consists of three (6)Li-glass scintillators embedded in a liquid organic scintillator BC-501A, which can detect selectively neutrons that deposit the total energy in the BC-501A using a coincidence signal generated from the capture event of thermalised neutrons in the (6)Li-glass scintillators. The relative efficiency and the energy response were measured using 4.7, 7.2 and 9.0 MeV monoenergetic neutrons. The measured ones were compared with the Monte Carlo calculations performed by combining the neutron transport code PHITS and the scintillator response calculation code SCINFUL. The experimental light output spectra were in good agreement with the calculated ones in shape. The energy dependence of the detection efficiency was reproduced by the calculation. The response matrices for 1-10 MeV neutrons were finally obtained.

Introducing single-crystal scattering and optical potentials into MCNPX: Predicting neutron emission from a convoluted moderator

DOE PAGES

Gallmeier, F. X.; Iverson, E. B.; Lu, W.; ...

2016-01-08

Neutron transport simulation codes are an indispensable tool used for the design and construction of modern neutron scattering facilities and instrumentation. It has become increasingly clear that some neutron instrumentation has started to exploit physics that is not well-modelled by the existing codes. Particularly, the transport of neutrons through single crystals and across interfaces in MCNP(X), Geant4 and other codes ignores scattering from oriented crystals and refractive effects, and yet these are essential ingredients for the performance of monochromators and ultra-cold neutron transport respectively (to mention but two examples). In light of these developments, we have extended the MCNPX codemore » to include a single-crystal neutron scattering model and neutron reflection/refraction physics. Furthermore, we have also generated silicon scattering kernels for single crystals of definable orientation with respect to an incoming neutron beam. As a first test of these new tools, we have chosen to model the recently developed convoluted moderator concept, in which a moderating material is interleaved with layers of perfect crystals to provide an exit path for neutrons moderated to energies below the crystal s Bragg cut off at locations deep within the moderator. Studies of simple cylindrical convoluted moderator systems of 100 mm diameter and composed of polyethylene and single crystal silicon were performed with the upgraded MCNPX code and reproduced the magnitude of effects seen in experiments compared to homogeneous moderator systems. Applying different material properties for refraction and reflection, and by replacing the silicon in the models with voids, we show that the emission enhancements seen in recent experiments are primarily caused by the transparency of the silicon/void layers. Finally the convoluted moderator experiments described by Iverson et al. were simulated and we find satisfactory agreement between the measurement and the results of simulations performed using the tools we have developed.« less

Calculation and benchmarking of an azimuthal pressure vessel neutron fluence distribution using the BOXER code and scraping experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Holzgrewe, F.; Hegedues, F.; Paratte, J.M.

1995-03-01

The light water reactor BOXER code was used to determine the fast azimuthal neutron fluence distribution at the inner surface of the reactor pressure vessel after the tenth cycle of a pressurized water reactor (PWR). Using a cross-section library in 45 groups, fixed-source calculations in transport theory and x-y geometry were carried out to determine the fast azimuthal neutron flux distribution at the inner surface of the pressure vessel for four different cycles. From these results, the fast azimuthal neutron fluence after the tenth cycle was estimated and compared with the results obtained from scraping test experiments. In these experiments,more » small samples of material were taken from the inner surface of the pressure vessel. The fast neutron fluence was then determined form the measured activity of the samples. Comparing the BOXER and scraping test results have maximal differences of 15%, which is very good, considering the factor of 10{sup 3} neutron attenuation between the reactor core and the pressure vessel. To compare the BOXER results with an independent code, the 21st cycle of the PWR was also calculated with the TWODANT two-dimensional transport code, using the same group structure and cross-section library. Deviations in the fast azimuthal flux distribution were found to be <3%, which verifies the accuracy of the BOXER results.« less

ANITA-IEAF activation code package - updating of the decay and cross section data libraries and validation on the experimental data from the Karlsruhe Isochronous Cyclotron

NASA Astrophysics Data System (ADS)

Frisoni, Manuela

2017-09-01

ANITA-IEAF is an activation package (code and libraries) developed in the past in ENEA-Bologna in order to assess the activation of materials exposed to neutrons with energies greater than 20 MeV. An updated version of the ANITA-IEAF activation code package has been developed. It is suitable to be applied to the study of the irradiation effects on materials in facilities like the International Fusion Materials Irradiation Facility (IFMIF) and the DEMO Oriented Neutron Source (DONES), in which a considerable amount of neutrons with energies above 20 MeV is produced. The present paper summarizes the main characteristics of the updated version of ANITA-IEAF, able to use decay and cross section data based on more recent evaluated nuclear data libraries, i.e. the JEFF-3.1.1 Radioactive Decay Data Library and the EAF-2010 neutron activation cross section library. In this paper the validation effort related to the comparison between the code predictions and the activity measurements obtained from the Karlsruhe Isochronous Cyclotron is presented. In this integral experiment samples of two different steels, SS-316 and F82H, pure vanadium and a vanadium alloy, structural materials of interest in fusion technology, were activated in a neutron spectrum similar to the IFMIF neutron field.

In situ calibration of neutron activation system on the large helical device

NASA Astrophysics Data System (ADS)

Pu, N.; Nishitani, T.; Isobe, M.; Ogawa, K.; Kawase, H.; Tanaka, T.; Li, S. Y.; Yoshihashi, S.; Uritani, A.

2017-11-01

In situ calibration of the neutron activation system on the Large Helical Device (LHD) was performed by using an intense 252Cf neutron source. To simulate a ring-shaped neutron source, we installed a railway inside the LHD vacuum vessel and made a train loaded with the 252Cf source run along a typical magnetic axis position. Three activation capsules loaded with thirty pieces of indium foils stacked with total mass of approximately 18 g were prepared. Each capsule was irradiated over 15 h while the train was circulating. The activation response coefficient (9.4 ± 1.2) × 10-8 of 115In(n, n')115mIn reaction obtained from the experiment is in good agreement with results from three-dimensional neutron transport calculations using the Monte Carlo neutron transport simulation code 6. The activation response coefficients of 2.45 MeV birth neutron and secondary 14.1 MeV neutron from deuterium plasma were evaluated from the activation response coefficient obtained in this calibration experiment with results from three-dimensional neutron calculations using the Monte Carlo neutron transport simulation code 6.

KINETICS OF LOW SOURCE REACTOR STARTUPS. PART II

DOE Office of Scientific and Technical Information (OSTI.GOV)

hurwitz, H. Jr.; MacMillan, D.B.; Smith, J.H.

1962-06-01

A computational technique is described for computation of the probability distribution of power level for a low source reactor startup. The technique uses a mathematical model, for the time-dependent probability distribution of neutron and precursor concentration, having finite neutron lifetime, one group of delayed neutron precursors, and no spatial dependence. Results obtained by the technique are given. (auth)

Neutronic safety parameters and transient analyses for Poland's MARIA research reactor.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bretscher, M. M.; Hanan, N. A.; Matos, J. E.

1999-09-27

Reactor kinetic parameters, reactivity feedback coefficients, and control rod reactivity worths have been calculated for the MARIA Research Reactor (Swierk, Poland) for M6-type fuel assemblies with {sup 235}U enrichments of 80% and 19.7%. Kinetic parameters were evaluated for family-dependent effective delayed neutron fractions, decay constants, and prompt neutron lifetimes and neutron generation times. Reactivity feedback coefficients were determined for fuel Doppler coefficients, coolant (H{sub 2}O) void and temperature coefficients, and for in-core and ex-core beryllium temperature coefficients. Total and differential control rod worths and safety rod worths were calculated for each fuel type. These parameters were used to calculate genericmore » transients for fast and slow reactivity insertions with both HEU and LEU fuels. The analyses show that the HEU and LEU cores have very similar responses to these transients.« less

Two-dimensional implosion simulations with a kinetic particle code [2D implosion simulations with a kinetic particle code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sagert, Irina; Even, Wesley Paul; Strother, Terrance Timothy

Here, we perform two-dimensional implosion simulations using a Monte Carlo kinetic particle code. The application of a kinetic transport code is motivated, in part, by the occurrence of nonequilibrium effects in inertial confinement fusion capsule implosions, which cannot be fully captured by hydrodynamic simulations. Kinetic methods, on the other hand, are able to describe both continuum and rarefied flows. We perform simple two-dimensional disk implosion simulations using one-particle species and compare the results to simulations with the hydrodynamics code rage. The impact of the particle mean free path on the implosion is also explored. In a second study, we focusmore » on the formation of fluid instabilities from induced perturbations. We find good agreement with hydrodynamic studies regarding the location of the shock and the implosion dynamics. Differences are found in the evolution of fluid instabilities, originating from the higher resolution of rage and statistical noise in the kinetic studies.« less

Two-dimensional implosion simulations with a kinetic particle code [2D implosion simulations with a kinetic particle code

DOE PAGES

Sagert, Irina; Even, Wesley Paul; Strother, Terrance Timothy

2017-05-17

Here, we perform two-dimensional implosion simulations using a Monte Carlo kinetic particle code. The application of a kinetic transport code is motivated, in part, by the occurrence of nonequilibrium effects in inertial confinement fusion capsule implosions, which cannot be fully captured by hydrodynamic simulations. Kinetic methods, on the other hand, are able to describe both continuum and rarefied flows. We perform simple two-dimensional disk implosion simulations using one-particle species and compare the results to simulations with the hydrodynamics code rage. The impact of the particle mean free path on the implosion is also explored. In a second study, we focusmore » on the formation of fluid instabilities from induced perturbations. We find good agreement with hydrodynamic studies regarding the location of the shock and the implosion dynamics. Differences are found in the evolution of fluid instabilities, originating from the higher resolution of rage and statistical noise in the kinetic studies.« less

Simulated and measured neutron/gamma light output distribution for poly-energetic neutron/gamma sources

NASA Astrophysics Data System (ADS)

Hosseini, S. A.; Zangian, M.; Aghabozorgi, S.

2018-03-01

In the present paper, the light output distribution due to poly-energetic neutron/gamma (neutron or gamma) source was calculated using the developed MCNPX-ESUT-PE (MCNPX-Energy engineering of Sharif University of Technology-Poly Energetic version) computational code. The simulation of light output distribution includes the modeling of the particle transport, the calculation of scintillation photons induced by charged particles, simulation of the scintillation photon transport and considering the light resolution obtained from the experiment. The developed computational code is able to simulate the light output distribution due to any neutron/gamma source. In the experimental step of the present study, the neutron-gamma discrimination based on the light output distribution was performed using the zero crossing method. As a case study, 241Am-9Be source was considered and the simulated and measured neutron/gamma light output distributions were compared. There is an acceptable agreement between the discriminated neutron/gamma light output distributions obtained from the simulation and experiment.

Some Notes on Neutron Up-Scattering and the Doppler-Broadening of High-Z Scattering Resonances

DOE Office of Scientific and Technical Information (OSTI.GOV)

Parsons, Donald Kent

When neutrons are scattered by target nuclei at elevated temperatures, it is entirely possible that the neutron will actually gain energy (i.e., up-scatter) from the interaction. This phenomenon is in addition to the more usual case of the neutron losing energy (i.e., down-scatter). Furthermore, the motion of the target nuclei can also cause extended neutron down-scattering, i.e., the neutrons can and do scatter to energies lower than predicted by the simple asymptotic models. In recent years, more attention has been given to temperature-dependent scattering cross sections for materials in neutron multiplying systems. This has led to the inclusion of neutronmore » up-scatter in deterministic codes like Partisn and to free gas scattering models for material temperature effects in Monte Carlo codes like MCNP and cross section processing codes like NJOY. The free gas scattering models have the effect of Doppler Broadening the scattering cross section output spectra in energy and angle. The current state of Doppler-Broadening numerical techniques used at Los Alamos for scattering resonances will be reviewed, and suggestions will be made for further developments. The focus will be on the free gas scattering models currently in use and the development of new models to include high-Z resonance scattering effects. These models change the neutron up-scattering behavior.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Du, T. F.; Chen, Z. J.; Peng, X. Y.

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 spectrometermore » at EAST are studied for future data interpretation.« less

α-Phase transformation kinetics of U – 8 wt% Mo established by in situ neutron diffraction

DOE PAGES

Garlea, Elena; Steiner, M. A.; Calhoun, C. A.; ...

2016-05-08

The α-phase transformation kinetics of as-cast U - 8 wt% Mo below the eutectoid temperature have been established by in situ neutron diffraction. α-phase weight fraction data acquired through Rietveld refinement at five different isothermal hold temperatures can be modeled accurately utilizing a simple Johnson-Mehl-Avrami-Kolmogorov impingement-based theory, and the results are validated by a corresponding evolution in the γ-phase lattice parameter during transformation that follows Vegard’s law. Neutron diffraction data is used to produce a detailed Time-Temperature-Transformation diagram that improves upon inconsistencies in the current literature, exhibiting a minimum transformation start time of 40 min at temperatures between 500 °Cmore » and 510 °C. Lastly, the transformation kinetics of U – 8 wt% Mo can vary significantly from as-cast conditions after extensive heat treatments, due to homogenization of the typical dendritic microstructure which possesses non-negligible solute segregation.« less

Calculations of the thermal and fast neutron fluxes in the Syrian miniature neutron source reactor using the MCNP-4C code.

PubMed

Khattab, K; Sulieman, I

2009-04-01

The MCNP-4C code, based on the probabilistic approach, was used to model the 3D configuration of the core of the Syrian miniature neutron source reactor (MNSR). The continuous energy neutron cross sections from the ENDF/B-VI library were used to calculate the thermal and fast neutron fluxes in the inner and outer irradiation sites of MNSR. The thermal fluxes in the MNSR inner irradiation sites were also measured experimentally by the multiple foil activation method ((197)Au (n, gamma) (198)Au and (59)Co (n, gamma) (60)Co). The foils were irradiated simultaneously in each of the five MNSR inner irradiation sites to measure the thermal neutron flux and the epithermal index in each site. The calculated and measured results agree well.

Experimental investigation of neutronic characteristics of the IR-8 reactor to confirm the results of calculations by MCU-PTR code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Surkov, A. V., E-mail: surkov.andrew@gmail.com; Kochkin, V. N.; Pesnya, Yu. E.

2015-12-15

A comparison of measured and calculated neutronic characteristics (fast neutron flux and fission rate of {sup 235}U) in the core and reflector of the IR-8 reactor is presented. The irradiation devices equipped with neutron activation detectors were prepared. The determination of fast neutron flux was performed using the {sup 54}Fe (n, p) and {sup 58}Ni (n, p) reactions. The {sup 235}U fission rate was measured using uranium dioxide with 10% enrichment in {sup 235}U. The determination of specific activities of detectors was carried out by measuring the intensity of characteristic gamma peaks using the ORTEC gamma spectrometer. Neutron fields inmore » the core and reflector of the IR-8 reactor were calculated using the MCU-PTR code.« less

Spallation neutron production and the current intra-nuclear cascade and transport codes

NASA Astrophysics Data System (ADS)

Filges, D.; Goldenbaum, F.; Enke, M.; Galin, J.; Herbach, C.-M.; Hilscher, D.; Jahnke, U.; Letourneau, A.; Lott, B.; Neef, R.-D.; Nünighoff, K.; Paul, N.; Péghaire, A.; Pienkowski, L.; Schaal, H.; Schröder, U.; Sterzenbach, G.; Tietze, A.; Tishchenko, V.; Toke, J.; Wohlmuther, M.

A recent renascent interest in energetic proton-induced production of neutrons originates largely from the inception of projects for target stations of intense spallation neutron sources, like the planned European Spallation Source (ESS), accelerator-driven nuclear reactors, nuclear waste transmutation, and also from the application for radioactive beams. In the framework of such a neutron production, of major importance is the search for ways for the most efficient conversion of the primary beam energy into neutron production. Although the issue has been quite successfully addressed experimentally by varying the incident proton energy for various target materials and by covering a huge collection of different target geometries --providing an exhaustive matrix of benchmark data-- the ultimate challenge is to increase the predictive power of transport codes currently on the market. To scrutinize these codes, calculations of reaction cross-sections, hadronic interaction lengths, average neutron multiplicities, neutron multiplicity and energy distributions, and the development of hadronic showers are confronted with recent experimental data of the NESSI collaboration. Program packages like HERMES, LCS or MCNPX master the prevision of reaction cross-sections, hadronic interaction lengths, averaged neutron multiplicities and neutron multiplicity distributions in thick and thin targets for a wide spectrum of incident proton energies, geometrical shapes and materials of the target generally within less than 10% deviation, while production cross-section measurements for light charged particles on thin targets point out that appreciable distinctions exist within these models.

Microdosimetric investigation of the spectra from YAYOI by use of the Monte Carlo code PHITS.

PubMed

Nakao, Minoru; Baba, Hiromi; Oishi, Ayumu; Onizuka, Yoshihiko

2010-07-01

The purpose of this study was to obtain the neutron energy spectrum on the surface of the moderator of the Tokyo University reactor YAYOI and to investigate the origins of peaks observed in the neutron energy spectrum by use of the Monte Carlo Code PHITS for evaluating biological studies. The moderator system was modeled with the use of details from an article that reported a calculation result and a measurement result for a neutron spectrum on the surface of the moderator of the reactor. Our calculation results with PHITS were compared to those obtained with the discrete ordinate code ANISN described in the article. In addition, the changes in the neutron spectrum at the boundaries of materials in the moderator system were examined with PHITS. Also, microdosimetric energy distributions of secondary charged particles from neutron recoil or reaction were calculated by use of PHITS and compared with a microdosimetric experiment. Our calculations of the neutron energy spectrum with PHITS showed good agreement with the results of ANISN in terms of the energy and structure of the peaks. However, the microdosimetric dose distribution spectrum with PHITS showed a remarkable discrepancy with the experimental one. The experimental spectrum could not be explained by PHITS when we used neutron beams of two mono-energies.

Kinetic Simulations of Dense Plasma Focus Breakdown

NASA Astrophysics Data System (ADS)

Schmidt, A.; Higginson, D. P.; Jiang, S.; Link, A.; Povilus, A.; Sears, J.; Bennett, N.; Rose, D. V.; Welch, D. R.

2015-11-01

A dense plasma focus (DPF) device is a type of plasma gun that drives current through a set of coaxial electrodes to assemble gas inside the device and then implode that gas on axis to form a Z-pinch. This implosion drives hydrodynamic and kinetic instabilities that generate strong electric fields, which produces a short intense pulse of x-rays, high-energy (>100 keV) electrons and ions, and (in deuterium gas) neutrons. A strong factor in pinch performance is the initial breakdown and ionization of the gas along the insulator surface separating the two electrodes. The smoothness and isotropy of this ionized sheath are imprinted on the current sheath that travels along the electrodes, thus making it an important portion of the DPF to both understand and optimize. Here we use kinetic simulations in the Particle-in-cell code LSP to model the breakdown. Simulations are initiated with neutral gas and the breakdown modeled self-consistently as driven by a charged capacitor system. We also investigate novel geometries for the insulator and electrodes to attempt to control the electric field profile. The initial ionization fraction of gas is explored computationally to gauge possible advantages of pre-ionization which could be created experimentally via lasers or a glow-discharge. Prepared by LLNL under Contract DE-AC52-07NA27344.

Time and Energy Characterization of a Neutron time of Flight Detector for Re-designing Line of Sight 270 at the Z Pulsed Power Facility.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Styron, Jedediah D.

2016-11-01

This work will focus on the characterization of NTOF detectors fielded on ICF experiments conducted at the Z-experimental facility with emphasis on the MagLif and gas puff campaigns. Three experiments have been proposed. The first experiment will characterize the response of the PMT with respect to the amplitude and width of signals produced by single neutron events. A second experiment will characterize the neutron transit time through the scintillator and the third is to characterize the pulse amplitude for a very specific range of neutron induced charged particle interactions within the scintillator. These experiments will cover incident neutron energies relevantmore » to D-D and D-T fusion reactions. These measurements will be taken as a function of detector bias to cover the entire dynamic range of the detector. Throughout the characterization process, the development of a predictive capability is desired. A new post processing code has been proposed that will calculate a neutron time-of-flight spectrum in units of MeVee. This code will couple the experimentally obtained values and the results obtained with the Monte Carlo code MCNP6. The motivation of this code is to correct for geometry issues when transferring the calibration results from a light lab setting to the Zenvironment. This capability will be used to develop a hypothetical design of LOS270 such that more favorable neutron measurements, requiring less correction, can be made in the future.« less

JOZSO, a computer code for calculating broad neutron resonances in phenomenological nuclear potentials

NASA Astrophysics Data System (ADS)

Baran, Á.; Noszály, Cs.; Vertse, T.

2018-07-01

A renewed version of the computer code GAMOW (Vertse et al., 1982) is given in which the difficulties in calculating broad neutron resonances are amended. New types of phenomenological neutron potentials with strict finite range are built in. Landscape of the S-matrix can be generated on a given domain of the complex wave number plane and S-matrix poles in the domain are localized. Normalized Gamow wave functions and trajectories of given poles can be calculated optionally.

Geometry Survey of the Time-of-Flight Neutron-Elastic Scattering (Antonella) Experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Oshinowo, Babatunde O.; Izraelevitch, Federico

The Antonella experiment is a measurement of the ionization efficiency of nuclear recoils in silicon at low energies [1]. It is a neutron elastic scattering experiment motivated by the search for dark matter particles. In this experiment, a proton beam hits a lithium target and neutrons are produced. The neutron shower passes through a collimator that produces a neutron beam. The beam illuminates a silicon detector. With a certain probability, a neutron interacts with a silicon nucleus of the detector producing elastic scattering. After the interaction, a fraction of the neutron energy is transferred to the silicon nucleus which acquiresmore » kinetic energy and recoils. This kinetic energy is then dissipated in the detector producing ionization and thermal energy. The ionization produced is measured with the silicon detector electronics. On the other hand, the neutron is scattered out of the beam. A neutron-detector array (made of scintillator bars) registers the neutron arrival time and the scattering angle to reconstruct the kinematics of the neutron-nucleus interaction with the time-of-flight technique [2]. In the reconstruction equations, the energy of the nuclear recoil is a function of the scattering angle with respect to the beam direction, the time-of-flight of the neutron and the geometric distances between components of the setup (neutron-production target, silicon detector, scintillator bars). This paper summarizes the survey of the different components of the experiment that made possible the off-line analysis of the collected data. Measurements were made with the API Radian Laser Tracker and I-360 Probe Wireless. The survey was completed at the University of Notre Dame, Indiana, USA in February 2015.« less

Shielding materials for highly penetrating space radiations

NASA Technical Reports Server (NTRS)

Kiefer, Richard L.; Orwoll, Robert A.

1995-01-01

Interplanetary travel involves the transfer from an Earth orbit to a solar orbit. Once outside the Earth's magnetosphere, the major sources of particulate radiation are solar cosmic rays (SCR's) and galactic cosmic rays (GCR's). Intense fluxes of SCR's come from solar flares and consist primarily of protons with energies up to 1 GeV. The GCR consists of a low flux of nuclei with energies up to 10(exp 10) GeV. About 70 percent of the GCR are protons, but a small amount (0.6 percent) are nuclei with atomic numbers greater than 10. High energy charged particles (HZE) interact with matter by transferring energy to atomic electrons in a Coulomb process and by reacting with an atomic nucleus. Energy transferred in the first process increases with the square of the atomic number, so particles with high atomic numbers would be expected to lose large amounts of energy by this process. Nuclear reactions produced by (HZE) particles produce high-energy secondary particles which in turn lose energy to the material. The HZE nuclei are a major concern for radiation protection of humans during interplanetary missions because of the very high specific ionization of both primary and secondary particles. Computer codes have been developed to calculate the deposition of energy by very energetic charged particles in various materials. Calculations show that there is a significant buildup of secondary particles from nuclear fragmentation and Coulomb dissociation processes. A large portion of these particles are neutrons. Since neutrons carry no charge, they only lose energy by collision or reaction with a nucleus. Neutrons with high energies transfer large amounts of energy by inelastic collisions with nuclei. However, as the neutron energy decreases, elastic collisions become much more effective for energy loss. The lighter the nucleus, the greater the fraction of the neutron's kinetic energy that can be lost in an elastic collision. Thus, hydrogen-containing materials such as polymers are most effective in reducing the energy of neutrons. Once neutrons are reduced to very low energies, the probability for undergoing a reaction with a nucleus (the cross section) becomes very high. The product of such a reaction is often radioactive and can involve the release of a significant amount of energy. Thus, it is important to provide protection from low energy neutrons during a long duration space flight. Among the light elements, lithium and boron each have an isotope with a large thermal neutron capture cross section, Li-6 and B-10. However, B-10 is more abundant in the naturally-occurring element than Li-6, has a thermal neutron capture cross section four times that of Li-6, and produces the stable products, He-4 and Li-7 in the interaction while Li-6 produces radioactive tritium (H-3). Thus, boron is the best light-weight material for thermal neutron absorption in spacecraft. The work on this project was focused in two areas: computer design where existing computer codes were used, and in some cases modified, to calculate the propagation and interactions of high energy charged particles through various media, and materials development where boron was incorporated into high performance materials.

VENTURE: a code block for solving multigroup neutronics problems applying the finite-difference diffusion-theory approximation to neutron transport

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.

1975-10-01

The computer code block VENTURE, designed to solve multigroup neutronics problems with application of the finite-difference diffusion-theory approximation to neutron transport (or alternatively simple P$sub 1$) in up to three- dimensional geometry is described. A variety of types of problems may be solved: the usual eigenvalue problem, a direct criticality search on the buckling, on a reciprocal velocity absorber (prompt mode), or on nuclide concentrations, or an indirect criticality search on nuclide concentrations, or on dimensions. First- order perturbation analysis capability is available at the macroscopic cross section level. (auth)

Computer codes for checking, plotting and processing of neutron cross-section covariance data and their application

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sartori, E.; Roussin, R.W.

This paper presents a brief review of computer codes concerned with checking, plotting, processing and using of covariances of neutron cross-section data. It concentrates on those available from the computer code information centers of the United States and the OECD/Nuclear Energy Agency. Emphasis will be placed also on codes using covariances for specific applications such as uncertainty analysis, data adjustment and data consistency analysis. Recent evaluations contain neutron cross section covariance information for all isotopes of major importance for technological applications of nuclear energy. It is therefore important that the available software tools needed for taking advantage of this informationmore » are widely known as hey permit the determination of better safety margins and allow the optimization of more economic, I designs of nuclear energy systems.« less

The use of the SRIM code for calculation of radiation damage induced by neutrons

NASA Astrophysics Data System (ADS)

Mohammadi, A.; Hamidi, S.; Asadabad, Mohsen Asadi

2017-12-01

Materials subjected to neutron irradiation will being evolve to structural changes by the displacement cascades initiated by nuclear reaction. This study discusses a methodology to compute primary knock-on atoms or PKAs information that lead to radiation damage. A program AMTRACK has been developed for assessing of the PKAs information. This software determines the specifications of recoil atoms (using PTRAC card of MCNPX code) and also the kinematics of interactions. The deterministic method was used for verification of the results of (MCNPX+AMTRACK). The SRIM (formely TRIM) code is capable to compute neutron radiation damage. The PKAs information was extracted by AMTRACK program, which can be used as an input of SRIM codes for systematic analysis of primary radiation damage. Then the Bushehr Nuclear Power Plant (BNPP) radiation damage on reactor pressure vessel is calculated.

Simulations of inspiraling and merging double neutron stars using the Spectral Einstein Code

NASA Astrophysics Data System (ADS)

Haas, Roland; Ott, Christian D.; Szilagyi, Bela; Kaplan, Jeffrey D.; Lippuner, Jonas; Scheel, Mark A.; Barkett, Kevin; Muhlberger, Curran D.; Dietrich, Tim; Duez, Matthew D.; Foucart, Francois; Pfeiffer, Harald P.; Kidder, Lawrence E.; Teukolsky, Saul A.

2016-06-01

We present results on the inspiral, merger, and postmerger evolution of a neutron star-neutron star (NSNS) system. Our results are obtained using the hybrid pseudospectral-finite volume Spectral Einstein Code (SpEC). To test our numerical methods, we evolve an equal-mass system for ≈22 orbits before merger. This waveform is the longest waveform obtained from fully general-relativistic simulations for NSNSs to date. Such long (and accurate) numerical waveforms are required to further improve semianalytical models used in gravitational wave data analysis, for example, the effective one body models. We discuss in detail the improvements to SpEC's ability to simulate NSNS mergers, in particular mesh refined grids to better resolve the merger and postmerger phases. We provide a set of consistency checks and compare our results to NSNS merger simulations with the independent bam code. We find agreement between them, which increases confidence in results obtained with either code. This work paves the way for future studies using long waveforms and more complex microphysical descriptions of neutron star matter in SpEC.

Application of a Java-based, univel geometry, neutral particle Monte Carlo code to the searchlight problem

DOE Office of Scientific and Technical Information (OSTI.GOV)

Charles A. Wemple; Joshua J. Cogliati

2005-04-01

A univel geometry, neutral particle Monte Carlo transport code, written entirely in the Java programming language, is under development for medical radiotherapy applications. The code uses ENDF-VI based continuous energy cross section data in a flexible XML format. Full neutron-photon coupling, including detailed photon production and photonuclear reactions, is included. Charged particle equilibrium is assumed within the patient model so that detailed transport of electrons produced by photon interactions may be neglected. External beam and internal distributed source descriptions for mixed neutron-photon sources are allowed. Flux and dose tallies are performed on a univel basis. A four-tap, shift-register-sequence random numbermore » generator is used. Initial verification and validation testing of the basic neutron transport routines is underway. The searchlight problem was chosen as a suitable first application because of the simplicity of the physical model. Results show excellent agreement with analytic solutions. Computation times for similar numbers of histories are comparable to other neutron MC codes written in C and FORTRAN.« less

Mass Yields and Average Total Kinetic Energy Release in Fission for 235U, 238U, and 239Pu

NASA Astrophysics Data System (ADS)

Duke, Dana

2015-10-01

Mass yield distributions and average total kinetic energy (TKE) in neutron induced fission of 235U, 238U, and 239Pu targets were measured with a gridded ionization chamber. Despite decades of fission research, our understanding of how fragment mass yields and TKE depend on incident neutron energy is limited, especially at higher energies (above 5-10 MeV). Improved accuracy in these quantities is important for nuclear technology as it enhances our simulation capabilities and increases the confidence in diagnostic tools. The data can also guide and validate theoretical fission models where the correlation between the fragment mass and TKE is of particular value for constraining models. The Los Alamos Neutron Science Center - Weapons Neutron Research (LANSCE - WNR) provides a neutron beam with energies from thermal to hundreds of MeV, well-suited for filling in the gaps in existing data and exploring fission behavior in the fast neutron region. The results of the studies on target nuclei 235U, 238U, and 239Pu will be presented with a focus on exploring data trends as a function of neutron energy from thermal through 30 MeV. Results indicate clear evidence of structure due to multi-chance fission in the TKE . LA-UR-15-24761.

Thermal neutron filter design for the neutron radiography facility at the LVR-15 reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Soltes, Jaroslav; Faculty of Nuclear Sciences and Physical Engineering, CTU in Prague,; Viererbl, Ladislav

2015-07-01

In 2011 a decision was made to build a neutron radiography facility at one of the unused horizontal channels of the LVR-15 research reactor in Rez, Czech Republic. One of the key conditions for operating an effective radiography facility is the delivery of a high intensity, homogeneous and collimated thermal neutron beam at the sample location. Additionally the intensity of fast neutrons has to be kept as low as possible as the fast neutrons may damage the detectors used for neutron imaging. As the spectrum in the empty horizontal channel roughly copies the spectrum in the reactor core, which hasmore » a high ratio of fast neutrons, neutron filter components have to be installed inside the channel in order to achieve desired beam parameters. As the channel design does not allow the instalment of complex filters and collimators, an optimal solution represent neutron filters made of large single-crystal ingots of proper material composition. Single-crystal silicon was chosen as a favorable filter material for its wide availability in sufficient dimensions. Besides its ability to reasonably lower the ratio of fast neutrons while still keeping high intensities of thermal neutrons, due to its large dimensions, it suits as a shielding against gamma radiation from the reactor core. For designing the necessary filter dimensions the Monte-Carlo MCNP transport code was used. As the code does not provide neutron cross-section libraries for thermal neutron transport through single-crystalline silicon, these had to be created by approximating the theory of thermal neutron scattering and modifying the original cross-section data which are provided with the code. Carrying out a series of calculations the filter thickness of 1 m proved good for gaining a beam with desired parameters and a low gamma background. After mounting the filter inside the channel several measurements of the neutron field were realized at the beam exit. The results have justified the expected calculated values. After the successful filter installing and a series of measurements, first test neutron radiography attempts with test samples could been carried out. (authors)« less

The EUCLID/V1 Integrated Code for Safety Assessment of Liquid Metal Cooled Fast Reactors. Part 1: Basic Models

NASA Astrophysics Data System (ADS)

Mosunova, N. A.

2018-05-01

The article describes the basic models included in the EUCLID/V1 integrated code intended for safety analysis of liquid metal (sodium, lead, and lead-bismuth) cooled fast reactors using fuel rods with a gas gap and pellet dioxide, mixed oxide or nitride uranium-plutonium fuel under normal operation, under anticipated operational occurrences and accident conditions by carrying out interconnected thermal-hydraulic, neutronics, and thermal-mechanical calculations. Information about the Russian and foreign analogs of the EUCLID/V1 integrated code is given. Modeled objects, equation systems in differential form solved in each module of the EUCLID/V1 integrated code (the thermal-hydraulic, neutronics, fuel rod analysis module, and the burnup and decay heat calculation modules), the main calculated quantities, and also the limitations on application of the code are presented. The article also gives data on the scope of functions performed by the integrated code's thermal-hydraulic module, using which it is possible to describe both one- and twophase processes occurring in the coolant. It is shown that, owing to the availability of the fuel rod analysis module in the integrated code, it becomes possible to estimate the performance of fuel rods in different regimes of the reactor operation. It is also shown that the models implemented in the code for calculating neutron-physical processes make it possible to take into account the neutron field distribution over the fuel assembly cross section as well as other features important for the safety assessment of fast reactors.

Modeling and simulation of CANDU reactor and its regulating system

NASA Astrophysics Data System (ADS)

Javidnia, Hooman

Analytical computer codes are indispensable tools in design, optimization, and control of nuclear power plants. Numerous codes have been developed to perform different types of analyses related to the nuclear power plants. A large number of these codes are designed to perform safety analyses. In the context of safety analyses, the control system is often neglected. Although there are good reasons for such a decision, that does not mean that the study of control systems in the nuclear power plants should be neglected altogether. In this thesis, a proof of concept code is developed as a tool that can be used in the design. optimization. and operation stages of the control system. The main objective in the design of this computer code is providing a tool that is easy to use by its target audience and is capable of producing high fidelity results that can be trusted to design the control system and optimize its performance. Since the overall plant control system covers a very wide range of processes, in this thesis the focus has been on one particular module of the the overall plant control system, namely, the reactor regulating system. The center of the reactor regulating system is the CANDU reactor. A nodal model for the reactor is used to represent the spatial neutronic kinetics of the core. The nodal model produces better results compared to the point kinetics model which is often used in the design and analysis of control system for nuclear reactors. The model can capture the spatial effects to some extent. although it is not as detailed as the finite difference methods. The criteria for choosing a nodal model of the core are: (1) the model should provide more detail than point kinetics and capture spatial effects, (2) it should not be too complex or overly detailed to slow down the simulation and provide details that are extraneous or unnecessary for a control engineer. Other than the reactor itself, there are auxiliary models that describe dynamics of different phenomena related to the transfer of the energy from the core. The main function of the reactor regulating system is to control the power of the reactor. This is achieved by using a set of detectors. reactivity devices. and digital control algorithms. Three main reactivity devices that are activated during short-term or intermediate-term transients are modeled in this thesis. The main elements of the digital control system are implemented in accordance to the program specifications for the actual control system in CANDU reactors. The simulation results are validated against requirements of the reactor regulating system. actual plant data. and pre-validated data from other computer codes. The validation process shows that the simulation results can be trusted in making engineering decisions regarding the reactor regulating system and prediction of the system performance in response to upset conditions or disturbances. KEYWORDS: CANDU reactors. reactor regulating system. nodal model. spatial kinetics. reactivity devices. simulation.

Source terms, shielding calculations and soil activation for a medical cyclotron.

PubMed

Konheiser, J; Naumann, B; Ferrari, A; Brachem, C; Müller, S E

2016-12-01

Calculations of the shielding and estimates of soil activation for a medical cyclotron are presented in this work. Based on the neutron source term from the 18 O(p,n) 18 F reaction produced by a 28 MeV proton beam, neutron and gamma dose rates outside the building were estimated with the Monte Carlo code MCNP6 (Goorley et al 2012 Nucl. Technol. 180 298-315). The neutron source term was calculated with the MCNP6 code and FLUKA (Ferrari et al 2005 INFN/TC_05/11, SLAC-R-773) code as well as with supplied data by the manufacturer. MCNP and FLUKA calculations yielded comparable results, while the neutron yield obtained using the manufacturer-supplied information is about a factor of 5 smaller. The difference is attributed to the missing channels in the manufacturer-supplied neutron source terms which considers only the 18 O(p,n) 18 F reaction, whereas the MCNP and FLUKA calculations include additional neutron reaction channels. Soil activation was performed using the FLUKA code. The estimated dose rate based on MCNP6 calculations in the public area is about 0.035 µSv h -1 and thus significantly below the reference value of 0.5 µSv h -1 (2011 Strahlenschutzverordnung, 9 Auflage vom 01.11.2011, Bundesanzeiger Verlag). After 5 years of continuous beam operation and a subsequent decay time of 30 d, the activity concentration of the soil is about 0.34 Bq g -1 .

MCNP capabilities for nuclear well logging calculations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Forster, R.A.; Little, R.C.; Briesmeister, J.F.

The Los Alamos Radiation Transport Code System (LARTCS) consists of state-of-the-art Monte Carlo and discrete ordinates transport codes and data libraries. This paper discusses how the general-purpose continuous-energy Monte Carlo code MCNP ({und M}onte {und C}arlo {und n}eutron {und p}hoton), part of the LARTCS, provides a computational predictive capability for many applications of interest to the nuclear well logging community. The generalized three-dimensional geometry of MCNP is well suited for borehole-tool models. SABRINA, another component of the LARTCS, is a graphics code that can be used to interactively create a complex MCNP geometry. Users can define many source and tallymore » characteristics with standard MCNP features. The time-dependent capability of the code is essential when modeling pulsed sources. Problems with neutrons, photons, and electrons as either single particle or coupled particles can be calculated with MCNP. The physics of neutron and photon transport and interactions is modeled in detail using the latest available cross-section data.« less

Comparison of the LLNL ALE3D and AKTS Thermal Safety Computer Codes for Calculating Times to Explosion in ODTX and STEX Thermal Cookoff Experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wemhoff, A P; Burnham, A K

2006-04-05

Cross-comparison of the results of two computer codes for the same problem provides a mutual validation of their computational methods. This cross-validation exercise was performed for LLNL's ALE3D code and AKTS's Thermal Safety code, using the thermal ignition of HMX in two standard LLNL cookoff experiments: the One-Dimensional Time to Explosion (ODTX) test and the Scaled Thermal Explosion (STEX) test. The chemical kinetics model used in both codes was the extended Prout-Tompkins model, a relatively new addition to ALE3D. This model was applied using ALE3D's new pseudospecies feature. In addition, an advanced isoconversional kinetic approach was used in the AKTSmore » code. The mathematical constants in the Prout-Tompkins code were calibrated using DSC data from hermetically sealed vessels and the LLNL optimization code Kinetics05. The isoconversional kinetic parameters were optimized using the AKTS Thermokinetics code. We found that the Prout-Tompkins model calculations agree fairly well between the two codes, and the isoconversional kinetic model gives very similar results as the Prout-Tompkins model. We also found that an autocatalytic approach in the beta-delta phase transition model does affect the times to explosion for some conditions, especially STEX-like simulations at ramp rates above 100 C/hr, and further exploration of that effect is warranted.« less

Image enhancement using MCNP5 code and MATLAB in neutron radiography.

PubMed

Tharwat, Montaser; Mohamed, Nader; Mongy, T

2014-07-01

This work presents a method that can be used to enhance the neutron radiography (NR) image for objects with high scattering materials like hydrogen, carbon and other light materials. This method used Monte Carlo code, MCNP5, to simulate the NR process and get the flux distribution for each pixel of the image and determines the scattered neutron distribution that caused image blur, and then uses MATLAB to subtract this scattered neutron distribution from the initial image to improve its quality. This work was performed before the commissioning of digital NR system in Jan. 2013. The MATLAB enhancement method is quite a good technique in the case of static based film neutron radiography, while in neutron imaging (NI) technique, image enhancement and quantitative measurement were efficient by using ImageJ software. The enhanced image quality and quantitative measurements were presented in this work. Copyright © 2014 Elsevier Ltd. All rights reserved.

The crystal acceleration effect for cold neutrons

DOE Office of Scientific and Technical Information (OSTI.GOV)

Braginetz, Yu. P., E-mail: aiver@pnpi.spb.ru; Berdnikov, Ya. A.; Fedorov, V. V., E-mail: vfedorov@pnpi.spb.ru

A new mechanism of neutron acceleration is discussed and studied experimentally in detail for cold neutrons passing through the accelerated perfect crystal with the energies close to the Bragg one. The effect arises due to the following reason. The crystal refraction index (neutron-crystal interaction potential) for neutron in the vicinity of the Bragg resonance sharply depends on the parameter of deviation from the exact Bragg condition, i.e. on the crystal-neutron relative velocity. Therefore the neutrons enter into accelerated crystal with one neutron-crystal interaction potential and exit with the other. Neutron kinetic energy cannot vary inside the crystal due to itsmore » homogeneity. So after passage through such a crystal neutrons will be accelerated or decelerated because of the different energy change at the entrance and exit crystal boundaries.« less

Reduced Equations for Calculating the Combustion Rates of Jet-A and Methane Fuel

NASA Technical Reports Server (NTRS)

Molnar, Melissa; Marek, C. John

2003-01-01

Simplified kinetic schemes for Jet-A and methane fuels were developed to be used in numerical combustion codes, such as the National Combustor Code (NCC) that is being developed at Glenn. These kinetic schemes presented here result in a correlation that gives the chemical kinetic time as a function of initial overall cell fuel/air ratio, pressure, and temperature. The correlations would then be used with the turbulent mixing times to determine the limiting properties and progress of the reaction. A similar correlation was also developed using data from NASA's Chemical Equilibrium Applications (CEA) code to determine the equilibrium concentration of carbon monoxide as a function of fuel air ratio, pressure, and temperature. The NASA Glenn GLSENS kinetics code calculates the reaction rates and rate constants for each species in a kinetic scheme for finite kinetic rates. These reaction rates and the values obtained from the equilibrium correlations were then used to calculate the necessary chemical kinetic times. Chemical kinetic time equations for fuel, carbon monoxide, and NOx were obtained for both Jet-A fuel and methane.

Performance and accuracy of criticality calculations performed using WARP – A framework for continuous energy Monte Carlo neutron transport in general 3D geometries on GPUs

DOE PAGES

Bergmann, Ryan M.; Rowland, Kelly L.; Radnović, Nikola; ...

2017-05-01

In this companion paper to "Algorithmic Choices in WARP - A Framework for Continuous Energy Monte Carlo Neutron Transport in General 3D Geometries on GPUs" (doi:10.1016/j.anucene.2014.10.039), the WARP Monte Carlo neutron transport framework for graphics processing units (GPUs) is benchmarked against production-level central processing unit (CPU) Monte Carlo neutron transport codes for both performance and accuracy. We compare neutron flux spectra, multiplication factors, runtimes, speedup factors, and costs of various GPU and CPU platforms running either WARP, Serpent 2.1.24, or MCNP 6.1. WARP compares well with the results of the production-level codes, and it is shown that on the newestmore » hardware considered, GPU platforms running WARP are between 0.8 to 7.6 times as fast as CPU platforms running production codes. Also, the GPU platforms running WARP were between 15% and 50% as expensive to purchase and between 80% to 90% as expensive to operate as equivalent CPU platforms performing at an equal simulation rate.« less

Performance and accuracy of criticality calculations performed using WARP – A framework for continuous energy Monte Carlo neutron transport in general 3D geometries on GPUs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bergmann, Ryan M.; Rowland, Kelly L.; Radnović, Nikola

In this companion paper to "Algorithmic Choices in WARP - A Framework for Continuous Energy Monte Carlo Neutron Transport in General 3D Geometries on GPUs" (doi:10.1016/j.anucene.2014.10.039), the WARP Monte Carlo neutron transport framework for graphics processing units (GPUs) is benchmarked against production-level central processing unit (CPU) Monte Carlo neutron transport codes for both performance and accuracy. We compare neutron flux spectra, multiplication factors, runtimes, speedup factors, and costs of various GPU and CPU platforms running either WARP, Serpent 2.1.24, or MCNP 6.1. WARP compares well with the results of the production-level codes, and it is shown that on the newestmore » hardware considered, GPU platforms running WARP are between 0.8 to 7.6 times as fast as CPU platforms running production codes. Also, the GPU platforms running WARP were between 15% and 50% as expensive to purchase and between 80% to 90% as expensive to operate as equivalent CPU platforms performing at an equal simulation rate.« less

Monte Carol-based validation of neutronic methodology for EBR-II analyses

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liaw, J.R.; Finck, P.J.

1993-01-01

The continuous-energy Monte Carlo code VIM (Ref. 1) has been validated extensively over the years against fast critical experiments and other neutronic analysis codes. A high degree of confidence in VIM for predicting reactor physics parameters has been firmly established. This paper presents a numerical validation of two conventional multigroup neutronic analysis codes, DIF3D (Ref. 4) and VARIANT (Ref. 5), against VIM for two Experimental Breeder Reactor II (EBR-II) core loadings in detailed three-dimensional hexagonal-z geometry. The DIF3D code is based on nodal diffusion theory, and it is used in calculations for day-today reactor operations, whereas the VARIANT code ismore » based on nodal transport theory and is used with increasing frequency for specific applications. Both DIF3D and VARIANT rely on multigroup cross sections generated from ENDF/B-V by the ETOE-2/MC[sup 2]-II/SDX (Ref. 6) code package. Hence, this study also validates the multigroup cross-section processing methodology against the continuous-energy approach used in VIM.« less

Investigation of the heavy nuclei fission with anomalously high values of the fission fragments total kinetic energy

NASA Astrophysics Data System (ADS)

Khryachkov, Vitaly; Goverdovskii, Andrei; Ketlerov, Vladimir; Mitrofanov, Vecheslav; Sergachev, Alexei

2018-03-01

Binary fission of 232Th and 238U induced by fast neutrons were under intent investigation in the IPPE during recent years. These measurements were performed with a twin ionization chamber with Frisch grids. Signals from the detector were digitized for further processing with a specially developed software. It results in information of kinetic energies, masses, directions and Bragg curves of registered fission fragments. Total statistics of a few million fission events were collected during each experiment. It was discovered that for several combinations of fission fragment masses their total kinetic energy was very close to total free energy of the fissioning system. The probability of such fission events for the fast neutron induced fission was found to be much higher than for spontaneous fission of 252Cf and thermal neutron induced fission of 235U. For experiments with 238U target the energy of incident neutrons were 5 MeV and 6.5 MeV. Close analysis of dependence of fission fragment distribution on compound nucleus excitation energy gave us some explanation of the phenomenon. It could be a process in highly excited compound nucleus which leads the fissioning system from the scission point into the fusion valley with high probability.

Neutron dose measurements of Varian and Elekta linacs by TLD600 and TLD700 dosimeters and comparison with MCNP calculations

PubMed Central

Nedaie, Hassan Ali; Darestani, Hoda; Banaee, Nooshin; Shagholi, Negin; Mohammadi, Kheirollah; Shahvar, Arjang; Bayat, Esmaeel

2014-01-01

High-energy linacs produce secondary particles such as neutrons (photoneutron production). The neutrons have the important role during treatment with high energy photons in terms of protection and dose escalation. In this work, neutron dose equivalents of 18 MV Varian and Elekta accelerators are measured by thermoluminescent dosimeter (TLD) 600 and TLD700 detectors and compared with the Monte Carlo calculations. For neutron and photon dose discrimination, first TLDs were calibrated separately by gamma and neutron doses. Gamma calibration was carried out in two procedures; by standard 60Co source and by 18 MV linac photon beam. For neutron calibration by 241Am-Be source, irradiations were performed in several different time intervals. The Varian and Elekta linac heads and the phantom were simulated by the MCNPX code (v. 2.5). Neutron dose equivalent was calculated in the central axis, on the phantom surface and depths of 1, 2, 3.3, 4, 5, and 6 cm. The maximum photoneutron dose equivalents which calculated by the MCNPX code were 7.06 and 2.37 mSv.Gy-1 for Varian and Elekta accelerators, respectively, in comparison with 50 and 44 mSv.Gy-1 achieved by TLDs. All the results showed more photoneutron production in Varian accelerator compared to Elekta. According to the results, it seems that TLD600 and TLD700 pairs are not suitable dosimeters for neutron dosimetry inside the linac field due to high photon flux, while MCNPX code is an appropriate alternative for studying photoneutron production. PMID:24600167

Neutron dose measurements of Varian and Elekta linacs by TLD600 and TLD700 dosimeters and comparison with MCNP calculations.

PubMed

Nedaie, Hassan Ali; Darestani, Hoda; Banaee, Nooshin; Shagholi, Negin; Mohammadi, Kheirollah; Shahvar, Arjang; Bayat, Esmaeel

2014-01-01

High-energy linacs produce secondary particles such as neutrons (photoneutron production). The neutrons have the important role during treatment with high energy photons in terms of protection and dose escalation. In this work, neutron dose equivalents of 18 MV Varian and Elekta accelerators are measured by thermoluminescent dosimeter (TLD) 600 and TLD700 detectors and compared with the Monte Carlo calculations. For neutron and photon dose discrimination, first TLDs were calibrated separately by gamma and neutron doses. Gamma calibration was carried out in two procedures; by standard 60Co source and by 18 MV linac photon beam. For neutron calibration by (241)Am-Be source, irradiations were performed in several different time intervals. The Varian and Elekta linac heads and the phantom were simulated by the MCNPX code (v. 2.5). Neutron dose equivalent was calculated in the central axis, on the phantom surface and depths of 1, 2, 3.3, 4, 5, and 6 cm. The maximum photoneutron dose equivalents which calculated by the MCNPX code were 7.06 and 2.37 mSv.Gy(-1) for Varian and Elekta accelerators, respectively, in comparison with 50 and 44 mSv.Gy(-1) achieved by TLDs. All the results showed more photoneutron production in Varian accelerator compared to Elekta. According to the results, it seems that TLD600 and TLD700 pairs are not suitable dosimeters for neutron dosimetry inside the linac field due to high photon flux, while MCNPX code is an appropriate alternative for studying photoneutron production.

Maximum proton kinetic energy and patient-generated neutron fluence considerations in proton beam arc delivery radiation therapy.

PubMed

Sengbusch, E; Pérez-Andújar, A; DeLuca, P M; Mackie, T R

2009-02-01

Several compact proton accelerator systems for use in proton therapy have recently been proposed. Of paramount importance to the development of such an accelerator system is the maximum kinetic energy of protons, immediately prior to entry into the patient, that must be reached by the treatment system. The commonly used value for the maximum kinetic energy required for a medical proton accelerator is 250 MeV, but it has not been demonstrated that this energy is indeed necessary to treat all or most patients eligible for proton therapy. This article quantifies the maximum kinetic energy of protons, immediately prior to entry into the patient, necessary to treat a given percentage of patients with rotational proton therapy, and examines the impact of this energy threshold on the cost and feasibility of a compact, gantry-mounted proton accelerator treatment system. One hundred randomized treatment plans from patients treated with IMRT were analyzed. The maximum radiological pathlength from the surface of the patient to the distal edge of the treatment volume was obtained for 180 degrees continuous arc proton therapy and for 180 degrees split arc proton therapy (two 90 degrees arcs) using CT# profiles from the Pinnacle (Philips Medical Systems, Madison, WI) treatment planning system. In each case, the maximum kinetic energy of protons, immediately prior to entry into the patient, that would be necessary to treat the patient was calculated using proton range tables for various media. In addition, Monte Carlo simulations were performed to quantify neutron production in a water phantom representing a patient as a function of the maximum proton kinetic energy achievable by a proton treatment system. Protons with a kinetic energy of 240 MeV, immediately prior to entry into the patient, were needed to treat 100% of patients in this study. However, it was shown that 90% of patients could be treated at 198 MeV, and 95% of patients could be treated at 207 MeV. Decreasing the proton kinetic energy from 250 to 200 MeV decreases the total neutron energy fluence produced by stopping a monoenergetic pencil beam in a water phantom by a factor of 2.3. It is possible to significantly lower the requirements on the maximum kinetic energy of a compact proton accelerator if the ability to treat a small percentage of patients with rotational therapy is sacrificed. This decrease in maximum kinetic energy, along with the corresponding decrease in neutron production, could lower the cost and ease the engineering constraints on a compact proton accelerator treatment facility.

Study of an External Neutron Source for an Accelerator-Driven System using the PHITS Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sugawara, Takanori; Iwasaki, Tomohiko; Chiba, Takashi

A code system for the Accelerator Driven System (ADS) has been under development for analyzing dynamic behaviors of a subcritical core coupled with an accelerator. This code system named DSE (Dynamics calculation code system for a Subcritical system with an External neutron source) consists of an accelerator part and a reactor part. The accelerator part employs a database, which is calculated by using PHITS, for investigating the effect related to the accelerator such as the changes of beam energy, beam diameter, void generation, and target level. This analysis method using the database may introduce some errors into dynamics calculations sincemore » the neutron source data derived from the database has some errors in fitting or interpolating procedures. In this study, the effects of various events are investigated to confirm that the method based on the database is appropriate.« less

Status Report on NEAMS PROTEUS/ORIGEN Integration

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wieselquist, William A

2016-02-18

The US Department of Energy’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) Program has contributed significantly to the development of the PROTEUS neutron transport code at Argonne National Laboratory and to the Oak Ridge Isotope Generation and Depletion Code (ORIGEN) depletion/decay code at Oak Ridge National Laboratory. PROTEUS’s key capability is the efficient and scalable (up to hundreds of thousands of cores) neutron transport solver on general, unstructured, three-dimensional finite-element-type meshes. The scalability and mesh generality enable the transfer of neutron and power distributions to other codes in the NEAMS toolkit for advanced multiphysics analysis. Recently, ORIGEN has received considerablemore » modernization to provide the high-performance depletion/decay capability within the NEAMS toolkit. This work presents a description of the initial integration of ORIGEN in PROTEUS, mainly performed during FY 2015, with minor updates in FY 2016.« less

Development of Monte Carlo based real-time treatment planning system with fast calculation algorithm for boron neutron capture therapy.

PubMed

Takada, Kenta; Kumada, Hiroaki; Liem, Peng Hong; Sakurai, Hideyuki; Sakae, Takeji

2016-12-01

We simulated the effect of patient displacement on organ doses in boron neutron capture therapy (BNCT). In addition, we developed a faster calculation algorithm (NCT high-speed) to simulate irradiation more efficiently. We simulated dose evaluation for the standard irradiation position (reference position) using a head phantom. Cases were assumed where the patient body is shifted in lateral directions compared to the reference position, as well as in the direction away from the irradiation aperture. For three groups of neutron (thermal, epithermal, and fast), flux distribution using NCT high-speed with a voxelized homogeneous phantom was calculated. The three groups of neutron fluxes were calculated for the same conditions with Monte Carlo code. These calculated results were compared. In the evaluations of body movements, there were no significant differences even with shifting up to 9mm in the lateral directions. However, the dose decreased by about 10% with shifts of 9mm in a direction away from the irradiation aperture. When comparing both calculations in the phantom surface up to 3cm, the maximum differences between the fluxes calculated by NCT high-speed with those calculated by Monte Carlo code for thermal neutrons and epithermal neutrons were 10% and 18%, respectively. The time required for NCT high-speed code was about 1/10th compared to Monte Carlo calculation. In the evaluation, the longitudinal displacement has a considerable effect on the organ doses. We also achieved faster calculation of depth distribution of thermal neutron flux using NCT high-speed calculation code. Copyright © 2016 Associazione Italiana di Fisica Medica. Published by Elsevier Ltd. All rights reserved.

Extension of applicable neutron energy of DARWIN up to 1 GeV.

PubMed

Satoh, D; Sato, T; Endo, A; Matsufuji, N; Takada, M

2007-01-01

The radiation-dose monitor, DARWIN, needs a set of response functions of the liquid organic scintillator to assess a neutron dose. SCINFUL-QMD is a Monte Carlo based computer code to evaluate the response functions. In order to improve the accuracy of the code, a new light-output function based on the experimental data was developed for the production and transport of protons deuterons, tritons, (3)He nuclei and alpha particles, and incorporated into the code. The applicable energy of DARWIN was extended to 1 GeV using the response functions calculated by the modified SCINFUL-QMD code.

Dual neutral particle induced transmutation in CINDER2008

NASA Astrophysics Data System (ADS)

Martin, W. J.; de Oliveira, C. R. E.; Hecht, A. A.

2014-12-01

Although nuclear transmutation methods for fission have existed for decades, the focus has been on neutron-induced reactions. Recent novel concepts have sought to use both neutrons and photons for purposes such as active interrogation of cargo to detect the smuggling of highly enriched uranium, a concept that would require modeling the transmutation caused by both incident particles. As photonuclear transmutation has yet to be modeled alongside neutron-induced transmutation in a production code, new methods need to be developed. The CINDER2008 nuclear transmutation code from Los Alamos National Laboratory is extended from neutron applications to dual neutral particle applications, allowing both neutron- and photon-induced reactions for this modeling with a focus on fission. Following standard reaction modeling, the induced fission reaction is understood as a two-part reaction, with an entrance channel to the excited compound nucleus, and an exit channel from the excited compound nucleus to the fission fragmentation. Because photofission yield data-the exit channel from the compound nucleus-are sparse, neutron fission yield data are used in this work. With a different compound nucleus and excitation, the translation to the excited compound state is modified, as appropriate. A verification and validation of these methods and data has been performed. This has shown that the translation of neutron-induced fission product yield sets, and their use in photonuclear applications, is appropriate, and that the code has been extended correctly.

Neutron stars at the dark matter direct detection frontier

NASA Astrophysics Data System (ADS)

Raj, Nirmal; Tanedo, Philip; Yu, Hai-Bo

2018-02-01

Neutron stars capture dark matter efficiently. The kinetic energy transferred during capture heats old neutron stars in the local galactic halo to temperatures detectable by upcoming infrared telescopes. We derive the sensitivity of this probe in the framework of effective operators. For dark matter heavier than a GeV, we find that neutron star heating can set limits on the effective operator cutoff that are orders of magnitude stronger than possible from terrestrial direct detection experiments in the case of spin-dependent and velocity-suppressed scattering.

Coupled neutronics and thermal-hydraulics numerical simulations of a Molten Fast Salt Reactor (MFSR)

NASA Astrophysics Data System (ADS)

Laureau, A.; Rubiolo, P. R.; Heuer, D.; Merle-Lucotte, E.; Brovchenko, M.

2014-06-01

Coupled neutronics and thermalhydraulic numerical analyses of a molten salt fast reactor are presented. These preliminary numerical simulations are carried-out using the Monte Carlo code MCNP and the Computation Fluid Dynamic code OpenFOAM. The main objectives of this analysis performed at steady-reactor conditions are to confirm the acceptability of the current neutronic and thermalhydraulic designs of the reactor, to study the effects of the reactor operating conditions on some of the key MSFR design parameters such as the temperature peaking factor. The effects of the precursor's motion on the reactor safety parameters such as the effective fraction of delayed neutrons have been evaluated.

Improved neutron activation prediction code system development

NASA Technical Reports Server (NTRS)

Saqui, R. M.

1971-01-01

Two integrated neutron activation prediction code systems have been developed by modifying and integrating existing computer programs to perform the necessary computations to determine neutron induced activation gamma ray doses and dose rates in complex geometries. Each of the two systems is comprised of three computational modules. The first program module computes the spatial and energy distribution of the neutron flux from an input source and prepares input data for the second program which performs the reaction rate, decay chain and activation gamma source calculations. A third module then accepts input prepared by the second program to compute the cumulative gamma doses and/or dose rates at specified detector locations in complex, three-dimensional geometries.

Neutron-induced reaction cross-sections of 93Nb with fast neutron based on 9Be(p,n) reaction

NASA Astrophysics Data System (ADS)

Naik, H.; Kim, G. N.; Kim, K.; Zaman, M.; Nadeem, M.; Sahid, M.

2018-02-01

The cross-sections of the 93Nb (n , 2 n)92mNb, 93Nb (n , 3 n)91mNb and 93Nb (n , 4 n)90Nb reactions with the average neutron energies of 14.4 to 34.0 MeV have been determined by using an activation and off-line γ-ray spectrometric technique. The fast neutrons were produced using the 9Be (p , n) reaction with the proton energies of 25-, 35- and 45-MeV from the MC-50 Cyclotron at the Korea Institute of Radiological and Medical Sciences (KIRAMS). The neutron flux-weighted average cross-sections of the 93Nb(n , xn ; x = 2- 4) reactions were also obtained from the mono-energetic neutron-induced reaction cross-sections of 93Nb calculated using the TALYS 1.8 code, and the neutron flux spectrum based on the MCNPX 2.6.0 code. The present results for the 93Nb(n , xn ; x = 2- 4) reactions are compared with the calculated neutron flux-weighted average values and found to be in good agreement.

Development of a NRSE Spectrometer with the Help of McStas - Application to the Design of Present and Future Instruments

NASA Astrophysics Data System (ADS)

Kredler, L.; Häußler, W.; Martin, N.; Böni, P.

The flux is still a major limiting factor in neutron research. For instruments being supplied by cold neutrons using neutron guides, both at present steady-state and at new spallation neutron sources, it is therefore important to optimize the instrumental setup and the neutron guidance. Optimization of neutron guide geometry and of the instrument itself can be performed by numerical ray-tracing simulations using existing open-access codes. In this paper, we discuss how such Monte Carlo simulations have been employed in order to plan improvements of the Neutron Resonant Spin Echo spectrometer RESEDA (FRM II, Germany) as well as the neutron guides before and within the instrument. The essential components have been represented with the help of the McStas ray-tracing package. The expected intensity has been tested by means of several virtual detectors, implemented in the simulation code. Comparison between simulations and preliminary measurements results shows good agreement and demonstrates the reliability of the numerical approach. These results will be taken into account in the planning of new components installed in the guide system.

Neutron beam characterization measurements at the Manuel Lujan Jr. neutron scattering center

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mocko, Michal; Muhrer, Guenter; Daemen, Luke L

We have measured the neutron beam characteristics of neutron moderators at the Manuel Lujan Jr. Neutron Scattering Center at LANSCE. The absolute thermal neutron flux, energy spectra and time emission spectra were measured for the high resolution and high intensity decoupled water, partially coupled liquid hydrogen and partially coupled water moderators. The results of our experimental study will provide an insight into aging of different target-moderator-reflector-shield components as well as new experimental data for benchmarking of neutron transport codes.

A scintillator-based approach to monitor secondary neutron production during proton therapy.

PubMed

Clarke, S D; Pryser, E; Wieger, B M; Pozzi, S A; Haelg, R A; Bashkirov, V A; Schulte, R W

2016-11-01

The primary objective of this work is to measure the secondary neutron field produced by an uncollimated proton pencil beam impinging on different tissue-equivalent phantom materials using organic scintillation detectors. Additionally, the Monte Carlo code mcnpx-PoliMi was used to simulate the detector response for comparison to the measured data. Comparison of the measured and simulated data will validate this approach for monitoring secondary neutron dose during proton therapy. Proton beams of 155- and 200-MeV were used to irradiate a variety of phantom materials and secondary particles were detected using organic liquid scintillators. These detectors are sensitive to fast neutrons and gamma rays: pulse shape discrimination was used to classify each detected pulse as either a neutron or a gamma ray. The mcnpx-PoliMi code was used to simulate the secondary neutron field produced during proton irradiation of the same tissue-equivalent phantom materials. An experiment was performed at the Loma Linda University Medical Center proton therapy research beam line and corresponding models were created using the mcnpx-PoliMi code. The authors' analysis showed agreement between the simulations and the measurements. The simulated detector response can be used to validate the simulations of neutron and gamma doses on a particular beam line with or without a phantom. The authors have demonstrated a method of monitoring the neutron component of the secondary radiation field produced by therapeutic protons. The method relies on direct detection of secondary neutrons and gamma rays using organic scintillation detectors. These detectors are sensitive over the full range of biologically relevant neutron energies above 0.5 MeV and allow effective discrimination between neutron and photon dose. Because the detector system is portable, the described system could be used in the future to evaluate secondary neutron and gamma doses on various clinical beam lines for commissioning and prospective data collection in pediatric patients treated with proton therapy.

CFD and Neutron codes coupling on a computational platform

NASA Astrophysics Data System (ADS)

Cerroni, D.; Da Vià, R.; Manservisi, S.; Menghini, F.; Scardovelli, R.

2017-01-01

In this work we investigate the thermal-hydraulics behavior of a PWR nuclear reactor core, evaluating the power generation distribution taking into account the local temperature field. The temperature field, evaluated using a self-developed CFD module, is exchanged with a neutron code, DONJON-DRAGON, which updates the macroscopic cross sections and evaluates the new neutron flux. From the updated neutron flux the new peak factor is evaluated and the new temperature field is computed. The exchange of data between the two codes is obtained thanks to their inclusion into the computational platform SALOME, an open-source tools developed by the collaborative project NURESAFE. The numerical libraries MEDmem, included into the SALOME platform, are used in this work, for the projection of computational fields from one problem to another. The two problems are driven by a common supervisor that can access to the computational fields of both systems, in every time step, the temperature field, is extracted from the CFD problem and set into the neutron problem. After this iteration the new power peak factor is projected back into the CFD problem and the new time step can be computed. Several computational examples, where both neutron and thermal-hydraulics quantities are parametrized, are finally reported in this work.

Studies of fission fragment properties at the Los Alamos Neutron Science Center (LANSCE)

NASA Astrophysics Data System (ADS)

Tovesson, Fredrik; Mayorov, Dmitriy; Duke, Dana; Manning, Brett; Geppert-Kleinrath, Verena

2017-09-01

Nuclear data related to the fission process are needed for a wide variety of research areas, including fundamental science, nuclear energy and non-proliferation. While some of the relevant data have been measured to the required accuracies there are still many aspects of fission that need further investigation. One such aspect is how Total Kinetic Energy (TKE), fragment yields, angular distributions and other fission observables depend on excitation energy of the fissioning system. Another question is the correlation between mass, charge and energy of fission fragments. At the Los Alamos Neutron Science Center (LANSCE) we are studying neutron-induced fission at incident energies from thermal up to hundreds of MeV using the Lujan Center and Weapons Neutron Research (WNR) facilities. Advanced instruments such as SPIDER (time-of-flight and kinetic energy spectrometer), the NIFFTE Time Projection Chamber (TPC), and Frisch grid Ionization Chambers (FGIC) are used to investigate the properties of fission fragments, and some important results for the major actinides have been obtained.

Extending Measurements to En=30 MeV and Beyond

DOE Office of Scientific and Technical Information (OSTI.GOV)

Duke, Dana Lynn

The majority of energy release in the fission process is due to the kinetic energy of the fission fragments. Average Total Kinetic Energy measurements for the major actinides over a wide range of incident neutron energies were performed at LANSCE using a Frisch-gridded ionization chamber. The experiments and results of the 238U(n,f) and 235U(n,f) will be presented, including (En), (A), and mass yield distributions as a function of neutron energy. A preliminary (En) for 239Pu(n,f) will also be shown. The (En) shows a clear structure at multichance fission thresholds for all the reactions that we studied. The fragment masses aremore » determined using the iterative double energy (2E) method, with a resolution of A = 4 - 5 amu. The correction for the prompt fission neutrons is the main source of uncertainty, especially at high incident neutron energies, since the behavior of nubar(A,En) is largely unknown. Different correction methods will be discussed.« less

Studies of fission fragment properties at the Los Alamos Neutron Science Center (LANSCE)

DOE PAGES

Tovesson, Fredrik; Mayorov, Dmitriy; Duke, Dana; ...

2017-09-13

Nuclear data related to the fission process are needed for a wide variety of research areas, including fundamental science, nuclear energy and non-proliferation. While some of the relevant data have been measured to the required accuracies there are still many aspects of fission that need further investigation. One such aspect is how Total Kinetic Energy (TKE), fragment yields, angular distributions and other fission observables depend on excitation energy of the fissioning system. Another question is the correlation between mass, charge and energy of fission fragments. At the Los Alamos Neutron Science Center (LANSCE) we are studying neutron-induced fission at incidentmore » energies from thermal up to hundreds of MeV using the Lujan Center and Weapons Neutron Research (WNR) facilities. Advanced instruments such as SPIDER (time-of-flight and kinetic energy spectrometer), the NIFFTE Time Projection Chamber (TPC), and Frisch grid Ionization Chambers (FGIC) are used to investigate the properties of fission fragments, and some important results for the major actinides have been obtained.« less

Studies of fission fragment properties at the Los Alamos Neutron Science Center (LANSCE)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tovesson, Fredrik; Mayorov, Dmitriy; Duke, Dana

Nuclear data related to the fission process are needed for a wide variety of research areas, including fundamental science, nuclear energy and non-proliferation. While some of the relevant data have been measured to the required accuracies there are still many aspects of fission that need further investigation. One such aspect is how Total Kinetic Energy (TKE), fragment yields, angular distributions and other fission observables depend on excitation energy of the fissioning system. Another question is the correlation between mass, charge and energy of fission fragments. At the Los Alamos Neutron Science Center (LANSCE) we are studying neutron-induced fission at incidentmore » energies from thermal up to hundreds of MeV using the Lujan Center and Weapons Neutron Research (WNR) facilities. Advanced instruments such as SPIDER (time-of-flight and kinetic energy spectrometer), the NIFFTE Time Projection Chamber (TPC), and Frisch grid Ionization Chambers (FGIC) are used to investigate the properties of fission fragments, and some important results for the major actinides have been obtained.« less

Microdosimetric Modeling of Biological Effectiveness for Boron Neutron Capture Therapy Considering Intra- and Intercellular Heterogeneity in 10B Distribution.

PubMed

Sato, Tatsuhiko; Masunaga, Shin-Ichiro; Kumada, Hiroaki; Hamada, Nobuyuki

2018-01-17

We here propose a new model for estimating the biological effectiveness for boron neutron capture therapy (BNCT) considering intra- and intercellular heterogeneity in 10 B distribution. The new model was developed from our previously established stochastic microdosimetric kinetic model that determines the surviving fraction of cells irradiated with any radiations. In the model, the probability density of the absorbed doses in microscopic scales is the fundamental physical index for characterizing the radiation fields. A new computational method was established to determine the probability density for application to BNCT using the Particle and Heavy Ion Transport code System PHITS. The parameters used in the model were determined from the measured surviving fraction of tumor cells administrated with two kinds of 10 B compounds. The model quantitatively highlighted the indispensable need to consider the synergetic effect and the dose dependence of the biological effectiveness in the estimate of the therapeutic effect of BNCT. The model can predict the biological effectiveness of newly developed 10 B compounds based on their intra- and intercellular distributions, and thus, it can play important roles not only in treatment planning but also in drug discovery research for future BNCT.

A time-dependent neutron transport method of characteristics formulation with time derivative propagation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hoffman, Adam J., E-mail: adamhoff@umich.edu; Lee, John C., E-mail: jcl@umich.edu

2016-02-15

A new time-dependent Method of Characteristics (MOC) formulation for nuclear reactor kinetics was developed utilizing angular flux time-derivative propagation. This method avoids the requirement of storing the angular flux at previous points in time to represent a discretized time derivative; instead, an equation for the angular flux time derivative along 1D spatial characteristics is derived and solved concurrently with the 1D transport characteristic equation. This approach allows the angular flux time derivative to be recast principally in terms of the neutron source time derivatives, which are approximated to high-order accuracy using the backward differentiation formula (BDF). This approach, called Sourcemore » Derivative Propagation (SDP), drastically reduces the memory requirements of time-dependent MOC relative to methods that require storing the angular flux. An SDP method was developed for 2D and 3D applications and implemented in the computer code DeCART in 2D. DeCART was used to model two reactor transient benchmarks: a modified TWIGL problem and a C5G7 transient. The SDP method accurately and efficiently replicated the solution of the conventional time-dependent MOC method using two orders of magnitude less memory.« less

Neutron kinetics in moderators and SNM detection through epithermal-neutron-induced fissions

NASA Astrophysics Data System (ADS)

Gozani, Tsahi; King, Michael J.

2016-01-01

Extension of the well-established Differential Die Away Analysis (DDAA) into a faster time domain, where more penetrating epithermal neutrons induce fissions, is proposed and demonstrated via simulations and experiments. In the proposed method the fissions stimulated by thermal, epithermal and even higher-energy neutrons are measured after injection of a narrow pulse of high-energy 14 MeV (d,T) or 2.5 MeV (d,D) source neutrons, appropriately moderated. The ability to measure these fissions stems from the inherent correlation of neutron energy and time ("E-T" correlation) during the process of slowing down of high-energy source neutrons in common moderating materials such as hydrogenous compounds (e.g., polyethylene), heavy water, beryllium and graphite. The kinetic behavior following injection of a delta-function-shaped pulse (in time) of 14 MeV neutrons into such moderators is studied employing MCNPX simulations and, when applicable, some simple "one-group" models. These calculations served as a guide for the design of a source moderator which was used in experiments. Qualitative relationships between slowing-down time after the pulse and the prevailing neutron energy are discussed. A laboratory system consisting of a 14 MeV neutron generator, a polyethylene-reflected Be moderator, a liquid scintillator with pulse-shape discrimination (PSD) and a two-parameter E-T data acquisition system was set up to measure prompt neutron and delayed gamma-ray fission signatures in a 19.5% enriched LEU sample. The measured time behavior of thermal and epithermal neutron fission signals agreed well with the detailed simulations. The laboratory system can readily be redesigned and deployed as a mobile inspection system for SNM in, e.g., cars and vans. A strong pulsed neutron generator with narrow pulse (<75 ns) at a reasonably high pulse frequency could make the high-energy neutron induced fission modality a realizable SNM detection technique.

The effects of nuclear data library processing on Geant4 and MCNP simulations of the thermal neutron scattering law

NASA Astrophysics Data System (ADS)

Hartling, K.; Ciungu, B.; Li, G.; Bentoumi, G.; Sur, B.

2018-05-01

Monte Carlo codes such as MCNP and Geant4 rely on a combination of physics models and evaluated nuclear data files (ENDF) to simulate the transport of neutrons through various materials and geometries. The grid representation used to represent the final-state scattering energies and angles associated with neutron scattering interactions can significantly affect the predictions of these codes. In particular, the default thermal scattering libraries used by MCNP6.1 and Geant4.10.3 do not accurately reproduce the ENDF/B-VII.1 model in simulations of the double-differential cross section for thermal neutrons interacting with hydrogen nuclei in a thin layer of water. However, agreement between model and simulation can be achieved within the statistical error by re-processing ENDF/B-VII.I thermal scattering libraries with the NJOY code. The structure of the thermal scattering libraries and sampling algorithms in MCNP and Geant4 are also reviewed.

SUGGEL: A Program Suggesting the Orbital Angular Momentum of a Neutron Resonance from the Magnitude of its Neutron Width

DOE Office of Scientific and Technical Information (OSTI.GOV)

Oh, S.Y.

2001-02-02

The SUGGEL computer code has been developed to suggest a value for the orbital angular momentum of a neutron resonance that is consistent with the magnitude of its neutron width. The suggestion is based on the probability that a resonance having a certain value of g{Gamma}{sub n} is an l-wave resonance. The probability is calculated by using Bayes' theorem on the conditional probability. The probability density functions (pdf's) of g{Gamma}{sub n} for up to d-wave (l=2) have been derived from the {chi}{sup 2} distribution of Porter and Thomas. The pdf's take two possible channel spins into account. This code ismore » a tool which evaluators will use to construct resonance parameters and help to assign resonance spin. The use of this tool is expected to reduce time and effort in the evaluation procedure, since the number of repeated runs of the fitting code (e.g., SAMMY) may be reduced.« less

Indoor Fast Neutron Generator for Biophysical and Electronic Applications

NASA Astrophysics Data System (ADS)

Cannuli, A.; Caccamo, M. T.; Marchese, N.; Tomarchio, E. A.; Pace, C.; Magazù, S.

2018-05-01

This study focuses the attention on an indoor fast neutron generator for biophysical and electronic applications. More specifically, the findings obtained by several simulations with the MCNP Monte Carlo code, necessary for the realization of a shield for indoor measurements, are presented. Furthermore, an evaluation of the neutron spectrum modification caused by the shielding is reported. Fast neutron generators are a valid and interesting available source of neutrons, increasingly employed in a wide range of research fields, such as science and engineering. The employed portable pulsed neutron source is a MP320 Thermo Scientific neutron generator, able to generate 2.5 MeV neutrons with a neutron yield of 2.0 x 106 n/s, a pulse rate of 250 Hz to 20 KHz and a duty factor varying from 5% to 100%. The neutron generator, based on Deuterium-Deuterium nuclear fusion reactions, is employed in conjunction with a solid-state photon detector, made of n-type high-purity germanium (PINS-GMX by ORTEC) and it is mainly addressed to biophysical and electronic studies. The present study showed a proposal for the realization of a shield necessary for indoor applications for MP320 neutron generator, with a particular analysis of the transport of neutrons simulated with Monte Carlo code and described the two main lines of research in which the source will be used.

Evaluation of a new neutron energy spectrum unfolding code based on an Adaptive Neuro-Fuzzy Inference System (ANFIS).

PubMed

Hosseini, Seyed Abolfazl; Esmaili Paeen Afrakoti, Iman

2018-01-17

The purpose of the present study was to reconstruct the energy spectrum of a poly-energetic neutron source using an algorithm developed based on an Adaptive Neuro-Fuzzy Inference System (ANFIS). ANFIS is a kind of artificial neural network based on the Takagi-Sugeno fuzzy inference system. The ANFIS algorithm uses the advantages of both fuzzy inference systems and artificial neural networks to improve the effectiveness of algorithms in various applications such as modeling, control and classification. The neutron pulse height distributions used as input data in the training procedure for the ANFIS algorithm were obtained from the simulations performed by MCNPX-ESUT computational code (MCNPX-Energy engineering of Sharif University of Technology). Taking into account the normalization condition of each energy spectrum, 4300 neutron energy spectra were generated randomly. (The value in each bin was generated randomly, and finally a normalization of each generated energy spectrum was performed). The randomly generated neutron energy spectra were considered as output data of the developed ANFIS computational code in the training step. To calculate the neutron energy spectrum using conventional methods, an inverse problem with an approximately singular response matrix (with the determinant of the matrix close to zero) should be solved. The solution of the inverse problem using the conventional methods unfold neutron energy spectrum with low accuracy. Application of the iterative algorithms in the solution of such a problem, or utilizing the intelligent algorithms (in which there is no need to solve the problem), is usually preferred for unfolding of the energy spectrum. Therefore, the main reason for development of intelligent algorithms like ANFIS for unfolding of neutron energy spectra is to avoid solving the inverse problem. In the present study, the unfolded neutron energy spectra of 252Cf and 241Am-9Be neutron sources using the developed computational code were found to have excellent agreement with the reference data. Also, the unfolded energy spectra of the neutron sources as obtained using ANFIS were more accurate than the results reported from calculations performed using artificial neural networks in previously published papers. © The Author(s) 2018. Published by Oxford University Press on behalf of The Japan Radiation Research Society and Japanese Society for Radiation Oncology.

A development and integration of database code-system with a compilation of comparator, k0 and absolute methods for INAA using microsoft access

NASA Astrophysics Data System (ADS)

Hoh, Siew Sin; Rapie, Nurul Nadiah; Lim, Edwin Suh Wen; Tan, Chun Yuan; Yavar, Alireza; Sarmani, Sukiman; Majid, Amran Ab.; Khoo, Kok Siong

2013-05-01

Instrumental Neutron Activation Analysis (INAA) is often used to determine and calculate the elemental concentrations of a sample at The National University of Malaysia (UKM) typically in Nuclear Science Programme, Faculty of Science and Technology. The objective of this study was to develop a database code-system based on Microsoft Access 2010 which could help the INAA users to choose either comparator method, k0-method or absolute method for calculating the elemental concentrations of a sample. This study also integrated k0data, Com-INAA, k0Concent, k0-Westcott and Abs-INAA to execute and complete the ECC-UKM database code-system. After the integration, a study was conducted to test the effectiveness of the ECC-UKM database code-system by comparing the concentrations between the experiments and the code-systems. 'Triple Bare Monitor' Zr-Au and Cr-Mo-Au were used in k0Concent, k0-Westcott and Abs-INAA code-systems as monitors to determine the thermal to epithermal neutron flux ratio (f). Calculations involved in determining the concentration were net peak area (Np), measurement time (tm), irradiation time (tirr), k-factor (k), thermal to epithermal neutron flux ratio (f), parameters of the neutron flux distribution epithermal (α) and detection efficiency (ɛp). For Com-INAA code-system, certified reference material IAEA-375 Soil was used to calculate the concentrations of elements in a sample. Other CRM and SRM were also used in this database codesystem. Later, a verification process to examine the effectiveness of the Abs-INAA code-system was carried out by comparing the sample concentrations between the code-system and the experiment. The results of the experimental concentration values of ECC-UKM database code-system were performed with good accuracy.

Energy spectra unfolding of fast neutron sources using the group method of data handling and decision tree algorithms

NASA Astrophysics Data System (ADS)

Hosseini, Seyed Abolfazl; Afrakoti, Iman Esmaili Paeen

2017-04-01

Accurate unfolding of the energy spectrum of a neutron source gives important information about unknown neutron sources. The obtained information is useful in many areas like nuclear safeguards, nuclear nonproliferation, and homeland security. In the present study, the energy spectrum of a poly-energetic fast neutron source is reconstructed using the developed computational codes based on the Group Method of Data Handling (GMDH) and Decision Tree (DT) algorithms. The neutron pulse height distribution (neutron response function) in the considered NE-213 liquid organic scintillator has been simulated using the developed MCNPX-ESUT computational code (MCNPX-Energy engineering of Sharif University of Technology). The developed computational codes based on the GMDH and DT algorithms use some data for training, testing and validation steps. In order to prepare the required data, 4000 randomly generated energy spectra distributed over 52 bins are used. The randomly generated energy spectra and the simulated neutron pulse height distributions by MCNPX-ESUT for each energy spectrum are used as the output and input data. Since there is no need to solve the inverse problem with an ill-conditioned response matrix, the unfolded energy spectrum has the highest accuracy. The 241Am-9Be and 252Cf neutron sources are used in the validation step of the calculation. The unfolded energy spectra for the used fast neutron sources have an excellent agreement with the reference ones. Also, the accuracy of the unfolded energy spectra obtained using the GMDH is slightly better than those obtained from the DT. The results obtained in the present study have good accuracy in comparison with the previously published paper based on the logsig and tansig transfer functions.

Neutron displacement cross-sections for tantalum and tungsten at energies up to 1 GeV

NASA Astrophysics Data System (ADS)

Broeders, C. H. M.; Konobeyev, A. Yu.; Villagrasa, C.

2005-06-01

The neutron displacement cross-section has been evaluated for tantalum and tungsten at energies from 10 -5 eV up to 1 GeV. The nuclear optical model, the intranuclear cascade model combined with the pre-equilibrium and evaporation models were used for the calculations. The number of defects produced by recoil atoms nuclei in materials was calculated by the Norgett, Robinson, Torrens model and by the approach combining calculations using the binary collision approximation model and the results of the molecular dynamics simulation. The numerical calculations were done using the NJOY code, the ECIS96 code, the MCNPX code and the IOTA code.

Nodal Green’s Function Method Singular Source Term and Burnable Poison Treatment in Hexagonal Geometry

DOE Office of Scientific and Technical Information (OSTI.GOV)

A.A. Bingham; R.M. Ferrer; A.M. ougouag

2009-09-01

An accurate and computationally efficient two or three-dimensional neutron diffusion model will be necessary for the development, safety parameters computation, and fuel cycle analysis of a prismatic Very High Temperature Reactor (VHTR) design under Next Generation Nuclear Plant Project (NGNP). For this purpose, an analytical nodal Green’s function solution for the transverse integrated neutron diffusion equation is developed in two and three-dimensional hexagonal geometry. This scheme is incorporated into HEXPEDITE, a code first developed by Fitzpatrick and Ougouag. HEXPEDITE neglects non-physical discontinuity terms that arise in the transverse leakage due to the transverse integration procedure application to hexagonal geometry andmore » cannot account for the effects of burnable poisons across nodal boundaries. The test code being developed for this document accounts for these terms by maintaining an inventory of neutrons by using the nodal balance equation as a constraint of the neutron flux equation. The method developed in this report is intended to restore neutron conservation and increase the accuracy of the code by adding these terms to the transverse integrated flux solution and applying the nodal Green’s function solution to the resulting equation to derive a semi-analytical solution.« less

Stochastic analog neutron transport with TRIPOLI-4 and FREYA: Bayesian uncertainty quantification for neutron multiplicity counting

DOE PAGES

Verbeke, J. M.; Petit, O.

2016-06-01

From nuclear safeguards to homeland security applications, the need for the better modeling of nuclear interactions has grown over the past decades. Current Monte Carlo radiation transport codes compute average quantities with great accuracy and performance; however, performance and averaging come at the price of limited interaction-by-interaction modeling. These codes often lack the capability of modeling interactions exactly: for a given collision, energy is not conserved, energies of emitted particles are uncorrelated, and multiplicities of prompt fission neutrons and photons are uncorrelated. Many modern applications require more exclusive quantities than averages, such as the fluctuations in certain observables (e.g., themore » neutron multiplicity) and correlations between neutrons and photons. In an effort to meet this need, the radiation transport Monte Carlo code TRIPOLI-4® was modified to provide a specific mode that models nuclear interactions in a full analog way, replicating as much as possible the underlying physical process. Furthermore, the computational model FREYA (Fission Reaction Event Yield Algorithm) was coupled with TRIPOLI-4 to model complete fission events. As a result, FREYA automatically includes fluctuations as well as correlations resulting from conservation of energy and momentum.« less

ANITA-2000 activation code package - updating of the decay data libraries and validation on the experimental data of the 14 MeV Frascati Neutron Generator

NASA Astrophysics Data System (ADS)

Frisoni, Manuela

2016-03-01

ANITA-2000 is a code package for the activation characterization of materials exposed to neutron irradiation released by ENEA to OECD-NEADB and ORNL-RSICC. The main component of the package is the activation code ANITA-4M that computes the radioactive inventory of a material exposed to neutron irradiation. The code requires the decay data library (file fl1) containing the quantities describing the decay properties of the unstable nuclides and the library (file fl2) containing the gamma ray spectra emitted by the radioactive nuclei. The fl1 and fl2 files of the ANITA-2000 code package, originally based on the evaluated nuclear data library FENDL/D-2.0, were recently updated on the basis of the JEFF-3.1.1 Radioactive Decay Data Library. This paper presents the results of the validation of the new fl1 decay data library through the comparison of the ANITA-4M calculated values with the measured electron and photon decay heats and activities of fusion material samples irradiated at the 14 MeV Frascati Neutron Generator (FNG) of the NEA-Frascati Research Centre. Twelve material samples were considered, namely: Mo, Cu, Hf, Mg, Ni, Cd, Sn, Re, Ti, W, Ag and Al. The ratios between calculated and experimental values (C/E) are shown and discussed in this paper.

Measurement of pion induced neutron-production double-differential cross sections on Fe and Pb at 870 MeV and 2.1 GeV

NASA Astrophysics Data System (ADS)

Iwamoto, Y.; Shigyo, N.; Satoh, D.; Kunieda, S.; Watanabe, T.; Ishimoto, S.; Tenzou, H.; Maehata, K.; Ishibashi, K.; Nakamoto, T.; Numajiri, M.; Meigo, S.; Takada, H.

2004-08-01

Neutron-production double-differential cross sections for 870 MeV π+ and π- and 2.1 GeV π+ mesons incident on iron and lead targets were measured with NE213 liquid scintillators by time-of-flight technique. NE213 liquid scintillators 12.7 cm in diameter and 12.7 cm thick were placed in directions of 15, 30, 60, 90, 120, and 150° . The typical flight path length was 1.5 m . Neutron detection efficiencies were evaluated by calculation results of SCINFUL and CECIL codes. The experimental results were compared with JAERI quantum molecular dynamics code. For the meson incident reactions, adoption of NN in-medium effects was slightly useful for reproducing 870 MeV π+ -incident neutron yields at neutron energies of 10 30 MeV , as was the case for proton incident reactions. The π- incident reaction generates more neutrons than π+ incidence as the number of nucleons in targets decrease.

Measurement of in-phantom neutron flux and gamma dose in Tehran research reactor boron neutron capture therapy beam line.

PubMed

Bavarnegin, Elham; Sadremomtaz, Alireza; Khalafi, Hossein; Kasesaz, Yaser

2016-01-01

Determination of in-phantom quality factors of Tehran research reactor (TRR) boron neutron capture therapy (BNCT) beam. The doses from thermal neutron reactions with 14N and 10B are calculated by kinetic energy released per unit mass approach, after measuring thermal neutron flux using neutron activation technique. Gamma dose is measured using TLD-700 dosimeter. Different dose components have been measured in a head phantom which has been designed and constructed for BNCT purpose in TRR. Different in-phantom beam quality factors have also been determined. This study demonstrates that the TRR BNCT beam line has potential for treatment of superficial tumors.

Comparisons of dense-plasma-focus kinetic simulations with experimental measurements.

PubMed

Schmidt, A; Link, A; Welch, D; Ellsworth, J; Falabella, S; Tang, V

2014-06-01

Dense-plasma-focus (DPF) Z-pinch devices are sources of copious high-energy electrons and ions, x rays, and neutrons. The mechanisms through which these physically simple devices generate such high-energy beams in a relatively short distance are not fully understood and past optimization efforts of these devices have been largely empirical. Previously we reported on fully kinetic simulations of a DPF and compared them with hybrid and fluid simulations of the same device. Here we present detailed comparisons between fully kinetic simulations and experimental data on a 1.2 kJ DPF with two electrode geometries, including neutron yield and ion beam energy distributions. A more intensive third calculation is presented which examines the effects of a fully detailed pulsed power driver model. We also compare simulated electromagnetic fluctuations with direct measurement of radiofrequency electromagnetic fluctuations in a DPF plasma. These comparisons indicate that the fully kinetic model captures the essential physics of these plasmas with high fidelity, and provide further evidence that anomalous resistivity in the plasma arises due to a kinetic instability near the lower hybrid frequency.

A new three-tier architecture design for multi-sphere neutron spectrometer with the FLUKA code

NASA Astrophysics Data System (ADS)

Huang, Hong; Yang, Jian-Bo; Tuo, Xian-Guo; Liu, Zhi; Wang, Qi-Biao; Wang, Xu

2016-07-01

The current commercially, available Bonner sphere neutron spectrometer (BSS) has high sensitivity to neutrons below 20 MeV, which causes it to be poorly placed to measure neutrons ranging from a few MeV to 100 MeV. The paper added moderator layers and the auxiliary material layer upon 3He proportional counters with FLUKA code, with a view to improve. The results showed that the responsive peaks to neutrons below 20 MeV gradually shift to higher energy region and decrease slightly with the increasing moderator thickness. On the contrary, the response for neutrons above 20 MeV was always very low until we embed auxiliary materials such as copper (Cu), lead (Pb), tungsten (W) into moderator layers. This paper chose the most suitable auxiliary material Pb to design a three-tier architecture multi-sphere neutron spectrometer (NBSS). Through calculating and comparing, the NBSS was advantageous in terms of response for 5-100 MeV and the highest response was 35.2 times the response of polyethylene (PE) ball with the same PE thickness.

Study of neutron spectra in a water bath from a Pb target irradiated by 250 MeV protons

NASA Astrophysics Data System (ADS)

Li, Yan-Yan; Zhang, Xue-Ying; Ju, Yong-Qin; Ma, Fei; Zhang, Hong-Bin; Chen, Liang; Ge, Hong-Lin; Wan, Bo; Luo, Peng; Zhou, Bin; Zhang, Yan-Bin; Li, Jian-Yang; Xu, Jun-Kui; Wang, Song-Lin; Yang, Yong-Wei; Yang, Lei

2015-04-01

Spallation neutrons were produced by the irradiation of Pb with 250 MeV protons. The Pb target was surrounded by water which was used to slow down the emitted neutrons. The moderated neutrons in the water bath were measured by using the resonance detectors of Au, Mn and In with a cadmium (Cd) cover. According to the measured activities of the foils, the neutron flux at different resonance energies were deduced and the epithermal neutron spectra were proposed. Corresponding results calculated with the Monte Carlo code MCNPX were compared with the experimental data to check the validity of the code. The comparison showed that the simulation could give a good prediction for the neutron spectra above 50 eV, while the finite thickness of the foils greatly effected the experimental data in low energy. It was also found that the resonance detectors themselves had great impact on the simulated energy spectra. Supported by National Natural Science Foundation and Strategic Priority Research Program of the Chinese Academy of Sciences (11305229, 11105186, 91226107, 91026009, XDA03030300)

Neutron-Encoded Protein Quantification by Peptide Carbamylation

NASA Astrophysics Data System (ADS)

Ulbrich, Arne; Merrill, Anna E.; Hebert, Alexander S.; Westphall, Michael S.; Keller, Mark P.; Attie, Alan D.; Coon, Joshua J.

2014-01-01

We describe a chemical tag for duplex proteome quantification using neutron encoding (NeuCode). The method utilizes the straightforward, efficient, and inexpensive carbamylation reaction. We demonstrate the utility of NeuCode carbamylation by accurately measuring quantitative ratios from tagged yeast lysates mixed in known ratios and by applying this method to quantify differential protein expression in mice fed a either control or high-fat diet.

Radiation transport codes for potential applications related to radiobiology and radiotherapy using protons, neutrons, and negatively charged pions

NASA Technical Reports Server (NTRS)

Armstrong, T. W.

1972-01-01

Several Monte Carlo radiation transport computer codes are used to predict quantities of interest in the fields of radiotherapy and radiobiology. The calculational methods are described and comparisions of calculated and experimental results are presented for dose distributions produced by protons, neutrons, and negatively charged pions. Comparisons of calculated and experimental cell survival probabilities are also presented.

Radioactive ion beams produced by neutron-induced fission at ISOLDE

NASA Astrophysics Data System (ADS)

Catherall, R.; Lettry, J.; Gilardoni, S.; Köster, U.; Isolde Collaboration

2003-05-01

The production rates of neutron-rich fission products for the next-generation radioactive beam facility EURISOL [EU-RTD Project EURISOL (HPRI-CT-1999-50001)] are mainly limited by the maximum amount of power deposited by protons in the target. An alternative approach is to use neutron beams to induce fission in actinide targets. This has the advantage of reducing: the energy deposited by the proton beam in the target; contamination from neutron-deficient isobars that would be produced by spallation; and mechanical stress on the target. At ISOLDE CERN [E. Kugler, Hyperfine Interact. 129 (2000) 23], tests have been made on standard ISOLDE actinide targets using fast-neutron bunches produced by bombarding thick, high- Z metal converters with 1 and 1.4 GeV proton pulses. This paper reviews the first applications of converters used at ISOLDE. It highlights the different geometries and the techniques used to compare fission yields produced by the proton beam directly on the target with neutron-induced fission. Results from the six targets already tested, namely UC 2/graphite and ThO 2 targets with tungsten and tantalum converters, are presented. To gain further knowledge for the design of a dedicated target as required by the TARGISOL project [EU-RTD Project TARGISOL (HPRI-CT-2001-50033)], the results are compared to simulations, using the MARS [N.V. Mokhov, S.I. Striganov, A. Van Ginneken, S.G. Mashnik, A.J. Sierk, J. Ranft, MARS code developments, in: 4th Workshop on Simulating Accelerator Radiation Environments, SARE-4, Knoxville, USA, 14-15.9.1998, FERMILAB-PUB-98-379, nucl-th/9812038; N.V. Mokhov, The Mars Code System User's Guide, Fermilab-FN-628, 1995; N.V. Mokhov, MARS Code Developments, Benchmarking and Applications, Fermilab-Conf-00-066, 2000; O.E. Krivosheev, N.V. Mokhov, A New MARS and its Applications, Fermilab-Conf-98/43, 1998] code interfaced with MCNP [J.S. Hendrics, MCNP4C LANL Memo X-5; JSH-2000-3; J.F. Briemesteir (Ed.), MCNP - A General Montecarlo N-Particle Transport Code, Version 4C, LA-13709-M] libraries, of the neutron flux from the converters interacting with the actinide targets.

Radioactive ion beams produced by neutron-induced fission at ISOLDE

NASA Astrophysics Data System (ADS)

Isolde Collaboration; Catherall, R.; Lettry, J.; Gilardoni, S.; Köster, U.

2003-05-01

The production rates of neutron-rich fission products for the next-generation radioactive beam facility EURISOL [EU-RTD Project EURISOL (HPRI-CT-1999-50001)] are mainly limited by the maximum amount of power deposited by protons in the target. An alternative approach is to use neutron beams to induce fission in actinide targets. This has the advantage of reducing: the energy deposited by the proton beam in the target; contamination from neutron-deficient isobars that would be produced by spallation; and mechanical stress on the target. At ISOLDE CERN [E. Kugler, Hyperfine Interact. 129 (2000) 23], tests have been made on standard ISOLDE actinide targets using fast-neutron bunches produced by bombarding thick, high-/Z metal converters with 1 and 1.4 GeV proton pulses. This paper reviews the first applications of converters used at ISOLDE. It highlights the different geometries and the techniques used to compare fission yields produced by the proton beam directly on the target with neutron-induced fission. Results from the six targets already tested, namely UC2/graphite and ThO2 targets with tungsten and tantalum converters, are presented. To gain further knowledge for the design of a dedicated target as required by the TARGISOL project [EU-RTD Project TARGISOL (HPRI-CT-2001-50033)], the results are compared to simulations, using the MARS [N.V. Mokhov, S.I. Striganov, A. Van Ginneken, S.G. Mashnik, A.J. Sierk, J. Ranft, MARS code developments, in: 4th Workshop on Simulating Accelerator Radiation Environments, SARE-4, Knoxville, USA, 14-15.9.1998, FERMILAB-PUB-98-379, nucl-th/9812038; N.V. Mokhov, The Mars Code System User's Guide, Fermilab-FN-628, 1995; N.V. Mokhov, MARS Code Developments, Benchmarking and Applications, Fermilab-Conf-00-066, 2000; O.E. Krivosheev, N.V. Mokhov, A New MARS and its Applications, Fermilab-Conf-98/43, 1998] code interfaced with MCNP [J.S. Hendrics, MCNP4C LANL Memo X-5; JSH-2000-3; J.F. Briemesteir (Ed.), MCNP - A General Montecarlo N-Particle Transport Code, Version 4C, LA-13709-M] libraries, of the neutron flux from the converters interacting with the actinide targets.

Erratum: Binary neutron stars with arbitrary spins in numerical relativity [Phys. Rev. D 92, 124012 (2015)

NASA Astrophysics Data System (ADS)

Tacik, Nick; Foucart, Francois; Pfeiffer, Harald P.; Haas, Roland; Ossokine, Serguei; Kaplan, Jeff; Muhlberger, Curran; Duez, Matt D.; Kidder, Lawrence E.; Scheel, Mark A.; Szilágyi, Béla

2016-08-01

The code used in [Phys. Rev. D 92, 124012 (2015)] erroneously computed the enthalpy at the center of the neutron stars. Upon correcting this error, density oscillations in evolutions of rotating neutron stars are significantly reduced (from ˜20 % to ˜0.5 % ). Furthermore, it is possible to construct neutron stars with faster rotation rates.

Physical and numerical sources of computational inefficiency in integration of chemical kinetic rate equations: Etiology, treatment and prognosis

NASA Technical Reports Server (NTRS)

Pratt, D. T.; Radhakrishnan, K.

1986-01-01

The design of a very fast, automatic black-box code for homogeneous, gas-phase chemical kinetics problems requires an understanding of the physical and numerical sources of computational inefficiency. Some major sources reviewed in this report are stiffness of the governing ordinary differential equations (ODE's) and its detection, choice of appropriate method (i.e., integration algorithm plus step-size control strategy), nonphysical initial conditions, and too frequent evaluation of thermochemical and kinetic properties. Specific techniques are recommended (and some advised against) for improving or overcoming the identified problem areas. It is argued that, because reactive species increase exponentially with time during induction, and all species exhibit asymptotic, exponential decay with time during equilibration, exponential-fitted integration algorithms are inherently more accurate for kinetics modeling than classical, polynomial-interpolant methods for the same computational work. But current codes using the exponential-fitted method lack the sophisticated stepsize-control logic of existing black-box ODE solver codes, such as EPISODE and LSODE. The ultimate chemical kinetics code does not exist yet, but the general characteristics of such a code are becoming apparent.

Developing Discontinuous Galerkin Methods for Solving Multiphysics Problems in General Relativity

NASA Astrophysics Data System (ADS)

Kidder, Lawrence; Field, Scott; Teukolsky, Saul; Foucart, Francois; SXS Collaboration

2016-03-01

Multi-messenger observations of the merger of black hole-neutron star and neutron star-neutron star binaries, and of supernova explosions will probe fundamental physics inaccessible to terrestrial experiments. Modeling these systems requires a relativistic treatment of hydrodynamics, including magnetic fields, as well as neutrino transport and nuclear reactions. The accuracy, efficiency, and robustness of current codes that treat all of these problems is not sufficient to keep up with the observational needs. We are building a new numerical code that uses the Discontinuous Galerkin method with a task-based parallelization strategy, a promising combination that will allow multiphysics applications to be treated both accurately and efficiently on petascale and exascale machines. The code will scale to more than 100,000 cores for efficient exploration of the parameter space of potential sources and allowed physics, and the high-fidelity predictions needed to realize the promise of multi-messenger astronomy. I will discuss the current status of the development of this new code.

Kinetics of Slow Neutrons in a Time-of-flight Spectrometer. II. Probability of Transmission Across a Rotating Slit and Distribution after the Flight of Neutrons with Velocity Spectrum F (v); CINETICA DEI NEUTRONI LENTI IN UNO SPETTROMETRO A TEMPO DI VOLO. II. PROBABILITA DI TRANSMISSIONE ATTRAVERSO UNA FENDITURA RUOTANTE E DISTRIBUZIONE DOPO IL VOLO DI NEUTRONI CON SPETTRO DI VELOCITA F (V)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marsequerra, M.; Pauli, G.

1958-12-01

On the basis of the results obtained in Part I (CNC-1), expressions are derived for the transmission probability through a revolving curved slit for neutrons having a velocity distribution f(v), the distribution shown by the neutrons after the flight, and the uncertainty in the energy of neutrons detected in an infinitesimal time interval. (auth)

WARP

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bergmann, Ryan M.; Rowland, Kelly L.

2017-04-12

WARP, which can stand for ``Weaving All the Random Particles,'' is a three-dimensional (3D) continuous energy Monte Carlo neutron transport code developed at UC Berkeley to efficiently execute on NVIDIA graphics processing unit (GPU) platforms. WARP accelerates Monte Carlo simulations while preserving the benefits of using the Monte Carlo method, namely, that very few physical and geometrical simplifications are applied. WARP is able to calculate multiplication factors, neutron flux distributions (in both space and energy), and fission source distributions for time-independent neutron transport problems. It can run in both criticality or fixed source modes, but fixed source mode is currentlymore » not robust, optimized, or maintained in the newest version. WARP can transport neutrons in unrestricted arrangements of parallelepipeds, hexagonal prisms, cylinders, and spheres. The goal of developing WARP is to investigate algorithms that can grow into a full-featured, continuous energy, Monte Carlo neutron transport code that is accelerated by running on GPUs. The crux of the effort is to make Monte Carlo calculations faster while producing accurate results. Modern supercomputers are commonly being built with GPU coprocessor cards in their nodes to increase their computational efficiency and performance. GPUs execute efficiently on data-parallel problems, but most CPU codes, including those for Monte Carlo neutral particle transport, are predominantly task-parallel. WARP uses a data-parallel neutron transport algorithm to take advantage of the computing power GPUs offer.« less

The r-Java 2.0 code: nuclear physics

NASA Astrophysics Data System (ADS)

Kostka, M.; Koning, N.; Shand, Z.; Ouyed, R.; Jaikumar, P.

2014-08-01

Aims: We present r-Java 2.0, a nucleosynthesis code for open use that performs r-process calculations, along with a suite of other analysis tools. Methods: Equipped with a straightforward graphical user interface, r-Java 2.0 is capable of simulating nuclear statistical equilibrium (NSE), calculating r-process abundances for a wide range of input parameters and astrophysical environments, computing the mass fragmentation from neutron-induced fission and studying individual nucleosynthesis processes. Results: In this paper we discuss enhancements to this version of r-Java, especially the ability to solve the full reaction network. The sophisticated fission methodology incorporated in r-Java 2.0 that includes three fission channels (beta-delayed, neutron-induced, and spontaneous fission), along with computation of the mass fragmentation, is compared to the upper limit on mass fission approximation. The effects of including beta-delayed neutron emission on r-process yield is studied. The role of Coulomb interactions in NSE abundances is shown to be significant, supporting previous findings. A comparative analysis was undertaken during the development of r-Java 2.0 whereby we reproduced the results found in the literature from three other r-process codes. This code is capable of simulating the physical environment of the high-entropy wind around a proto-neutron star, the ejecta from a neutron star merger, or the relativistic ejecta from a quark nova. Likewise the users of r-Java 2.0 are given the freedom to define a custom environment. This software provides a platform for comparing proposed r-process sites.

Development of a reliable estimation procedure of radioactivity inventory in a BWR plant due to neutron irradiation for decommissioning

NASA Astrophysics Data System (ADS)

Tanaka, Ken-ichi; Ueno, Jun

2017-09-01

Reliable information of radioactivity inventory resulted from the radiological characterization is important in order to plan decommissioning planning and is also crucial in order to promote decommissioning in effectiveness and in safe. The information is referred to by planning of decommissioning strategy and by an application to regulator. Reliable information of radioactivity inventory can be used to optimize the decommissioning processes. In order to perform the radiological characterization reliably, we improved a procedure of an evaluation of neutron-activated materials for a Boiling Water Reactor (BWR). Neutron-activated materials are calculated with calculation codes and their validity should be verified with measurements. The evaluation of neutron-activated materials can be divided into two processes. One is a distribution calculation of neutron-flux. Another is an activation calculation of materials. The distribution calculation of neutron-flux is performed with neutron transport calculation codes with appropriate cross section library to simulate neutron transport phenomena well. Using the distribution of neutron-flux, we perform distribution calculations of radioactivity concentration. We also estimate a time dependent distribution of radioactivity classification and a radioactive-waste classification. The information obtained from the evaluation is utilized by other tasks in the preparatory tasks to make the decommissioning plan and the activity safe and rational.

ALICE: A non-LTE plasma atomic physics, kinetics and lineshape package

NASA Astrophysics Data System (ADS)

Hill, E. G.; Pérez-Callejo, G.; Rose, S. J.

2018-03-01

All three parts of an atomic physics, atomic kinetics and lineshape code, ALICE, are described. Examples of the code being used to model the emissivity and opacity of plasmas are discussed and interesting features of the code which build on the existing corpus of models are shown throughout.

Analytical three-dimensional neutron transport benchmarks for verification of nuclear engineering codes. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ganapol, B.D.; Kornreich, D.E.

Because of the requirement of accountability and quality control in the scientific world, a demand for high-quality analytical benchmark calculations has arisen in the neutron transport community. The intent of these benchmarks is to provide a numerical standard to which production neutron transport codes may be compared in order to verify proper operation. The overall investigation as modified in the second year renewal application includes the following three primary tasks. Task 1 on two dimensional neutron transport is divided into (a) single medium searchlight problem (SLP) and (b) two-adjacent half-space SLP. Task 2 on three-dimensional neutron transport covers (a) pointmore » source in arbitrary geometry, (b) single medium SLP, and (c) two-adjacent half-space SLP. Task 3 on code verification, includes deterministic and probabilistic codes. The primary aim of the proposed investigation was to provide a suite of comprehensive two- and three-dimensional analytical benchmarks for neutron transport theory applications. This objective has been achieved. The suite of benchmarks in infinite media and the three-dimensional SLP are a relatively comprehensive set of one-group benchmarks for isotropically scattering media. Because of time and resource limitations, the extensions of the benchmarks to include multi-group and anisotropic scattering are not included here. Presently, however, enormous advances in the solution for the planar Green`s function in an anisotropically scattering medium have been made and will eventually be implemented in the two- and three-dimensional solutions considered under this grant. Of particular note in this work are the numerical results for the three-dimensional SLP, which have never before been presented. The results presented were made possible only because of the tremendous advances in computing power that have occurred during the past decade.« less

A 3DHZETRN Code in a Spherical Uniform Sphere with Monte Carlo Verification

NASA Technical Reports Server (NTRS)

Wilson, John W.; Slaba, Tony C.; Badavi, Francis F.; Reddell, Brandon D.; Bahadori, Amir A.

2014-01-01

The computationally efficient HZETRN code has been used in recent trade studies for lunar and Martian exploration and is currently being used in the engineering development of the next generation of space vehicles, habitats, and extra vehicular activity equipment. A new version (3DHZETRN) capable of transporting High charge (Z) and Energy (HZE) and light ions (including neutrons) under space-like boundary conditions with enhanced neutron and light ion propagation is under development. In the present report, new algorithms for light ion and neutron propagation with well-defined convergence criteria in 3D objects is developed and tested against Monte Carlo simulations to verify the solution methodology. The code will be available through the software system, OLTARIS, for shield design and validation and provides a basis for personal computer software capable of space shield analysis and optimization.

Development of a new EMP code at LANL

NASA Astrophysics Data System (ADS)

Colman, J. J.; Roussel-Dupré, R. A.; Symbalisty, E. M.; Triplett, L. A.; Travis, B. J.

2006-05-01

A new code for modeling the generation of an electromagnetic pulse (EMP) by a nuclear explosion in the atmosphere is being developed. The source of the EMP is the Compton current produced by the prompt radiation (γ-rays, X-rays, and neutrons) of the detonation. As a first step in building a multi- dimensional EMP code we have written three kinetic codes, Plume, Swarm, and Rad. Plume models the transport of energetic electrons in air. The Plume code solves the relativistic Fokker-Planck equation over a specified energy range that can include ~ 3 keV to 50 MeV and computes the resulting electron distribution function at each cell in a two dimensional spatial grid. The energetic electrons are allowed to transport, scatter, and experience Coulombic drag. Swarm models the transport of lower energy electrons in air, spanning 0.005 eV to 30 keV. The swarm code performs a full 2-D solution to the Boltzmann equation for electrons in the presence of an applied electric field. Over this energy range the relevant processes to be tracked are elastic scattering, three body attachment, two body attachment, rotational excitation, vibrational excitation, electronic excitation, and ionization. All of these occur due to collisions between the electrons and neutral bodies in air. The Rad code solves the full radiation transfer equation in the energy range of 1 keV to 100 MeV. It includes effects of photo-absorption, Compton scattering, and pair-production. All of these codes employ a spherical coordinate system in momentum space and a cylindrical coordinate system in configuration space. The "z" axis of the momentum and configuration spaces is assumed to be parallel and we are currently also assuming complete spatial symmetry around the "z" axis. Benchmarking for each of these codes will be discussed as well as the way forward towards an integrated modern EMP code.

Three-dimensional modeling of the neutron spectrum to infer plasma conditions in cryogenic inertial confinement fusion implosions

NASA Astrophysics Data System (ADS)

Weilacher, F.; Radha, P. B.; Forrest, C.

2018-04-01

Neutron-based diagnostics are typically used to infer compressed core conditions such as areal density and ion temperature in deuterium-tritium (D-T) inertial confinement fusion (ICF) implosions. Asymmetries in the observed neutron-related quantities are important to understanding failure modes in these implosions. Neutrons from fusion reactions and their subsequent interactions including elastic scattering and neutron-induced deuteron breakup reactions are tracked to create spectra. It is shown that background subtraction is important for inferring areal density from backscattered neutrons and is less important for the forward-scattered neutrons. A three-dimensional hydrodynamic simulation of a cryogenic implosion on the OMEGA Laser System [Boehly et al., Opt. Commun. 133, 495 (1997)] using the hydrodynamic code HYDRA [Marinak et al., Phys. Plasmas 8, 2275 (2001)] is post-processed using the tracking code IRIS3D. It is shown that different parts of the neutron spectrum from the view can be mapped into different regions of the implosion, enabling an inference of an areal-density map. It is also shown that the average areal-density and an areal-density map of the compressed target can be reconstructed with a finite number of detectors placed around the target chamber. Ion temperatures are inferred from the width of the D-D and D-T fusion neutron spectra. Backgrounds can significantly alter the inferred ion temperatures from the D-D reaction, whereas they insignificantly influence the inferred D-T ion temperatures for the areal densities typical of OMEGA implosions. Asymmetries resulting in fluid flow in the core are shown to influence the absolute inferred ion temperatures from both reactions, although relative inferred values continue to reflect the underlying asymmetry pattern. The work presented here is part of the wide range of the first set of studies performed with IRIS3D. This code will continue to be used for post-processing detailed hydrodynamic simulations and interpreting observed neutron spectra in ICF implosions.

a Dosimetry Assessment for the Core Restraint of AN Advanced Gas Cooled Reactor

NASA Astrophysics Data System (ADS)

Thornton, D. A.; Allen, D. A.; Tyrrell, R. J.; Meese, T. C.; Huggon, A. P.; Whiley, G. S.; Mossop, J. R.

2009-08-01

This paper describes calculations of neutron damage rates within the core restraint structures of Advanced Gas Cooled Reactors (AGRs). Using advanced features of the Monte Carlo radiation transport code MCBEND, and neutron source data from core follow calculations performed with the reactor physics code PANTHER, a detailed model of the reactor cores of two of British Energy's AGR power plants has been developed for this purpose. Because there are no relevant neutron fluence measurements directly supporting this assessment, results of benchmark comparisons and successful validation of MCBEND for Magnox reactors have been used to estimate systematic and random uncertainties on the predictions. In particular, it has been necessary to address the known under-prediction of lower energy fast neutron responses associated with the penetration of large thicknesses of graphite.

DETECTORS AND EXPERIMENTAL METHODS: Measurement of the response function and the detection efficiency of an organic liquid scintillator for neutrons between 1 and 30 MeV

NASA Astrophysics Data System (ADS)

Huang, Han-Xiong; Ruan, Xi-Chao; Chen, Guo-Chang; Zhou, Zu-Ying; Li, Xia; Bao, Jie; Nie, Yang-Bo; Zhong, Qi-Ping

2009-08-01

The light output function of a varphi50.8 mm × 50.8 mm BC501A scintillation detector was measured in the neutron energy region of 1 to 30 MeV by fitting the pulse height (PH) spectra for neutrons with the simulations from the NRESP code at the edge range. Using the new light output function, the neutron detection efficiency was determined with two Monte-Carlo codes, NEFF and SCINFUL. The calculated efficiency was corrected by comparing the simulated PH spectra with the measured ones. The determined efficiency was verified at the near threshold region and normalized with a Proton-Recoil-Telescope (PRT) at the 8-14 MeV energy region.

Integral experiments on thorium assemblies with D-T neutron source

NASA Astrophysics Data System (ADS)

Liu, Rong; Yang, Yiwei; Feng, Song; Zheng, Lei; Lai, Caifeng; Lu, Xinxin; Wang, Mei; Jiang, Li

2017-09-01

To validate nuclear data and code in the neutronics design of a hybrid reactor with thorium, integral experiments in two kinds of benchmark thorium assemblies with a D-T fusion neutron source have been performed. The one kind of 1D assemblies consists of polyethylene and depleted uranium shells. The other kind of 2D assemblies consists of three thorium oxide cylinders. The capture reaction rates, fission reaction rates, and (n, 2n) reaction rates in 232Th in the assemblies are measured by ThO2 foils. The leakage neutron spectra from the ThO2 cylinders are measured by a liquid scintillation detector. The experimental uncertainties in all the results are analyzed. The measured results are compared to the calculated ones with MCNP code and ENDF/B-VII.0 library data.

Detecting neutrons by forward recoil protons at the Energy & Transmutation facility: Detector development and calibration with 14.1-MeV neutrons

NASA Astrophysics Data System (ADS)

Afanasev, S.; Vishnevskiy, A.; Vishnevskiy, D.; Rogachev, A.; Tyutyunnikov, S.

2017-05-01

As part of the Energy & Transmutation project, we are developing a detector for neutrons with energies in the 10-100 MeV range emitted from the target irradiated by a charged-particle beam. The neutron is detected by measuring the time-of-flight and total kinetic energy of the forward-going recoil proton [1] knocked out at a small angle from a thin layer of plastic scintillator, which has to be selected against an intense background created by γ quanta, scattered neutrons, and charged particles. On the other hand, neutron energy has to be measured over the full range with no extra tuning of the detector operation regime. Initial measurements with a source of 14.1-MeV neutrons are reported.

Coincident measurements of prompt fission γ rays and fission fragments at DANCE

NASA Astrophysics Data System (ADS)

Walker, C. L.; Baramsai, B.; Jandel, M.; Rusev, G.; Couture, A.; Mosby, S.; Ullmann, J.; Kawano, T.; Stetcu, I.; Talou, P.

2015-10-01

Modern statistical approaches to modeling fission involve the calculation of not only average quantities but also fully correlated distributions of all fission products. Applications such as those involving the detection of special nuclear materials also rely on fully correlated data of fission products. Experimental measurements of correlated data are thus critical to the validation of theory and the development of important applications. The goal of this experiment was to measure properties of prompt fission gamma-ray emission as a function of fission fragments' total kinetic energy in the spontaneous fission of 252Cf. The measurement was carried out at the Detector for Advanced Neutron Capture Experiments (DANCE), a 4 π γ-ray calorimeter. A prototype design consisting of two silicon detectors was installed in the center of DANCE, allowing simultaneous measurement of fission fragments and γ rays. Effort has been taken to simulate fragment kinetic energy losses as well as γ-ray attenuation in DANCE using such tools as GEANT4 and SRIM. Theoretical predictions generated by the code CGMF were also incorporated as input for these simulations. Results from the experiment and simulations will be presented, along with plans for future measurements.

Core Physics and Kinetics Calculations for the Fissioning Plasma Core Reactor

NASA Technical Reports Server (NTRS)

Butler, C.; Albright, D.

2007-01-01

Highly efficient, compact nuclear reactors would provide high specific impulse spacecraft propulsion. This analysis and numerical simulation effort has focused on the technical feasibility issues related to the nuclear design characteristics of a novel reactor design. The Fissioning Plasma Core Reactor (FPCR) is a shockwave-driven gaseous-core nuclear reactor, which uses Magneto Hydrodynamic effects to generate electric power to be used for propulsion. The nuclear design of the system depends on two major calculations: core physics calculations and kinetics calculations. Presently, core physics calculations have concentrated on the use of the MCNP4C code. However, initial results from other codes such as COMBINE/VENTURE and SCALE4a. are also shown. Several significant modifications were made to the ISR-developed QCALC1 kinetics analysis code. These modifications include testing the state of the core materials, an improvement to the calculation of the material properties of the core, the addition of an adiabatic core temperature model and improvement of the first order reactivity correction model. The accuracy of these modifications has been verified, and the accuracy of the point-core kinetics model used by the QCALC1 code has also been validated. Previously calculated kinetics results for the FPCR were described in the ISR report, "QCALC1: A code for FPCR Kinetics Model Feasibility Analysis" dated June 1, 2002.

Induction and disappearance of G2 chromatid breaks in lymphocytes after low doses of low-LET gamma-rays and high-LET fast neutrons.

PubMed

Vral, A; Thierens, H; Baeyens, A; De Ridder, L

2002-04-01

To determine by means of the G2 assay the number of chromatid breaks induced by low-LET gamma-rays and high-LET neutrons, and to compare the kinetics of chromatid break rejoining for radiations of different quality. The G2 assay was performed on blood samples of four healthy donors who were irradiated with low-LET gamma-rays and high-LET neutrons. In a first set of experiments a dose-response curve for the formation of chromatid breaks was carried out for gamma-rays and neutrons with doses ranging between 0.1 and 0.5 Gy. In a second set of experiments, the kinetics of chromatid break formation and disappearance were investigated after a dose of 0.5 Gy using post-irradiation times ranging between 0.5 and 3.5 h. For the highest dose of 0.5 Gy, the number of isochromatid breaks was also scored. No significant differences in the number of chromatid breaks were observed between low-LET gamma-rays and high-LET neutrons for the four donors at any of the doses given. The dose-response curves for the formation of chromatid breaks are linear for both radiation qualities and RBEs = 1 were obtained. Scoring of isochromatid breaks at the highest dose of 0.5 Gy revealed that high-LET neutrons were, however, more effective at inducing isochromatid breaks (RBE = 6.2). The rejoining experiments further showed that the kinetics of disappearance of chromatid breaks following irradiation with low-LET gamma-rays or high-LET neutrons were not significantly different. Half-times of 0.92 h for gamma-rays and 0.84 h for neutrons were obtained. Applying the G2 assay, the results demonstrate that at low doses of irradiation, the induction as well as the disappearance of chromatid breaks is independent of the LET of the radiation qualities used (0.24 keV x microm(-1) 60Co gamma-rays and 20 keV x microm(-1) fast neutrons). As these radiation qualities produce the same initial number of double-strand breaks, the results support the signal model that proposes that chromatid breaks are the result of an exchange process which is triggered by a single double-strand break.

Dosimetric and microdosimetric analyses for blood exposed to reactor-derived thermal neutrons.

PubMed

Ali, F; Atanackovic, J; Boyer, C; Festarini, A; Kildea, J; Paterson, L C; Rogge, R; Stuart, M; Richardson, R B

2018-06-06

Thermal neutrons are found in reactor, radiotherapy, aircraft, and space environments. The purpose of this study was to characterise the dosimetry and microdosimetry of thermal neutron exposures, using three simulation codes, as a precursor to quantitative radiobiological studies using blood samples. An irradiation line was designed employing a pyrolytic graphite crystal or-alternatively-a super mirror to expose blood samples to thermal neutrons from the National Research Universal reactor to determine radiobiological parameters. The crystal was used when assessing the relative biological effectiveness for dicentric chromosome aberrations, and other biomarkers, in lymphocytes over a low absorbed dose range of 1.2-14 mGy. Higher exposures using a super mirror will allow the additional quantification of mitochondrial responses. The physical size of the thermal neutron fields and their respective wavelength distribution was determined using the McStas Monte Carlo code. Spinning the blood samples produced a spatially uniform absorbed dose as determined from Monte Carlo N-Particle version 6 simulations. The major part (71%) of the total absorbed dose to blood was determined to be from the 14 N(n,p) 14 C reaction and the remainder from the 1 H(n,γ) 2 H reaction. Previous radiobiological experiments at Canadian Nuclear Laboratories involving thermal neutron irradiation of blood yielded a relative biological effectiveness of 26 ± 7. Using the Particle and Heavy Ion Transport Code System, a similar value of ∼19 for the quality factor of thermal neutrons initiating the 14 N(n,p) 14 C reaction in soft tissue was determined by microdosimetric simulations. This calculated quality factor is of similar high value to the experimentally-derived relative biological effectiveness, and indicates the potential of thermal neutrons to induce deleterious health effects in superficial organs such as cataracts of the eye lens.

Neutron-induced fission cross-section measurement of 234U with quasi-monoenergetic beams in the keV and MeV range using micromegas detectors

NASA Astrophysics Data System (ADS)

Tsinganis, A.; Kokkoris, M.; Vlastou, R.; Kalamara, A.; Stamatopoulos, A.; Kanellakopoulos, A.; Lagoyannis, A.; Axiotis, M.

2017-09-01

Accurate data on neutron-induced fission cross-sections of actinides are essential for the design of advanced nuclear reactors based either on fast neutron spectra or alternative fuel cycles, as well as for the reduction of safety margins of existing and future conventional facilities. The fission cross-section of 234U was measured at incident neutron energies of 560 and 660 keV and 7.5 MeV with a setup based on `microbulk' Micromegas detectors and the same samples previously used for the measurement performed at the CERN n_TOF facility (Karadimos et al., 2014). The 235U fission cross-section was used as reference. The (quasi-)monoenergetic neutron beams were produced via the 7Li(p,n) and the 2H(d,n) reactions at the neutron beam facility of the Institute of Nuclear and Particle Physics at the `Demokritos' National Centre for Scientific Research. A detailed study of the neutron spectra produced in the targets and intercepted by the samples was performed coupling the NeuSDesc and MCNPX codes, taking into account the energy spread, energy loss and angular straggling of the beam ions in the target assemblies, as well as contributions from competing reactions and neutron scattering in the experimental setup. Auxiliary Monte-Carlo simulations were performed with the FLUKA code to study the behaviour of the detectors, focusing particularly on the reproduction of the pulse height spectra of α-particles and fission fragments (using distributions produced with the GEF code) for the evaluation of the detector efficiency. An overview of the developed methodology and preliminary results are presented.

Benchmarking of Neutron Production of Heavy-Ion Transport Codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Remec, Igor; Ronningen, Reginald M.; Heilbronn, Lawrence

Accurate prediction of radiation fields generated by heavy ion interactions is important in medical applications, space missions, and in design and operation of rare isotope research facilities. In recent years, several well-established computer codes in widespread use for particle and radiation transport calculations have been equipped with the capability to simulate heavy ion transport and interactions. To assess and validate these capabilities, we performed simulations of a series of benchmark-quality heavy ion experiments with the computer codes FLUKA, MARS15, MCNPX, and PHITS. We focus on the comparisons of secondary neutron production. Results are encouraging; however, further improvements in models andmore » codes and additional benchmarking are required.« less

NASA Radiation Protection Research for Exploration Missions

NASA Technical Reports Server (NTRS)

Wilson, John W.; Cucinotta, Francis A.; Tripathi, Ram K.; Heinbockel, John H.; Tweed, John; Mertens, Christopher J.; Walker, Steve A.; Blattnig, Steven R.; Zeitlin, Cary J.

2006-01-01

The HZETRN code was used in recent trade studies for renewed lunar exploration and currently used in engineering development of the next generation of space vehicles, habitats, and EVA equipment. A new version of the HZETRN code capable of simulating high charge and energy (HZE) ions, light-ions and neutrons with either laboratory or space boundary conditions with enhanced neutron and light-ion propagation is under development. Atomic and nuclear model requirements to support that development will be discussed. Such engineering design codes require establishing validation processes using laboratory ion beams and space flight measurements in realistic geometries. We discuss limitations of code validation due to the currently available data and recommend priorities for new data sets.

Assessment and Mitigation of Radiation, EMP, Debris & Shrapnel Impacts at Megajoule-Class Laser Facilities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Eder, D C; Anderson, R W; Bailey, D S

2009-10-05

The generation of neutron/gamma radiation, electromagnetic pulses (EMP), debris and shrapnel at mega-Joule class laser facilities (NIF and LMJ) impacts experiments conducted at these facilities. The complex 3D numerical codes used to assess these impacts range from an established code that required minor modifications (MCNP - calculates neutron and gamma radiation levels in complex geometries), through a code that required significant modifications to treat new phenomena (EMSolve - calculates EMP from electrons escaping from laser targets), to a new code, ALE-AMR, that is being developed through a joint collaboration between LLNL, CEA, and UC (UCSD, UCLA, and LBL) for debrismore » and shrapnel modelling.« less

McStas event logger: Definition and applications

NASA Astrophysics Data System (ADS)

Bergbäck Knudsen, Erik; Bryndt Klinkby, Esben; Kjær Willendrup, Peter

2014-02-01

Functionality is added to the McStas neutron ray-tracing code, which allows individual neutron states before and after a scattering to be temporarily stored, and analysed. This logging mechanism has multiple uses, including studies of longitudinal intensity loss in neutron guides and guide coating design optimisations. Furthermore, the logging method enables the cold/thermal neutron induced gamma background along the guide to be calculated from the un-reflected neutron, using a recently developed MCNPX-McStas interface.

Study of muon-induced neutron production using accelerator muon beam at CERN

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nakajima, Y.; Lin, C. J.; Ochoa-Ricoux, J. P.

2015-08-17

Cosmogenic muon-induced neutrons are one of the most problematic backgrounds for various underground experiments for rare event searches. In order to accurately understand such backgrounds, experimental data with high-statistics and well-controlled systematics is essential. We performed a test experiment to measure muon-induced neutron production yield and energy spectrum using a high-energy accelerator muon beam at CERN. We successfully observed neutrons from 160 GeV/c muon interaction on lead, and measured kinetic energy distributions for various production angles. Works towards evaluation of absolute neutron production yield is underway. This work also demonstrates that the setup is feasible for a future large-scale experimentmore » for more comprehensive study of muon-induced neutron production.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Agafonov, A. I., E-mail: aai@isssph.kiae.ru

The inelastic scattering of cold neutrons by a ring leads to quantum jumps of a superconducting current which correspond to a decrease in the fluxoid quantum number by one or several units while the change in the ring energy is transferred to the kinetic energy of the scattered neutron. The scattering cross sections of transversely polarized neutrons have been calculated for a thin type-II superconductor ring, the thickness of which is smaller than the field penetration depth but larger than the electron mean free path.

Computed secondary-particle energy spectra following nonelastic neutron interactions with C-12 for E(n) between 15 and 60 MeV: Comparisons of results from two calculational methods

NASA Astrophysics Data System (ADS)

Dickens, J. K.

1991-04-01

The organic scintillation detector response code SCINFUL has been used to compute secondary-particle energy spectra, d(sigma)/dE, following nonelastic neutron interactions with C-12 for incident neutron energies between 15 and 60 MeV. The resulting spectra are compared with published similar spectra computed by Brenner and Prael who used an intranuclear cascade code, including alpha clustering, a particle pickup mechanism, and a theoretical approach to sequential decay via intermediate particle-unstable states. The similarities of and the differences between the results of the two approaches are discussed.

Object kinetic Monte Carlo model for neutron and ion irradiation in tungsten: Impact of transmutation and carbon impurities

NASA Astrophysics Data System (ADS)

Castin, N.; Bonny, G.; Bakaev, A.; Ortiz, C. J.; Sand, A. E.; Terentyev, D.

2018-03-01

We upgrade our object kinetic Monte Carlo (OKMC) model, aimed at describing the microstructural evolution in tungsten (W) under neutron and ion irradiation. Two main improvements are proposed based on recently published atomistic data: (a) interstitial carbon impurities, and their interaction with radiation-induced defects (point defect clusters and loops), are more accurately parameterized thanks to ab initio findings; (b) W transmutation to rhenium (Re) upon neutron irradiation, impacting the diffusivity of radiation defects, is included, also relying on recent atomistic data. These essential amendments highly improve the portability of our OKMC model, providing a description for the formation of SIA-type loops under different irradiation conditions. The model is applied to simulate neutron and ion irradiation in pure W samples, in a wide range of fluxes and temperatures. We demonstrate that it performs a realistic prediction of the population of TEM-visible voids and loops, as compared to experimental evidence. The impact of the transmutation of W to Re, and of carbon trapping, is assessed.

Fission Fragment Mass Distributions and Total Kinetic Energy Release of 235-Uranium and 238-Uranium in Neutron-Induced Fission at Intermediate and Fast Neutron Energies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Duke, Dana Lynn

2015-11-12

This Ph.D. dissertation describes a measurement of the change in mass distributions and average total kinetic energy (TKE) release with increasing incident neutron energy for fission of 235U and 238U. Although fission was discovered over seventy-five years ago, open questions remain about the physics of the fission process. The energy of the incident neutron, En, changes the division of energy release in the resulting fission fragments, however, the details of energy partitioning remain ambiguous because the nucleus is a many-body quantum system. Creating a full theoretical model is difficult and experimental data to validate existing models are lacking. Additional fissionmore » measurements will lead to higher-quality models of the fission process, therefore improving applications such as the development of next-generation nuclear reactors and defense. This work also paves the way for precision experiments such as the Time Projection Chamber (TPC) for fission cross section measurements and the Spectrometer for Ion Determination in Fission (SPIDER) for precision mass yields.« less

Correlations of neutron multiplicity and γ -ray multiplicity with fragment mass and total kinetic energy in spontaneous fission of Cf 252

DOE PAGES

Wang, Taofeng; Li, Guangwu; Zhu, Liping; ...

2016-01-08

The dependence of correlations of neutron multiplicity ν and γ-ray multiplicity M γ in spontaneous fission of 252Cf on fragment mass A* and total kinetic energy (TKE) have been investigated by employing the ratio of M γ/ν and the form of M γ(ν). We show for the first time that M γ and ν have a complex correlation for heavy fragment masses, while there is a positive dependence of Mγ for light fragment masses and for near-symmetric mass splits. The ratio M γ/ν exhibits strong shell effects for neutron magic number N=50 and near doubly magic number shell closure atmore » Z=50 and N=82. The γ-ray multiplicity Mγ has a maximum for TKE=165-170 MeV. Above 170 MeV M γ(TKE) is approximately linear, while it deviates significantly from a linear dependence at lower TKE. The correlation between the average neutron and γ-ray multiplicities can be partly reproduced by model calculations.« less

LSENS, The NASA Lewis Kinetics and Sensitivity Analysis Code

NASA Technical Reports Server (NTRS)

Radhakrishnan, K.

2000-01-01

A general chemical kinetics and sensitivity analysis code for complex, homogeneous, gas-phase reactions is described. The main features of the code, LSENS (the NASA Lewis kinetics and sensitivity analysis code), are its flexibility, efficiency and convenience in treating many different chemical reaction models. The models include: static system; steady, one-dimensional, inviscid flow; incident-shock initiated reaction in a shock tube; and a perfectly stirred reactor. In addition, equilibrium computations can be performed for several assigned states. An implicit numerical integration method (LSODE, the Livermore Solver for Ordinary Differential Equations), which works efficiently for the extremes of very fast and very slow reactions, is used to solve the "stiff" ordinary differential equation systems that arise in chemical kinetics. For static reactions, the code uses the decoupled direct method to calculate sensitivity coefficients of the dependent variables and their temporal derivatives with respect to the initial values of dependent variables and/or the rate coefficient parameters. Solution methods for the equilibrium and post-shock conditions and for perfectly stirred reactor problems are either adapted from or based on the procedures built into the NASA code CEA (Chemical Equilibrium and Applications).

An Improved Neutron Transport Algorithm for HZETRN2006

NASA Astrophysics Data System (ADS)

Slaba, Tony

NASA's new space exploration initiative includes plans for long term human presence in space thereby placing new emphasis on space radiation analyses. In particular, a systematic effort of verification, validation and uncertainty quantification of the tools commonly used for radiation analysis for vehicle design and mission planning has begun. In this paper, the numerical error associated with energy discretization in HZETRN2006 is addressed; large errors in the low-energy portion of the neutron fluence spectrum are produced due to a numerical truncation error in the transport algorithm. It is shown that the truncation error results from the narrow energy domain of the neutron elastic spectral distributions, and that an extremely fine energy grid is required in order to adequately resolve the problem under the current formulation. Since adding a sufficient number of energy points will render the code computationally inefficient, we revisit the light-ion transport theory developed for HZETRN2006 and focus on neutron elastic interactions. The new approach that is developed numerically integrates with adequate resolution in the energy domain without affecting the run-time of the code and is easily incorporated into the current code. Efforts were also made to optimize the computational efficiency of the light-ion propagator; a brief discussion of the efforts is given along with run-time comparisons between the original and updated codes. Convergence testing is then completed by running the code for various environments and shielding materials with many different energy grids to ensure stability of the proposed method.

Neutron production by cosmic-ray muons in various materials

NASA Astrophysics Data System (ADS)

Manukovsky, K. V.; Ryazhskaya, O. G.; Sobolevsky, N. M.; Yudin, A. V.

2016-07-01

The results obtained by studying the background of neutrons produced by cosmic-raymuons in underground experimental facilities intended for rare-event searches and in surrounding rock are presented. The types of this rock may include granite, sedimentary rock, gypsum, and rock salt. Neutron production and transfer were simulated using the Geant4 and SHIELD transport codes. These codes were tuned via a comparison of the results of calculations with experimental data—in particular, with data of the Artemovsk research station of the Institute for Nuclear Research (INR, Moscow, Russia)—as well as via an intercomparison of results of calculations with the Geant4 and SHIELD codes. It turns out that the atomic-number dependence of the production and yield of neutrons has an irregular character and does not allow a description in terms of a universal function of the atomic number. The parameters of this dependence are different for two groups of nuclei—nuclei consisting of alpha particles and all of the remaining nuclei. Moreover, there are manifest exceptions from a power-law dependence—for example, argon. This may entail important consequences both for the existing underground experimental facilities and for those under construction. Investigation of cosmic-ray-induced neutron production in various materials is of paramount importance for the interpretation of experiments conducted at large depths under the Earth's surface.

TRIPOLI-4® - MCNP5 ITER A-lite neutronic model benchmarking

NASA Astrophysics Data System (ADS)

Jaboulay, J.-C.; Cayla, P.-Y.; Fausser, C.; Lee, Y.-K.; Trama, J.-C.; Li-Puma, A.

2014-06-01

The aim of this paper is to present the capability of TRIPOLI-4®, the CEA Monte Carlo code, to model a large-scale fusion reactor with complex neutron source and geometry. In the past, numerous benchmarks were conducted for TRIPOLI-4® assessment on fusion applications. Experiments (KANT, OKTAVIAN, FNG) analysis and numerical benchmarks (between TRIPOLI-4® and MCNP5) on the HCLL DEMO2007 and ITER models were carried out successively. In this previous ITER benchmark, nevertheless, only the neutron wall loading was analyzed, its main purpose was to present MCAM (the FDS Team CAD import tool) extension for TRIPOLI-4®. Starting from this work a more extended benchmark has been performed about the estimation of neutron flux, nuclear heating in the shielding blankets and tritium production rate in the European TBMs (HCLL and HCPB) and it is presented in this paper. The methodology to build the TRIPOLI-4® A-lite model is based on MCAM and the MCNP A-lite model (version 4.1). Simplified TBMs (from KIT) have been integrated in the equatorial-port. Comparisons of neutron wall loading, flux, nuclear heating and tritium production rate show a good agreement between the two codes. Discrepancies are mainly included in the Monte Carlo codes statistical error.

Kinetics of methane-ethane gas replacement in clathrate-hydrates studied by time-resolved neutron diffraction and Raman spectroscopy.

PubMed

Murshed, M Mangir; Schmidt, Burkhard C; Kuhs, Werner F

2010-01-14

The kinetics of CH(4)-C(2)H(6) replacement in gas hydrates has been studied by in situ neutron diffraction and Raman spectroscopy. Deuterated ethane structure type I (C(2)H(6) sI) hydrates were transformed in a closed volume into methane-ethane mixed structure type II (CH(4)-C(2)H(6) sII) hydrates at 5 MPa and various temperatures in the vicinity of 0 degrees C while followed by time-resolved neutron powder diffraction on D20 at ILL, Grenoble. The role of available surface area of the sI starting material on the formation kinetics of sII hydrates was studied. Ex situ Raman spectroscopic investigations were carried out to crosscheck the gas composition and the distribution of the gas species over the cages as a function of structure type and compared to the in situ neutron results. Raman micromapping on single hydrate grains showed compositional and structural gradients between the surface and core of the transformed hydrates. Moreover, the observed methane-ethane ratio is very far from the one expected for a formation from a constantly equilibrated gas phase. The results also prove that gas replacement in CH(4)-C(2)H(6) hydrates is a regrowth process involving the nucleation of new crystallites commencing at the surface of the parent C(2)H(6) sI hydrate with a progressively shrinking core of unreacted material. The time-resolved neutron diffraction results clearly indicate an increasing diffusion limitation of the exchange process. This diffusion limitation leads to a progressive slowing down of the exchange reaction and is likely to be responsible for the incomplete exchange of the gases.

Deep inelastic neutron scattering on 207Pb and NaHF 2 as a test of a detectors array on the VESUVIO spectrometer

NASA Astrophysics Data System (ADS)

Pietropaolo, A.; Senesi, R.

2008-01-01

A prototype array of resonance detectors for deep inelastic neutron scattering experiments has been installed on the VESUVIO spectrometer, at the ISIS spallation neutron source. Deep inelastic neutron scattering measurements on a reference lead sample and on NaHF 2 molecular system are presented. Despite on an explorative level, the results obtained for the values of mean kinetic energy are found in good agreement with the theoretical predictions, thus assessing the potential capability of the device for a routine use on the instrument.

Development and preliminary verification of the 3D core neutronic code: COCO

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lu, H.; Mo, K.; Li, W.

As the recent blooming economic growth and following environmental concerns (China)) is proactively pushing forward nuclear power development and encouraging the tapping of clean energy. Under this situation, CGNPC, as one of the largest energy enterprises in China, is planning to develop its own nuclear related technology in order to support more and more nuclear plants either under construction or being operation. This paper introduces the recent progress in software development for CGNPC. The focus is placed on the physical models and preliminary verification results during the recent development of the 3D Core Neutronic Code: COCO. In the COCO code,more » the non-linear Green's function method is employed to calculate the neutron flux. In order to use the discontinuity factor, the Neumann (second kind) boundary condition is utilized in the Green's function nodal method. Additionally, the COCO code also includes the necessary physical models, e.g. single-channel thermal-hydraulic module, burnup module, pin power reconstruction module and cross-section interpolation module. The preliminary verification result shows that the COCO code is sufficient for reactor core design and analysis for pressurized water reactor (PWR). (authors)« less

Ex-vessel neutron dosimetry analysis for westinghouse 4-loop XL pressurized water reactor plant using the RadTrack{sup TM} Code System with the 3D parallel discrete ordinates code RAPTOR-M3G

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chen, J.; Alpan, F. A.; Fischer, G.A.

2011-07-01

Traditional two-dimensional (2D)/one-dimensional (1D) SYNTHESIS methodology has been widely used to calculate fast neutron (>1.0 MeV) fluence exposure to reactor pressure vessel in the belt-line region. However, it is expected that this methodology cannot provide accurate fast neutron fluence calculation at elevations far above or below the active core region. A three-dimensional (3D) parallel discrete ordinates calculation for ex-vessel neutron dosimetry on a Westinghouse 4-Loop XL Pressurized Water Reactor has been done. It shows good agreement between the calculated results and measured results. Furthermore, the results show very different fast neutron flux values at some of the former plate locationsmore » and elevations above and below an active core than those calculated by a 2D/1D SYNTHESIS method. This indicates that for certain irregular reactor internal structures, where the fast neutron flux has a very strong local effect, it is required to use a 3D transport method to calculate accurate fast neutron exposure. (authors)« less

Monte Carlo calculation for the development of a BNCT neutron source (1eV-10KeV) using MCNP code.

PubMed

El Moussaoui, F; El Bardouni, T; Azahra, M; Kamili, A; Boukhal, H

2008-09-01

Different materials have been studied in order to produce the epithermal neutron beam between 1eV and 10KeV, which are extensively used to irradiate patients with brain tumors such as GBM. For this purpose, we have studied three different neutrons moderators (H(2)O, D(2)O and BeO) and their combinations, four reflectors (Al(2)O(3), C, Bi, and Pb) and two filters (Cd and Bi). Results of calculation showed that the best obtained assembly configuration corresponds to the combination of the three moderators H(2)O, BeO and D(2)O jointly to Al(2)O(3) reflector and two filter Cd+Bi optimize the spectrum of the epithermal neutron at 72%, and minimize the thermal neutron to 4% and thus it can be used to treat the deep tumor brain. The calculations have been performed by means of the Monte Carlo N (particle code MCNP 5C). Our results strongly encourage further studying of irradiation of the head with epithermal neutron fields.

Digital pile-up rejection for plutonium experiments with solution-grown stilbene

NASA Astrophysics Data System (ADS)

Bourne, M. M.; Clarke, S. D.; Paff, M.; DiFulvio, A.; Norsworthy, M.; Pozzi, S. A.

2017-01-01

A solution-grown stilbene detector was used in several experiments with plutonium samples including plutonium oxide, mixed oxide, and plutonium metal samples. Neutrons from different reactions and plutonium isotopes are accompanied by numerous gamma rays especially by the 59-keV gamma ray of 241Am. Identifying neutrons correctly is important for nuclear nonproliferation applications and makes neutron/gamma discrimination and pile-up rejection necessary. Each experimental dataset is presented with and without pile-up filtering using a previously developed algorithm. The experiments were simulated using MCNPX-PoliMi, a Monte Carlo code designed to accurately model scintillation detector response. Collision output from MCNPX-PoliMi was processed using the specialized MPPost post-processing code to convert neutron energy depositions event-by-event into light pulses. The model was compared to experimental data after pulse-shape discrimination identified waveforms as gamma ray or neutron interactions. We show that the use of the digital pile-up rejection algorithm allows for accurate neutron counting with stilbene to within 2% even when not using lead shielding.

Interfacing MCNPX and McStas for simulation of neutron transport

NASA Astrophysics Data System (ADS)

Klinkby, Esben; Lauritzen, Bent; Nonbøl, Erik; Kjær Willendrup, Peter; Filges, Uwe; Wohlmuther, Michael; Gallmeier, Franz X.

2013-02-01

Simulations of target-moderator-reflector system at spallation sources are conventionally carried out using Monte Carlo codes such as MCNPX (Waters et al., 2007 [1]) or FLUKA (Battistoni et al., 2007; Ferrari et al., 2005 [2,3]) whereas simulations of neutron transport from the moderator and the instrument response are performed by neutron ray tracing codes such as McStas (Lefmann and Nielsen, 1999; Willendrup et al., 2004, 2011a,b [4-7]). The coupling between the two simulation suites typically consists of providing analytical fits of MCNPX neutron spectra to McStas. This method is generally successful but has limitations, as it e.g. does not allow for re-entry of neutrons into the MCNPX regime. Previous work to resolve such shortcomings includes the introduction of McStas inspired supermirrors in MCNPX. In the present paper different approaches to interface MCNPX and McStas are presented and applied to a simple test case. The direct coupling between MCNPX and McStas allows for more accurate simulations of e.g. complex moderator geometries, backgrounds, interference between beam-lines as well as shielding requirements along the neutron guides.

Determination of neutron flux distribution in an Am-Be irradiator using the MCNP.

PubMed

Shtejer-Diaz, K; Zamboni, C B; Zahn, G S; Zevallos-Chávez, J Y

2003-10-01

A neutron irradiator has been assembled at IPEN facilities to perform qualitative-quantitative analysis of many materials using thermal and fast neutrons outside the nuclear reactor premises. To establish the prototype specifications, the neutron flux distribution and the absorbed dose rates were calculated using the MCNP computer code. These theoretical predictions then allow one to discuss the optimum irradiator design and its performance.

MCNP-REN - A Monte Carlo Tool for Neutron Detector Design Without Using the Point Model

DOE Office of Scientific and Technical Information (OSTI.GOV)

Abhold, M.E.; Baker, M.C.

1999-07-25

The development of neutron detectors makes extensive use of the predictions of detector response through the use of Monte Carlo techniques in conjunction with the point reactor model. Unfortunately, the point reactor model fails to accurately predict detector response in common applications. For this reason, the general Monte Carlo N-Particle code (MCNP) was modified to simulate the pulse streams that would be generated by a neutron detector and normally analyzed by a shift register. This modified code, MCNP - Random Exponentially Distributed Neutron Source (MCNP-REN), along with the Time Analysis Program (TAP) predict neutron detector response without using the pointmore » reactor model, making it unnecessary for the user to decide whether or not the assumptions of the point model are met for their application. MCNP-REN is capable of simulating standard neutron coincidence counting as well as neutron multiplicity counting. Measurements of MOX fresh fuel made using the Underwater Coincidence Counter (UWCC) as well as measurements of HEU reactor fuel using the active neutron Research Reactor Fuel Counter (RRFC) are compared with calculations. The method used in MCNP-REN is demonstrated to be fundamentally sound and shown to eliminate the need to use the point model for detector performance predictions.« less

VESUVIO: a novel instrument for performing spectroscopic studies in condensed matter with eV neutrons at the ISIS facility

NASA Astrophysics Data System (ADS)

Senesi, R.; Andreani, C.; Bowden, Z.; Colognesi, D.; Degiorgi, E.; Fielding, A. L.; Mayers, J.; Nardone, M.; Norris, J.; Praitano, M.; Rhodes, N. J.; Stirling, W. G.; Tomkinson, J.; Uden, C.

2000-03-01

The VESUVIO project aims to provide unique prototype instrumentation at the ISIS-pulsed neutron source and to establish a routine experimental and theoretical program in neutron scattering spectroscopy at eV energies. This instrumentation will be specifically designed for high momentum, (20 Å-11 eV) inelastic neutron scattering studies of microscopic dynamical processes in materials and will represent a unique facility for EU researchers. It will allow to derive single-particle kinetic energies and single-particle momentum distributions, n(p), providing additional and/or complementary information to other neutron inelastic spectroscopic techniques.

Studies of Neutron-Induced Fission of 235U, 238U, and 239Pu

NASA Astrophysics Data System (ADS)

Duke, Dana; TKE Team

2014-09-01

A Frisch-gridded ionization chamber and the double energy (2E) analysis method were used to study mass yield distributions and average total kinetic energy (TKE) release from neutron-induced fission of 235U, 238U, and 239Pu. Despite decades of fission research, little or no TKE data exist for high incident neutron energies. Additional average TKE information at incident neutron energies relevant to defense- and energy-related applications will provide a valuable observable for benchmarking simulations. The data can also be used as inputs in theoretical fission models. The Los Alamos Neutron Science Center-Weapons Neutron Research (LANSCE - WNR) provides a neutron beam from thermal to hundreds of MeV, well-suited for filling in the gaps in existing data and exploring fission behavior in the fast neutron region. The results of the studies on 238U, 235U, and 239Pu will be presented. LA-UR-14-24921.

The VESUVIO electron volt neutron spectrometer

NASA Astrophysics Data System (ADS)

Mayers, J.; Reiter, G.

2012-04-01

This paper describes the VESUVIO electron volt neutron spectrometer at the ISIS pulsed neutron source and its data analysis routines. VESUVIO is used primarily for the measurement of proton momentum distributions in condensed matter systems, but can also be used to measure the kinetic energies of heavier masses and bulk in-situ sample compositions. A series of VESUVIO runs on the same zirconium hydride sample over the past two years show that (1) kinetic energies of protons can be measured to an absolute accuracy of ˜1%. (2) Measurements of the proton momentum distribution n(p) are highly reproducible from run to run. This shows that small changes in kinetic energy and the detailed shape of n(p) with parameters such as temperature, pressure and sample composition can be reliably extracted from VESUVIO data. (3) The impulse approximation (IA) is well satisfied on VESUVIO. (4) The small deviations from the IA due to the finite momentum transfer of measurement are well understood. (5) There is an anomaly in the magnitude of the inelastic neutron cross-section of the protons in zirconium hydride, with an observed reduction of 10% ± 0.3% from that given in standard tables. This anomaly is independent of energy transfer to within experimental error. Future instrument developments are discussed. These would allow the measurement of n(p) in other light atoms, D, 3He, 4He, Li, C and O and measurement of eV electronic and magnetic excitations.

Triaxial instabilities in rapidly rotating neutron stars

NASA Astrophysics Data System (ADS)

Basak, Arkadip

2018-06-01

Viscosity driven bar mode secular instabilities of rapidly rotating neutron stars are studied using LORENE/Nrotstar code. These instabilities set a more rigorous limit to the rotation frequency of a neutron star than the Kepler frequency/mass-shedding limit. The procedure employed in the code comprises of perturbing an axisymmetric and stationary configuration of a neutron star and studying its evolution by constructing a series of triaxial quasi-equilibrium configurations. Symmetry breaking point was found out for Polytropic as well as 10 realistic equations of states (EOS) from the CompOSE data base. The concept of piecewise polytropic EOSs has been used to comprehend the rotational instability of Realistic EOSs and validated with 19 different Realistic EOSs from CompOSE. The possibility of detecting quasi-periodic gravitational waves from viscosity driven instability with ground-based LIGO/VIRGO interferometers is also discussed very briefly.

Kinetic physics in ICF: present understanding and future directions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rinderknecht, Hans G.; Amendt, P. A.; Wilks, S. C.

Kinetic physics has the potential to impact the performance of indirect-drive inertial confinement fusion (ICF) experiments. Systematic anomalies in the National Ignition Facility implosion dataset have been identified in which kinetic physics may play a role, including inferred missing energy in the hohlraum, drive asymmetry in near-vacuum hohlraums, low areal density and high burn-averaged ion temperatures (T i ) compared with mainline simulations, and low ratios of the DD-neutron and DT-neutron yields and inferred T i . Several components of ICF implosions are likely to be influenced or dominated by kinetic physics: laser-plasma interactions in the LEH and hohlraum interior;more » the hohlraum wall blowoff, blowoff/gas and blowoff/ablator interfaces; the ablator and ablator/ice interface; and the DT fuel all present conditions in which kinetic physics can significantly affect the dynamics. This review presents the assembled experimental data and simulation results to date, which indicate that the effects of long mean-free-path plasma phenomena and self-generated electromagnetic fields may have a significant impact in ICF targets. Finally, simulation and experimental efforts are proposed to definitively quantify the importance of these effects at ignition-relevant conditions, including priorities for ongoing study.« less

Kinetic physics in ICF: present understanding and future directions

DOE PAGES

Rinderknecht, Hans G.; Amendt, P. A.; Wilks, S. C.; ...

2018-03-19

Kinetic physics has the potential to impact the performance of indirect-drive inertial confinement fusion (ICF) experiments. Systematic anomalies in the National Ignition Facility implosion dataset have been identified in which kinetic physics may play a role, including inferred missing energy in the hohlraum, drive asymmetry in near-vacuum hohlraums, low areal density and high burn-averaged ion temperatures (T i ) compared with mainline simulations, and low ratios of the DD-neutron and DT-neutron yields and inferred T i . Several components of ICF implosions are likely to be influenced or dominated by kinetic physics: laser-plasma interactions in the LEH and hohlraum interior;more » the hohlraum wall blowoff, blowoff/gas and blowoff/ablator interfaces; the ablator and ablator/ice interface; and the DT fuel all present conditions in which kinetic physics can significantly affect the dynamics. This review presents the assembled experimental data and simulation results to date, which indicate that the effects of long mean-free-path plasma phenomena and self-generated electromagnetic fields may have a significant impact in ICF targets. Finally, simulation and experimental efforts are proposed to definitively quantify the importance of these effects at ignition-relevant conditions, including priorities for ongoing study.« less

Kinetic physics in ICF: present understanding and future directions

NASA Astrophysics Data System (ADS)

Rinderknecht, Hans G.; Amendt, P. A.; Wilks, S. C.; Collins, G.

2018-06-01

Kinetic physics has the potential to impact the performance of indirect-drive inertial confinement fusion (ICF) experiments. Systematic anomalies in the National Ignition Facility implosion dataset have been identified in which kinetic physics may play a role, including inferred missing energy in the hohlraum, drive asymmetry in near-vacuum hohlraums, low areal density and high burn-averaged ion temperatures (〈Ti 〉) compared with mainline simulations, and low ratios of the DD-neutron and DT-neutron yields and inferred 〈Ti 〉. Several components of ICF implosions are likely to be influenced or dominated by kinetic physics: laser-plasma interactions in the LEH and hohlraum interior; the hohlraum wall blowoff, blowoff/gas and blowoff/ablator interfaces; the ablator and ablator/ice interface; and the DT fuel all present conditions in which kinetic physics can significantly affect the dynamics. This review presents the assembled experimental data and simulation results to date, which indicate that the effects of long mean-free-path plasma phenomena and self-generated electromagnetic fields may have a significant impact in ICF targets. Simulation and experimental efforts are proposed to definitively quantify the importance of these effects at ignition-relevant conditions, including priorities for ongoing study.

A neutron spectrometer for studying giant resonances with (p,n) reactions in inverse kinematics

NASA Astrophysics Data System (ADS)

Stuhl, L.; Krasznahorkay, A.; Csatlós, M.; Algora, A.; Gulyás, J.; Kalinka, G.; Timár, J.; Kalantar-Nayestanaki, N.; Rigollet, C.; Bagchi, S.; Najafi, M. A.

2014-02-01

A neutron spectrometer, the European Low-Energy Neutron Spectrometer (ELENS), has been constructed to study exotic nuclei in inverse-kinematics experiments. The spectrometer, which consists of plastic scintillator bars, can be operated in the neutron energy range of 100 keV-10 MeV. The neutron energy is determined using the time-of-flight technique, while the position of the neutron detection is deduced from the time-difference information from photomultipliers attached to both ends of each bar. A novel wrapping method has been developed for the plastic scintillators. The array has a larger than 25% detection efficiency for neutrons of approximately 500 keV in kinetic energy and an angular resolution of less than 1°. Details of the design, construction and experimental tests of the spectrometer will be presented.

Radionuclide production and dose rate estimation during the commissioning of the W-Ta spallation target

NASA Astrophysics Data System (ADS)

Yu, Q. Z.; Liang, T. J.

2018-06-01

China Spallation Neutron Source (CSNS) is intended to begin operation in 2018. CSNS is an accelerator-base multidisciplinary user facility. The pulsed neutrons are produced by a 1.6GeV short-pulsed proton beam impinging on a W-Ta spallation target, at a beam power of100 kW and a repetition rate of 25 Hz. 20 neutron beam lines are extracted for the neutron scattering and neutron irradiation research. During the commissioning and maintenance scenarios, the gamma rays induced from the W-Ta target can cause the dose threat to the personal and the environment. In this paper, the gamma dose rate distributions for the W-Ta spallation are calculated, based on the engineering model of the target-moderator-reflector system. The shipping cask is analyzed to satisfy the dose rate limit that less than 2 mSv/h at the surface of the shipping cask. All calculations are performed by the Monte carlo code MCNPX2.5 and the activation code CINDER’90.

Neutron Capture gamma ENDF libraries for modeling and identification of neutron sources

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sleaford, B

2007-10-29

There are a number of inaccuracies and data omissions with respect to gammas from neutron capture in the ENDF libraries used as field reference information and by modeling codes used in JTOT. As the use of Active Neutron interrogation methods is expanded, these shortfalls become more acute. A new, more accurate and complete evaluated experimental database of gamma rays (over 35,000 lines for 262 isotopes up to U so far) from thermal neutron capture has recently become available from the IAEA. To my knowledge, none of this new data has been installed in ENDF libraries and disseminated. I propose tomore » upgrade libraries of {sup 184,186}W, {sup 56}Fe, {sup 204,206,207}Pb, {sup 104}Pd, and {sup 19}F the 1st year. This will involve collaboration with Richard Firestone at LBL in evaluating the data and installing it in the libraries. I will test them with the transport code MCNP5.« less

Fast neutron counting in a mobile, trailer-based search platform

NASA Astrophysics Data System (ADS)

Hayward, Jason P.; Sparger, John; Fabris, Lorenzo; Newby, Robert J.

2017-12-01

Trailer-based search platforms for detection of radiological and nuclear threats are often based upon coded aperture gamma-ray imaging, because this method can be rendered insensitive to local variations in gamma background while still localizing the source well. Since gamma source emissions are rather easily shielded, in this work we consider the addition of fast neutron counting to a mobile platform for detection of sources containing Pu. A proof-of-concept system capable of combined gamma and neutron coded-aperture imaging was built inside of a trailer and used to detect a 252Cf source while driving along a roadway. Neutron detector types employed included EJ-309 in a detector plane and EJ-299-33 in a front mask plane. While the 252Cf gamma emissions were not readily detectable while driving by at 16.9 m standoff, the neutron emissions can be detected while moving. Mobile detection performance for this system and a scaled-up system design are presented, along with implications for threat sensing.

Benchmarking of neutron production of heavy-ion transport codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Remec, I.; Ronningen, R. M.; Heilbronn, L.

Document available in abstract form only, full text of document follows: Accurate prediction of radiation fields generated by heavy ion interactions is important in medical applications, space missions, and in design and operation of rare isotope research facilities. In recent years, several well-established computer codes in widespread use for particle and radiation transport calculations have been equipped with the capability to simulate heavy ion transport and interactions. To assess and validate these capabilities, we performed simulations of a series of benchmark-quality heavy ion experiments with the computer codes FLUKA, MARS15, MCNPX, and PHITS. We focus on the comparisons of secondarymore » neutron production. Results are encouraging; however, further improvements in models and codes and additional benchmarking are required. (authors)« less

Verification of a neutronic code for transient analysis in reactors with Hex-z geometry

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gonzalez-Pintor, S.; Verdu, G.; Ginestar, D.

Due to the geometry of the fuel bundles, to simulate reactors such as VVER reactors it is necessary to develop methods that can deal with hexagonal prisms as basic elements of the spatial discretization. The main features of a code based on a high order finite element method for the spatial discretization of the neutron diffusion equation and an implicit difference method for the time discretization of this equation are presented and the performance of the code is tested solving the first exercise of the AER transient benchmark. The obtained results are compared with the reference results of the benchmarkmore » and with the results provided by PARCS code. (authors)« less

Simulation of Nuclear Reactor Kinetics by the Monte Carlo Method

NASA Astrophysics Data System (ADS)

Gomin, E. A.; Davidenko, V. D.; Zinchenko, A. S.; Kharchenko, I. K.

2017-12-01

The KIR computer code intended for calculations of nuclear reactor kinetics using the Monte Carlo method is described. The algorithm implemented in the code is described in detail. Some results of test calculations are given.

Fast-neutron, coded-aperture imager

NASA Astrophysics Data System (ADS)

Woolf, Richard S.; Phlips, Bernard F.; Hutcheson, Anthony L.; Wulf, Eric A.

2015-06-01

This work discusses a large-scale, coded-aperture imager for fast neutrons, building off a proof-of concept instrument developed at the U.S. Naval Research Laboratory (NRL). The Space Science Division at the NRL has a heritage of developing large-scale, mobile systems, using coded-aperture imaging, for long-range γ-ray detection and localization. The fast-neutron, coded-aperture imaging instrument, designed for a mobile unit (20 ft. ISO container), consists of a 32-element array of 15 cm×15 cm×15 cm liquid scintillation detectors (EJ-309) mounted behind a 12×12 pseudorandom coded aperture. The elements of the aperture are composed of 15 cm×15 cm×10 cm blocks of high-density polyethylene (HDPE). The arrangement of the aperture elements produces a shadow pattern on the detector array behind the mask. By measuring of the number of neutron counts per masked and unmasked detector, and with knowledge of the mask pattern, a source image can be deconvolved to obtain a 2-d location. The number of neutrons per detector was obtained by processing the fast signal from each PMT in flash digitizing electronics. Digital pulse shape discrimination (PSD) was performed to filter out the fast-neutron signal from the γ background. The prototype instrument was tested at an indoor facility at the NRL with a 1.8-μCi and 13-μCi 252Cf neutron/γ source at three standoff distances of 9, 15 and 26 m (maximum allowed in the facility) over a 15-min integration time. The imaging and detection capabilities of the instrument were tested by moving the source in half- and one-pixel increments across the image plane. We show a representative sample of the results obtained at one-pixel increments for a standoff distance of 9 m. The 1.8-μCi source was not detected at the 26-m standoff. In order to increase the sensitivity of the instrument, we reduced the fastneutron background by shielding the top, sides and back of the detector array with 10-cm-thick HDPE. This shielding configuration led to a reduction in the background by a factor of 1.7 and thus allowed for the detection and localization of the 1.8 μCi. The detection significance for each source at different standoff distances will be discussed.

Fragment emission from the mass-symmetric reactions 58Fe,58Ni +58Fe,58Ni at Ebeam=30 MeV/nucleon

NASA Astrophysics Data System (ADS)

Ramakrishnan, E.; Johnston, H.; Gimeno-Nogues, F.; Rowland, D. J.; Laforest, R.; Lui, Y.-W.; Ferro, S.; Vasal, S.; Yennello, S. J.

1998-04-01

The mass-symmetric reactions 58Fe,58Ni +58Fe,58Ni were studied at a beam energy of Ebeam=30 MeV/nucleon in order to investigate the isospin dependence of fragment emission. Ratios of inclusive yields of isotopic fragments from hydrogen through nitrogen were extracted as a function of laboratory angle. A moving source analysis of the data indicates that at laboratory angles around 40° the yield of intermediate mass fragments (IMF's) beyond Z=3 is predominantly from a midrapidity source. The angular dependence of the relative yields of isotopes beyond Z=3 indicates that the IMF's at more central angles originate from a source which is more neutron deficient than the source responsible for fragments emitted at forward angles. The charge distributions and kinetic energy spectra of the IMF's at various laboratory angles were well reproduced by calculations employing a quantum molecular-dynamics code followed by a statistical multifragmentation model for generating fragments. The calculations indicate that the measured IMF's originate mainly from a single source. The isotopic composition of the emitted fragments is, however, not reproduced by the same calculation. The measured isotopic and isobaric ratios indicate an emitting source that is more neutron rich in comparison to the source predicted by model calculations.

Natural Circulation Level Optimization and the Effect during ULOF Accident in the SPINNOR Reactors

NASA Astrophysics Data System (ADS)

Abdullah, Ade Gafar; Su'ud, Zaki; Kurniadi, Rizal; Kurniasih, Neny; Yulianti, Yanti

2010-12-01

Natural circulation level optimization and the effect during loss of flow accident in the 250 MWt MOX fuelled small Pb-Bi Cooled non-refueling nuclear reactors (SPINNOR) have been performed. The simulation was performed using FI-ITB safety code which has been developed in ITB. The simulation begins with steady state calculation of neutron flux, power distribution and temperature distribution across the core, hot pool and cool pool, and also steam generator. When the accident is started due to the loss of pumping power the power distribution and the temperature distribution of core, hot pool and cool pool, and steam generator change. Then the feedback reactivity calculation is conducted, followed by kinetic calculation. The process is repeated until the optimum power distribution is achieved. The results show that the SPINNOR reactor has inherent safety capability against this accident.

Identification of limiting case between DBA and SBDBA (CL break area sensitivity): A new model for the boron injection system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gonzalez Gonzalez, R.; Petruzzi, A.; D'Auria, F.

2012-07-01

Atucha-2 is a Siemens-designed PHWR reactor under construction in the Republic of Argentina. Its geometrical complexity and (e.g., oblique Control Rods, Positive Void coefficient) required a developed and validated complex three dimensional (3D) neutron kinetics (NK) coupled thermal hydraulic (TH) model. Reactor shut-down is obtained by oblique CRs and, during accidental conditions, by an emergency shut-down system (JDJ) injecting a highly concentrated boron solution (boron clouds) in the moderator tank, the boron clouds reconstruction is obtained using a CFD (CFX) code calculation. A complete LBLOCA calculation implies the application of the RELAP5-3D{sup C} system code. Within the framework of themore » third Agreement 'NA-SA - Univ. of Pisa' a new RELAP5-3D control system for the boron injection system was developed and implemented in the validated coupled RELAP5-3D/NESTLE model of the Atucha 2 NPP. The aim of this activity is to find out the limiting case (maximum break area size) for the Peak Cladding Temperature for LOCAs under fixed boundary conditions. (authors)« less

Space Radiation Transport Code Development: 3DHZETRN

NASA Technical Reports Server (NTRS)

Wilson, John W.; Slaba, Tony C.; Badavi, Francis F.; Reddell, Brandon D.; Bahadori, Amir A.

2015-01-01

The space radiation transport code, HZETRN, has been used extensively for research, vehicle design optimization, risk analysis, and related applications. One of the simplifying features of the HZETRN transport formalism is the straight-ahead approximation, wherein all particles are assumed to travel along a common axis. This reduces the governing equation to one spatial dimension allowing enormous simplification and highly efficient computational procedures to be implemented. Despite the physical simplifications, the HZETRN code is widely used for space applications and has been found to agree well with fully 3D Monte Carlo simulations in many circumstances. Recent work has focused on the development of 3D transport corrections for neutrons and light ions (Z < 2) for which the straight-ahead approximation is known to be less accurate. Within the development of 3D corrections, well-defined convergence criteria have been considered, allowing approximation errors at each stage in model development to be quantified. The present level of development assumes the neutron cross sections have an isotropic component treated within N explicit angular directions and a forward component represented by the straight-ahead approximation. The N = 1 solution refers to the straight-ahead treatment, while N = 2 represents the bi-directional model in current use for engineering design. The figure below shows neutrons, protons, and alphas for various values of N at locations in an aluminum sphere exposed to a solar particle event (SPE) spectrum. The neutron fluence converges quickly in simple geometry with N > 14 directions. The improved code, 3DHZETRN, transports neutrons, light ions, and heavy ions under space-like boundary conditions through general geometry while maintaining a high degree of computational efficiency. A brief overview of the 3D transport formalism for neutrons and light ions is given, and extensive benchmarking results with the Monte Carlo codes Geant4, FLUKA, and PHITS are provided for a variety of boundary conditions and geometries. Improvements provided by the 3D corrections are made clear in the comparisons. Developments needed to connect 3DHZETRN to vehicle design and optimization studies will be discussed. Future theoretical development will relax the forward plus isotropic interaction assumption to more general angular dependence.

Toward a New Evaluation of Neutron Standards

DOE PAGES

Carlson, Allan D.; Pronyaev, Vladimir G.; Capote, Roberto; ...

2016-02-03

Measurements related to neutron cross section standards and certain prompt neutron fission spectra are being evaluated. In addition to the standard cross sections, investigations of reference data that are not as well known as the standards are being considered. We discuss procedures and codes for performing this work. A number of libraries will use the results of this standards evaluation for new versions of their libraries. Most of these data have applications in neutron dosimetry.

Comparison between calculation and measured data on secondary neutron energy spectra by heavy ion reactions from different thick targets.

PubMed

Iwase, H; Wiegel, B; Fehrenbacher, G; Schardt, D; Nakamura, T; Niita, K; Radon, T

2005-01-01

Measured neutron energy fluences from high-energy heavy ion reactions through targets several centimeters to several hundred centimeters thick were compared with calculations made using the recently developed general-purpose particle and heavy ion transport code system (PHITS). It was confirmed that the PHITS represented neutron production by heavy ion reactions and neutron transport in thick shielding with good overall accuracy.

COMPTEL neutron response at 17 MeV

NASA Technical Reports Server (NTRS)

Oneill, Terrence J.; Ait-Ouamer, Farid; Morris, Joann; Tumer, O. Tumay; White, R. Stephen; Zych, Allen D.

1992-01-01

The Compton imaging telescope (COMPTEL) instrument of the Gamma Ray Observatory was exposed to 17 MeV d,t neutrons prior to launch. These data were analyzed and compared with Monte Carlo calculations using the MCNP(LANL) code. Energy and angular resolutions are compared and absolute efficiencies are calculated at 0 and 30 degrees incident angle. The COMPTEL neutron responses at 17 MeV and higher energies are needed to understand solar flare neutron data.

Measurement of DT and DD components in neutron spectrum with a double-crystal time-of-flight spectrometer

DOE Office of Scientific and Technical Information (OSTI.GOV)

Okada, K.; Okamoto, A.; Kitajima, S.

To investigate the deuteron and triton density ratio in core plasmas, a new methodology with measurement of tritium (DT) and deuterium (DD) neutron count rate ratio using a double-crystal time-of-flight (TOF) spectrometer is proposed. Multi-discriminator electronic circuits for the first and second detectors are used in addition to the TOF technique. The optimum arrangement of the detectors and discrimination window were examined considering the relations between the geometrical arrangement and deposited energy using a Monte Carlo Code, PHITS (Particle and Heavy Ion Transport Code System). An experiment to verify the calculations was performed using DD neutrons from an accelerator.

Nuclear Reaction Models Responsible for Simulation of Neutron-induced Soft Errors in Microelectronics

NASA Astrophysics Data System (ADS)

Watanabe, Y.; Abe, S.

2014-06-01

Terrestrial neutron-induced soft errors in MOSFETs from a 65 nm down to a 25 nm design rule are analyzed by means of multi-scale Monte Carlo simulation using the PHITS-HyENEXSS code system. Nuclear reaction models implemented in PHITS code are validated by comparisons with experimental data. From the analysis of calculated soft error rates, it is clarified that secondary He and H ions provide a major impact on soft errors with decreasing critical charge. It is also found that the high energy component from 10 MeV up to several hundreds of MeV in secondary cosmic-ray neutrons has the most significant source of soft errors regardless of design rule.

Multigroup computation of the temperature-dependent Resonance Scattering Model (RSM) and its implementation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ghrayeb, S. Z.; Ouisloumen, M.; Ougouag, A. M.

2012-07-01

A multi-group formulation for the exact neutron elastic scattering kernel is developed. This formulation is intended for implementation into a lattice physics code. The correct accounting for the crystal lattice effects influences the estimated values for the probability of neutron absorption and scattering, which in turn affect the estimation of core reactivity and burnup characteristics. A computer program has been written to test the formulation for various nuclides. Results of the multi-group code have been verified against the correct analytic scattering kernel. In both cases neutrons were started at various energies and temperatures and the corresponding scattering kernels were tallied.more » (authors)« less

Preliminary investigation of parasitic radioisotope production using the LANL IPF secondary neutron flux

NASA Astrophysics Data System (ADS)

Engle, J. W.; Kelsey, C. T.; Bach, H.; Ballard, B. D.; Fassbender, M. E.; John, K. D.; Birnbaum, E. R.; Nortier, F. M.

2012-12-01

In order to ascertain the potential for radioisotope production and material science studies using the Isotope Production Facility at Los Alamos National Lab, a two-pronged investigation has been initiated. The Monte Carlo for Neutral Particles eXtended (MCNPX) code has been used in conjunction with the CINDER 90 burnup code to predict neutron flux energy distributions as a result of routine irradiations and to estimate yields of radioisotopes of interest for hypothetical irradiation conditions. A threshold foil activation experiment is planned to study the neutron flux using measured yields of radioisotopes, quantified by HPGe gamma spectroscopy, from representative nuclear reactions with known thresholds up to 50 MeV.

Studies of neutron and proton nuclear activation in low-Earth orbit

NASA Technical Reports Server (NTRS)

Laird, C. E.

1982-01-01

The expected induced radioactivity of experimental material in low Earth orbit was studied for characteristics of activating particles such as cosmic rays, high energy Earth albedo neutrons, trapped protons, and secondary protons and neutrons. The activation cross sections for the production of long lived radioisotopes and other existing nuclear data appropriate to the study of these reactions were compiled. Computer codes which are required to calculate the expected activation of orbited materials were developed. The decreased computer code used to predict the activation of trapped protons of materials placed in the expected orbits of LDEF and Spacelab II. Techniques for unfolding the fluxes of activating particles from the measured activation of orbited materials are examined.

A User Guide to PARET/ANL

DOE Office of Scientific and Technical Information (OSTI.GOV)

Olson, A. P.; Dionne, B.; Marin-Lafleche, A.

2015-01-01

PARET was originally created in 1969 at what is now Idaho National Laboratory (INL), to analyze reactivity insertion events in research and test reactor cores cooled by light or heavy water, with fuel composed of either plates or pins. The use of PARET is also appropriate for fuel assemblies with curved fuel plates when their radii of curvatures are large with respect to the fuel plate thickness. The PARET/ANL version of the code has been developed at Argonne National Laboratory (ANL) under the sponsorship of the U.S. Department of Energy/NNSA, and has been used by the Reactor Conversion Program tomore » determine the expected transient behavior of a large number of reactors. PARET/ANL models the various fueled regions of a reactor core as channels. Each of these channels consists of a single flat fuel plate/pin (including cladding and, optionally, a gap) with water coolant on each side. In slab geometry the coolant channels for a given fuel plate are of identical dimensions (mirror symmetry), but they can be of different thickness in each channel. There can be many channels, but each channel is independent and coupled only through reactivity feedback effects to the whole core. The time-dependent differential equations that represent the system are replaced by an equivalent set of finite-difference equations in space and time, which are integrated numerically. PARET/ANL uses fundamentally the same numerical scheme as RELAP5 for the time-integration of the point-kinetics equations. The one-dimensional thermal-hydraulic model includes temperature-dependent thermal properties of the solid materials, such as heat capacity and thermal conductivity, as well as the transient heat production and heat transfer from the fuel meat to the coolant. Temperature- and pressure-dependent thermal properties of the coolant such as enthalpy, density, thermal conductivity, and viscosity are also used in determining parameters such as friction factors and heat transfer coefficients. The code first determines the steady-state solution for the initial state. Then the solution of the transient is obtained by integration in time and space. Multiple heat transfer, DNB and flow instability correlations are available. The code was originally developed to model reactors cooled by an open loop, which was adequate for rapid transients in pool-type cores. An external loop model appropriate for Miniature Neutron Source Reactors (MNSR’s) was also added to PARET/ANL to model natural circulation within the vessel, heat transfer from the vessel to pool and heat loss by evaporation from the pool. PARET/ANL also contains models for decay heat after shutdown, control rod reactivity versus time or position, time-dependent pump flow, and loss-of-flow event with flow reversal as well as logic for trips on period, power, and flow. Feedback reactivity effects from coolant density changes and temperature changes are represented by tables. Feedback reactivity from fuel heat-up (Doppler Effect) is represented by a four-term polynomial in powers of fuel temperature. Photo-neutrons produced in beryllium or in heavy water may be included in the point-kinetics equations by using additional delayed neutron groups.« less

Input data requirements for special processors in the computation system containing the VENTURE neutronics code. [LMFBR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.

1979-07-01

User input data requirements are presented for certain special processors in a nuclear reactor computation system. These processors generally read data in formatted form and generate binary interface data files. Some data processing is done to convert from the user oriented form to the interface file forms. The VENTURE diffusion theory neutronics code and other computation modules in this system use the interface data files which are generated.

Reliability assessment of MVP-BURN and JENDL-4.0 related to nuclear transmutation of light platinum group elements

NASA Astrophysics Data System (ADS)

Terashima, Atsunori; Nilsson, Mikael; Ozawa, Masaki; Chiba, Satoshi

2017-09-01

The Aprés ORIENT research program, as a concept of advanced nuclear fuel cycle, was initiated in FY2011 aiming at creating stable, highly-valuable elements by nuclear transmutation from ↓ssion products. In order to simulate creation of such elements by (n, γ) reaction succeeded by β- decay in reactors, a continuous-energy Monte Carlo burnup calculation code MVP-BURN was employed. Then, it is one of the most important tasks to con↓rm the reliability of MVP-BURN code and evaluated neutron cross section library. In this study, both an experiment of neutron activation analysis in TRIGA Mark I reactor at University of California, Irvine and the corresponding burnup calculation using MVP-BURN code were performed for validation of the simulation on transmutation of light platinum group elements. Especially, some neutron capture reactions such as 102Ru(n, γ)103Ru, 104Ru(n, γ)105Ru, and 108Pd(n, γ)109Pd were dealt with in this study. From a comparison between the calculation (C) and the experiment (E) about 102Ru(n, γ)103Ru, the deviation (C/E-1) was signi↓cantly large. Then, it is strongly suspected that not MVP-BURN code but the neutron capture cross section of 102Ru belonging to JENDL-4.0 used in this simulation have made the big di↑erence as (C/E-1) >20%.

A Monte Carlo simulation and setup optimization of output efficiency to PGNAA thermal neutron using 252Cf neutrons

NASA Astrophysics Data System (ADS)

Zhang, Jin-Zhao; Tuo, Xian-Guo

2014-07-01

We present the design and optimization of a prompt γ-ray neutron activation analysis (PGNAA) thermal neutron output setup based on Monte Carlo simulations using MCNP5 computer code. In these simulations, the moderator materials, reflective materials, and structure of the PGNAA 252Cf neutrons of thermal neutron output setup are optimized. The simulation results reveal that the thin layer paraffin and the thick layer of heavy water moderating effect work best for the 252Cf neutron spectrum. Our new design shows a significantly improved performance of the thermal neutron flux and flux rate, that are increased by 3.02 times and 3.27 times, respectively, compared with the conventional neutron source design.

Clustering of water molecules in ultramicroporous carbon: In-situ small-angle neutron scattering

DOE PAGES

Bahadur, Jitendra; Contescu, Cristian I.; Rai, Durgesh K.; ...

2016-10-19

The adsorption of water is central to most of the applications of microporous carbon as adsorbent material. We report early kinetics of water adsorption in the microporous carbon using in-situ small-angle neutron scattering. It is observed that adsorption of water occurs via cluster formation of molecules. Interestingly, the cluster size remains constant throughout the adsorption process whereas number density of clusters increases with time. The role of surface chemistry of microporous carbon on the early kinetics of adsorption process was also investigated. Lastly, the present study provides direct experimental evidence for cluster assisted adsorption of water molecules in microporous carbonmore » (Do-Do model).« less

Measurement of the argon-38(n,2n)argon-37 and calcium- 40(n,alpha)argon-37 cross sections, and National Ignition Facility concrete activation using the rotating target neutron source. The design of an experiment to measure the beryllium-9(n,gamma)beryllium-10 cross section at 14 MeV

NASA Astrophysics Data System (ADS)

Belian, Anthony Paul

The Rotating Target Neutron Source (RTNS) was used in experiments to measure neutron induced cross sections at 14 MeV, and the activation properties of a specific mix of concrete. The RTNS is an accelerator based DT fusion neutron source located at the University of California, Berkeley. Two of the experiments performed for this thesis were specifically of interest for the construction and operation of the National Ignition Facility (NIF), they were the 38Ar(n,2n)37Ar cross section measurement, and the concrete activation measurement. The NIF is a large multi-beam laser facility that will study the effects of age on the nation's stockpile of nuclear weapons. The NIF, when fully operational, will focus the energy of 192 Neodymium glass lasers onto a 1 mm diameter pellet filled with deuterium and tritium fuel. This pellet is compressed by the laser energy giving some of the individual atoms of deuterium and tritium enough kinetic energy to overcome the coulomb barrier and fuse. The energy output from these pellet implosions will be in the range of tens of mega-joules (MJ). The 38Ar(n,2n)37Ar reaction will be useful to NIF scientists to measure important parameters such as target energy yield and areal density. In order to make these measurements precise, an accurate 38Ar(n,2n)37Ar cross section was necessary. The cross sections measured were: 74.9 +/- 3.8 millibarns (mb) at 13.3 +/- 0.01 MeV, 89.2 +/- 4.0 mb at 14.0 +/- 0.03 MeV, and 123.57 +/- 6.4 mb at 15.0 +/- 0.06 MeV. With anticipated energy yields in the tens of mega-joules per pellet implosion, the number of neutrons released is in the range of 1019 to 1020 neutrons per implosion. With such a large number of neutrons, minimizing the activation of the surrounding structure is very much of interest for the sake of personnel radiation safety. To benchmark the computer codes used to calculate the anticipated neutron activation of target bay concrete, samples were irradiated at the RTNS. Dose rates from each sample were recorded as a function of time after irradiation. These dose rates were compared to those calculated using the Monte Carlo code TART and the activation code ACAB. It was found that 95.8% of the comparisons agreed within the experimental uncertainty. The 40Ca(n,α)37Ar reaction was of interest for the detection of clandestine underground nuclear detonations. Since calcium is naturally abundant in the earth's crust, and since 37Ar is an inert gas and is not found naturally, the 40Ca(n, α) 37Ar reaction is a good candidate for detecting a nuclear detonation. An accurate cross section is needed to estimate the yield of the nuclear device. The average cross sections measured were: 175.6 +/- 9.2 millibarns (mb) at 13.2 +/- 0.6 MeV and 122.1 +/- 4.6 mb at 15.2 +/- 0.12 MeV. One of the current NIF pellet designs uses beryllium as the ablation layer, and the target positioner will be made of a beryllium/copper alloy. The reaction product, 10Be, from the 9Be(n,γ) 10Be reaction will be generated, although probably in very small quantities, during the lifetime of the NIF. This cross section has not been measured at 14 MeV, but should be measured to estimate the amount of 10Be produced at the NIF.

Ion absorption of the high harmonic fast wave in the National Spherical Torus Experiment

NASA Astrophysics Data System (ADS)

Rosenberg, Adam Lewis

Ion absorption of the high harmonic fast wave in a spherical torus is of critical importance to assessing the viability of the wave as a means of heating and driving current. Analysis of recent NSTX shots has revealed that under some conditions when neutral beam and RF power are injected into the plasma simultaneously, a fast ion population with energy above the beam injection energy is sustained by the wave. In agreement with modeling, these experiments find the RF-induced fast ion tail strength and neutron rate at lower B-fields to be less enhanced, likely due to a larger β profile, which promotes greater off-axis absorption where the fast ion population is small. Ion loss codes find the increased loss fraction with decreased B insufficient to account for the changes in tail strength, providing further evidence that this is an RF interaction effect. Though greater ion absorption is predicted with lower k∥, surprisingly little variation in the tail was observed, along with a neutron rate enhancement with higher k∥. Data from the neutral particle analyzer, neutron detectors, x-ray crystal spectrometer, and Thomson scattering is presented, along with results from the TRANSP transport analysis code, ray-tracing codes HPRT and CURRAY, full-wave code and AORSA, quasilinear code CQL3D, and ion loss codes EIGOL and CONBEAM.

Three-dimensional modeling of the neutron spectrum to infer plasma conditions in cryogenic inertial confinement fusion implosions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Weilacher, F.; Radha, P. B.; Forrest, C.

Neutron-based diagnostics are typically used to infer compressed core conditions such as areal density and ion temperature in deuterium–tritium (D–T) inertial confinement fusion (ICF) implosions. Asymmetries in the observed neutron-related quantities are important to understanding failure modes in these implosions. Neutrons from fusion reactions and their subsequent interactions including elastic scattering and neutron-induced deuteron breakup reactions are tracked to create spectra. Here, it is shown that background subtraction is important for inferring areal density from backscattered neutrons and is less important for the forward-scattered neutrons. A three-dimensional hydrodynamic simulation of a cryogenic implosion on the OMEGA Laser System [T. R.more » Boehly et al., Opt. Commun. 133, 495 (1997)] using the hydrodynamic code HYDRA [M. M. Marinak et al., Phys. Plasmas 8, 2275 (2001)] is post-processed using the tracking code IRIS3D. It is shown that different parts of the neutron spectrum from the view can be mapped into different regions of the implosion, enabling an inference of an areal-density map. It is also shown that the average areal-density and an areal-density map of the compressed target can be reconstructed with a finite number of detectors placed around the target chamber. Ion temperatures are inferred from the width of the D–D and D–T fusion neutron spectra. Backgrounds can significantly alter the inferred ion temperatures from the D–D reaction, whereas they insignificantly influence the inferred D–T ion temperatures for the areal densities typical of OMEGA implosions. Asymmetries resulting in fluid flow in the core are shown to influence the absolute inferred ion temperatures from both reactions, although relative inferred values continue to reflect the underlying asymmetry pattern. The work presented here is part of the wide range of the first set of studies performed with IRIS3D. Finally, this code will continue to be used for post-processing detailed hydrodynamic simulations and interpreting observed neutron spectra in ICF implosions.« less

Three-dimensional modeling of the neutron spectrum to infer plasma conditions in cryogenic inertial confinement fusion implosions

DOE PAGES

Weilacher, F.; Radha, P. B.; Forrest, C.

2018-04-26

Neutron-based diagnostics are typically used to infer compressed core conditions such as areal density and ion temperature in deuterium–tritium (D–T) inertial confinement fusion (ICF) implosions. Asymmetries in the observed neutron-related quantities are important to understanding failure modes in these implosions. Neutrons from fusion reactions and their subsequent interactions including elastic scattering and neutron-induced deuteron breakup reactions are tracked to create spectra. Here, it is shown that background subtraction is important for inferring areal density from backscattered neutrons and is less important for the forward-scattered neutrons. A three-dimensional hydrodynamic simulation of a cryogenic implosion on the OMEGA Laser System [T. R.more » Boehly et al., Opt. Commun. 133, 495 (1997)] using the hydrodynamic code HYDRA [M. M. Marinak et al., Phys. Plasmas 8, 2275 (2001)] is post-processed using the tracking code IRIS3D. It is shown that different parts of the neutron spectrum from the view can be mapped into different regions of the implosion, enabling an inference of an areal-density map. It is also shown that the average areal-density and an areal-density map of the compressed target can be reconstructed with a finite number of detectors placed around the target chamber. Ion temperatures are inferred from the width of the D–D and D–T fusion neutron spectra. Backgrounds can significantly alter the inferred ion temperatures from the D–D reaction, whereas they insignificantly influence the inferred D–T ion temperatures for the areal densities typical of OMEGA implosions. Asymmetries resulting in fluid flow in the core are shown to influence the absolute inferred ion temperatures from both reactions, although relative inferred values continue to reflect the underlying asymmetry pattern. The work presented here is part of the wide range of the first set of studies performed with IRIS3D. Finally, this code will continue to be used for post-processing detailed hydrodynamic simulations and interpreting observed neutron spectra in ICF implosions.« less

Simulations of GCR interactions within planetary bodies using GEANT4

NASA Astrophysics Data System (ADS)

Mesick, K.; Feldman, W. C.; Stonehill, L. C.; Coupland, D. D. S.

2017-12-01

On planetary bodies with little to no atmosphere, Galactic Cosmic Rays (GCRs) can hit the body and produce neutrons primarily through nuclear spallation within the top few meters of the surfaces. These neutrons undergo further nuclear interactions with elements near the planetary surface and some will escape the surface and can be detected by landed or orbiting neutron radiation detector instruments. The neutron leakage signal at fast neutron energies provides a measure of average atomic mass of the near-surface material and in the epithermal and thermal energy ranges is highly sensitive to the presence of hydrogen. Gamma-rays can also escape the surface, produced at characteristic energies depending on surface composition, and can be detected by gamma-ray instruments. The intra-nuclear cascade (INC) that occurs when high-energy GCRs interact with elements within a planetary surface to produce the leakage neutron and gamma-ray signals is highly complex, and therefore Monte Carlo based radiation transport simulations are commonly used for predicting and interpreting measurements from planetary neutron and gamma-ray spectroscopy instruments. In the past, the simulation code that has been widely used for this type of analysis is MCNPX [1], which was benchmarked against data from the Lunar Neutron Probe Experiment (LPNE) on Apollo 17 [2]. In this work, we consider the validity of the radiation transport code GEANT4 [3], another widely used but open-source code, by benchmarking simulated predictions of the LPNE experiment to the Apollo 17 data. We consider the impact of different physics model options on the results, and show which models best describe the INC based on agreement with the Apollo 17 data. The success of this validation then gives us confidence in using GEANT4 to simulate GCR-induced neutron leakage signals on Mars in relevance to a re-analysis of Mars Odyssey Neutron Spectrometer data. References [1] D.B. Pelowitz, Los Alamos National Laboratory, LA-CP-05-0369, 2005. [2] G.W. McKinney et al, Journal of Geophysics Research, 111, E06004, 2006. [3] S. Agostinelli et al, Nuclear Instrumentation and Methods A, 506, 2003.

High-energy neutron depth-dose distribution experiment.

PubMed

Ferenci, M S; Hertel, N E

2003-01-01

A unique set of high-energy neutron depth-dose benchmark experiments were performed at the Los Alamos Neutron Science Center/Weapons Neutron Research (LANSCE/WNR) complex. The experiments consisted of filtered neutron beams with energies up to 800 MeV impinging on a 30 x 30 x 30 cm3 liquid, tissue-equivalent phantom. The absorbed dose was measured in the phantom at various depths with tissue-equivalent ion chambers. This experiment is intended to serve as a benchmark experiment for the testing of high-energy radiation transport codes for the international radiation protection community.

Monte Carlo analysis of a time-dependent neutron and secondary gamma-ray integral experiment on a thick concrete and steel shield

NASA Astrophysics Data System (ADS)

Cramer, S. N.; Roussin, R. W.

1981-11-01

A Monte Carlo analysis of a time-dependent neutron and secondary gamma-ray integral experiment on a thick concrete and steel shield is presented. The energy range covered in the analysis is 15-2 MeV for neutron source energies. The multigroup MORSE code was used with the VITAMIN C 171-36 neutron-gamma-ray cross-section data set. Both neutron and gamma-ray count rates and unfolded energy spectra are presented and compared, with good general agreement, with experimental results.

Neutron production by cosmic-ray muons in various materials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Manukovsky, K. V.; Ryazhskaya, O. G.; Sobolevsky, N. M.

The results obtained by studying the background of neutrons produced by cosmic-raymuons in underground experimental facilities intended for rare-event searches and in surrounding rock are presented. The types of this rock may include granite, sedimentary rock, gypsum, and rock salt. Neutron production and transfer were simulated using the Geant4 and SHIELD transport codes. These codes were tuned via a comparison of the results of calculations with experimental data—in particular, with data of the Artemovsk research station of the Institute for Nuclear Research (INR, Moscow, Russia)—as well as via an intercomparison of results of calculations with the Geant4 and SHIELD codes.more » It turns out that the atomic-number dependence of the production and yield of neutrons has an irregular character and does not allow a description in terms of a universal function of the atomic number. The parameters of this dependence are different for two groups of nuclei—nuclei consisting of alpha particles and all of the remaining nuclei. Moreover, there are manifest exceptions from a power-law dependence—for example, argon. This may entail important consequences both for the existing underground experimental facilities and for those under construction. Investigation of cosmic-ray-induced neutron production in various materials is of paramount importance for the interpretation of experiments conducted at large depths under the Earth’s surface.« less

Measurement of Thick Target Neutron Yields at 0-Degree Bombarded With 140-MeV, 250-MeV And 350-MeV Protons

DOE Office of Scientific and Technical Information (OSTI.GOV)

Iwamoto, Yosuke; /JAERI, Kyoto; Taniguchi, Shingo

Neutron energy spectra at 0{sup o} produced from stopping-length graphite, aluminum, iron and lead targets bombarded with 140, 250 and 350 MeV protons were measured at the neutron TOF course in RCNP of Osaka University. The neutron energy spectra were obtained by using the time-of-flight technique in the energy range from 10 MeV to incident proton energy. To compare the experimental results, Monte Carlo calculations with the PHITS and MCNPX codes were performed using the JENDL-HE and the LA150 evaluated nuclear data files, the ISOBAR model implemented in PHITS, and the LAHET code in MCNPX. It was found that thesemore » calculated results at 0{sup o} generally agreed with the experimental results in the energy range above 20 MeV except for graphite at 250 and 350 MeV.« less

A multi-detector neutron spectrometer with nearly isotropic response for environmental and workplace monitoring

NASA Astrophysics Data System (ADS)

Gómez-Ros, J. M.; Bedogni, R.; Moraleda, M.; Delgado, A.; Romero, A.; Esposito, A.

2010-01-01

This communication describes an improved design for a neutron spectrometer consisting of 6Li thermoluminescent dosemeters located at selected positions within a single moderating polyethylene sphere. The spatial arrangement of the dosemeters has been designed using the MCNPX Monte Carlo code to calculate the response matrix for 56 log-equidistant energies from 10 -9 to 100 MeV, looking for a configuration that permits to obtain a nearly isotropic response for neutrons in the energy range from thermal to 20 MeV. The feasibility of the proposed spectrometer and the isotropy of its response have been evaluated by simulating exposures to different reference and workplace neutron fields. The FRUIT code has been used for unfolding purposes. The results of the simulations as well as the experimental tests confirm the suitability of the prototype for environmental and workplace monitoring applications.

Shutdown Dose Rate Analysis for the long-pulse D-D Operation Phase in KSTAR

NASA Astrophysics Data System (ADS)

Park, Jin Hun; Han, Jung-Hoon; Kim, D. H.; Joo, K. S.; Hwang, Y. S.

2017-09-01

KSTAR is a medium size fully superconducting tokamak. The deuterium-deuterium (D-D) reaction in the KSTAR tokamak generates neutrons with a peak yield of 3.5x1016 per second through a pulse operation of 100 seconds. The effect of neutron generation from full D-D high power KSTAR operation mode to the machine, such as activation, shutdown dose rate, and nuclear heating, are estimated for an assurance of safety during operation, maintenance, and machine upgrade. The nuclear heating of the in-vessel components, and neutron activation of the surrounding materials have been investigated. The dose rates during operation and after shutdown of KSTAR have been calculated by a 3D CAD model of KSTAR with the Monte Carlo code MCNP5 (neutron flux and decay photon), the inventory code FISPACT (activation and decay photon) and the FENDL 2.1 nuclear data library.

Improved Convergence Rate of Multi-Group Scattering Moment Tallies for Monte Carlo Neutron Transport Codes

NASA Astrophysics Data System (ADS)

Nelson, Adam

Multi-group scattering moment matrices are critical to the solution of the multi-group form of the neutron transport equation, as they are responsible for describing the change in direction and energy of neutrons. These matrices, however, are difficult to correctly calculate from the measured nuclear data with both deterministic and stochastic methods. Calculating these parameters when using deterministic methods requires a set of assumptions which do not hold true in all conditions. These quantities can be calculated accurately with stochastic methods, however doing so is computationally expensive due to the poor efficiency of tallying scattering moment matrices. This work presents an improved method of obtaining multi-group scattering moment matrices from a Monte Carlo neutron transport code. This improved method of tallying the scattering moment matrices is based on recognizing that all of the outgoing particle information is known a priori and can be taken advantage of to increase the tallying efficiency (therefore reducing the uncertainty) of the stochastically integrated tallies. In this scheme, the complete outgoing probability distribution is tallied, supplying every one of the scattering moment matrices elements with its share of data. In addition to reducing the uncertainty, this method allows for the use of a track-length estimation process potentially offering even further improvement to the tallying efficiency. Unfortunately, to produce the needed distributions, the probability functions themselves must undergo an integration over the outgoing energy and scattering angle dimensions. This integration is too costly to perform during the Monte Carlo simulation itself and therefore must be performed in advance by way of a pre-processing code. The new method increases the information obtained from tally events and therefore has a significantly higher efficiency than the currently used techniques. The improved method has been implemented in a code system containing a new pre-processor code, NDPP, and a Monte Carlo neutron transport code, OpenMC. This method is then tested in a pin cell problem and a larger problem designed to accentuate the importance of scattering moment matrices. These tests show that accuracy was retained while the figure-of-merit for generating scattering moment matrices and fission energy spectra was significantly improved.

Designing an extended energy range single-sphere multi-detector neutron spectrometer

NASA Astrophysics Data System (ADS)

Gómez-Ros, J. M.; Bedogni, R.; Moraleda, M.; Esposito, A.; Pola, A.; Introini, M. V.; Mazzitelli, G.; Quintieri, L.; Buonomo, B.

2012-06-01

This communication describes the design specifications for a neutron spectrometer consisting of 31 thermal neutron detectors, namely Dysprosium activation foils, embedded in a 25 cm diameter polyethylene sphere which includes a 1 cm thick lead shell insert that degrades the energy of neutrons through (n,xn) reactions, thus allowing to extension of the energy range of the response up to hundreds of MeV neutrons. The new spectrometer, called SP2 (SPherical SPectrometer), relies on the same detection mechanism as that of the Bonner Sphere Spectrometer, but with the advantage of determining the whole neutron spectrum in a single exposure. The Monte Carlo transport code MCNPX was used to design the spectrometer in terms of sphere diameter, number and position of the detectors, position and thickness of the lead shell, as well as to obtain the response matrix for the final configuration. This work focuses on evaluating the spectrometric capabilities of the SP2 design by simulating the exposure of SP2 in neutron fields representing different irradiation conditions (test spectra). The simulated SP2 readings were then unfolded with the FRUIT unfolding code, in the absence of detailed pre-information, and the unfolded spectra were compared with the known test spectra. The results are satisfactory and allowed approving the production of a prototypal spectrometer.

Benchmarking the MCNP code for Monte Carlo modelling of an in vivo neutron activation analysis system.

PubMed

Natto, S A; Lewis, D G; Ryde, S J

1998-01-01

The Monte Carlo computer code MCNP (version 4A) has been used to develop a personal computer-based model of the Swansea in vivo neutron activation analysis (IVNAA) system. The model included specification of the neutron source (252Cf), collimators, reflectors and shielding. The MCNP model was 'benchmarked' against fast neutron and thermal neutron fluence data obtained experimentally from the IVNAA system. The Swansea system allows two irradiation geometries using 'short' and 'long' collimators, which provide alternative dose rates for IVNAA. The data presented here relate to the short collimator, although results of similar accuracy were obtained using the long collimator. The fast neutron fluence was measured in air at a series of depths inside the collimator. The measurements agreed with the MCNP simulation within the statistical uncertainty (5-10%) of the calculations. The thermal neutron fluence was measured and calculated inside the cuboidal water phantom. The depth of maximum thermal fluence was 3.2 cm (measured) and 3.0 cm (calculated). The width of the 50% thermal fluence level across the phantom at its mid-depth was found to be the same by both MCNP and experiment. This benchmarking exercise has given us a high degree of confidence in MCNP as a tool for the design of IVNAA systems.

AMPX: a modular code system for generating coupled multigroup neutron-gamma libraries from ENDF/B

DOE Office of Scientific and Technical Information (OSTI.GOV)

Greene, N.M.; Lucius, J.L.; Petrie, L.M.

1976-03-01

AMPX is a modular system for producing coupled multigroup neutron-gamma cross section sets. Basic neutron and gamma cross-section data for AMPX are obtained from ENDF/B libraries. Most commonly used operations required to generate and collapse multigroup cross-section sets are provided in the system. AMPX is flexibly dimensioned; neutron group structures, and gamma group structures, and expansion orders to represent anisotropic processes are all arbitrary and limited only by available computer core and budget. The basic processes provided will (1) generate multigroup neutron cross sections; (2) generate multigroup gamma cross sections; (3) generate gamma yields for gamma-producing neutron interactions; (4) combinemore » neutron cross sections, gamma cross sections, and gamma yields into final ''coupled sets''; (5) perform one-dimensional discrete ordinates transport or diffusion theory calculations for neutrons and gammas and, on option, collapse the cross sections to a broad-group structure, using the one-dimensional results as weighting functions; (6) plot cross sections, on option, to facilitate the ''evaluation'' of a particular multigroup set of data; (7) update and maintain multigroup cross section libraries in such a manner as to make it not only easy to combine new data with previously processed data but also to do it in a single pass on the computer; and (8) output multigroup cross sections in convenient formats for other codes. (auth)« less

Slow neutron total cross-section, transmission and reflection calculation for poly- and mono-NaCl and PbF2 crystals

NASA Astrophysics Data System (ADS)

Mansy, Muhammad S.; Adib, M.; Habib, N.; Bashter, I. I.; Morcos, H. N.; El-Mesiry, M. S.

2016-10-01

A detailed study about the calculation of total neutron cross-section, transmission and reflection from crystalline materials was performed. The developed computer code is approved to be sufficient for the required calculations, also an excellent agreement has been shown when comparing the code results with the other calculated and measured values. The optimal monochromator and filter parameters were discussed in terms of crystal orientation, mosaic spread, and thickness. Calculations show that 30 cm thick of PbF2 poly-crystal is an excellent cold neutron filter producing neutron wavelengths longer than 0.66 nm needed for the investigation of magnetic structure experiments. While mono-crystal filter PbF2 cut along its (1 1 1), having mosaic spread (η = 0.5°) and thickness 10 cm can only transmit thermal neutrons of the desired wavelengths and suppress epithermal and γ-rays forming unwanted background, when it is cooled to liquid nitrogen temperature. NaCl (2 0 0) and PbF2 (1 1 1) monochromator crystals having mosaic spread (η = 0.5°) and thickness 10 mm shows high neutron reflectivity for neutron wavelengths (λ = 0.114 nm and λ = 0.43 nm) when they used as a thermal and cold neutron monochromators respectively with very low contamination from higher order reflections.

Light transport feature for SCINFUL.

PubMed

Etaati, G R; Ghal-Eh, N

2008-03-01

An extended version of the scintillator response function prediction code SCINFUL has been developed by incorporating PHOTRACK, a Monte Carlo light transport code. Comparisons of calculated and experimental results for organic scintillators exposed to neutrons show that the extended code improves the predictive capability of SCINFUL.

Test case for VVER-1000 complex modeling using MCU and ATHLET

NASA Astrophysics Data System (ADS)

Bahdanovich, R. B.; Bogdanova, E. V.; Gamtsemlidze, I. D.; Nikonov, S. P.; Tikhomirov, G. V.

2017-01-01

The correct modeling of processes occurring in the fuel core of the reactor is very important. In the design and operation of nuclear reactors it is necessary to cover the entire range of reactor physics. Very often the calculations are carried out within the framework of only one domain, for example, in the framework of structural analysis, neutronics (NT) or thermal hydraulics (TH). However, this is not always correct, as the impact of related physical processes occurring simultaneously, could be significant. Therefore it is recommended to spend the coupled calculations. The paper provides test case for the coupled neutronics-thermal hydraulics calculation of VVER-1000 using the precise neutron code MCU and system engineering code ATHLET. The model is based on the fuel assembly (type 2M). Test case for calculation of power distribution, fuel and coolant temperature, coolant density, etc. has been developed. It is assumed that the test case will be used for simulation of VVER-1000 reactor and in the calculation using other programs, for example, for codes cross-verification. The detailed description of the codes (MCU, ATHLET), geometry and material composition of the model and an iterative calculation scheme is given in the paper. Script in PERL language was written to couple the codes.

Preliminary results of Malaysian nuclear agency plasma focus (MNA-PF) as a slow focus mode device for argon and deuterium filling gas in correlation with Lee model code

NASA Astrophysics Data System (ADS)

Zin, M. F. M.; Baijan, A. H.; Damideh, V.; Hashim, S. A.; Sabri, R. M.

2017-03-01

In this work, preliminary results of MNA-PF device as a Slow Focus Mode device are presented. Four different kinds of Rogowski Coils which have been designed and constructed for dI/dt signals measurements show that response frequency of Rogowski Coil can affect signal time resolution and delay which can change the discharge circuit inductance. Experimental results for 10 to 20 mbar Deuterium and 0.5 mbar to 6 mbar Argon which are captured by 630 MHz Rogowski coil in correlation with Lee Model Code are presented. Proper current fitting using Lee Model Code shows that the speed factor for MNA-PF device working with 13 mbar Deuterium is 30 kA/cm.torr1/2 at 14 kV which indicates that the device is operating at slow focus mode. Model parameters fm and fmr predicted by Lee Model Code during current fitting for 13 mbar Deuterium at 14kV were 0.025 and 0.31 respectively. Microspec-4 Neutron Detector was used to obtain the dose rate which was found to be maximum at 4.78 uSv/hr and also the maximum neutron yield calculated from Lee Model Code is 7.5E+03 neutron per shot.

Current and anticipated uses of thermal hydraulic codes at the Japan Atomic Energy Research Institute

DOE Office of Scientific and Technical Information (OSTI.GOV)

Akimoto, Hajime; Kukita; Ohnuki, Akira

1997-07-01

The Japan Atomic Energy Research Institute (JAERI) is conducting several research programs related to thermal-hydraulic and neutronic behavior of light water reactors (LWRs). These include LWR safety research projects, which are conducted in accordance with the Nuclear Safety Commission`s research plan, and reactor engineering projects for the development of innovative reactor designs or core/fuel designs. Thermal-hydraulic and neutronic codes are used for various purposes including experimental analysis, nuclear power plant (NPP) safety analysis, and design assessment.

Hardening neutron spectrum for advanced actinide transmutation experiments in the ATR.

PubMed

Chang, G S; Ambrosek, R G

2005-01-01

The most effective method for transmuting long-lived isotopes contained in spent nuclear fuel into shorter-lived fission products is in a fast neutron spectrum reactor. In the absence of a fast test reactor in the United States, initial irradiation testing of candidate fuels can be performed in a thermal test reactor that has been modified to produce a test region with a hardened neutron spectrum. Such a test facility, with a spectrum similar but somewhat softer than that of the liquid-metal fast breeder reactor (LMFBR), has been constructed in the INEEL's Advanced Test Reactor (ATR). The radial fission power distribution of the actinide fuel pin, which is an important parameter in fission gas release modelling, needs to be accurately predicted and the hardened neutron spectrum in the ATR and the LMFBR fast neutron spectrum is compared. The comparison analyses in this study are performed using MCWO, a well-developed tool that couples the Monte Carlo transport code MCNP with the isotope depletion and build-up code ORIGEN-2. MCWO analysis yields time-dependent and neutron-spectrum-dependent minor actinide and Pu concentrations and detailed radial fission power profile calculations for a typical fast reactor (LMFBR) neutron spectrum and the hardened neutron spectrum test region in the ATR. The MCWO-calculated results indicate that the cadmium basket used in the advanced fuel test assembly in the ATR can effectively depress the linear heat generation rate in the experimental fuels and harden the neutron spectrum in the test region.

Benchmark test of neutron transport calculations: indium, nickel, gold, europium, and cobalt activation with and without energy moderated fission neutrons by iron simulating the Hiroshima atomic bomb casing.

PubMed

Iwatani, K; Hoshi, M; Shizuma, K; Hiraoka, M; Hayakawa, N; Oka, T; Hasai, H

1994-10-01

A benchmark test of the Monte Carlo neutron and photon transport code system (MCNP) was performed using a bare- and energy-moderated 252Cf fission neutron source which was obtained by transmission through 10-cm-thick iron. An iron plate was used to simulate the effect of the Hiroshima atomic bomb casing. This test includes the activation of indium and nickel for fast neutrons and gold, europium, and cobalt for thermal and epithermal neutrons, which were inserted in the moderators. The latter two activations are also to validate 152Eu and 60Co activity data obtained from the atomic bomb-exposed specimens collected at Hiroshima and Nagasaki, Japan. The neutron moderators used were Lucite and Nylon 6 and the total thickness of each moderator was 60 cm or 65 cm. Measured activity data (reaction yield) of the neutron-irradiated detectors in these moderators decreased to about 1/1,000th or 1/10,000th, which corresponds to about 1,500 m ground distance from the hypocenter in Hiroshima. For all of the indium, nickel, and gold activity data, the measured and calculated values agreed within 25%, and the corresponding values for europium and cobalt were within 40%. From this study, the MCNP code was found to be accurate enough for the bare- and energy-moderated 252Cf neutron activation calculations of these elements using moderators containing hydrogen, carbon, nitrogen, and oxygen.

The origin of neutron biological effectiveness as a function of energy.

PubMed

Baiocco, G; Barbieri, S; Babini, G; Morini, J; Alloni, D; Friedland, W; Kundrát, P; Schmitt, E; Puchalska, M; Sihver, L; Ottolenghi, A

2016-09-22

The understanding of the impact of radiation quality in early and late responses of biological targets to ionizing radiation exposure necessarily grounds on the results of mechanistic studies starting from physical interactions. This is particularly true when, already at the physical stage, the radiation field is mixed, as it is the case for neutron exposure. Neutron Relative Biological Effectiveness (RBE) is energy dependent, maximal for energies ~1 MeV, varying significantly among different experiments. The aim of this work is to shed light on neutron biological effectiveness as a function of field characteristics, with a comprehensive modeling approach: this brings together transport calculations of neutrons through matter (with the code PHITS) and the predictive power of the biophysical track structure code PARTRAC in terms of DNA damage evaluation. Two different energy dependent neutron RBE models are proposed: the first is phenomenological and based only on the characterization of linear energy transfer on a microscopic scale; the second is purely ab-initio and based on the induction of complex DNA damage. Results for the two models are compared and found in good qualitative agreement with current standards for radiation protection factors, which are agreed upon on the basis of RBE data.

The origin of neutron biological effectiveness as a function of energy

NASA Astrophysics Data System (ADS)

Baiocco, G.; Barbieri, S.; Babini, G.; Morini, J.; Alloni, D.; Friedland, W.; Kundrát, P.; Schmitt, E.; Puchalska, M.; Sihver, L.; Ottolenghi, A.

2016-09-01

The understanding of the impact of radiation quality in early and late responses of biological targets to ionizing radiation exposure necessarily grounds on the results of mechanistic studies starting from physical interactions. This is particularly true when, already at the physical stage, the radiation field is mixed, as it is the case for neutron exposure. Neutron Relative Biological Effectiveness (RBE) is energy dependent, maximal for energies ~1 MeV, varying significantly among different experiments. The aim of this work is to shed light on neutron biological effectiveness as a function of field characteristics, with a comprehensive modeling approach: this brings together transport calculations of neutrons through matter (with the code PHITS) and the predictive power of the biophysical track structure code PARTRAC in terms of DNA damage evaluation. Two different energy dependent neutron RBE models are proposed: the first is phenomenological and based only on the characterization of linear energy transfer on a microscopic scale; the second is purely ab-initio and based on the induction of complex DNA damage. Results for the two models are compared and found in good qualitative agreement with current standards for radiation protection factors, which are agreed upon on the basis of RBE data.

The origin of neutron biological effectiveness as a function of energy

PubMed Central

Baiocco, G.; Barbieri, S.; Babini, G.; Morini, J.; Alloni, D.; Friedland, W.; Kundrát, P.; Schmitt, E.; Puchalska, M.; Sihver, L.; Ottolenghi, A.

2016-01-01

The understanding of the impact of radiation quality in early and late responses of biological targets to ionizing radiation exposure necessarily grounds on the results of mechanistic studies starting from physical interactions. This is particularly true when, already at the physical stage, the radiation field is mixed, as it is the case for neutron exposure. Neutron Relative Biological Effectiveness (RBE) is energy dependent, maximal for energies ~1 MeV, varying significantly among different experiments. The aim of this work is to shed light on neutron biological effectiveness as a function of field characteristics, with a comprehensive modeling approach: this brings together transport calculations of neutrons through matter (with the code PHITS) and the predictive power of the biophysical track structure code PARTRAC in terms of DNA damage evaluation. Two different energy dependent neutron RBE models are proposed: the first is phenomenological and based only on the characterization of linear energy transfer on a microscopic scale; the second is purely ab-initio and based on the induction of complex DNA damage. Results for the two models are compared and found in good qualitative agreement with current standards for radiation protection factors, which are agreed upon on the basis of RBE data. PMID:27654349

Applicability of a Bonner Shere technique for pulsed neutron in 120 GeV proton facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sanami, T.; Hagiwara, M.; Iwase, H.

2008-02-01

The data on neutron spectra and intensity behind shielding are important for radiation safety design of high-energy accelerators since neutrons are capable of penetrating thick shielding and activating materials. Corresponding particle transport codes--that involve physics models of neutron and other particle production, transportation, and interaction--have been developed and used world-wide [1-8]. The results of these codes have been ensured through plenty of comparisons with experimental results taken in simple geometries. For neutron generation and transport, several related experiments have been performed to measure neutron spectra, attenuation length and reaction rates behind shielding walls of various thicknesses and materials in energymore » range up to several hundred of MeV [9-11]. The data have been used to benchmark--and modify if needed--the simulation modes and parameters in the codes, as well as the reference data for radiation safety design. To obtain such kind of data above several hundred of MeV, Japan-Fermi National Accelerator Laboratory (FNAL) collaboration for shielding experiments has been started in 2007, based on suggestion from the specialist meeting of shielding, Shielding Aspects of Target, Irradiation Facilities (SATIF), because of very limited data available in high-energy region (see, for example, [12]). As a part of this shielding experiment, a set of Bonner sphere (BS) was tested at the antiproton production target facility (pbar target station) at FNAL to obtain neutron spectra induced by a 120-GeV proton beam in concrete and iron shielding. Generally, utilization of an active detector around high-energy accelerators requires an improvement on its readout to overcome burst of secondary radiation since the accelerator delivers an intense beam to a target in a short period after relatively long acceleration period. In this paper, we employ BS for a spectrum measurement of neutrons that penetrate the shielding wall of the pbar target station in FNAL.« less

Study of neutron generation in the compact tokamak TUMAN-3M in support of a tokamak-based fusion neutron source

NASA Astrophysics Data System (ADS)

Kornev, V. A.; Askinazi, L. G.; Belokurov, A. A.; Chernyshev, F. V.; Lebedev, S. V.; Melnik, A. D.; Shabelsky, A. A.; Tukachinsky, A. S.; Zhubr, N. A.

2017-12-01

The paper presents DD neutron flux measurements in neutron beam injection (NBI) experiments aimed at the optimization of target plasma and heating beam parameters to achieve maximum neutron flux in the TUMAN-3M compact tokamak. Two ion sources of different design were used, which allowed the separation of the beam’s energy and power influence on the neutron rate. Using the database of experiments performed with the two ion sources, an empirical scaling was derived describing the neutron rate dependence on the target plasma and heating beam parameters. Numerical modeling of the neutron rate in the NBI experiments performed using the ASTRA transport code showed good agreement with the scaling.

Spallation yield of neutrons produced in thick lead target bombarded with 250 MeV protons

NASA Astrophysics Data System (ADS)

Chen, L.; Ma, F.; Zhanga, X. Y.; Ju, Y. Q.; Zhang, H. B.; Ge, H. L.; Wang, J. G.; Zhou, B.; Li, Y. Y.; Xu, X. W.; Luo, P.; Yang, L.; Zhang, Y. B.; Li, J. Y.; Xu, J. K.; Liang, T. J.; Wang, S. L.; Yang, Y. W.; Gu, L.

2015-01-01

The neutron yield from thick target of Pb irradiated with 250 MeV protons has been studied experimentally. The neutron production was measured with the water-bath gold method. The thermal neutron distributions in the water were determined according to the measured activities of Au foils. Corresponding results calculated with the Monte Carlo code MCNPX were compared with the experimental data. It was found out that the Au foils with cadmium cover significantly changed the spacial distribution of the thermal neutron field. The corrected neutron yield was deduced to be 2.23 ± 0.19 n/proton by considering the influence of the Cd cover on the thermal neutron flux.

On the onset of void swelling in pure tungsten under neutron irradiation: An object kinetic Monte Carlo approach

NASA Astrophysics Data System (ADS)

Castin, N.; Bakaev, A.; Bonny, G.; Sand, A. E.; Malerba, L.; Terentyev, D.

2017-09-01

We propose an object kinetic Monte Carlo (OKMC) model for describing the microstructural evolution in pure tungsten under neutron irradiation. We here focus on low doses (under 1 dpa), and we neglect transmutation in first approximation. The emphasis is mainly centred on an adequate description of neutron irradiation, the subsequent introduction of primary defects, and their thermal diffusion properties. Besides grain boundaries and the dislocation network, our model includes the contribution of carbon impurities, which are shown to have a strong influence on the onset of void swelling. Our parametric study analyses the quality of our model in detail, and confronts its predictions with experimental microstructural observations with satisfactory agreement. We highlight the importance for an accurate determination of the dissolved carbon content in the tungsten matrix, and we advocate for an accurate description of atomic collision cascades, in light of the sensitivity of our results with respect to correlated recombination.

KINETIC-J: A computational kernel for solving the linearized Vlasov equation applied to calculations of the kinetic, configuration space plasma current for time harmonic wave electric fields

NASA Astrophysics Data System (ADS)

Green, David L.; Berry, Lee A.; Simpson, Adam B.; Younkin, Timothy R.

2018-04-01

We present the KINETIC-J code, a computational kernel for evaluating the linearized Vlasov equation with application to calculating the kinetic plasma response (current) to an applied time harmonic wave electric field. This code addresses the need for a configuration space evaluation of the plasma current to enable kinetic full-wave solvers for waves in hot plasmas to move beyond the limitations of the traditional Fourier spectral methods. We benchmark the kernel via comparison with the standard k →-space forms of the hot plasma conductivity tensor.

Improvement of gross theory of beta-decay for application to nuclear data

NASA Astrophysics Data System (ADS)

Koura, Hiroyuki; Yoshida, Tadashi; Tachibana, Takahiro; Chiba, Satoshi

2017-09-01

A theoretical study of β decay and delayed neutron has been carried out with a global β-decay model, the gross theory. The gross theory is based on a consideration of the sum rule of the β-strength function, and gives reasonable results of β-decay rates and delayed neutron in the entire nuclear mass region. In a fissioning nucleus, neutrons are produced by β decay of neutron-rich fission fragments from actinides known as delayed neutrons. The average number of delayed neutrons is estimated based on the sum of the β-delayed neutron-emission probabilities multiplied by the cumulative fission yield for each nucleus. Such a behavior is important to manipulate nuclear reactors, and when we adopt some new high-burn-up reactors, properties of minor actinides will play an important roll in the system, but these data have not been sufficient. We re-analyze and improve the gross theory. For example, we considered the parity of neutrons and protons at the Fermi surface, and treat a suppression for the allowed transitions in the framework of the gross theory. By using the improved gross theory, underestimated half-lives in the neutron-rich indium isotopes and neighboring region increase, and consequently follow experimental trend. The ability of reproduction (and also prediction) of the β-decay rates, delayed-neutron emission probabilities is discussed. With this work, we have described the development of a programming code of the gross theory of β-decay including the improved parts. After preparation finished, this code can be released for the nuclear data community.

Measurement and Interpretation of DT Neutron Emission from Tftr.

NASA Astrophysics Data System (ADS)

McCauley, John Scott, Jr.

A fast-ion diffusion coefficient of 0.1 +/- 0.1 m^2s ^{-1} has been deduced from the triton burnup neutron emission profile measured by a collimated array of helium-4 spectrometers. The experiment was performed with high-power deuterium discharges produced by Princeton University's Tokamak Fusion Test Reactor (TFTR). The fast ions monitored were the 1.0 MeV tritons produced from the d(d,t)p triton burnup reaction. These tritons "burn up" with deuterons and emit a 14 MeV neutron by the d(t, alpha)n reaction. The measured radial profiles of DT emission were compared with the predictions of a computer transport code. The ratio of the measured-to -calculated DT yield is typically 70%. The measured DT profile width is typically 5 cm larger than predicted by the transport code. The radial 14 MeV neutron profile was measured by a radial array of helium-4 recoil neutron spectrometers installed in the TFTR Multichannel Neutron Collimator (MCNC). The spectrometers are capable of measuring the primary and secondary neutron fluxes from deuterium discharges. The response to 14 MeV neutrons of the array has been measured by cross calibrating with the MCNC ZnS detector array when the emission from TFTR is predominantly DT neutrons. The response was also checked by comparing a model of the recoil spectrum based on nuclear physics data to the observed spectrum from ^{252 }Cf, ^{238}Pu -Be, and DT neutron sources. Extensions of this diagnostic to deuterium-tritium plasma and the implications for fusion research are discussed.

Advanced plastic scintillators for fast neutron discrimination

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feng, Patrick L; Anstey, Mitchell; Doty, F. Patrick

2014-09-01

The present work addresses the need for solid-state, fast neutron discriminating scintillators that possess higher light yields and faster decay kinetics than existing organic scintillators. These respective attributes are of critical importance for improving the gamma-rejection capabilities and increasing the neutron discrimination performance under high-rate conditions. Two key applications that will benefit from these improvements include large-volume passive detection scenarios as well as active interrogation search for special nuclear materials. Molecular design principles were employed throughout this work, resulting in synthetically tailored materials that possess the targeted scintillation properties.

Simulation of the neutron response matrix of an EJ309 liquid scintillator

NASA Astrophysics Data System (ADS)

Bai, Huaiyong; Wang, Zhimin; Zhang, Luyu; Jiang, Haoyu; Lu, Yi; Chen, Jinxiang; Zhang, Guohui

2018-04-01

The neutron response matrix is the basis for measuring the neutron energy spectrum through unfolding the pulse height spectrum detected with a liquid scintillator. Based on the light output of the EJ309 liquid scintillator and the related reaction cross sections, a Monte Carlo code is developed to obtain the neutron response matrix. The effects of the related reactions, the contributions of different number of neutron interactions and the wall effect of the recoil proton are discussed. With the obtained neutron response matrix and the GRAVEL iterative unfolding method, the neutron energy spectra of the 252Cf and the 241AmBe neutron sources are measured, and the results are respectively compared with the theoretical prediction of the 252Cf neutron energy spectrum and the previous results of the 241AmBe neutron energy spectra.

Neutron spectra from beam-target reactions in dense Z-pinches

NASA Astrophysics Data System (ADS)

Appelbe, B.; Chittenden, J.

2015-10-01

The energy spectrum of neutrons emitted by a range of deuterium and deuterium-tritium Z-pinch devices is investigated computationally using a hybrid kinetic-MHD model. 3D MHD simulations are used to model the implosion, stagnation, and break-up of dense plasma focus devices at currents of 70 kA, 500 kA, and 2 MA and also a 15 MA gas puff. Instabilities in the MHD simulations generate large electric and magnetic fields, which accelerate ions during the stagnation and break-up phases. A kinetic model is used to calculate the trajectories of these ions and the neutron spectra produced due to the interaction of these ions with the background plasma. It is found that these beam-target neutron spectra are sensitive to the electric and magnetic fields at stagnation resulting in significant differences in the spectra emitted by each device. Most notably, magnetization of the accelerated ions causes the beam-target spectra to be isotropic for the gas puff simulations. It is also shown that beam-target spectra can have a peak intensity located at a lower energy than the peak intensity of a thermonuclear spectrum. A number of other differences in the shapes of beam-target and thermonuclear spectra are also observed for each device. Finally, significant differences between the shapes of beam-target DD and DT neutron spectra, due to differences in the reaction cross-sections, are illustrated.

Nuclear Reaction Models Responsible for Simulation of Neutron-induced Soft Errors in Microelectronics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Watanabe, Y., E-mail: watanabe@aees.kyushu-u.ac.jp; Abe, S.

Terrestrial neutron-induced soft errors in MOSFETs from a 65 nm down to a 25 nm design rule are analyzed by means of multi-scale Monte Carlo simulation using the PHITS-HyENEXSS code system. Nuclear reaction models implemented in PHITS code are validated by comparisons with experimental data. From the analysis of calculated soft error rates, it is clarified that secondary He and H ions provide a major impact on soft errors with decreasing critical charge. It is also found that the high energy component from 10 MeV up to several hundreds of MeV in secondary cosmic-ray neutrons has the most significant sourcemore » of soft errors regardless of design rule.« less

Computed secondary-particle energy spectra following nonelastic neutron interactions with sup 12 C for E sub n between 15 and 60 MeV: Comparisons of results from two calculational methods

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dickens, J.K.

1991-04-01

The organic scintillation detector response code SCINFUL has been used to compute secondary-particle energy spectra, d{sigma}/dE, following nonelastic neutron interactions with {sup 12}C for incident neutron energies between 15 and 60 MeV. The resulting spectra are compared with published similar spectra computed by Brenner and Prael who used an intranuclear cascade code, including alpha clustering, a particle pickup mechanism, and a theoretical approach to sequential decay via intermediate particle-unstable states. The similarities of and the differences between the results of the two approaches are discussed. 16 refs., 44 figs., 2 tabs.

EBR-II Static Neutronic Calculations by PHISICS / MCNP6 codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Paolo Balestra; Carlo Parisi; Andrea Alfonsi

2016-02-01

The International Atomic Energy Agency (IAEA) launched a Coordinated Research Project (CRP) on the Shutdown Heat Removal Tests (SHRT) performed in the '80s at the Experimental fast Breeder Reactor EBR-II, USA. The scope of the CRP is to improve and validate the simulation tools for the study and the design of the liquid metal cooled fast reactors. Moreover, training of the next generation of fast reactor analysts is being also considered the other scope of the CRP. In this framework, a static neutronic model was developed, using state-of-the art neutron transport codes like SCALE/PHISICS (deterministic solution) and MCNP6 (stochastic solution).more » Comparison between both solutions is briefly illustrated in this summary.« less

MMAPDNG: A new, fast code backed by a memory-mapped database for simulating delayed γ-ray emission with MCNPX package

NASA Astrophysics Data System (ADS)

Lou, Tak Pui; Ludewigt, Bernhard

2015-09-01

The simulation of the emission of beta-delayed gamma rays following nuclear fission and the calculation of time-dependent energy spectra is a computational challenge. The widely used radiation transport code MCNPX includes a delayed gamma-ray routine that is inefficient and not suitable for simulating complex problems. This paper describes the code "MMAPDNG" (Memory-Mapped Delayed Neutron and Gamma), an optimized delayed gamma module written in C, discusses usage and merits of the code, and presents results. The approach is based on storing required Fission Product Yield (FPY) data, decay data, and delayed particle data in a memory-mapped file. When compared to the original delayed gamma-ray code in MCNPX, memory utilization is reduced by two orders of magnitude and the ray sampling is sped up by three orders of magnitude. Other delayed particles such as neutrons and electrons can be implemented in future versions of MMAPDNG code using its existing framework.

Delayed neutron spectral data for Hansen-Roach energy group structure

DOE Office of Scientific and Technical Information (OSTI.GOV)

Campbell, J.M.; Spriggs, G.D.

A detailed knowledge of delayed neutron spectra is important in reactor physics. It not only allows for an accurate estimate of the effective delayed neutron fraction {beta}{sub eff} but also is essential to calculating important reactor kinetic parameters, such as effective group abundances and the ratio of {beta}{sub eff} to the prompt neutron generation time. Numerous measurements of delayed neutron spectra for various delayed neutron precursors have been performed and reported in the literature. However, for application in reactor physics calculations, these spectra are usually lumped into one of the traditional six groups of delayed neutrons in accordance to theirmore » half-lives. Subsequently, these six-group spectra are binned into energy intervals corresponding to the energy intervals of a chosen nuclear cross-section set. In this work, the authors present a set of delayed neutron spectra that were formulated specifically to match Keepin`s six-group parameters and the 16-energy-group Hansen-Roach cross sections.« less

Simulation of Charge Collection in Diamond Detectors Irradiated with Deuteron-Triton Neutron Sources

DOE Office of Scientific and Technical Information (OSTI.GOV)

Milocco, Alberto; Trkov, Andrej; Pillon, Mario

2011-12-13

Diamond-based neutron spectrometers exhibit outstanding properties such as radiation hardness, low sensitivity to gamma rays, fast response and high-energy resolution. They represent a very promising application of diamonds for plasma diagnostics in fusion devices. The measured pulse height spectrum is obtained from the collection of helium and beryllium ions produced by the reactions on {sup 12}C. An original code is developed to simulate the production and the transport of charged particles inside the diamond detector. The ion transport methodology is based on the well-known TRIM code. The reactions of interest are triggered using the ENDF/B-VII.0 nuclear data for the neutronmore » interactions on carbon. The model is implemented in the TALLYX subroutine of the MCNP5 and MCNPX codes. Measurements with diamond detectors in a {approx}14 MeV neutron field have been performed at the FNG (Rome, Italy) and IRMM (Geel, Belgium) facilities. The comparison of the experimental data with the simulations validates the proposed model.« less

Energy spectrum of 208Pb(n,x) reactions

NASA Astrophysics Data System (ADS)

Tel, E.; Kavun, Y.; Özdoǧan, H.; Kaplan, A.

2018-02-01

Fission and fusion reactor technologies have been investigated since 1950's on the world. For reactor technology, fission and fusion reaction investigations are play important role for improve new generation technologies. Especially, neutron reaction studies have an important place in the development of nuclear materials. So neutron effects on materials should study as theoretically and experimentally for improve reactor design. For this reason, Nuclear reaction codes are very useful tools when experimental data are unavailable. For such circumstances scientists created many nuclear reaction codes such as ALICE/ASH, CEM95, PCROSS, TALYS, GEANT, FLUKA. In this study we used ALICE/ASH, PCROSS and CEM95 codes for energy spectrum calculation of outgoing particles from Pb bombardment by neutron. While Weisskopf-Ewing model has been used for the equilibrium process in the calculations, full exciton, hybrid and geometry dependent hybrid nuclear reaction models have been used for the pre-equilibrium process. The calculated results have been discussed and compared with the experimental data taken from EXFOR.

A comparative study of history-based versus vectorized Monte Carlo methods in the GPU/CUDA environment for a simple neutron eigenvalue problem

NASA Astrophysics Data System (ADS)

Liu, Tianyu; Du, Xining; Ji, Wei; Xu, X. George; Brown, Forrest B.

2014-06-01

For nuclear reactor analysis such as the neutron eigenvalue calculations, the time consuming Monte Carlo (MC) simulations can be accelerated by using graphics processing units (GPUs). However, traditional MC methods are often history-based, and their performance on GPUs is affected significantly by the thread divergence problem. In this paper we describe the development of a newly designed event-based vectorized MC algorithm for solving the neutron eigenvalue problem. The code was implemented using NVIDIA's Compute Unified Device Architecture (CUDA), and tested on a NVIDIA Tesla M2090 GPU card. We found that although the vectorized MC algorithm greatly reduces the occurrence of thread divergence thus enhancing the warp execution efficiency, the overall simulation speed is roughly ten times slower than the history-based MC code on GPUs. Profiling results suggest that the slow speed is probably due to the memory access latency caused by the large amount of global memory transactions. Possible solutions to improve the code efficiency are discussed.

Benchmarking kinetic calculations of resistive wall mode stability

NASA Astrophysics Data System (ADS)

Berkery, J. W.; Liu, Y. Q.; Wang, Z. R.; Sabbagh, S. A.; Logan, N. C.; Park, J.-K.; Manickam, J.; Betti, R.

2014-05-01

Validating the calculations of kinetic resistive wall mode (RWM) stability is important for confidently predicting RWM stable operating regions in ITER and other high performance tokamaks for disruption avoidance. Benchmarking the calculations of the Magnetohydrodynamic Resistive Spectrum—Kinetic (MARS-K) [Y. Liu et al., Phys. Plasmas 15, 112503 (2008)], Modification to Ideal Stability by Kinetic effects (MISK) [B. Hu et al., Phys. Plasmas 12, 057301 (2005)], and Perturbed Equilibrium Nonambipolar Transport PENT) [N. Logan et al., Phys. Plasmas 20, 122507 (2013)] codes for two Solov'ev analytical equilibria and a projected ITER equilibrium has demonstrated good agreement between the codes. The important particle frequencies, the frequency resonance energy integral in which they are used, the marginally stable eigenfunctions, perturbed Lagrangians, and fluid growth rates are all generally consistent between the codes. The most important kinetic effect at low rotation is the resonance between the mode rotation and the trapped thermal particle's precession drift, and MARS-K, MISK, and PENT show good agreement in this term. The different ways the rational surface contribution was treated historically in the codes is identified as a source of disagreement in the bounce and transit resonance terms at higher plasma rotation. Calculations from all of the codes support the present understanding that RWM stability can be increased by kinetic effects at low rotation through precession drift resonance and at high rotation by bounce and transit resonances, while intermediate rotation can remain susceptible to instability. The applicability of benchmarked kinetic stability calculations to experimental results is demonstrated by the prediction of MISK calculations of near marginal growth rates for experimental marginal stability points from the National Spherical Torus Experiment (NSTX) [M. Ono et al., Nucl. Fusion 40, 557 (2000)].

Benchmark of neutron production cross sections with Monte Carlo codes

NASA Astrophysics Data System (ADS)

Tsai, Pi-En; Lai, Bo-Lun; Heilbronn, Lawrence H.; Sheu, Rong-Jiun

2018-02-01

Aiming to provide critical information in the fields of heavy ion therapy, radiation shielding in space, and facility design for heavy-ion research accelerators, the physics models in three Monte Carlo simulation codes - PHITS, FLUKA, and MCNP6, were systematically benchmarked with comparisons to fifteen sets of experimental data for neutron production cross sections, which include various combinations of 12C, 20Ne, 40Ar, 84Kr and 132Xe projectiles and natLi, natC, natAl, natCu, and natPb target nuclides at incident energies between 135 MeV/nucleon and 600 MeV/nucleon. For neutron energies above 60% of the specific projectile energy per nucleon, the LAQGMS03.03 in MCNP6, the JQMD/JQMD-2.0 in PHITS, and the RQMD-2.4 in FLUKA all show a better agreement with data in heavy-projectile systems than with light-projectile systems, suggesting that the collective properties of projectile nuclei and nucleon interactions in the nucleus should be considered for light projectiles. For intermediate-energy neutrons whose energies are below the 60% projectile energy per nucleon and above 20 MeV, FLUKA is likely to overestimate the secondary neutron production, while MCNP6 tends towards underestimation. PHITS with JQMD shows a mild tendency for underestimation, but the JQMD-2.0 model with a modified physics description for central collisions generally improves the agreement between data and calculations. For low-energy neutrons (below 20 MeV), which are dominated by the evaporation mechanism, PHITS (which uses GEM linked with JQMD and JQMD-2.0) and FLUKA both tend to overestimate the production cross section, whereas MCNP6 tends to underestimate more systems than to overestimate. For total neutron production cross sections, the trends of the benchmark results over the entire energy range are similar to the trends seen in the dominate energy region. Also, the comparison of GEM coupled with either JQMD or JQMD-2.0 in the PHITS code indicates that the model used to describe the first stage of a nucleus-nucleus collision also affects the low-energy neutron production. Thus, in this case, a proper combination of two physics models is desired to reproduce the measured results. In addition, code users should be aware that certain models consistently produce secondary neutrons within a constant fraction of another model in certain energy regions, which might be correlated to different physics treatments in different models.

Development of a technique using MCNPX code for determination of nitrogen content of explosive materials using prompt gamma neutron activation analysis method

NASA Astrophysics Data System (ADS)

Nasrabadi, M. N.; Bakhshi, F.; Jalali, M.; Mohammadi, A.

2011-12-01

Nuclear-based explosive detection methods can detect explosives by identifying their elemental components, especially nitrogen. Thermal neutron capture reactions have been used for detecting prompt gamma 10.8 MeV following radioactive neutron capture by 14N nuclei. We aimed to study the feasibility of using field-portable prompt gamma neutron activation analysis (PGNAA) along with improved nuclear equipment to detect and identify explosives, illicit substances or landmines. A 252Cf radio-isotopic source was embedded in a cylinder made of high-density polyethylene (HDPE) and the cylinder was then placed in another cylindrical container filled with water. Measurements were performed on high nitrogen content compounds such as melamine (C3H6N6). Melamine powder in a HDPE bottle was placed underneath the vessel containing water and the neutron source. Gamma rays were detected using two NaI(Tl) crystals. The results were simulated with MCNP4c code calculations. The theoretical calculations and experimental measurements were in good agreement indicating that this method can be used for detection of explosives and illicit drugs.

Neutron Productions from thin Be target irradiated by 50 MeV/u 238U beam

NASA Astrophysics Data System (ADS)

Lee, Hee-Seock; Oh, Joo-Hee; Jung, Nam-Suk; Oranj, Leila Mokhtari; Nakao, Noriaki; Uwamino, Yoshitomo

2017-09-01

Neutrons generated from thin beryllium target by 50 MeV/u 238U beam were measured using activation analysis at 15, 30, 45, and 90 degrees from the beam direction. A 0.085 mm-thick Be stripper of RIBF was used as the neutron generating target. Activation detectors of bismuth, cobalt, and aluminum were placed out of the stripper chamber. The threshold reactions of 209Bi(n, xn)210-xBi(x=4 8), 59Co(n, xn)60-xCO(x=2 5), 59Co(n, 2nα)54Mn, 27Al(n, α)24Na, and 27Al(n,2nα)22Na were applied to measure the production rates of radionuclides. The neutron spectra were obtained using an unfolding method with the SAND-II code. All of production rates and neutron spectra were compared with the calculated results using Monte Carlo codes, the PHITS and the FLUKA. The FLUKA results showed better agreement with the measurements than the PHITS. The discrepancy between the measurements and the calculations were discussed.

Neutron spectrometry with a monolithic silicon telescope.

PubMed

Agosteo, S; D'Angelo, G; Fazzi, A; Para, A Foglio; Pola, A; Zotto, P

2007-01-01

A neutron spectrometer was set-up by coupling a polyethylene converter with a monolithic silicon telescope, consisting of a DeltaE and an E stage-detector (about 2 and 500 microm thick, respectively). The detection system was irradiated with monoenergetic neutrons at INFN-Laboratori Nazionali di Legnaro (Legnaro, Italy). The maximum detectable energy, imposed by the thickness of the E stage, is about 8 MeV for the present detector. The scatter plots of the energy deposited in the two stages were acquired using two independent electronic chains. The distributions of the recoil-protons are well-discriminated from those due to secondary electrons for energies above 0.350 MeV. The experimental spectra of the recoil-protons were compared with the results of Monte Carlo simulations using the FLUKA code. An analytical model that takes into account the geometrical structure of the silicon telescope was developed, validated and implemented in an unfolding code. The capability of reproducing continuous neutron spectra was investigated by irradiating the detector with neutrons from a thick beryllium target bombarded with protons. The measured spectra were compared with data taken from the literature. Satisfactory agreement was found.

Prompt fission neutron spectra from fission induced by 1 to 8 MeV neutrons on U235 and Pu239 using the double time-of-flight technique

NASA Astrophysics Data System (ADS)

Noda, S.; Haight, R. C.; Nelson, R. O.; Devlin, M.; O'Donnell, J. M.; Chatillon, A.; Granier, T.; Bélier, G.; Taieb, J.; Kawano, T.; Talou, P.

2011-03-01

Prompt fission neutron spectra from U235 and Pu239 were measured for incident neutron energies from 1 to 200 MeV at the Weapons Neutron Research facility (WNR) of the Los Alamos Neutron Science Center, and the experimental data were analyzed with the Los Alamos model for the incident neutron energies of 1-8 MeV. A CEA multiple-foil fission chamber containing deposits of 100 mg U235 and 90 mg Pu239 detected fission events. Outgoing neutrons were detected by the Fast Neutron-Induced γ-Ray Observer array of 20 liquid organic scintillators. A double time-of-flight technique was used to deduce the neutron incident energies from the spallation target and the outgoing energies from the fission chamber. These data were used for testing the Los Alamos model, and the total kinetic energy parameters were optimized to obtain a best fit to the data. The prompt fission neutron spectra were also compared with the Evaluated Nuclear Data File (ENDF/B-VII.0). We calculate average energies from both experimental and calculated fission neutron spectra.

Measurement of Continuous-Energy Neutron-Incident Neutron-Production Cross Section

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shigyo, Nobuhiro; Kunieda, Satoshi; Watanabe, Takehito

Continuous energy neutron-incident neutron-production double differential cross sections were measured at the Weapons Neutron Research (WNR) facility of the Los Alamos Neutron Science Center. The energy of emitted neutrons was derived from the energy deposition in a detector. The incident-neutron energy was obtained by the time-of-flight method between the spallation target of WNR and the emitted neutron detector. Two types of detectors were adopted to measure the wide energy range of neutrons. The liquid organic scintillators covered up to 100 MeV. The recoil proton detectors that constitute the recoil proton radiator and phoswich type NaI (Tl) scintillators were used formore » neutrons above several tens of MeV. Iron and lead were used as sample materials. The experimental data were compared with the evaluated nuclear data, the results of GNASH, JQMD, and PHITS codes.« less

LIGKA: A linear gyrokinetic code for the description of background kinetic and fast particle effects on the MHD stability in tokamaks

NASA Astrophysics Data System (ADS)

Lauber, Ph.; Günter, S.; Könies, A.; Pinches, S. D.

2007-09-01

In a plasma with a population of super-thermal particles generated by heating or fusion processes, kinetic effects can lead to the additional destabilisation of MHD modes or even to additional energetic particle modes. In order to describe these modes, a new linear gyrokinetic MHD code has been developed and tested, LIGKA (linear gyrokinetic shear Alfvén physics) [Ph. Lauber, Linear gyrokinetic description of fast particle effects on the MHD stability in tokamaks, Ph.D. Thesis, TU München, 2003; Ph. Lauber, S. Günter, S.D. Pinches, Phys. Plasmas 12 (2005) 122501], based on a gyrokinetic model [H. Qin, Gyrokinetic theory and computational methods for electromagnetic perturbations in tokamaks, Ph.D. Thesis, Princeton University, 1998]. A finite Larmor radius expansion together with the construction of some fluid moments and specification to the shear Alfvén regime results in a self-consistent, electromagnetic, non-perturbative model, that allows not only for growing or damped eigenvalues but also for a change in mode-structure of the magnetic perturbation due to the energetic particles and background kinetic effects. Compared to previous implementations [H. Qin, mentioned above], this model is coded in a more general and comprehensive way. LIGKA uses a Fourier decomposition in the poloidal coordinate and a finite element discretisation in the radial direction. Both analytical and numerical equilibria can be treated. Integration over the unperturbed particle orbits is performed with the drift-kinetic HAGIS code [S.D. Pinches, Ph.D. Thesis, The University of Nottingham, 1996; S.D. Pinches et al., CPC 111 (1998) 131] which accurately describes the particles' trajectories. This allows finite-banana-width effects to be implemented in a rigorous way since the linear formulation of the model allows the exchange of the unperturbed orbit integration and the discretisation of the perturbed potentials in the radial direction. Successful benchmarks for toroidal Alfvén eigenmodes (TAEs) and kinetic Alfvén waves (KAWs) with analytical results, ideal MHD codes, drift-kinetic codes and other codes based on kinetic models are reported.

Simplified Two-Time Step Method for Calculating Combustion and Emission Rates of Jet-A and Methane Fuel With and Without Water Injection

NASA Technical Reports Server (NTRS)

Molnar, Melissa; Marek, C. John

2005-01-01

A simplified kinetic scheme for Jet-A, and methane fuels with water injection was developed to be used in numerical combustion codes, such as the National Combustor Code (NCC) or even simple FORTRAN codes. The two time step method is either an initial time averaged value (step one) or an instantaneous value (step two). The switch is based on the water concentration in moles/cc of 1x10(exp -20). The results presented here results in a correlation that gives the chemical kinetic time as two separate functions. This two time step method is used as opposed to a one step time averaged method previously developed to determine the chemical kinetic time with increased accuracy. The first time averaged step is used at the initial times for smaller water concentrations. This gives the average chemical kinetic time as a function of initial overall fuel air ratio, initial water to fuel mass ratio, temperature, and pressure. The second instantaneous step, to be used with higher water concentrations, gives the chemical kinetic time as a function of instantaneous fuel and water mole concentration, pressure and temperature (T4). The simple correlations would then be compared to the turbulent mixing times to determine the limiting rates of the reaction. The NASA Glenn GLSENS kinetics code calculates the reaction rates and rate constants for each species in a kinetic scheme for finite kinetic rates. These reaction rates are used to calculate the necessary chemical kinetic times. Chemical kinetic time equations for fuel, carbon monoxide and NOx are obtained for Jet-A fuel and methane with and without water injection to water mass loadings of 2/1 water to fuel. A similar correlation was also developed using data from NASA's Chemical Equilibrium Applications (CEA) code to determine the equilibrium concentrations of carbon monoxide and nitrogen oxide as functions of overall equivalence ratio, water to fuel mass ratio, pressure and temperature (T3). The temperature of the gas entering the turbine (T4) was also correlated as a function of the initial combustor temperature (T3), equivalence ratio, water to fuel mass ratio, and pressure.

COMPLETE DETERMINATION OF POLARIZATION FOR A HIGH-ENERGY DEUTERON BEAM (thesis)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Button, J

1959-05-01

please delete the no. 17076<>13:017077The P/sub 1/ multigroup code was written for the IBM-704 in order to determine the accuracy of the few- group diffusion scheme with various imposed conditions and also to provide an alternate computational method when this scheme fails to be sufficiently accurate. The code solves for the spatially dependent multigroup flux, taking into account such nuclear phenomena is slowing down of neutrons resulting from elastic and inelastic scattering, the removal of neutrons resulting from epithermal capture and fission resonances, and the regeneration of fist neutrons resulting from fissioning which may occur in any of as manymore » as 80 fast multigroups or in the one thermal group. The code will accept as input a physical description of the reactor (that is: slab, cylindrical, or spherical geometry, number of points and regions, composition description group dependent boundary condition, transverse buckling, and mesh sizes) and a prepared library of nuclear properties of all the isotopes in each composition. The code will produce as output multigroup fluxes, currents, and isotopic slowing-down densities, in addition to pointwise and regionwise few-group macroscopic cross sections. (auth)« less

THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vondy, D.R.

1984-07-01

The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures inmore » the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations.« less

Benchmark studies of the gyro-Landau-fluid code and gyro-kinetic codes on kinetic ballooning modes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tang, T. F.; Lawrence Livermore National Laboratory, Livermore, California 94550; Xu, X. Q.

2016-03-15

A Gyro-Landau-Fluid (GLF) 3 + 1 model has been recently implemented in BOUT++ framework, which contains full Finite-Larmor-Radius effects, Landau damping, and toroidal resonance [Ma et al., Phys. Plasmas 22, 055903 (2015)]. A linear global beta scan has been conducted using the JET-like circular equilibria (cbm18 series), showing that the unstable modes are kinetic ballooning modes (KBMs). In this work, we use the GYRO code, which is a gyrokinetic continuum code widely used for simulation of the plasma microturbulence, to benchmark with GLF 3 + 1 code on KBMs. To verify our code on the KBM case, we first perform the beta scan basedmore » on “Cyclone base case parameter set.” We find that the growth rate is almost the same for two codes, and the KBM mode is further destabilized as beta increases. For JET-like global circular equilibria, as the modes localize in peak pressure gradient region, a linear local beta scan using the same set of equilibria has been performed at this position for comparison. With the drift kinetic electron module in the GYRO code by including small electron-electron collision to damp electron modes, GYRO generated mode structures and parity suggest that they are kinetic ballooning modes, and the growth rate is comparable to the GLF results. However, a radial scan of the pedestal for a particular set of cbm18 equilibria, using GYRO code, shows different trends for the low-n and high-n modes. The low-n modes show that the linear growth rate peaks at peak pressure gradient position as GLF results. However, for high-n modes, the growth rate of the most unstable mode shifts outward to the bottom of pedestal and the real frequency of what was originally the KBMs in ion diamagnetic drift direction steadily approaches and crosses over to the electron diamagnetic drift direction.« less

Object-oriented code SUR for plasma kinetic simulation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Levchenko, V.D.; Sigov, Y.S.

1995-12-31

We have developed a self-consistent simulation code based on object-oriented model of plasma (OOMP) for solving the Vlasov/Poisson (V/P), Vlasov/Maxwell (V/M), Bhatnagar-Gross-Krook (BGK) as well as Fokker-Planck (FP) kinetic equations. The application of an object-oriented approach (OOA) to simulation of plasmas and plasma-like media by means of splitting methods permits to uniformly describe and solve the wide circle of plasma kinetics problems, including those being very complicated: many-dimensional, relativistic, with regard for collisions, specific boundary conditions etc. This paper gives the brief description of possibilities of the SUR code, as a concrete realization of OOMP.

VENTURE/PC manual: A multidimensional multigroup neutron diffusion code system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shapiro, A.; Huria, H.C.; Cho, K.W.

1991-12-01

VENTURE/PC is a recompilation of part of the Oak Ridge BOLD VENTURE code system, which will operate on an IBM PC or compatible computer. Neutron diffusion theory solutions are obtained for multidimensional, multigroup problems. This manual contains information associated with operating the code system. The purpose of the various modules used in the code system, and the input for these modules are discussed. The PC code structure is also given. Version 2 included several enhancements not given in the original version of the code. In particular, flux iterations can be done in core rather than by reading and writing tomore » disk, for problems which allow sufficient memory for such in-core iterations. This speeds up the iteration process. Version 3 does not include any of the special processors used in the previous versions. These special processors utilized formatted input for various elements of the code system. All such input data is now entered through the Input Processor, which produces standard interface files for the various modules in the code system. In addition, a Standard Interface File Handbook is included in the documentation which is distributed with the code, to assist in developing the input for the Input Processor.« less

MCNP (Monte Carlo Neutron Photon) capabilities for nuclear well logging calculations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Forster, R.A.; Little, R.C.; Briesmeister, J.F.

The Los Alamos Radiation Transport Code System (LARTCS) consists of state-of-the-art Monte Carlo and discrete ordinates transport codes and data libraries. The general-purpose continuous-energy Monte Carlo code MCNP (Monte Carlo Neutron Photon), part of the LARTCS, provides a computational predictive capability for many applications of interest to the nuclear well logging community. The generalized three-dimensional geometry of MCNP is well suited for borehole-tool models. SABRINA, another component of the LARTCS, is a graphics code that can be used to interactively create a complex MCNP geometry. Users can define many source and tally characteristics with standard MCNP features. The time-dependent capabilitymore » of the code is essential when modeling pulsed sources. Problems with neutrons, photons, and electrons as either single particle or coupled particles can be calculated with MCNP. The physics of neutron and photon transport and interactions is modeled in detail using the latest available cross-section data. A rich collections of variance reduction features can greatly increase the efficiency of a calculation. MCNP is written in FORTRAN 77 and has been run on variety of computer systems from scientific workstations to supercomputers. The next production version of MCNP will include features such as continuous-energy electron transport and a multitasking option. Areas of ongoing research of interest to the well logging community include angle biasing, adaptive Monte Carlo, improved discrete ordinates capabilities, and discrete ordinates/Monte Carlo hybrid development. Los Alamos has requested approval by the Department of Energy to create a Radiation Transport Computational Facility under their User Facility Program to increase external interactions with industry, universities, and other government organizations. 21 refs.« less

EXCALIBUR-at-CALIBAN: a neutron transmission experiment for {sup 238}U(n,n'{sub continuum}γ) nuclear data validation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bernard, David; Leconte, Pierre; Destouches, Christophe

2015-07-01

Two recent papers justified a new experimental program to give a new basis for the validation of {sup 238}U nuclear data, namely neutron induced inelastic scattering and transport codes at neutron fission energies. The general idea is to perform a neutron transmission experiment through natural uranium material. As shown by Hans Bethe, neutron transmissions measured by dosimetric responses are linked to inelastic cross sections. This paper describes the principle and the results of such an experience called EXCALIBUR performed recently (January and October 2014) at the CALIBAN reactor facility. (authors)

Design of a setup for 252Cf neutron source for storage and analysis purpose

NASA Astrophysics Data System (ADS)

Hei, Daqian; Zhuang, Haocheng; Jia, Wenbao; Cheng, Can; Jiang, Zhou; Wang, Hongtao; Chen, Da

2016-11-01

252Cf is a reliable isotopic neutron source and widely used in the prompt gamma ray neutron activation analysis (PGNAA) technique. A cylindrical barrel made by polymethyl methacrylate contained with the boric acid solution was designed for storage and application of a 5 μg 252Cf neutron source. The size of the setup was optimized with Monte Carlo code. The experiments were performed and the results showed the doses were reduced with the setup and less than the allowable limit. The intensity and collimating radius of the neutron beam could also be adjusted through different collimator.

Mock-up experiment at Birmingham University for BNCT project of Osaka University--Neutron flux measurement with gold foil.

PubMed

Tamaki, S; Sakai, M; Yoshihashi, S; Manabe, M; Zushi, N; Murata, I; Hoashi, E; Kato, I; Kuri, S; Oshiro, S; Nagasaki, M; Horiike, H

2015-12-01

Mock-up experiment for development of accelerator based neutron source for Osaka University BNCT project was carried out at Birmingham University, UK. In this paper, spatial distribution of neutron flux intensity was evaluated by foil activation method. Validity of the design code system was confirmed by comparing measured gold foil activities with calculations. As a result, it was found that the epi-thermal neutron beam was well collimated by our neutron moderator assembly. Also, the design accuracy was evaluated to have less than 20% error. Copyright © 2015 Elsevier Ltd. All rights reserved.

Nuclear instrumentation in VENUS-F

NASA Astrophysics Data System (ADS)

Wagemans, J.; Borms, L.; Kochetkov, A.; Krása, A.; Van Grieken, C.; Vittiglio, G.

2018-01-01

VENUS-F is a fast zero power reactor with 30 wt% U fuel and Pb/Bi as a coolant simulator. Depending on the experimental configuration, various neutron spectra (fast, epithermal, and thermal islands) are present. This paper gives a review of the nuclear instrumentation that is applied for reactor control and in a large variety of physics experiments. Activation foils and fission chambers are used to measure spatial neutron flux profiles, spectrum indices, reactivity effects (with positive period and compensation method or the MSM method) and kinetic parameters (with the Rossi-alpha method). Fission chamber calibrations are performed in the standard irradiation fields of the BR1 reactor (prompt fission neutron spectrum and Maxwellian thermal neutron spectrum).

Proceedings of the OECD/CSNI workshop on transient thermal-hydraulic and neutronic codes requirements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ebert, D.

1997-07-01

This is a report on the CSNI Workshop on Transient Thermal-Hydraulic and Neutronic Codes Requirements held at Annapolis, Maryland, USA November 5-8, 1996. This experts` meeting consisted of 140 participants from 21 countries; 65 invited papers were presented. The meeting was divided into five areas: (1) current and prospective plans of thermal hydraulic codes development; (2) current and anticipated uses of thermal-hydraulic codes; (3) advances in modeling of thermal-hydraulic phenomena and associated additional experimental needs; (4) numerical methods in multi-phase flows; and (5) programming language, code architectures and user interfaces. The workshop consensus identified the following important action items tomore » be addressed by the international community in order to maintain and improve the calculational capability: (a) preserve current code expertise and institutional memory, (b) preserve the ability to use the existing investment in plant transient analysis codes, (c) maintain essential experimental capabilities, (d) develop advanced measurement capabilities to support future code validation work, (e) integrate existing analytical capabilities so as to improve performance and reduce operating costs, (f) exploit the proven advances in code architecture, numerics, graphical user interfaces, and modularization in order to improve code performance and scrutibility, and (g) more effectively utilize user experience in modifying and improving the codes.« less

Towards neutron scattering experiments with sub-millisecond time resolution

DOE PAGES

Adlmann, F. A.; Gutfreund, Phillip; Ankner, John Francis; ...

2015-02-01

Neutron scattering techniques offer several unique opportunities in materials research. However, most neutron scattering experiments suffer from the limited flux available at current facilities. This limitation becomes even more severe if time-resolved or kinetic experiments are performed. A new method has been developed which overcomes these limitations when a reversible process is studied, without any compromise on resolution or beam intensity. We demonstrate that, by recording in absolute time the neutron detector events linked to an excitation, information can be resolved on sub-millisecond timescales. Specifically, the concept of the method is demonstrated by neutron reflectivity measurements in time-of-flight mode atmore » the Liquids Reflectometer located at the Spallation Neutron Source, Oak Ridge National Laboratory, Tennessee, USA, combined with in situ rheometry. Finally, the opportunities and limitations of this new technique are evaluated by investigations of a micellar polymer solution offering excellent scattering contrast combined with high sensitivity to shear.« less

Radiative neutron capture cross sections on 176Lu at DANCE

NASA Astrophysics Data System (ADS)

Roig, O.; Jandel, M.; Méot, V.; Bond, E. M.; Bredeweg, T. A.; Couture, A. J.; Haight, R. C.; Keksis, A. L.; Rundberg, R. S.; Ullmann, J. L.; Vieira, D. J.

2016-03-01

The cross section of the neutron capture reaction 176Lu(n ,γ ) has been measured for a wide incident neutron energy range with the Detector for Advanced Neutron Capture Experiments at the Los Alamos Neutron Science Center. The thermal neutron capture cross section was determined to be (1912 ±132 ) b for one of the Lu natural isotopes, 176Lu. The resonance part was measured and compared to the Mughabghab's atlas using the R -matrix code, sammy. At higher neutron energies the measured cross sections are compared to ENDF/B-VII.1, JEFF-3.2, and BRC evaluated nuclear data. The Maxwellian averaged cross sections in a stellar plasma for thermal energies between 5 keV and 100 keV were extracted using these data.

2D Implosion Simulations with a Kinetic Particle Code

NASA Astrophysics Data System (ADS)

Sagert, Irina; Even, Wesley; Strother, Terrance

2017-10-01

Many problems in laboratory and plasma physics are subject to flows that move between the continuum and the kinetic regime. We discuss two-dimensional (2D) implosion simulations that were performed using a Monte Carlo kinetic particle code. The application of kinetic transport theory is motivated, in part, by the occurrence of non-equilibrium effects in inertial confinement fusion (ICF) capsule implosions, which cannot be fully captured by hydrodynamics simulations. Kinetic methods, on the other hand, are able to describe both, continuum and rarefied flows. We perform simple 2D disk implosion simulations using one particle species and compare the results to simulations with the hydrodynamics code RAGE. The impact of the particle mean-free-path on the implosion is also explored. In a second study, we focus on the formation of fluid instabilities from induced perturbations. I.S. acknowledges support through the Director's fellowship from Los Alamos National Laboratory. This research used resources provided by the LANL Institutional Computing Program.

Secondary Neutron Doses to Pediatric Patients During Intracranial Proton Therapy: Monte Carlo Simulation of the Neutron Energy Spectrum and its Organ Doses.

PubMed

Matsumoto, Shinnosuke; Koba, Yusuke; Kohno, Ryosuke; Lee, Choonsik; Bolch, Wesley E; Kai, Michiaki

2016-04-01

Proton therapy has the physical advantage of a Bragg peak that can provide a better dose distribution than conventional x-ray therapy. However, radiation exposure of normal tissues cannot be ignored because it is likely to increase the risk of secondary cancer. Evaluating secondary neutrons generated by the interaction of the proton beam with the treatment beam-line structure is necessary; thus, performing the optimization of radiation protection in proton therapy is required. In this research, the organ dose and energy spectrum were calculated from secondary neutrons using Monte Carlo simulations. The Monte Carlo code known as the Particle and Heavy Ion Transport code System (PHITS) was used to simulate the transport proton and its interaction with the treatment beam-line structure that modeled the double scattering body of the treatment nozzle at the National Cancer Center Hospital East. The doses of the organs in a hybrid computational phantom simulating a 5-y-old boy were calculated. In general, secondary neutron doses were found to decrease with increasing distance to the treatment field. Secondary neutron energy spectra were characterized by incident neutrons with three energy peaks: 1×10, 1, and 100 MeV. A block collimator and a patient collimator contributed significantly to organ doses. In particular, the secondary neutrons from the patient collimator were 30 times higher than those from the first scatter. These results suggested that proactive protection will be required in the design of the treatment beam-line structures and that organ doses from secondary neutrons may be able to be reduced.

Assay of the Martian Regolith with Neutrons

NASA Technical Reports Server (NTRS)

Drake, Darrell M.

1997-01-01

The purpose of the research is to combine experiments and Monte Carlo transport of neutrons through volume of soil in an attempt to model neutron leakage from planetary surfaces. Emphasis is given to the change of neutron spectra as a function of water content and location. During the first stage of effort, two experiments were conducted in which leakage of neutrons from a Pu-Be source through about 30 g/cm(exp 2) of soil were measured with several counters. A Monte Carlo code, MCNP, has been used to model many of the 100 individual runs of the experiment. Hydrogen is the element that has the most dramatic effect on the neutron spectrum and its effect on the neutron spectrum is almost the same whether it is in the form of water or polyethylene. In order to simulate various water configurations, sheets of polyethylene have been used between layers of soil as well as water in several concentrations up to 18%. Comparison of experimental results to theoretical predictions made with the MCNP code were disappointing for low concentrations of water. We have made extensive calculations to see if room return could be the cause of the discrepancies. Water concentrations of the 'dry' soil were measured by two different laboratories and differed only by 0.5%. We have made calculations to optimize the next experiment and are investigating other methods of determining the water content of 'dry' soil.

Magnetic neutron star cooling and microphysics

NASA Astrophysics Data System (ADS)

Potekhin, A. Y.; Chabrier, G.

2018-01-01

Aims: We study the relative importance of several recent updates of microphysics input to the neutron star cooling theory and the effects brought about by superstrong magnetic fields of magnetars, including the effects of the Landau quantization in their crusts. Methods: We use a finite-difference code for simulation of neutron-star thermal evolution on timescales from hours to megayears with an updated microphysics input. The consideration of short timescales (≲1 yr) is made possible by a treatment of the heat-blanketing envelope without the quasistationary approximation inherent to its treatment in traditional neutron-star cooling codes. For the strongly magnetized neutron stars, we take into account the effects of Landau quantization on thermodynamic functions and thermal conductivities. We simulate cooling of ordinary neutron stars and magnetars with non-accreted and accreted crusts and compare the results with observations. Results: Suppression of radiative and conductive opacities in strongly quantizing magnetic fields and formation of a condensed radiating surface substantially enhance the photon luminosity at early ages, making the life of magnetars brighter but shorter. These effects together with the effect of strong proton superfluidity, which slows down the cooling of kiloyear-aged neutron stars, can explain thermal luminosities of about a half of magnetars without invoking heating mechanisms. Observed thermal luminosities of other magnetars are still higher than theoretical predictions, which implies heating, but the effects of quantizing magnetic fields and baryon superfluidity help to reduce the discrepancy.

A neutron spectrum unfolding code based on generalized regression artificial neural networks.

PubMed

Del Rosario Martinez-Blanco, Ma; Ornelas-Vargas, Gerardo; Castañeda-Miranda, Celina Lizeth; Solís-Sánchez, Luis Octavio; Castañeda-Miranada, Rodrigo; Vega-Carrillo, Héctor René; Celaya-Padilla, Jose M; Garza-Veloz, Idalia; Martínez-Fierro, Margarita; Ortiz-Rodríguez, José Manuel

2016-11-01

The most delicate part of neutron spectrometry, is the unfolding process. The derivation of the spectral information is not simple because the unknown is not given directly as a result of the measurements. Novel methods based on Artificial Neural Networks have been widely investigated. In prior works, back propagation neural networks (BPNN) have been used to solve the neutron spectrometry problem, however, some drawbacks still exist using this kind of neural nets, i.e. the optimum selection of the network topology and the long training time. Compared to BPNN, it's usually much faster to train a generalized regression neural network (GRNN). That's mainly because spread constant is the only parameter used in GRNN. Another feature is that the network will converge to a global minimum, provided that the optimal values of spread has been determined and that the dataset adequately represents the problem space. In addition, GRNN are often more accurate than BPNN in the prediction. These characteristics make GRNNs to be of great interest in the neutron spectrometry domain. This work presents a computational tool based on GRNN capable to solve the neutron spectrometry problem. This computational code, automates the pre-processing, training and testing stages using a k-fold cross validation of 3 folds, the statistical analysis and the post-processing of the information, using 7 Bonner spheres rate counts as only entrance data. The code was designed for a Bonner Spheres System based on a 6 LiI(Eu) neutron detector and a response matrix expressed in 60 energy bins taken from an International Atomic Energy Agency compilation. Copyright © 2016 Elsevier Ltd. All rights reserved.

Rates for neutron-capture reactions on tungsten isotopes in iron meteorites. [Abstract only

NASA Technical Reports Server (NTRS)

Masarik, J.; Reedy, R. C.

1994-01-01

High-precision W isotopic analyses by Harper and Jacobsen indicate the W-182/W-183 ratio in the Toluca iron meteorite is shifted by -(3.0 +/- 0.9) x 10(exp -4) relative to a terrestrial standard. Possible causes of this shift are neutron-capture reactions on W during Toluca's approximately 600-Ma exposure to cosmic ray particles or radiogenic growth of W-182 from 9-Ma Hf-182 in the silicate portion of the Earth after removal of W to the Earth's core. Calculations for the rates of neutron-capture reactions on W isotopes were done to study the first possibility. The LAHET Code System (LCS) which consists of the Los Alamos High Energy Transport (LAHET) code and the Monte Carlo N-Particle(MCNP) transport code was used to numerically simulate the irradiation of the Toluca iron meteorite by galactic-cosmic-ray (GCR) particles and to calculate the rates of W(n, gamma) reactions. Toluca was modeled as a 3.9-m-radius sphere with the composition of a typical IA iron meteorite. The incident GCR protons and their interactions were modeled with LAHET, which also handled the interactions of neutrons with energies above 20 MeV. The rates for the capture of neutrons by W-182, W-183, and W-186 were calculated using the detailed library of (n, gamma) cross sections in MCNP. For this study of the possible effect of W(n, gamma) reactions on W isotope systematics, we consider the peak rates. The calculated maximum change in the normalized W-182/W-183 ratio due to neutron-capture reactions cannot account for more than 25% of the mass 182 deficit observed in Toluca W.

GAMSOR: Gamma Source Preparation and DIF3D Flux Solution

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, M. A.; Lee, C. H.; Hill, R. N.

2017-06-28

Nuclear reactors that rely upon the fission reaction have two modes of thermal energy deposition in the reactor system: neutron absorption and gamma absorption. The gamma rays are typically generated by neutron capture reactions or during the fission process which means the primary driver of energy production is of course the neutron interaction. In conventional reactor physics methods, the gamma heating component is ignored such that the gamma absorption is forced to occur at the gamma emission site. For experimental reactor systems like EBR-II and FFTF, the placement of structural pins and assemblies internal to the core leads to problemsmore » with power heating predictions because there is no fission power source internal to the assembly to dictate a spatial distribution of the power. As part of the EBR-II support work in the 1980s, the GAMSOR code was developed to assist analysts in calculating the gamma heating. The GAMSOR code is a modified version of DIF3D and actually functions within a sequence of DIF3D calculations. The gamma flux in a conventional fission reactor system does not perturb the neutron flux and thus the gamma flux calculation can be cast as a fixed source problem given a solution to the steady state neutron flux equation. This leads to a sequence of DIF3D calculations, called the GAMSOR sequence, which involves solving the neutron flux, then the gamma flux, and then combining the results to do a summary edit. In this manuscript, we go over the GAMSOR code and detail how it is put together and functions. We also discuss how to setup the GAMSOR sequence and input for each DIF3D calculation in the GAMSOR sequence.« less

Earth and Planetary Science Letters

NASA Technical Reports Server (NTRS)

Nishiizumi, K.; Klein, J.; Middleton, R.; Masarik, J.; Reedy, R. C.; Arnold, J. R.; Fink, D.

1997-01-01

Systematic measurements of the concentrations of cosmogen Ca-41 (half-life = 1.04 x 10(exp 5) yr) in the Apollo 15 long core 15001-15006 were performed by accelerator mass spectroscopy. Earlier measurements of cosmogenic Be-10, C-14, Al-26, Cl-36, and Mn-53 in the same core have provided confirmation and improvement of theoretical models for predicting production profiles of nuclides by cosmic ray induced spallation in the Moon and large meteorites. Unlike these nuclides, Ca-40 in the lunar surface is produced mainly by thermal neutron capture reactions on Ca-40. The maximum production of Ca-41, about 1 dpm/g Ca, was observed at a depth in the Moon of about 150 g/sq cm. For depths below about 300 g/sq cm, Ca-41 production falls off exponentially with an e-folding length of 175 g/sq cm. Neutron production in the Moon was modeled with the Los Alamos High Energy Transport Code System, and yields of nuclei produced by low-energy thermal and epithermal neutrons were calculated with the Monte Carlo N-Particle code. The new theoretical calculations using these codes are in good agreement with our measured Ca-41 concentrations as well as with Co-60 and direct neutron fluence measurements in the Moon.

Current Mode Neutron Noise Measurements in the Zero Power Reactor CROCUS

NASA Astrophysics Data System (ADS)

Pakari, O.; Lamirand, V.; Perret, G.; Braun, L.; Frajtag, P.; Pautz, A.

2018-01-01

The present article is an overview of developments and results regarding neutron noise measurements in current mode at the CROCUS zero power facility. Neutron noise measurements offer a non-invasive method to determine kinetic reactor parameters such as the prompt decay constant at criticality α = βeff / λ, the effective delayed neutron fraction βeff, and the mean generation time λ for code validation efforts. At higher detection rates, i.e. above 2×104 cps in the used configuration at 0.1 W, the previously employed pulse charge amplification electronics with BF3 detectors yielded erroneous results due to dead time effects. Future experimental needs call for higher sensitivity in detectors, higher detection rates or higher reactor powers, and thus a generally more versatile measurement system. We, therefore, explored detectors operated with current mode acquisition electronics to accommodate the need. We approached the matter in two ways: 1) By using the two compensated 10B-coated ionization chambers available in CROCUS as operational monitors. The compensated current signal of these chambers was extracted from coremonitoring output channels. 2) By developing a new current mode amplification station to be used with other available detectors in core. Characteristics and first noise measurements of the new current system are presented. We implemented post-processing of the current signals from 1)and 2) with the APSD/CPSD method to determine α. At two critical states (0.5 and 1.5 W), using the 10B ionization chambers and their CPSD estimate, the prompt decay constant was measured after 1.5 hours to be α=(156.9 ± 4.3) s-1 (1σ). This result is within 1σ of statistical uncertainties of previous experiments and MCNPv5-1.6 predictions using the ENDF/B-7.1 library. The newsystem connected to a CFUL01 fission chamber using the APSDestimate at 100 mW after 33 min yielded α = (160.8 ± 6.3) s-1, also within 1σ agreement. The improvements to previous neutron noise measurementsinclude shorter measurement durations that can achievecomparable statistical uncertainties and measurements at higherdetection rates.

LSENS, a general chemical kinetics and sensitivity analysis code for homogeneous gas-phase reactions. 2: Code description and usage

NASA Technical Reports Server (NTRS)

Radhakrishnan, Krishnan; Bittker, David A.

1994-01-01

LSENS, the Lewis General Chemical Kinetics Analysis Code, has been developed for solving complex, homogeneous, gas-phase chemical kinetics problems and contains sensitivity analysis for a variety of problems, including nonisothermal situations. This report is part 2 of a series of three reference publications that describe LSENS, provide a detailed guide to its usage, and present many example problems. Part 2 describes the code, how to modify it, and its usage, including preparation of the problem data file required to execute LSENS. Code usage is illustrated by several example problems, which further explain preparation of the problem data file and show how to obtain desired accuracy in the computed results. LSENS is a flexible, convenient, accurate, and efficient solver for chemical reaction problems such as static system; steady, one-dimensional, inviscid flow; reaction behind incident shock wave, including boundary layer correction; and perfectly stirred (highly backmixed) reactor. In addition, the chemical equilibrium state can be computed for the following assigned states: temperature and pressure, enthalpy and pressure, temperature and volume, and internal energy and volume. For static problems the code computes the sensitivity coefficients of the dependent variables and their temporal derivatives with respect to the initial values of the dependent variables and/or the three rate coefficient parameters of the chemical reactions. Part 1 (NASA RP-1328) derives the governing equations describes the numerical solution procedures for the types of problems that can be solved by lSENS. Part 3 (NASA RP-1330) explains the kinetics and kinetics-plus-sensitivity-analysis problems supplied with LSENS and presents sample results.

A possible approach to 14MeV neutron moderation: A preliminary study case.

PubMed

Flammini, D; Pilotti, R; Pietropaolo, A

2017-07-01

Deuterium-Tritium (D-T) interactions produce almost monochromatic neutrons with about 14MeV energy. These neutrons are used in benchmark experiments as well as for neutron cross sections assessment in fusion reactors technology. The possibility to moderate 14MeV neutrons for purposes beyond fusion is worth to be studied in relation to projects of intense D-T sources. In this preliminary study, carried out using the MCNP Monte Carlo code, the moderation of 14MeV neutrons is approached foreseeing the use of combination of metallic materials as pre-moderator and reflectors coupled to standard water moderators. Copyright © 2017 Elsevier Ltd. All rights reserved.

Improvement of Modeling HTGR Neutron Physics by Uncertainty Analysis with the Use of Cross-Section Covariance Information

NASA Astrophysics Data System (ADS)

Boyarinov, V. F.; Grol, A. V.; Fomichenko, P. A.; Ternovykh, M. Yu

2017-01-01

This work is aimed at improvement of HTGR neutron physics design calculations by application of uncertainty analysis with the use of cross-section covariance information. Methodology and codes for preparation of multigroup libraries of covariance information for individual isotopes from the basic 44-group library of SCALE-6 code system were developed. A 69-group library of covariance information in a special format for main isotopes and elements typical for high temperature gas cooled reactors (HTGR) was generated. This library can be used for estimation of uncertainties, associated with nuclear data, in analysis of HTGR neutron physics with design codes. As an example, calculations of one-group cross-section uncertainties for fission and capture reactions for main isotopes of the MHTGR-350 benchmark, as well as uncertainties of the multiplication factor (k∞) for the MHTGR-350 fuel compact cell model and fuel block model were performed. These uncertainties were estimated by the developed technology with the use of WIMS-D code and modules of SCALE-6 code system, namely, by TSUNAMI, KENO-VI and SAMS. Eight most important reactions on isotopes for MHTGR-350 benchmark were identified, namely: 10B(capt), 238U(n,γ), ν5, 235U(n,γ), 238U(el), natC(el), 235U(fiss)-235U(n,γ), 235U(fiss).

Benchmarking PARTISN with Analog Monte Carlo: Moments of the Neutron Number and the Cumulative Fission Number Probability Distributions

DOE Office of Scientific and Technical Information (OSTI.GOV)

O'Rourke, Patrick Francis

The purpose of this report is to provide the reader with an understanding of how a Monte Carlo neutron transport code was written, developed, and evolved to calculate the probability distribution functions (PDFs) and their moments for the neutron number at a final time as well as the cumulative fission number, along with introducing several basic Monte Carlo concepts.

Current and anticipated use of thermal-hydraulic codes for BWR transient and accident analyses in Japan

DOE Office of Scientific and Technical Information (OSTI.GOV)

Arai, Kenji; Ebata, Shigeo

1997-07-01

This paper summarizes the current and anticipated use of the thermal-hydraulic and neutronic codes for the BWR transient and accident analyses in Japan. The codes may be categorized into the licensing codes and the best estimate codes for the BWR transient and accident analyses. Most of the licensing codes have been originally developed by General Electric. Some codes have been updated based on the technical knowledge obtained in the thermal hydraulic study in Japan, and according to the BWR design changes. The best estimates codes have been used to support the licensing calculations and to obtain the phenomenological understanding ofmore » the thermal hydraulic phenomena during a BWR transient or accident. The best estimate codes can be also applied to a design study for a next generation BWR to which the current licensing model may not be directly applied. In order to rationalize the margin included in the current BWR design and develop a next generation reactor with appropriate design margin, it will be required to improve the accuracy of the thermal-hydraulic and neutronic model. In addition, regarding the current best estimate codes, the improvement in the user interface and the numerics will be needed.« less

Measurement of the 234U(n, f ) cross-section with quasi-monoenergetic beams in the keV and MeV range using a Micromegas detector assembly

NASA Astrophysics Data System (ADS)

Stamatopoulos, A.; Kanellakopoulos, A.; Kalamara, A.; Diakaki, M.; Tsinganis, A.; Kokkoris, M.; Michalopoulou, V.; Axiotis, M.; Lagoyiannis, A.; Vlastou, R.

2018-01-01

The 234U neutron-induced fission cross-section has been measured at incident neutron energies of 452, 550, 651 keV and 7.5, 8.7, 10 MeV using the 7Li ( p, n) and the 2H( d, n) reactions, respectively, relative to the 235U( n, f ) and 238U( n, f ) reference reactions. The measurement was performed at the neutron beam facility of the National Center for Scientific Research "Demokritos", using a set-up based on Micromegas detectors. The active mass of the actinide samples and the corresponding impurities were determined via α-spectroscopy using a surface barrier silicon detector. The neutron spectra intercepted by the actinide samples have been thoroughly studied by coupling the NeuSDesc and MCNP5 codes, taking into account the energy and angular straggling of the primary ion beams in the neutron source targets in addition to contributions from competing reactions ( e.g. deuteron break-up) and neutron scattering in the surrounding materials. Auxiliary Monte Carlo simulations were performed making combined use of the FLUKA and GEF codes, focusing particularly on the determination of the fission fragment detection efficiency. The developed methodology and the final results are presented.

On formation of the asymptotic spectrum of delayed neutron emitters in measuring the VVER-1000 scram system effectiveness

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shishkov, L. K., E-mail: slk@vver.kiae.ru; Zizin, M. N., E-mail: zizin_m@mail.ru

The process of formation of an asymptotic distribution of the neutron flux density in the reactor systems after introducing different negative reactivities is considered. The impact of two factors after the reactivity introduction is evaluated: (1) nonuniformity of perturbation of core properties, on one hand, and (2) a sharp reduction in the density of prompt neutrons, which prevents the appearance of new delayed neutron emitters distributed in accordance with the “new” prompt neutron distribution, on the other hand. The results of calculations show that the errors of measuring the scram system effectiveness using the method of inverse solution of themore » kinetics equation are caused by the fact that, after the negative reactivity insertion, the sources of prompt and delayed neutrons have different spatial distributions. In the case of high negative reactivities, this difference remains while the system still has neutrons, which can be measured.« less

Performance of the Versatile Array of Neutron Detectors at Low Energy (VANDLE)

DOE PAGES

Peters, W. A.; Ilyushkin, S.; Madurga, M.; ...

2016-08-26

The Versatile Array of Neutron Detectors at Low Energy (VANDLE) is a new, highly efficient plastic-scintillator array constructed for decay and transfer reaction experimental setups that require neutron detection. The versatile and modular design allows for customizable experimental setups including beta-delayed neutron spectroscopy and (d,n) transfer reactions in normal and inverse kinematics. The neutron energy and prompt-photon discrimination is determined through the time of flight technique. Fully digital data acquisition electronics and integrated triggering logic enables some VANDLE modules to achieve an intrinsic efficiency over 70% for 300-keV neutrons, measured through two different methods. A custom Geant4 simulation models aspectsmore » of the detector array and the experimental setups to determine efficiency and detector response. Lastly, a low detection threshold, due to the trigger logic and digitizing data acquisition, allowed us to measure the light-yield response curve from elastically scattered carbon nuclei inside the scintillating plastic from incident neutrons with kinetic energies below 2 MeV.« less

Fission time scale from pre-scission neutron and α multiplicities in the 16O + 194Pt reaction

NASA Astrophysics Data System (ADS)

Kapoor, K.; Verma, S.; Sharma, P.; Mahajan, R.; Kaur, N.; Kaur, G.; Behera, B. R.; Singh, K. P.; Kumar, A.; Singh, H.; Dubey, R.; Saneesh, N.; Jhingan, A.; Sugathan, P.; Mohanto, G.; Nayak, B. K.; Saxena, A.; Sharma, H. P.; Chamoli, S. K.; Mukul, I.; Singh, V.

2017-11-01

Pre- and post-scission α -particle multiplicities have been measured for the reaction 16O+P194t at 98.4 MeV forming R210n compound nucleus. α particles were measured at various angles in coincidence with the fission fragments. Moving source technique was used to extract the pre- and post-scission contributions to the particle multiplicity. Study of the fission mechanism using the different probes are helpful in understanding the detailed reaction dynamics. The neutron multiplicities for this reaction have been reported earlier. The multiplicities of neutrons and α particles were reproduced using standard statistical model code joanne2 by varying the transient (τt r) and saddle to scission (τs s c) times. This code includes deformation dependent-particle transmission coefficients, binding energies and level densities. Fission time scales of the order of 50-65 ×10-21 s are required to reproduce the neutron and α -particle multiplicities.

Calculated organ doses for Mayak production association central hall using ICRP and MCNP.

PubMed

Choe, Dong-Ok; Shelkey, Brenda N; Wilde, Justin L; Walk, Heidi A; Slaughter, David M

2003-03-01

As part of an ongoing dose reconstruction project, equivalent organ dose rates from photons and neutrons were estimated using the energy spectra measured in the central hall above the graphite reactor core located in the Russian Mayak Production Association facility. Reconstruction of the work environment was necessary due to the lack of personal dosimeter data for neutrons in the time period prior to 1987. A typical worker scenario for the central hall was developed for the Monte Carlo Neutron Photon-4B (MCNP) code. The resultant equivalent dose rates for neutrons and photons were compared with the equivalent dose rates derived from calculations using the conversion coefficients in the International Commission on Radiological Protection Publications 51 and 74 in order to validate the model scenario for this Russian facility. The MCNP results were in good agreement with the results of the ICRP publications indicating the modeling scenario was consistent with actual work conditions given the spectra provided. The MCNP code will allow for additional orientations to accurately reflect source locations.

Neutron Environment Calculations for Low Earth Orbit

NASA Technical Reports Server (NTRS)

Clowdsley, M. S.; Wilson, J. W.; Shinn, J. L.; Badavi, F. F.; Heinbockel, J. H.; Atwell, W.

2001-01-01

The long term exposure of astronauts on the developing International Space Station (ISS) requires an accurate knowledge of the internal exposure environment for human risk assessment and other onboard processes. The natural environment is moderated by the solar wind, which varies over the solar cycle. The HZETRN high charge and energy transport code developed at NASA Langley Research Center can be used to evaluate the neutron environment on ISS. A time dependent model for the ambient environment in low earth orbit is used. This model includes GCR radiation moderated by the Earth's magnetic field, trapped protons, and a recently completed model of the albedo neutron environment formed through the interaction of galactic cosmic rays with the Earth's atmosphere. Using this code, the neutron environments for space shuttle missions were calculated and comparisons were made to measurements by the Johnson Space Center with onboard detectors. The models discussed herein are being developed to evaluate the natural and induced environment data for the Intelligence Synthesis Environment Project and eventual use in spacecraft optimization.

Coupled Neutron Transport for HZETRN

NASA Technical Reports Server (NTRS)

Slaba, Tony C.; Blattnig, Steve R.

2009-01-01

Exposure estimates inside space vehicles, surface habitats, and high altitude aircrafts exposed to space radiation are highly influenced by secondary neutron production. The deterministic transport code HZETRN has been identified as a reliable and efficient tool for such studies, but improvements to the underlying transport models and numerical methods are still necessary. In this paper, the forward-backward (FB) and directionally coupled forward-backward (DC) neutron transport models are derived, numerical methods for the FB model are reviewed, and a computationally efficient numerical solution is presented for the DC model. Both models are compared to the Monte Carlo codes HETC-HEDS, FLUKA, and MCNPX, and the DC model is shown to agree closely with the Monte Carlo results. Finally, it is found in the development of either model that the decoupling of low energy neutrons from the light particle transport procedure adversely affects low energy light ion fluence spectra and exposure quantities. A first order correction is presented to resolve the problem, and it is shown to be both accurate and efficient.

A Monte Carlo Simulation of Prompt Gamma Emission from Fission Fragments

NASA Astrophysics Data System (ADS)

Regnier, D.; Litaize, O.; Serot, O.

2013-03-01

The prompt fission gamma spectra and multiplicities are investigated through the Monte Carlo code FIFRELIN which is developed at the Cadarache CEA research center. Knowing the fully accelerated fragment properties, their de-excitation is simulated through a cascade of neutron, gamma and/or electron emissions. This paper presents the recent developments in the FIFRELIN code and the results obtained on the spontaneous fission of 252Cf. Concerning the decay cascades simulation, a fully Hauser-Feshbach model is compared with a previous one using a Weisskopf spectrum for neutron emission. A particular attention is paid to the treatment of the neutron/gamma competition. Calculations lead using different level density and gamma strength function models show significant discrepancies of the slope of the gamma spectra at high energy. The underestimation of the prompt gamma spectra obtained regardless our de-excitation cascade modeling choice is discussed. This discrepancy is probably linked to an underestimation of the post-neutron fragments spin in our calculation.

Neutron scattering from myelin revisited: bilayer asymmetry and water-exchange kinetics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Denninger, Andrew R.; Demé, Bruno; Cristiglio, Viviana

2014-12-01

The structure of internodal myelin in the rodent central and peripheral nervous systems has been determined using neutron diffraction. The kinetics of water exchange in these tissues is also described. Rapid nerve conduction in the central and peripheral nervous systems (CNS and PNS, respectively) of higher vertebrates is brought about by the ensheathment of axons with myelin, a lipid-rich, multilamellar assembly of membranes. The ability of myelin to electrically insulate depends on the regular stacking of these plasma membranes and on the presence of a number of specialized membrane-protein assemblies in the sheath, including the radial component, Schmidt–Lanterman incisures andmore » the axo–glial junctions of the paranodal loops. The disruption of this fine-structure is the basis for many demyelinating neuropathies in the CNS and PNS. Understanding the processes that govern myelin biogenesis, maintenance and destabilization requires knowledge of myelin structure; however, the tight packing of internodal myelin and the complexity of its junctional specializations make myelin a challenging target for comprehensive structural analysis. This paper describes an examination of myelin from the CNS and PNS using neutron diffraction. This investigation revealed the dimensions of the bilayers and aqueous spaces of myelin, asymmetry between the cytoplasmic and extracellular leaflets of the membrane, and the distribution of water and exchangeable hydrogen in internodal multilamellar myelin. It also uncovered differences between CNS and PNS myelin in their water-exchange kinetics.« less

A critical analysis of the accuracy of several numerical techniques for combustion kinetic rate equations

NASA Technical Reports Server (NTRS)

Radhadrishnan, Krishnan

1993-01-01

A detailed analysis of the accuracy of several techniques recently developed for integrating stiff ordinary differential equations is presented. The techniques include two general-purpose codes EPISODE and LSODE developed for an arbitrary system of ordinary differential equations, and three specialized codes CHEMEQ, CREK1D, and GCKP4 developed specifically to solve chemical kinetic rate equations. The accuracy study is made by application of these codes to two practical combustion kinetics problems. Both problems describe adiabatic, homogeneous, gas-phase chemical reactions at constant pressure, and include all three combustion regimes: induction, heat release, and equilibration. To illustrate the error variation in the different combustion regimes the species are divided into three types (reactants, intermediates, and products), and error versus time plots are presented for each species type and the temperature. These plots show that CHEMEQ is the most accurate code during induction and early heat release. During late heat release and equilibration, however, the other codes are more accurate. A single global quantity, a mean integrated root-mean-square error, that measures the average error incurred in solving the complete problem is used to compare the accuracy of the codes. Among the codes examined, LSODE is the most accurate for solving chemical kinetics problems. It is also the most efficient code, in the sense that it requires the least computational work to attain a specified accuracy level. An important finding is that use of the algebraic enthalpy conservation equation to compute the temperature can be more accurate and efficient than integrating the temperature differential equation.

A new hybrid code (CHIEF) implementing the inertial electron fluid equation without approximation

NASA Astrophysics Data System (ADS)

Muñoz, P. A.; Jain, N.; Kilian, P.; Büchner, J.

2018-03-01

We present a new hybrid algorithm implemented in the code CHIEF (Code Hybrid with Inertial Electron Fluid) for simulations of electron-ion plasmas. The algorithm treats the ions kinetically, modeled by the Particle-in-Cell (PiC) method, and electrons as an inertial fluid, modeled by electron fluid equations without any of the approximations used in most of the other hybrid codes with an inertial electron fluid. This kind of code is appropriate to model a large variety of quasineutral plasma phenomena where the electron inertia and/or ion kinetic effects are relevant. We present here the governing equations of the model, how these are discretized and implemented numerically, as well as six test problems to validate our numerical approach. Our chosen test problems, where the electron inertia and ion kinetic effects play the essential role, are: 0) Excitation of parallel eigenmodes to check numerical convergence and stability, 1) parallel (to a background magnetic field) propagating electromagnetic waves, 2) perpendicular propagating electrostatic waves (ion Bernstein modes), 3) ion beam right-hand instability (resonant and non-resonant), 4) ion Landau damping, 5) ion firehose instability, and 6) 2D oblique ion firehose instability. Our results reproduce successfully the predictions of linear and non-linear theory for all these problems, validating our code. All properties of this hybrid code make it ideal to study multi-scale phenomena between electron and ion scales such as collisionless shocks, magnetic reconnection and kinetic plasma turbulence in the dissipation range above the electron scales.

Proceedings of the Workshop on High Temperature Superconductivity

DTIC Science & Technology

1989-11-01

such magnetic excitations in neutron scattering studies of UPt3 and measured a corresponding Debye energy owc = 2 K, in excellent agreement with the...procedure of Budhani et al. Propylene carbonate has been found to be a suitable vehicle for direct painting, while poly (ethylene glycol methyl ether ...through neutron irradiation and chemical means will also be discussed. Specifically, results of comparative studies on the kinetics of flux motion in

A Pulsed Sphere Tutorial

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cullen, Dermott E.

2017-01-30

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).

Evaluated cross-section libraries and kerma factors for neutrons up to 100 MeV on {sup 12}C

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chadwick, M.B.; Blann, M.; Cox, L.

1995-04-11

A program is being carried out at Lawrence Livermore National Laboratory to develop high-energy evaluated nuclear data libraries for use in Monte Carlo simulations of cancer radiation therapy. In this report we describe evaluated cross sections and kerma factors for neutrons with incident energies up to 100 MeV on {sup 12}C. The aim of this effort is to incorporate advanced nuclear physics modeling methods, with new experimental measurements, to generate cross section libraries needed for an accurate simulation of dose deposition in fast neutron therapy. The evaluated libraries are based mainly on nuclear model calculations, benchmarked to experimental measurements wheremore » they exist. We use the GNASH code system, which includes Hauser-Feshbach, preequilibrium, and direct reaction mechanisms. The libraries tabulate elastic and nonelastic cross sections, angle-energy correlated production spectra for light ejectiles with A{le}and kinetic energies given to light ejectiles and heavy recoil fragments. The major steps involved in this effort are: (1) development and validation of nuclear models for incident energies up to 100 MeV; (2) collation of experimental measurements, including new results from Louvain-la-Nueve and Los Alamos; (3) extension of the Livermore ENDL formats for representing high-energy data; (4) calculation and evaluation of nuclear data; and (5) validation of the libraries. We describe the evaluations in detail, with particular emphasis on our new high-energy modeling developments. Our evaluations agree well with experimental measurements of integrated and differential cross sections. We compare our results with the recent ENDF/B-VI evaluation which extends up to 32 MeV.« less

Exposure calculation code module for reactor core analysis: BURNER

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vondy, D.R.; Cunningham, G.W.

1979-02-01

The code module BURNER for nuclear reactor exposure calculations is presented. The computer requirements are shown, as are the reference data and interface data file requirements, and the programmed equations and procedure of calculation are described. The operating history of a reactor is followed over the period between solutions of the space, energy neutronics problem. The end-of-period nuclide concentrations are determined given the necessary information. A steady state, continuous fueling model is treated in addition to the usual fixed fuel model. The control options provide flexibility to select among an unusually wide variety of programmed procedures. The code also providesmore » user option to make a number of auxiliary calculations and print such information as the local gamma source, cumulative exposure, and a fine scale power density distribution in a selected zone. The code is used locally in a system for computation which contains the VENTURE diffusion theory neutronics code and other modules.« less

Design study of multi-imaging plate system for BNCT irradiation field at Kyoto university reactor.

PubMed

Tanaka, Kenichi; Sakurai, Yoshinori; Kajimoto, Tsuyoshi; Tanaka, Hiroki; Takata, Takushi; Endo, Satoru

2016-09-01

The converter configuration for a multi-imaging plate system was investigated for the application of quality assurance in the irradiation field profile for boron neutron capture therapy. This was performed by the simulation calculation using the PHITS code in the fields at the Heavy Water Neutron Irradiation Facility of Kyoto University Reactor. The converter constituents investigated were carbon for gamma rays, and polyethylene with and without LiF at varied (6)Li concentration for thermal, epithermal, and fast neutrons. Consequently, potential combinations of the converters were found for two components, gamma rays and thermal neutrons, for the standard thermal neutron mode and three components of gamma rays, epithermal neutrons, and thermal or fast neutrons, for the standard mixed or epithermal neutron modes, respectively. Copyright © 2016 Elsevier Ltd. All rights reserved.

Determining organ dose conversion coefficients for external neutron irradiation by using a voxel mouse model

PubMed Central

Zhang, Xiaomin; Xie, Xiangdong; Qu, Decheng; Ning, Jing; Zhou, Hongmei; Pan, Jie; Yang, Guoshan

2016-01-01

A set of fluence-to-dose conversion coefficients has been calculated for neutrons with energies <20 MeV using a developed voxel mouse model and Monte Carlo N-particle code (MCNP), for the purpose of neutron radiation effect evaluation. The calculation used 37 monodirectional monoenergetic neutron beams in the energy range 10−9 MeV to 20 MeV, under five different source irradiation configurations: left lateral, right lateral, dorsal–ventral, ventral–dorsal, and isotropic. Neutron fluence-to-dose conversion coefficients for selected organs of the body were presented in the paper, and the effect of irradiation geometry conditions, neutron energy and the organ location on the organ dose was discussed. The results indicated that neutron dose conversion coefficients clearly show sensitivity to irradiation geometry at neutron energy below 1 MeV. PMID:26661852

Neutronics Phenomena Important in Modeling and Simulation of Liquid-Fuel Molten Salt Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Diamond, David J.

This paper discusses liquid-fuel molten salt reactors, how they will operate under normal, transient, and accident conditions, and the results of an expert elicitation to determine the corresponding neutronic phenomena important to understanding their behavior. Identifying these phenomena will enable the U.S. Nuclear Regulatory Commission (NRC) to develop or identify modeling functionalities and tools required to carry out confirmatory analyses that examine the validity and accuracy of applicants’ calculations and help determine the margin of safety in plant design. NRC frequently does an expert elicitation using a Phenomena Identification and Ranking Table (PIRT) to identify and evaluate the state ofmore » knowledge of important modeling phenomena. However, few details about the design of these reactors and the sequence of events during accidents are known, so the process used was considered a preliminary PIRT. A panel met to define phenomena that would need to be modeled and considered the impact/importance of each phenomenon with respect to specific figures-of-merit (FoMs) (e.g., power distribution, fluence, kinetics parameters and reactivity). Each FoM reflected a potential impact on radionuclide release or loss of a barrier to release. The panel considered what the path forward might be with respect to being able to model the phenomenon in a simulation code. Results are explained for both thermal and fast spectrum designs.« less

Prompt fission neutron emission in the reaction 235U(n,f)

NASA Astrophysics Data System (ADS)

Göök, Alf; Hambsch, Franz-Josef; Oberstedt, Stephan

2018-03-01

Experimental activities at JRC-Geel on prompt fission neutron (PFN) emission in response to OECD/NEA nuclear data requests are presented in this contribution. Specifically, on-going investigations of PFN emission from the reaction 235U(n,f) in the region of the resolved resonances, taking place at the GELINA facility, are presented. The focus of this contribution lies on studies of PFN correlations with fission fragment properties. The experiment employs a scintillation detector array for neutron detection, while fission fragment properties are determined via the double kinetic energy technique using a position sensitive twin ionization chamber. This setup allows us to study several correlations between properties of neutron and fission fragments simultaneously. Results on PFN correlations with fission fragment properties from the present study differ significantly from earlier studies on this reaction, induced by thermal neutrons.

Neutron dose rate analysis on HTGR-10 reactor using Monte Carlo code

NASA Astrophysics Data System (ADS)

Suwoto; Adrial, H.; Hamzah, A.; Zuhair; Bakhri, S.; Sunaryo, G. R.

2018-02-01

The HTGR-10 reactor is cylinder-shaped core fuelled with kernel TRISO coated fuel particles in the spherical pebble with helium cooling system. The outlet helium gas coolant temperature outputted from the reactor core is designed to 700 °C. One advantage HTGR type reactor is capable of co-generation, as an addition to generating electricity, the reactor was designed to produce heat at high temperature can be used for other processes. The spherical fuel pebble contains 8335 TRISO UO2 kernel coated particles with enrichment of 10% and 17% are dispersed in a graphite matrix. The main purpose of this study was to analysis the distribution of neutron dose rates generated from HTGR-10 reactors. The calculation and analysis result of neutron dose rate in the HTGR-10 reactor core was performed using Monte Carlo MCNP5v1.6 code. The problems of double heterogeneity in kernel fuel coated particles TRISO and spherical fuel pebble in the HTGR-10 core are modelled well with MCNP5v1.6 code. The neutron flux to dose conversion factors taken from the International Commission on Radiological Protection (ICRP-74) was used to determine the dose rate that passes through the active core, reflectors, core barrel, reactor pressure vessel (RPV) and a biological shield. The calculated results of neutron dose rate with MCNP5v1.6 code using a conversion factor of ICRP-74 (2009) for radiation workers in the radial direction on the outside of the RPV (radial position = 220 cm from the center of the patio HTGR-10) provides the respective value of 9.22E-4 μSv/h and 9.58E-4 μSv/h for enrichment 10% and 17%, respectively. The calculated values of neutron dose rates are compliant with BAPETEN Chairman’s Regulation Number 4 Year 2013 on Radiation Protection and Safety in Nuclear Energy Utilization which sets the limit value for the average effective dose for radiation workers 20 mSv/year or 10μSv/h. Thus the protection and safety for radiation workers to be safe from the radiation source has been fulfilled. From the result analysis, it can be concluded that the model of calculation result of neutron dose rate for HTGR-10 core has met the required radiation safety standards.

Hauser-Feshbach fission fragment de-excitation with calculated macroscopic-microscopic mass yields

NASA Astrophysics Data System (ADS)

Jaffke, Patrick; Möller, Peter; Talou, Patrick; Sierk, Arnold J.

2018-03-01

The Hauser-Feshbach statistical model is applied to the de-excitation of primary fission fragments using input mass yields calculated with macroscopic-microscopic models of the potential energy surface. We test the sensitivity of the prompt fission observables to the input mass yields for two important reactions, 235U(nth,f ) and 239Pu(nth,f ) , for which good experimental data exist. General traits of the mass yields, such as the location of the peaks and their widths, can impact both the prompt neutron and γ -ray multiplicities, as well as their spectra. Specifically, we use several mass yields to determine a linear correlation between the calculated prompt neutron multiplicity ν ¯ and the average heavy-fragment mass 〈Ah〉 of the input mass yields ∂ ν ¯/∂ 〈Ah〉 =±0.1 (n /f ) /u . The mass peak width influences the correlation between the total kinetic energy of the fission fragments and the total number of prompt neutrons emitted, ν¯T(TKE ) . Typical biases on prompt particle observables from using calculated mass yields instead of experimental ones are δ ν ¯=4 % for the average prompt neutron multiplicity, δ M ¯γ=1 % for the average prompt γ -ray multiplicity, δ ɛ¯nLAB=1 % for the average outgoing neutron energy, δ ɛ¯γ=1 % for the average γ -ray energy, and δ 〈TKE 〉=0.4 % for the average total kinetic energy of the fission fragments.

Reevaluation of secondary neutron spectra from thick targets upon heavy-ion bombardment

NASA Astrophysics Data System (ADS)

Satoh, D.; Kurosawa, T.; Sato, T.; Endo, A.; Takada, M.; Iwase, H.; Nakamura, T.; Niita, K.

2007-12-01

Previously published data of secondary neutron spectra from thick targets of C, Al, Cu and Pb bombarded with heavy ions from He to Xe are revised by using a new set of neutron-detection efficiency values for a liquid organic scintillator calculated with SCINFUL-QMD. Additional data have been measured for bombardment of C target by 400-MeV/nucleon C ions and 800-MeV/nucleon Si ions. The set of spectra are compared with the calculation results using a Monte-Carlo heavy-ion transport code, PHITS. It was found that PHITS is able to reproduce the secondary neutron spectra in a wide neutron-energy regime.

A Bonner Sphere Spectrometer with extended response matrix

NASA Astrophysics Data System (ADS)

Birattari, C.; Dimovasili, E.; Mitaroff, A.; Silari, M.

2010-08-01

This paper describes the design, calibration and applications at high-energy accelerators of an extended-range Bonner Sphere neutron Spectrometer (BSS). The BSS was designed by the FLUKA Monte Carlo code, investigating several combinations of materials and diameters of the moderators for the high-energy channels. The system was calibrated at PTB in Braunschweig, Germany, using monoenergetic neutron beams in the energy range 144 keV-19 MeV. It was subsequently tested with Am-Be source neutrons and in the simulated workplace neutron field at CERF (the CERN-EU high-energy reference field facility). Since 2002, it has been employed for neutron spectral measurements around CERN accelerators.

Filtered epithermal quasi-monoenergetic neutron beams at research reactor facilities.

PubMed

Mansy, M S; Bashter, I I; El-Mesiry, M S; Habib, N; Adib, M

2015-03-01

Filtered neutron techniques were applied to produce quasi-monoenergetic neutron beams in the energy range of 1.5-133keV at research reactors. A simulation study was performed to characterize the filter components and transmitted beam lines. The filtered beams were characterized in terms of the optimal thickness of the main and additive components. The filtered neutron beams had high purity and intensity, with low contamination from the accompanying thermal emission, fast neutrons and γ-rays. A computer code named "QMNB" was developed in the "MATLAB" programming language to perform the required calculations. Copyright © 2014 Elsevier Ltd. All rights reserved.

Recent skyshine calculations at Jefferson Lab

DOE Office of Scientific and Technical Information (OSTI.GOV)

Degtyarenko, P.

1997-12-01

New calculations of the skyshine dose distribution of neutrons and secondary photons have been performed at Jefferson Lab using the Monte Carlo method. The dose dependence on neutron energy, distance to the neutron source, polar angle of a source neutron, and azimuthal angle between the observation point and the momentum direction of a source neutron have been studied. The azimuthally asymmetric term in the skyshine dose distribution is shown to be important in the dose calculations around high-energy accelerator facilities. A parameterization formula and corresponding computer code have been developed which can be used for detailed calculations of the skyshinemore » dose maps.« less

Turbulence dissipation challenge: particle-in-cell simulations

NASA Astrophysics Data System (ADS)

Roytershteyn, V.; Karimabadi, H.; Omelchenko, Y.; Germaschewski, K.

2015-12-01

We discuss application of three particle in cell (PIC) codes to the problems relevant to turbulence dissipation challenge. VPIC is a fully kinetic code extensively used to study a variety of diverse problems ranging from laboratory plasmas to astrophysics. PSC is a flexible fully kinetic code offering a variety of algorithms that can be advantageous to turbulence simulations, including high order particle shapes, dynamic load balancing, and ability to efficiently run on Graphics Processing Units (GPUs). Finally, HYPERS is a novel hybrid (kinetic ions+fluid electrons) code, which utilizes asynchronous time advance and a number of other advanced algorithms. We present examples drawn both from large-scale turbulence simulations and from the test problems outlined by the turbulence dissipation challenge. Special attention is paid to such issues as the small-scale intermittency of inertial range turbulence, mode content of the sub-proton range of scales, the formation of electron-scale current sheets and the role of magnetic reconnection, as well as numerical challenges of applying PIC codes to simulations of astrophysical turbulence.

Progress on the Development of the hPIC Particle-in-Cell Code

NASA Astrophysics Data System (ADS)

Dart, Cameron; Hayes, Alyssa; Khaziev, Rinat; Marcinko, Stephen; Curreli, Davide; Laboratory of Computational Plasma Physics Team

2017-10-01

Advancements were made in the development of the kinetic-kinetic electrostatic Particle-in-Cell code, hPIC, designed for large-scale simulation of the Plasma-Material Interface. hPIC achieved a weak scaling efficiency of 87% using the Algebraic Multigrid Solver BoomerAMG from the PETSc library on more than 64,000 cores of the Blue Waters supercomputer at the University of Illinois at Urbana-Champaign. The code successfully simulates two-stream instability and a volume of plasma over several square centimeters of surface extending out to the presheath in kinetic-kinetic mode. Results from a parametric study of the plasma sheath in strongly magnetized conditions will be presented, as well as a detailed analysis of the plasma sheath structure at grazing magnetic angles. The distribution function and its moments will be reported for plasma species in the simulation domain and at the material surface for plasma sheath simulations. Membership Pending.

Monte Carlo simulations and benchmark measurements on the response of TE(TE) and Mg(Ar) ionization chambers in photon, electron and neutron beams

NASA Astrophysics Data System (ADS)

Lin, Yi-Chun; Huang, Tseng-Te; Liu, Yuan-Hao; Chen, Wei-Lin; Chen, Yen-Fu; Wu, Shu-Wei; Nievaart, Sander; Jiang, Shiang-Huei

2015-06-01

The paired ionization chambers (ICs) technique is commonly employed to determine neutron and photon doses in radiology or radiotherapy neutron beams, where neutron dose shows very strong dependence on the accuracy of accompanying high energy photon dose. During the dose derivation, it is an important issue to evaluate the photon and electron response functions of two commercially available ionization chambers, denoted as TE(TE) and Mg(Ar), used in our reactor based epithermal neutron beam. Nowadays, most perturbation corrections for accurate dose determination and many treatment planning systems are based on the Monte Carlo technique. We used general purposed Monte Carlo codes, MCNP5, EGSnrc, FLUKA or GEANT4 for benchmark verifications among them and carefully measured values for a precise estimation of chamber current from absorbed dose rate of cavity gas. Also, energy dependent response functions of two chambers were calculated in a parallel beam with mono-energies from 20 keV to 20 MeV photons and electrons by using the optimal simple spherical and detailed IC models. The measurements were performed in the well-defined (a) four primary M-80, M-100, M120 and M150 X-ray calibration fields, (b) primary 60Co calibration beam, (c) 6 MV and 10 MV photon, (d) 6 MeV and 18 MeV electron LINACs in hospital and (e) BNCT clinical trials neutron beam. For the TE(TE) chamber, all codes were almost identical over the whole photon energy range. In the Mg(Ar) chamber, MCNP5 showed lower response than other codes for photon energy region below 0.1 MeV and presented similar response above 0.2 MeV (agreed within 5% in the simple spherical model). With the increase of electron energy, the response difference between MCNP5 and other codes became larger in both chambers. Compared with the measured currents, MCNP5 had the difference from the measurement data within 5% for the 60Co, 6 MV, 10 MV, 6 MeV and 18 MeV LINACs beams. But for the Mg(Ar) chamber, the derivations reached 7.8-16.5% below 120 kVp X-ray beams. In this study, we were especially interested in BNCT doses where low energy photon contribution is less to ignore, MCNP model is recognized as the most suitable to simulate wide photon-electron and neutron energy distributed responses of the paired ICs. Also, MCNP provides the best prediction of BNCT source adjustment by the detector's neutron and photon responses.

Investigation on the reflector/moderator geometry and its effect on the neutron beam design in BNCT.

PubMed

Kasesaz, Y; Rahmani, F; Khalafi, H

2015-12-01

In order to provide an appropriate neutron beam for Boron Neutron Capture Therapy (BNCT), a special Beam Shaping Assembly (BSA) must be designed based on the neutron source specifications. A typical BSA includes moderator, reflector, collimator, thermal neutron filter, and gamma filter. In common BSA, the reflector is considered as a layer which covers the sides of the moderator materials. In this paper, new reflector/moderator geometries including multi-layer and hexagonal lattice have been suggested and the effect of them has been investigated by MCNP4C Monte Carlo code. It was found that the proposed configurations have a significant effect to improve the thermal to epithermal neutron flux ratio which is an important neutron beam parameter. Copyright © 2015 Elsevier Ltd. All rights reserved.

On the interpretation of the inverted kinetics equation and space-time calculations of the effectiveness of the VVER-1000 reactor scram system

NASA Astrophysics Data System (ADS)

Zizin, M. N.; Ivanov, L. D.

2013-12-01

In the present paper, an attempt is made to analyze the accuracy of calculating the effectiveness of the VVER-1000 reactor scram system by means of the inverted solution of the kinetics equation (ISKE). In the numerical studies in the intellectual ShIPR software system, the actuation of the reactor scram system with the possible jamming of one of the two most effective rods is simulated. First, the connection of functionals calculated in the space-time computation in different approximations with the kinetics equation is considered on the theoretical level. The formulas are presented in a manner facilitating their coding. Then, the results of processing of several such functions by the ISKE are presented. For estimating the effectiveness of the VVER-1000 reactor scram system, it is proposed to use the measured currents of ionization chambers (IC) jointly with calculated readings of IC imitators. In addition, the integral of the delayed neutron (DN) generation rate multiplied by the adjoint DN source over the volume of the reactor, calculated for the instant of time when insertion of safety rods ends, is used. This integral is necessary for taking into account the spatial reactivity effects. Reasonable agreement was attained for the considered example between the effectiveness of the scram system evaluated by this method and the values obtained by steady-state calculations as the difference of the reciprocal effective multiplication factors with withdrawn and inserted control rods. This agreement was attained with the use of eight-group DN parameters.

Temperature dependence of the kinetic energy in the Zr40Be60 amorphous alloy

NASA Astrophysics Data System (ADS)

Syrykh, G. F.; Stolyarov, A. A.; Krzystyniak, M.; Romanelli, G.; Sadykov, R. A.

2017-05-01

The average kinetic energy < E(T)> of the atomic nucleus for each element of the amorphous alloy Zr40Be60 in the temperature range 10-300 K has been measured for the first time using VESUVIO spectrometer (ISIS). The experimental values of < E(T)> have been compared to the partial ZrBe spectra refined by a recursion method based on the data obtained with thermal neutron scattering. The satisfactory agreement has been reached with the calculations using partial spectra based on thermal neutron spectra obtained with recursion method. In addition, the experimental data have been compared to the Debye model. The measurements at different temperatures (10, 200, and 300 K) will provide an opportunity to evaluate the significance of anharmonicity in the dynamics of metallic glasses.

Feasibility of creating a specialized reactimeter based on the inverse solution to kinetics equation with a current-mode neutron detector

DOE Office of Scientific and Technical Information (OSTI.GOV)

Koshelev, A. S., E-mail: alexsander.coshelev@yandex.ru; Arapov, A. V.; Ovchinnikov, M. A.

2016-12-15

The file-evaluation results of a reactimeter based on the inverse solution to the kinetics equation (ISKE) are presented, which were obtained using an operating hardware-measuring complex with a KNK-4 neutron detector working in the current mode. The processing of power-recording files of the BR-1M, BR-K1, and VIR-2M reactors of the Russian Federal Nuclear Center—All-Russian Research Institute of Experimental Physics, which was performed with the use of Excel simulation of the ISKE formalism, demonstrated the feasibility of implementation of the reactivity monitoring (during the operation of these reactors at stationary power) beginning from the level of ~5 × 10{sup –4}β{sub eff}.

SOC-DS computer code provides tool for design evaluation of homogeneous two-material nuclear shield

NASA Technical Reports Server (NTRS)

Disney, R. K.; Ricks, L. O.

1967-01-01

SOC-DS Code /Shield Optimization Code-Direc Search/, selects a nuclear shield material of optimum volume, weight, or cost to meet the requirments of a given radiation dose rate or energy transmission constraint. It is applicable to evaluating neutron and gamma ray shields for all nuclear reactors.

GCKP84-general chemical kinetics code for gas-phase flow and batch processes including heat transfer effects

NASA Technical Reports Server (NTRS)

Bittker, D. A.; Scullin, V. J.

1984-01-01

A general chemical kinetics code is described for complex, homogeneous ideal gas reactions in any chemical system. The main features of the GCKP84 code are flexibility, convenience, and speed of computation for many different reaction conditions. The code, which replaces the GCKP code published previously, solves numerically the differential equations for complex reaction in a batch system or one dimensional inviscid flow. It also solves numerically the nonlinear algebraic equations describing the well stirred reactor. A new state of the art numerical integration method is used for greatly increased speed in handling systems of stiff differential equations. The theory and the computer program, including details of input preparation and a guide to using the code are given.

Reactivity effects in VVER-1000 of the third unit of the kalinin nuclear power plant at physical start-up. Computations in ShIPR intellectual code system with library of two-group cross sections generated by UNK code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zizin, M. N.; Zimin, V. G.; Zizina, S. N., E-mail: zizin@adis.vver.kiae.ru

2010-12-15

The ShIPR intellectual code system for mathematical simulation of nuclear reactors includes a set of computing modules implementing the preparation of macro cross sections on the basis of the two-group library of neutron-physics cross sections obtained for the SKETCH-N nodal code. This library is created by using the UNK code for 3D diffusion computation of first VVER-1000 fuel loadings. Computation of neutron fields in the ShIPR system is performed using the DP3 code in the two-group diffusion approximation in 3D triangular geometry. The efficiency of all groups of control rods for the first fuel loading of the third unit ofmore » the Kalinin Nuclear Power Plant is computed. The temperature, barometric, and density effects of reactivity as well as the reactivity coefficient due to the concentration of boric acid in the reactor were computed additionally. Results of computations are compared with the experiment.« less

Reactivity effects in VVER-1000 of the third unit of the kalinin nuclear power plant at physical start-up. Computations in ShIPR intellectual code system with library of two-group cross sections generated by UNK code

NASA Astrophysics Data System (ADS)

Zizin, M. N.; Zimin, V. G.; Zizina, S. N.; Kryakvin, L. V.; Pitilimov, V. A.; Tereshonok, V. A.

2010-12-01

The ShIPR intellectual code system for mathematical simulation of nuclear reactors includes a set of computing modules implementing the preparation of macro cross sections on the basis of the two-group library of neutron-physics cross sections obtained for the SKETCH-N nodal code. This library is created by using the UNK code for 3D diffusion computation of first VVER-1000 fuel loadings. Computation of neutron fields in the ShIPR system is performed using the DP3 code in the two-group diffusion approximation in 3D triangular geometry. The efficiency of all groups of control rods for the first fuel loading of the third unit of the Kalinin Nuclear Power Plant is computed. The temperature, barometric, and density effects of reactivity as well as the reactivity coefficient due to the concentration of boric acid in the reactor were computed additionally. Results of computations are compared with the experiment.

THE UNREASONABLE WEAKNESS OF R -PROCESS COSMIC RAYS IN THE NEUTRON-STAR-MERGER NUCLEOSYNTHESIS SCENARIO

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kyutoku, Koutarou; Ioka, Kunihito, E-mail: koutarou.kyutoku@riken.jp

We reach the robust conclusion that, by combining the observed cosmic rays of r -process elements with the fact that the velocity of the neutron-star-merger ejecta is much higher than that of the supernova ejecta, either (1) the reverse shock in the neutron-star-merger ejecta is a very inefficient accelerator that converts less than 0.003% of the ejecta kinetic energy to the cosmic-ray energy or (2) the neutron star merger is not the origin of the Galactic r -process elements. We also find that the acceleration efficiency should be less than 0.1% for the reverse shock of the supernova ejecta withmore » observed cosmic rays lighter than the iron.« less

VENTURE/PC manual: A multidimensional multigroup neutron diffusion code system. Version 3

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shapiro, A.; Huria, H.C.; Cho, K.W.

1991-12-01

VENTURE/PC is a recompilation of part of the Oak Ridge BOLD VENTURE code system, which will operate on an IBM PC or compatible computer. Neutron diffusion theory solutions are obtained for multidimensional, multigroup problems. This manual contains information associated with operating the code system. The purpose of the various modules used in the code system, and the input for these modules are discussed. The PC code structure is also given. Version 2 included several enhancements not given in the original version of the code. In particular, flux iterations can be done in core rather than by reading and writing tomore » disk, for problems which allow sufficient memory for such in-core iterations. This speeds up the iteration process. Version 3 does not include any of the special processors used in the previous versions. These special processors utilized formatted input for various elements of the code system. All such input data is now entered through the Input Processor, which produces standard interface files for the various modules in the code system. In addition, a Standard Interface File Handbook is included in the documentation which is distributed with the code, to assist in developing the input for the Input Processor.« less

New Reduced Two-Time Step Method for Calculating Combustion and Emission Rates of Jet-A and Methane Fuel With and Without Water Injection

NASA Technical Reports Server (NTRS)

Molnar, Melissa; Marek, C. John

2004-01-01

A simplified kinetic scheme for Jet-A, and methane fuels with water injection was developed to be used in numerical combustion codes, such as the National Combustor Code (NCC) or even simple FORTRAN codes that are being developed at Glenn. The two time step method is either an initial time averaged value (step one) or an instantaneous value (step two). The switch is based on the water concentration in moles/cc of 1x10(exp -20). The results presented here results in a correlation that gives the chemical kinetic time as two separate functions. This two step method is used as opposed to a one step time averaged method previously developed to determine the chemical kinetic time with increased accuracy. The first time averaged step is used at the initial times for smaller water concentrations. This gives the average chemical kinetic time as a function of initial overall fuel air ratio, initial water to fuel mass ratio, temperature, and pressure. The second instantaneous step, to be used with higher water concentrations, gives the chemical kinetic time as a function of instantaneous fuel and water mole concentration, pressure and temperature (T4). The simple correlations would then be compared to the turbulent mixing times to determine the limiting properties of the reaction. The NASA Glenn GLSENS kinetics code calculates the reaction rates and rate constants for each species in a kinetic scheme for finite kinetic rates. These reaction rates were then used to calculate the necessary chemical kinetic times. Chemical kinetic time equations for fuel, carbon monoxide and NOx were obtained for Jet-A fuel and methane with and without water injection to water mass loadings of 2/1 water to fuel. A similar correlation was also developed using data from NASA's Chemical Equilibrium Applications (CEA) code to determine the equilibrium concentrations of carbon monoxide and nitrogen oxide as functions of overall equivalence ratio, water to fuel mass ratio, pressure and temperature (T3). The temperature of the gas entering the turbine (T4) was also correlated as a function of the initial combustor temperature (T3), equivalence ratio, water to fuel mass ratio, and pressure.

Design and feasibility of a multi-detector neutron spectrometer for radiation protection applications based on thermoluminescent 6LiF:Ti,Mg (TLD-600) detectors

NASA Astrophysics Data System (ADS)

Lis, M.; Gómez-Ros, J. M.; Bedogni, R.; Delgado, A.

2008-01-01

The design of a neutron detector with spectrometric capability based on thermoluminescent (TL) 6LiF:Ti,Mg (TLD-600) dosimeters located along three perpendicular axis within a single polyethylene (PE) sphere has been analyzed. The neutron response functions have been calculated in the energy range from 10 -8 to 100 MeV with the Monte Carlo (MC) code MCNPX 2.5 and their shape and behaviour have been used to discuss a suitable configuration for an actual instrument. The feasibility of such a device has been preliminary evaluated by the simulation of exposure to 241Am-Be, bare 252Cf and Fe-PE moderated 252Cf sources. The expected accuracy in the evaluation of energy quantities has been evaluated using the unfolding code FRUIT. The obtained results together with additional calculations performed using MAXED and GRAVEL codes show the spectrometric capability of the proposed design for radiation protection applications, especially in the range 1 keV-20 MeV.

Updated User's Guide for Sammy: Multilevel R-Matrix Fits to Neutron Data Using Bayes' Equations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Larson, Nancy M

2008-10-01

In 1980 the multilevel multichannel R-matrix code SAMMY was released for use in analysis of neutron-induced cross section data at the Oak Ridge Electron Linear Accelerator. Since that time, SAMMY has evolved to the point where it is now in use around the world for analysis of many different types of data. SAMMY is not limited to incident neutrons but can also be used for incident protons, alpha particles, or other charged particles; likewise, Coulomb exit hannels can be included. Corrections for a wide variety of experimental conditions are available in the code: Doppler and resolution broadening, multiple-scattering corrections formore » capture or reaction yields, normalizations and backgrounds, to name but a few. The fitting procedure is Bayes' method, and data and parameter covariance matrices are properly treated within the code. Pre- and post-processing capabilities are also available, including (but not limited to) connections with the Evaluated Nuclear Data Files. Though originally designed for use in the resolved resonance region, SAMMY also includes a treatment for data analysis in the unresolved resonance region.« less

ARCADIA{sup R} - A New Generation of Coupled Neutronics / Core Thermal- Hydraulics Code System at AREVA NP

DOE Office of Scientific and Technical Information (OSTI.GOV)

Curca-Tivig, Florin; Merk, Stephan; Pautz, Andreas

2007-07-01

Anticipating future needs of our customers and willing to concentrate synergies and competences existing in the company for the benefit of our customers, AREVA NP decided in 2002 to develop the next generation of coupled neutronics/ core thermal-hydraulic (TH) code systems for fuel assembly and core design calculations for both, PWR and BWR applications. The global CONVERGENCE project was born: after a feasibility study of one year (2002) and a conceptual phase of another year (2003), development was started at the beginning of 2004. The present paper introduces the CONVERGENCE project, presents the main feature of the new code systemmore » ARCADIA{sup R} and concludes on customer benefits. ARCADIA{sup R} is designed to meet AREVA NP market and customers' requirements worldwide. Besides state-of-the-art physical modeling, numerical performance and industrial functionality, the ARCADIA{sup R} system is featuring state-of-the-art software engineering. The new code system will bring a series of benefits for our customers: e.g. improved accuracy for heterogeneous cores (MOX/ UOX, Gd...), better description of nuclide chains, and access to local neutronics/ thermal-hydraulics and possibly thermal-mechanical information (3D pin by pin full core modeling). ARCADIA is a registered trademark of AREVA NP. (authors)« less

Effect of cationic substitution on the double-well hydrogen-bond potential in [K1-x(NH4)x]3H(SO4)2 proton conductors: a single-crystal neutron diffraction study.

PubMed

Choudhury, R R; Chitra, R; Selezneva, E V; Makarova, I P

2017-10-01

The structure of the mixed crystal [K 1-x (NH 4 ) x ] 3 H(SO 4 ) 2 as obtained from single-crystal neutron diffraction is compared with the previously reported room-temperature neutron structure of crystalline K 3 H(SO 4 ) 2 . The two structures are very similar, as indicated by the high value of their isostructurality index (94.8%). It was found that the replacement of even a small amount (3%) of K + with NH 4 + has a significant influence on the short strong hydrogen bond connecting the two SO 4 2- ions. Earlier optical measurements had revealed that the kinetics of the superionic transition in the solid solution [K 1-x (NH 4 ) x ] 3 H(SO 4 ) 2 are much faster than in K 3 H(SO 4 ) 2 ; this reported difference in the kinetics of the superionic phase transition in this class of crystal is explained on the basis of the difference in strength of the hydrogen-bond interactions in the two structures.

Extensions of the MCNP5 and TRIPOLI4 Monte Carlo Codes for Transient Reactor Analysis

NASA Astrophysics Data System (ADS)

Hoogenboom, J. Eduard; Sjenitzer, Bart L.

2014-06-01

To simulate reactor transients for safety analysis with the Monte Carlo method the generation and decay of delayed neutron precursors is implemented in the MCNP5 and TRIPOLI4 general purpose Monte Carlo codes. Important new variance reduction techniques like forced decay of precursors in each time interval and the branchless collision method are included to obtain reasonable statistics for the power production per time interval. For simulation of practical reactor transients also the feedback effect from the thermal-hydraulics must be included. This requires coupling of the Monte Carlo code with a thermal-hydraulics (TH) code, providing the temperature distribution in the reactor, which affects the neutron transport via the cross section data. The TH code also provides the coolant density distribution in the reactor, directly influencing the neutron transport. Different techniques for this coupling are discussed. As a demonstration a 3x3 mini fuel assembly with a moving control rod is considered for MCNP5 and a mini core existing of 3x3 PWR fuel assemblies with control rods and burnable poisons for TRIPOLI4. Results are shown for reactor transients due to control rod movement or withdrawal. The TRIPOLI4 transient calculation is started at low power and includes thermal-hydraulic feedback. The power rises about 10 decades and finally stabilises the reactor power at a much higher level than initial. The examples demonstrate that the modified Monte Carlo codes are capable of performing correct transient calculations, taking into account all geometrical and cross section detail.

Simulation study of accelerator based quasi-mono-energetic epithermal neutron beams for BNCT.

PubMed

Adib, M; Habib, N; Bashter, I I; El-Mesiry, M S; Mansy, M S

2016-01-01

Filtered neutron techniques were applied to produce quasi-mono-energetic neutron beams in the energy range of 1.5-7.5 keV at the accelerator port using the generated neutron spectrum from a Li (p, n) Be reaction. A simulation study was performed to characterize the filter components and transmitted beam lines. The feature of the filtered beams is detailed in terms of optimal thickness of the primary and additive components. A computer code named "QMNB-AS" was developed to carry out the required calculations. The filtered neutron beams had high purity and intensity with low contamination from the accompanying thermal, fast neutrons and γ-rays. Copyright © 2015 Elsevier Ltd. All rights reserved.

LSENS, a general chemical kinetics and sensitivity analysis code for gas-phase reactions: User's guide

NASA Technical Reports Server (NTRS)

Radhakrishnan, Krishnan; Bittker, David A.

1993-01-01

A general chemical kinetics and sensitivity analysis code for complex, homogeneous, gas-phase reactions is described. The main features of the code, LSENS, are its flexibility, efficiency and convenience in treating many different chemical reaction models. The models include static system, steady, one-dimensional, inviscid flow, shock initiated reaction, and a perfectly stirred reactor. In addition, equilibrium computations can be performed for several assigned states. An implicit numerical integration method, which works efficiently for the extremes of very fast and very slow reaction, is used for solving the 'stiff' differential equation systems that arise in chemical kinetics. For static reactions, sensitivity coefficients of all dependent variables and their temporal derivatives with respect to the initial values of dependent variables and/or the rate coefficient parameters can be computed. This paper presents descriptions of the code and its usage, and includes several illustrative example problems.

Hybrid model for simulation of plasma jet injection in tokamak

NASA Astrophysics Data System (ADS)

Galkin, Sergei A.; Bogatu, I. N.

2016-10-01

Hybrid kinetic model of plasma treats the ions as kinetic particles and the electrons as charge neutralizing massless fluid. The model is essentially applicable when most of the energy is concentrated in the ions rather than in the electrons, i.e. it is well suited for the high-density hyper-velocity C60 plasma jet. The hybrid model separates the slower ion time scale from the faster electron time scale, which becomes disregardable. That is why hybrid codes consistently outperform the traditional PIC codes in computational efficiency, still resolving kinetic ions effects. We discuss 2D hybrid model and code with exact energy conservation numerical algorithm and present some results of its application to simulation of C60 plasma jet penetration through tokamak-like magnetic barrier. We also examine the 3D model/code extension and its possible applications to tokamak and ionospheric plasmas. The work is supported in part by US DOE DE-SC0015776 Grant.

Impact of nonabsorbing control rod tips on kinetics feedback for BWR turbine trip without bypass RETRAN-03 analyses

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feltus, M.A.; Knerr, R.; Shoop, U.

1993-01-01

RETRAN-03 studies were performed for the boiling water reactor (BWR) turbine trip without bypass (TTWOB) event to investigate how the non-neutron-absorbing material on control rod tips affect scram delay timing and reactivity feedback. Scram delay, Doppler temperature, and moderator void (density) feedback were varied to assess their relative impact on kinetics behavior. Although a generic point-kinetics RETRAN-03 TTWOB model 2 was employed, actual plant information was used to develop the basic and parametric cases.

A Comparison of Monte Carlo and Deterministic Solvers for keff and Sensitivity Calculations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Haeck, Wim; Parsons, Donald Kent; White, Morgan Curtis

Verification and validation of our solutions for calculating the neutron reactivity for nuclear materials is a key issue to address for many applications, including criticality safety, research reactors, power reactors, and nuclear security. Neutronics codes solve variations of the Boltzmann transport equation. The two main variants are Monte Carlo versus deterministic solutions, e.g. the MCNP [1] versus PARTISN [2] codes, respectively. There have been many studies over the decades that examined the accuracy of such solvers and the general conclusion is that when the problems are well-posed, either solver can produce accurate results. However, the devil is always in themore » details. The current study examines the issue of self-shielding and the stress it puts on deterministic solvers. Most Monte Carlo neutronics codes use continuous-energy descriptions of the neutron interaction data that are not subject to this effect. The issue of self-shielding occurs because of the discretisation of data used by the deterministic solutions. Multigroup data used in these solvers are the average cross section and scattering parameters over an energy range. Resonances in cross sections can occur that change the likelihood of interaction by one to three orders of magnitude over a small energy range. Self-shielding is the numerical effect that the average cross section in groups with strong resonances can be strongly affected as neutrons within that material are preferentially absorbed or scattered out of the resonance energies. This affects both the average cross section and the scattering matrix.« less

LSENS, A General Chemical Kinetics and Sensitivity Analysis Code for Homogeneous Gas-Phase Reactions. Part 2; Code Description and Usage

NASA Technical Reports Server (NTRS)

Radhakrishnan, Krishnan; Bittker, David A.

1994-01-01

LSENS, the Lewis General Chemical Kinetics and Sensitivity Analysis Code, has been developed for solving complex, homogeneous, gas-phase chemical kinetics problems and contains sensitivity analysis for a variety of problems, including nonisothermal situations. This report is part II of a series of three reference publications that describe LSENS, provide a detailed guide to its usage, and present many example problems. Part II describes the code, how to modify it, and its usage, including preparation of the problem data file required to execute LSENS. Code usage is illustrated by several example problems, which further explain preparation of the problem data file and show how to obtain desired accuracy in the computed results. LSENS is a flexible, convenient, accurate, and efficient solver for chemical reaction problems such as static system; steady, one-dimensional, inviscid flow; reaction behind incident shock wave, including boundary layer correction; and perfectly stirred (highly backmixed) reactor. In addition, the chemical equilibrium state can be computed for the following assigned states: temperature and pressure, enthalpy and pressure, temperature and volume, and internal energy and volume. For static problems the code computes the sensitivity coefficients of the dependent variables and their temporal derivatives with respect to the initial values of the dependent variables and/or the three rate coefficient parameters of the chemical reactions. Part I (NASA RP-1328) derives the governing equations and describes the numerical solution procedures for the types of problems that can be solved by LSENS. Part III (NASA RP-1330) explains the kinetics and kinetics-plus-sensitivity-analysis problems supplied with LSENS and presents sample results.

GPU-accelerated atmospheric chemical kinetics in the ECHAM/MESSy (EMAC) Earth system model (version 2.52)

NASA Astrophysics Data System (ADS)

Alvanos, Michail; Christoudias, Theodoros

2017-10-01

This paper presents an application of GPU accelerators in Earth system modeling. We focus on atmospheric chemical kinetics, one of the most computationally intensive tasks in climate-chemistry model simulations. We developed a software package that automatically generates CUDA kernels to numerically integrate atmospheric chemical kinetics in the global climate model ECHAM/MESSy Atmospheric Chemistry (EMAC), used to study climate change and air quality scenarios. A source-to-source compiler outputs a CUDA-compatible kernel by parsing the FORTRAN code generated by the Kinetic PreProcessor (KPP) general analysis tool. All Rosenbrock methods that are available in the KPP numerical library are supported.Performance evaluation, using Fermi and Pascal CUDA-enabled GPU accelerators, shows achieved speed-ups of 4. 5 × and 20. 4 × , respectively, of the kernel execution time. A node-to-node real-world production performance comparison shows a 1. 75 × speed-up over the non-accelerated application using the KPP three-stage Rosenbrock solver. We provide a detailed description of the code optimizations used to improve the performance including memory optimizations, control code simplification, and reduction of idle time. The accuracy and correctness of the accelerated implementation are evaluated by comparing to the CPU-only code of the application. The median relative difference is found to be less than 0.000000001 % when comparing the output of the accelerated kernel the CPU-only code.The approach followed, including the computational workload division, and the developed GPU solver code can potentially be used as the basis for hardware acceleration of numerous geoscientific models that rely on KPP for atmospheric chemical kinetics applications.

Time Evolving Fission Chain Theory and Fast Neutron and Gamma-Ray Counting Distributions

DOE PAGES

Kim, K. S.; Nakae, L. F.; Prasad, M. K.; ...

2015-11-01

Here, we solve a simple theoretical model of time evolving fission chains due to Feynman that generalizes and asymptotically approaches the point model theory. The point model theory has been used to analyze thermal neutron counting data. This extension of the theory underlies fast counting data for both neutrons and gamma rays from metal systems. Fast neutron and gamma-ray counting is now possible using liquid scintillator arrays with nanosecond time resolution. For individual fission chains, the differential equations describing three correlated probability distributions are solved: the time-dependent internal neutron population, accumulation of fissions in time, and accumulation of leaked neutronsmore » in time. Explicit analytic formulas are given for correlated moments of the time evolving chain populations. The equations for random time gate fast neutron and gamma-ray counting distributions, due to randomly initiated chains, are presented. Correlated moment equations are given for both random time gate and triggered time gate counting. There are explicit formulas for all correlated moments are given up to triple order, for all combinations of correlated fast neutrons and gamma rays. The nonlinear differential equations for probabilities for time dependent fission chain populations have a remarkably simple Monte Carlo realization. A Monte Carlo code was developed for this theory and is shown to statistically realize the solutions to the fission chain theory probability distributions. Combined with random initiation of chains and detection of external quanta, the Monte Carlo code generates time tagged data for neutron and gamma-ray counting and from these data the counting distributions.« less

BOXER: Fine-flux Cross Section Condensation, 2D Few Group Diffusion and Transport Burnup Calculations

DOE Office of Scientific and Technical Information (OSTI.GOV)

2010-02-01

Neutron transport, calculation of multiplication factor and neutron fluxes in 2-D configurations: cell calculations, 2-D diffusion and transport, and burnup. Preparation of a cross section library for the code BOXER from a basic library in ENDF/B format (ETOBOX).

Neutron Fluence and Energy Reconstruction with the LNE-IRSN/MIMAC Recoil Detector MicroTPC at 27 keV

DOE Office of Scientific and Technical Information (OSTI.GOV)

Maire, D.; Lebreton, L.; Querre, Ph.

2015-07-01

The French Institute for Radiation protection and Nuclear Safety (IRSN), designated by the French Metrology Institute (LNE) for neutron metrology, is developing a time projection chamber using a Micromegas anode: microTPC. This work is carried out in collaboration with the Laboratory of Subatomic Physics and Cosmology (LPSC). The aim is to characterize the energy distribution of neutron fluence in the energy range 8 keV - 5 MeV with a primary procedure. The time projection chambers are gaseous detectors able to measure charged particles energy and to reconstruct their track if a pixelated anode is used. In our case, the gasmore » is used as a (n, p) converter in order to detect neutrons down to few keV. Coming from elastic collisions with neutrons, recoil protons lose a part of their kinetic energy by ionizing the gas. The ionization electrons are drifted toward a pixelated anode (2D projection), read at 50 MHz by a self-triggered electronic system to obtain the third track dimension. The neutron energy is reconstructed event by event thanks to proton scattering angle and proton energy measurements. The scattering angle is deduced from the 3D track. The proton energy is obtained by charge collection measurements, knowing the ionization quenching factor (i.e. the part of proton kinetic energy lost by ionizing the gas). The fluence is calculated thanks to the detected events number and the simulation of the detector response. The μTPC is a new reliable detector able to measure energy distribution of the neutron fluence without unfolding procedure or prior neutron calibration contrary to usual gaseous counters. The microTPC is still being developed and measurements have been carried out at the AMANDE facility, with neutrons energies going from 8 keV to 565 keV. After the context and the μ-TPC working principle presentation, measurements of the neutron energy and fluence at 27 keV and 144 keV are shown and compared to the complete detector response simulation. This work shows the first direct reconstruction of neutron energy and fluence, simultaneously, at 27.2 keV in a continuous irradiation mode. (authors)« less

GLSENS: A Generalized Extension of LSENS Including Global Reactions and Added Sensitivity Analysis for the Perfectly Stirred Reactor

NASA Technical Reports Server (NTRS)

Bittker, David A.

1996-01-01

A generalized version of the NASA Lewis general kinetics code, LSENS, is described. The new code allows the use of global reactions as well as molecular processes in a chemical mechanism. The code also incorporates the capability of performing sensitivity analysis calculations for a perfectly stirred reactor rapidly and conveniently at the same time that the main kinetics calculations are being done. The GLSENS code has been extensively tested and has been found to be accurate and efficient. Nine example problems are presented and complete user instructions are given for the new capabilities. This report is to be used in conjunction with the documentation for the original LSENS code.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, K. S.; Nakae, L. F.; Prasad, M. K.

Here, we solve a simple theoretical model of time evolving fission chains due to Feynman that generalizes and asymptotically approaches the point model theory. The point model theory has been used to analyze thermal neutron counting data. This extension of the theory underlies fast counting data for both neutrons and gamma rays from metal systems. Fast neutron and gamma-ray counting is now possible using liquid scintillator arrays with nanosecond time resolution. For individual fission chains, the differential equations describing three correlated probability distributions are solved: the time-dependent internal neutron population, accumulation of fissions in time, and accumulation of leaked neutronsmore » in time. Explicit analytic formulas are given for correlated moments of the time evolving chain populations. The equations for random time gate fast neutron and gamma-ray counting distributions, due to randomly initiated chains, are presented. Correlated moment equations are given for both random time gate and triggered time gate counting. There are explicit formulas for all correlated moments are given up to triple order, for all combinations of correlated fast neutrons and gamma rays. The nonlinear differential equations for probabilities for time dependent fission chain populations have a remarkably simple Monte Carlo realization. A Monte Carlo code was developed for this theory and is shown to statistically realize the solutions to the fission chain theory probability distributions. Combined with random initiation of chains and detection of external quanta, the Monte Carlo code generates time tagged data for neutron and gamma-ray counting and from these data the counting distributions.« less

Unsteady Plasma Ejections from Hollow Accretion Columns of Galactic Neutron Stars as a Trigger for Gamma-Ray Bursts

NASA Astrophysics Data System (ADS)

Gvaramadze, V. V.

1995-09-01

We propose a model of gamma-ray bursts (GRBs) based on close Galactic neutron stars with accretion disks. We outline a simple mechanism of unsteady plasma ejections during episodic accretion events. The relative kinetic energy of ejected blobs can be converted into gamma-rays by internal shocks. The beaming of gamma-ray emission can be responsible for the observed isotropic angular distribution of GRBs.

Development of a fully-consistent reduced order model to study instabilities in boiling water reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dykin, V.; Demaziere, C.

2012-07-01

A simple nonlinear Reduced Order Model to study global, regional and local instabilities in Boiling Water Reactors is described. The ROM consists of three submodels: neutron-kinetic, thermal-hydraulic and heat-transfer models. The neutron-kinetic model allows representing the time evolution of the three first neutron kinetic modes: the fundamental, the first and the second azimuthal modes. The thermal-hydraulic model describes four heated channels in order to correctly simulate out-of-phase behavior. The coupling between the different submodels is performed via both void and Doppler feedback mechanisms. After proper spatial homogenization, the governing equations are discretized in the time-domain. Several modifications, compared to othermore » existing ROMs, have been implemented, and are reported in this paper. One novelty of the ROM is the inclusion of both azimuthal modes, which allows to study combined instabilities (in-phase and out-of-phase), as well as to investigate the corresponding interference effects between them. The second modification concerns the precise estimation of so-called reactivity coefficients or C{sub mn}{sup *V,D} - coefficients by using direct cross-section data from SIMULATE-3 combined with the CORE SIM core simulator in order to calculate Eigenmodes. Furthermore, a non-uniform two-step axial power profile is introduced to simulate the separate heat production in the single and two-phase regions, respectively. An iterative procedure was developed to calculate the solution to the coupled neutron-kinetic/thermal-hydraulic static problem prior to solving the time-dependent problem. Besides, the possibility of taking into account the effect of local instabilities is demonstrated in a simplified manner. The present ROM is applied to the investigation of an actual instability that occurred at the Swedish Forsmark-1 BWR in 1996/1997. The results generated by the ROM are compared with real power plant measurements performed during stability tests and show a good qualitative agreement. The present study provides some insight in a deeper understanding of the physical principles which drive both core-wide and local instabilities. (authors)« less

Spectral unfolding of fast neutron energy distributions

NASA Astrophysics Data System (ADS)

Mosby, Michelle; Jackman, Kevin; Engle, Jonathan

2015-10-01

The characterization of the energy distribution of a neutron flux is difficult in experiments with constrained geometry where techniques such as time of flight cannot be used to resolve the distribution. The measurement of neutron fluxes in reactors, which often present similar challenges, has been accomplished using radioactivation foils as an indirect probe. Spectral unfolding codes use statistical methods to adjust MCNP predictions of neutron energy distributions using quantified radioactive residuals produced in these foils. We have applied a modification of this established neutron flux characterization technique to experimentally characterize the neutron flux in the critical assemblies at the Nevada National Security Site (NNSS) and the spallation neutron flux at the Isotope Production Facility (IPF) at Los Alamos National Laboratory (LANL). Results of the unfolding procedure are presented and compared with a priori MCNP predictions, and the implications for measurements using the neutron fluxes at these facilities are discussed.

Measurement and simulation for a complementary imaging with the neutron and X-ray beams

NASA Astrophysics Data System (ADS)

Hara, Kaoru Y.; Sato, Hirotaka; Kamiyama, Takashi; Shinohara, Takenao

2017-09-01

By using a composite source system, we measured radiographs of the thermal neutron and keV X-ray in the 45-MeV electron linear accelerator facility at Hokkaido University. The source system provides the alternative beam of neutron and X-ray by switching the production target onto the electron beam axis. In the measurement to demonstrate a complementary imaging, the detector based on a vacuum-tube type neutron color image intensifier was applied to the both beams for dual-purpose. On the other hand, for reducing background in a neutron transmission spectrum, test measurements using a gadolinium-type neutron grid were performed with a cold neutron source at Hokkaido University. In addition, the simulations of the neutron and X-ray transmissions for various substances were performed using the PHITS code. A data analysis procedure for estimating the substance of sample was investigated through the simulations.

PENTrack - a versatile Monte Carlo tool for ultracold neutron sources and experiments

NASA Astrophysics Data System (ADS)

Picker, Ruediger; Chahal, Sanmeet; Christopher, Nicolas; Losekamm, Martin; Marcellin, James; Paul, Stephan; Schreyer, Wolfgang; Yapa, Pramodh

2016-09-01

Ultracold neutrons have energies in the hundred nano eV region. They can be stored in traps for hundreds of seconds. This makes them the ideal tool to study the neutron itself. Measurements of neutron decay correlations, lifetime or electric dipole moment are ideally suited for ultracold neutrons, as well as experiments probing the neutron's gravitational levels in the earth's field. We have developed a Monte Carlo simulation tool that can serve to design and optimize these experiments, and possibly correct results: PENTrack is a C++ based simulation code that tracks neutrons, protons and electrons or atoms, as well as their spins, in gravitational and electromagnetic fields. In addition wall interactions of neutrons due to strong interaction are modeled with a Fermi-potential formalism and take surface roughness into account. The presentation will introduce the physics behind the simulation and provide examples of its application.

The fast neutron fluence and the activation detector activity calculations using the effective source method and the adjoint function

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hep, J.; Konecna, A.; Krysl, V.

2011-07-01

This paper describes the application of effective source in forward calculations and the adjoint method to the solution of fast neutron fluence and activation detector activities in the reactor pressure vessel (RPV) and RPV cavity of a VVER-440 reactor. Its objective is the demonstration of both methods on a practical task. The effective source method applies the Boltzmann transport operator to time integrated source data in order to obtain neutron fluence and detector activities. By weighting the source data by time dependent decay of the detector activity, the result of the calculation is the detector activity. Alternatively, if the weightingmore » is uniform with respect to time, the result is the fluence. The approach works because of the inherent linearity of radiation transport in non-multiplying time-invariant media. Integrated in this way, the source data are referred to as the effective source. The effective source in the forward calculations method thereby enables the analyst to replace numerous intensive transport calculations with a single transport calculation in which the time dependence and magnitude of the source are correctly represented. In this work, the effective source method has been expanded slightly in the following way: neutron source data were performed with few group method calculation using the active core calculation code MOBY-DICK. The follow-up neutron transport calculation was performed using the neutron transport code TORT to perform multigroup calculations. For comparison, an alternative method of calculation has been used based upon adjoint functions of the Boltzmann transport equation. Calculation of the three-dimensional (3-D) adjoint function for each required computational outcome has been obtained using the deterministic code TORT and the cross section library BGL440. Adjoint functions appropriate to the required fast neutron flux density and neutron reaction rates have been calculated for several significant points within the RPV and RPV cavity of the VVER-440 reacto rand located axially at the position of maximum power and at the position of the weld. Both of these methods (the effective source and the adjoint function) are briefly described in the present paper. The paper also describes their application to the solution of fast neutron fluence and detectors activities for the VVER-440 reactor. (authors)« less

Extension to Higher Mass Numbers of an Improved Knockout-Ablation-Coalescence Model for Secondary Neutron and Light Ion Production in Cosmic Ray Interactions

NASA Astrophysics Data System (ADS)

Indi Sriprisan, Sirikul; Townsend, Lawrence; Cucinotta, Francis A.; Miller, Thomas M.

Purpose: An analytical knockout-ablation-coalescence model capable of making quantitative predictions of the neutron spectra from high-energy nucleon-nucleus and nucleus-nucleus collisions is being developed for use in space radiation protection studies. The FORTRAN computer code that implements this model is called UBERNSPEC. The knockout or abrasion stage of the model is based on Glauber multiple scattering theory. The ablation part of the model uses the classical evaporation model of Weisskopf-Ewing. In earlier work, the knockout-ablation model has been extended to incorporate important coalescence effects into the formalism. Recently, alpha coalescence has been incorporated, and the ability to predict light ion spectra with the coalescence model added. The earlier versions were limited to nuclei with mass numbers less than 69. In this work, the UBERNSPEC code has been extended to make predictions of secondary neutrons and light ion production from the interactions of heavy charged particles with higher mass numbers (as large as 238). The predictions are compared with published measurements of neutron spectra and light ion energy for a variety of collision pairs. Furthermore, the predicted spectra from this work are compared with the predictions from the recently-developed heavy ion event generator incorporated in the Monte Carlo radiation transport code HETC-HEDS.

Microdosimetric evaluation of the neutron field for BNCT at Kyoto University reactor by using the PHITS code.

PubMed

Baba, H; Onizuka, Y; Nakao, M; Fukahori, M; Sato, T; Sakurai, Y; Tanaka, H; Endo, S

2011-02-01

In this study, microdosimetric energy distributions of secondary charged particles from the (10)B(n,α)(7)Li reaction in boron-neutron capture therapy (BNCT) field were calculated using the Particle and Heavy Ion Transport code System (PHITS). The PHITS simulation was performed to reproduce the geometrical set-up of an experiment that measured the microdosimetric energy distributions at the Kyoto University Reactor where two types of tissue-equivalent proportional counters were used, one with A-150 wall alone and another with a 50-ppm-boron-loaded A-150 wall. It was found that the PHITS code is a useful tool for the simulation of the energy deposited in tissue in BNCT based on the comparisons with experimental results.

Enhancements to the MCNP6 background source

DOE PAGES

McMath, Garrett E.; McKinney, Gregg W.

2015-10-19

The particle transport code MCNP has been used to produce a background radiation data file on a worldwide grid that can easily be sampled as a source in the code. Location-dependent cosmic showers were modeled by Monte Carlo methods to produce the resulting neutron and photon background flux at 2054 locations around Earth. An improved galactic-cosmic-ray feature was used to model the source term as well as data from multiple sources to model the transport environment through atmosphere, soil, and seawater. A new elevation scaling feature was also added to the code to increase the accuracy of the cosmic neutronmore » background for user locations with off-grid elevations. Furthermore, benchmarking has shown the neutron integral flux values to be within experimental error.« less

Verification of MCNP simulation of neutron flux parameters at TRIGA MK II reactor of Malaysia.

PubMed

Yavar, A R; Khalafi, H; Kasesaz, Y; Sarmani, S; Yahaya, R; Wood, A K; Khoo, K S

2012-10-01

A 3-D model for 1 MW TRIGA Mark II research reactor was simulated. Neutron flux parameters were calculated using MCNP-4C code and were compared with experimental results obtained by k(0)-INAA and absolute method. The average values of φ(th),φ(epi), and φ(fast) by MCNP code were (2.19±0.03)×10(12) cm(-2)s(-1), (1.26±0.02)×10(11) cm(-2)s(-1) and (3.33±0.02)×10(10) cm(-2)s(-1), respectively. These average values were consistent with the experimental results obtained by k(0)-INAA. The findings show a good agreement between MCNP code results and experimental results. Copyright © 2012 Elsevier Ltd. All rights reserved.

An integrated radiation physics computer code system.

NASA Technical Reports Server (NTRS)

Steyn, J. J.; Harris, D. W.

1972-01-01

An integrated computer code system for the semi-automatic and rapid analysis of experimental and analytic problems in gamma photon and fast neutron radiation physics is presented. Such problems as the design of optimum radiation shields and radioisotope power source configurations may be studied. The system codes allow for the unfolding of complex neutron and gamma photon experimental spectra. Monte Carlo and analytic techniques are used for the theoretical prediction of radiation transport. The system includes a multichannel pulse-height analyzer scintillation and semiconductor spectrometer coupled to an on-line digital computer with appropriate peripheral equipment. The system is geometry generalized as well as self-contained with respect to material nuclear cross sections and the determination of the spectrometer response functions. Input data may be either analytic or experimental.

Optimization of coupled multiphysics methodology for safety analysis of pebble bed modular reactor

NASA Astrophysics Data System (ADS)

Mkhabela, Peter Tshepo

The research conducted within the framework of this PhD thesis is devoted to the high-fidelity multi-physics (based on neutronics/thermal-hydraulics coupling) analysis of Pebble Bed Modular Reactor (PBMR), which is a High Temperature Reactor (HTR). The Next Generation Nuclear Plant (NGNP) will be a HTR design. The core design and safety analysis methods are considerably less developed and mature for HTR analysis than those currently used for Light Water Reactors (LWRs). Compared to LWRs, the HTR transient analysis is more demanding since it requires proper treatment of both slower and much longer transients (of time scale in hours and days) and fast and short transients (of time scale in minutes and seconds). There is limited operation and experimental data available for HTRs for validation of coupled multi-physics methodologies. This PhD work developed and verified reliable high fidelity coupled multi-physics models subsequently implemented in robust, efficient, and accurate computational tools to analyse the neutronics and thermal-hydraulic behaviour for design optimization and safety evaluation of PBMR concept The study provided a contribution to a greater accuracy of neutronics calculations by including the feedback from thermal hydraulics driven temperature calculation and various multi-physics effects that can influence it. Consideration of the feedback due to the influence of leakage was taken into account by development and implementation of improved buckling feedback models. Modifications were made in the calculation procedure to ensure that the xenon depletion models were accurate for proper interpolation from cross section tables. To achieve this, the NEM/THERMIX coupled code system was developed to create the system that is efficient and stable over the duration of transient calculations that last over several tens of hours. Another achievement of the PhD thesis was development and demonstration of full-physics, three-dimensional safety analysis methodology for the PBMR to provide reference solutions. Investigation of different aspects of the coupled methodology and development of efficient kinetics treatment for the PBMR were carried out, which accounts for all feedback phenomena in an efficient manner. The OECD/NEA PBMR-400 coupled code benchmark was used as a test matrix for the proposed investigations. The integrated thermal-hydraulics and neutronics (multi-physics) methods were extended to enable modeling of a wider range of transients pertinent to the PBMR. First, the effect of the spatial mapping schemes (spatial coupling) was studied and quantified for different types of transients, which resulted in implementation of improved mapping methodology based on user defined criteria. The second aspect that was studied and optimized is the temporal coupling and meshing schemes between the neutronics and thermal-hydraulics time step selection algorithms. The coupled code convergence was achieved supplemented by application of methods to accelerate it. Finally, the modeling of all feedback phenomena in PBMRs was investigated and a novel treatment of cross-section dependencies was introduced for improving the representation of cross-section variations. The added benefit was that in the process of studying and improving the coupled multi-physics methodology more insight was gained into the physics and dynamics of PBMR, which will help also to optimize the PBMR design and improve its safety. One unique contribution of the PhD research is the investigation of the importance of the correct representation of the three-dimensional (3-D) effects in the PBMR analysis. The performed studies demonstrated that explicit 3-D modeling of control rod movement is superior and removes the errors associated with the grey curtain (2-D homogenized) approximation.

Monte Carlo based dosimetry for neutron capture therapy of brain tumors

NASA Astrophysics Data System (ADS)

Zaidi, Lilia; Belgaid, Mohamed; Khelifi, Rachid

2016-11-01

Boron Neutron Capture Therapy (BNCT) is a biologically targeted, radiation therapy for cancer which combines neutron irradiation with a tumor targeting agent labeled with a boron10 having a high thermal neutron capture cross section. The tumor area is subjected to the neutron irradiation. After a thermal neutron capture, the excited 11B nucleus fissions into an alpha particle and lithium recoil nucleus. The high Linear Energy Transfer (LET) emitted particles deposit their energy in a range of about 10μm, which is of the same order of cell diameter [1], at the same time other reactions due to neutron activation with body component are produced. In-phantom measurement of physical dose distribution is very important for BNCT planning validation. Determination of total absorbed dose requires complex calculations which were carried out using the Monte Carlo MCNP code [2].

Advanced setup for high-pressure and low-temperature neutron diffraction at hydrostatic conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lokshin, Konstantin A.; Zhao Yusheng

2005-06-15

We describe a design of the experimental setup for neutron diffraction studies at low temperatures and hydrostatic pressure. The significant benefit of the setup, compared to the previous methods, is that it makes possible the simultaneous collection of neutrons diffracted at the 30 deg. -150 deg. range with no contamination by the primary scattering from the sample surroundings and without cutting out the incident and diffracted beams. The suggested design is most useful for third-generation time-of-flight diffractometers and constant wavelength instruments. Application of the setup expands the capabilities of high-pressure neutron diffraction, allowing time-resolved kinetics and structural studies, multihistogram Rietveld,more » and pair distribution function and texture analyses. The high efficiency of the setup was proven for the HIPPO diffractometer at Los Alamos Neutron Science Center under pressures up to 10 kbar and temperatures from 4 to 300 K.« less

Neutron skyshine from end stations of the Continuous Electron Beam Accelerator Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sun, Rai-Ko S.

1991-12-01

The MORSE{_}CG code from Oak Ridge National Laboratory was applied to the estimation of the neutron skyshine from three end stations of the Continuous Electron Beam Accelerator Facility (CEBAF), Newport News, VA. Calculations with other methods and an experiment had been directed at assessing the annual neutron dose equivalent at the site boundary. A comparison of results obtained with different methods is given, and the effect of different temperatures and humidities will be discussed.

Neutron skyshine from end stations of the Continuous Electron Beam Accelerator Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sun, Rai-Ko S.

1991-12-01

The MORSE{ }CG code from Oak Ridge National Laboratory was applied to the estimation of the neutron skyshine from three end stations of the Continuous Electron Beam Accelerator Facility (CEBAF), Newport News, VA. Calculations with other methods and an experiment had been directed at assessing the annual neutron dose equivalent at the site boundary. A comparison of results obtained with different methods is given, and the effect of different temperatures and humidities will be discussed.

Investigating Prompt Fission Neutron Emission from 235U(n,f) in the Resolved Resonance Region

NASA Astrophysics Data System (ADS)

Göök, Alf; Hambsch, Franz-Josef; Oberstedt, Stephan

2016-03-01

Investigations of prompt emission in fission is of importance in understanding the fission process in general and the sharing of excitation energy among the fission fragments in particular. Experimental activities at IRMM on prompt neutron emission from fission in response to OECD/NEA nuclear data requests is presented in this contribution. Main focus lies on currently on-going investigations of prompt neutron emission from the reaction 235U(n,f) in the region of the resolved resonances. For this reaction strong fluctuations of fission fragment mass distributions and mean total kinetic energy have been observed [Nucl. Phys. A 491, 56 (1989)] as a function of incident neutron energy in the resonance region. In addition fluctuations of prompt neutron multiplicities were also observed [Phys. Rev. C 13, 195 (1976)]. The goal of the present study is to verify the current knowledge of prompt neutron multiplicity fluctuations and to study correlations with fission fragment properties.

Neutrons on a surface of liquid helium

NASA Astrophysics Data System (ADS)

Grigoriev, P. D.; Zimmer, O.; Grigoriev, A. D.; Ziman, T.

2016-08-01

We investigate the possibility of ultracold neutron (UCN) storage in quantum states defined by the combined potentials of the Earth's gravity and the neutron optical repulsion by a horizontal surface of liquid helium. We analyze the stability of the lowest quantum state, which is most susceptible to perturbations due to surface excitations, against scattering by helium atoms in the vapor and by excitations of the liquid, comprised of ripplons, phonons, and surfons. This is an unusual scattering problem since the kinetic energy of the neutron parallel to the surface may be much greater than the binding energies perpendicular. The total scattering time of these UCNs at 0.7 K is found to exceed 1 h, and rapidly increases with decreasing temperature. Such low scattering rates should enable high-precision measurements of the sequence of discrete energy levels, thus providing improved tests of short-range gravity. The system might also be useful for neutron β -decay experiments. We also sketch new experimental propositions for level population and trapping of ultracold neutrons above a flat horizontal mirror.

Study on detecting spatial distribution of neutrons and gamma rays using a multi-imaging plate system.

PubMed

Tanaka, Kenichi; Sakurai, Yoshinori; Endo, Satoru; Takada, Jun

2014-06-01

In order to measure the spatial distributions of neutrons and gamma rays separately using the imaging plate, the requirement for the converter to enhance specific component was investigated with the PHITS code. Consequently, enhancing fast neutrons using recoil protons from epoxy resin was not effective due to high sensitivity of the imaging plate to gamma rays. However, the converter of epoxy resin doped with (10)B was found to have potential for thermal and epithermal neutrons, and graphite for gamma rays. Copyright © 2014 Elsevier Ltd. All rights reserved.

Modelisation and distribution of neutron flux in radium-beryllium source (226Ra-Be)

NASA Astrophysics Data System (ADS)

Didi, Abdessamad; Dadouch, Ahmed; Jai, Otman

2017-09-01

Using the Monte Carlo N-Particle code (MCNP-6), to analyze the thermal, epithermal and fast neutron fluxes, of 3 millicuries of radium-beryllium, for determine the qualitative and quantitative of many materials, using method of neutron activation analysis. Radium-beryllium source of neutron is established to practical work and research in nuclear field. The main objective of this work was to enable us harness the profile flux of radium-beryllium irradiation, this theoretical study permits to discuss the design of the optimal irradiation and performance for increased the facility research and education of nuclear physics.

Distributions of neutron yields and doses around a water phantom bombarded with 290-MeV/nucleon and 430-MeV/nucleon carbon ions

NASA Astrophysics Data System (ADS)

Satoh, D.; Kajimoto, T.; Shigyo, N.; Itashiki, Y.; Imabayashi, Y.; Koba, Y.; Matsufuji, N.; Sanami, T.; Nakao, N.; Uozumi, Y.

2016-11-01

Double-differential neutron yields from a water phantom bombarded with 290-MeV/nucleon and 430-MeV/nucleon carbon ions were measured at emission angles of 15°, 30°, 45°, 60°, 75°, and 90°, and angular distributions of neutron yields and doses around the phantom were obtained. The experimental data were compared with results of the Monte-Carlo simulation code PHITS. The PHITS results showed good agreement with the measured data. On the basis of the PHITS simulation, we estimated the angular distributions of neutron yields and doses from 0° to 180° including thermal neutrons.

NEUTRON PHYSICS DIVISION ANNUAL PROGRESS REPORT. Period Ending September 1, 1962

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1963-01-11

A total of 74 subsections are included in the report. The information in 4 subsections was previously abstracted in NSA. Separate abstracts were prepared for 38 of the subsections. Those sections for which no abstracts were prepared contain information on prompt neutron lifetime, Rover critical experiments, Pu/sup 239/ fission, neutron decay, the O5R code, alpha scattering, 8 and P wavelengths, proton scattering, deuteron scattering, local optical potentials, N. S. Savamah radiation leakage, reactor shielding, cross section data analysis, gamma transport, gamma energy deposition, gaussian integration, data interpolation, neutron scattering, neutron energy deposition, space vehicles, computer analyses, shielding, positron sources, andmore » secondary particles. (J.R.D.)« less