New Particle Transport Methods for Design and Optimization of Spherical-Shell Transmission Measurements
Alireza HaghighatShane R. GardnerVefa KucukboyaciMichael T. WennerJames M. AdamsA.D. CarlsonS. M. GrimesT. N. Massey
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This paper discusses new particle transport methods developed for accurate investigation of iron non-elastic scattering cross sections using the spherical-shell transmission method, employing iron shells with different thicknesses, and neutron time-of-flight spectroscopy of the scattered neutrons. New calculational techniques based on the deterministic and Monte Carlo methods have been developed for design and optimization of the experiment, and for identification and reduction of experimental uncertainties. The new methods include a new tallying option for the MCNP code and new quadrature sets for the 3-D parallel Sn code PENTRAN. The new tallying option is used to determine the optimum source energy versus target thickness, and among the new quadrature sets, the Pn-Tn with ordinate-splitting technique has resulted in accurate flux distributions as compared to Monte Carlo prediction.Keywords:
Neutron Transport
Abstract As of today, the standard method employed in tokamaks for the absolute measurement of the neutron flux (thus of the nuclear fusion power) is based on activation foils, being the most robust and unbiased technique for the absolute determination of neutron fluence. However, this technique is not able to provide real-time data useful for the control of future fusion plants like DEMO. In this paper, we present some preliminary results about the R&D activity aimed at developing the Single-crystal Diamond Detectors (SDD) used for fast neutron measurements into an absolute neutron flux monitor. Tests have been conducted at the new NILE neutron source of the Rutherford-Appleton Laboratory, a facility with compact neutron generators with a maximum yield of 10 9 n/s and 10 10 n/s for 2.5 MeV and 14 MeV neutrons, respectively. A series of neutron spectra and flux measurements have been taken with different SDD and associated DAQ. Comparisons with standard activation foils (and namely Fe, Zr, Al and Nb foils for 14 MeV neutrons and In for 2.5 MeV neutrons) and with other reference detectors are presented and discussed. Also discussed is the stability of the SDD over time when employed at high neutron rates in realistic neutron environment, and the effects of neutron irradiation on both the counting rate and detector resolution.
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To compute the breakdown thresholds of multipactor in microwave devices, a fast single particle Monte-Carlo (SP-MC) method is presented, which considers the random nature of secondary electrons and their initial energies, phases and angles. With Runge-Kutta method and Furman model, the motion of the electron and the secondary electron yield (SEY) of the wall of the device are computed. An effective SEY is regarded as a criterion to estimate whether multipactor occurs, which is computed by averaging the SEYs for all impacts. As an example, the multipactor in a transmission line composed of parallel plates is investigated with the presented SP-MC method, traditional Monte-Carlo method, statistical theory method and particle-in-cell method separately. The results obtained from the SP-MC method accord well with those from the statistical theory method and particle-in-cell method, including the results of the susceptibility zones, break thresholds on specific products of frequency and gap space. Moreover, the SP-MC method is more adaptive than the statistical theory method, more stable than the traditional Monte-Carlo method and much more efficient than the particle-in-cell method.
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This paper presents an overview of the investigations on the need for deterministic transport methods for the analysis of pebble-bed reactors. To account for the transport effects present in the PBMR design that cannot be modeled accurately with the diffusion theory, a two-dimensional neutronics solver based on transport theory is implemented in the Penn State NEM-THERMIX code system. The necessity of equipping neutronics analysis codes with neutron transport theory capability is investigated along with the challenges to accomplish this in an efficient and versatile manner. For this purpose a time-dependent version of the two-dimensional neutron transport code DORT is utilized as a first step. The developed benchmark test cases, based on the PBMR 268 MW design, are used for this work, and results from the comparative analyses of these test cases are presented. The results show clearly that even in steady-state calculations, the differences between diffusion and transport-based methods in analyzing the PBMR are observed and need to be addressed. (authors)
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NEUTRONICS ANALYSIS FOR CYLINDRICAL ASSEMBLY USING GREEN’S FUNCTIONS Jose E. Rivera, M.S. Department of Nuclear, Plasma, and Radiological Engineering University of Illinois at Urbana-Champaign, 2011 Professor Roy A. Axford, Adviser When designing a reactor a preliminary design is done in order to obtain a rough estimate of various reactor properties. These properties include the neutron flux, criticality condition, or distribution of material in the assembly. It is possible to obtain an analytic solution for the neutron flux for a reactor represented in cylindrical coordinates using a Green’s function or Green’s function matrix method for both one group and two group neutron diffusion. The analytic results for a two-dimensional cylindrical system are simplified by assuming a cosine flux shape in the horizontal direction. This assumption makes the flux in the horizontal direction a neutron sink term in the radial direction. From the neutron flux, expressions for the criticality condition and fuel distribution cross section can be obtained by specifying the form factor, which is equivalent to specifying the power shape. A specific form factor, such as a constant power shape or parabolic power shape allow for a comparision between the results obtained by using a one group diffusion model against a two group diffusion model.
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Abstract Operation of NPP has potential of nuclear hazard and radiation hazard. Therefore, all efforts are absolutely needed to be performed to ensure the safety of NPP operation. Neutronics analysis has important role in ensuring that safety aspect. In neutronics analysis, computer codes that could well simulate the neutronics condition in reactor core are needed. In this paper, neutronics analysis of 2652 MWt PWR core has been performed using DRAGON, TRIVAC, and DONJON computer codes. The purpose of simulation is to conduct the calculation of neutronics parameters which are important for safety. The neutronics analyses was conducted in two stages, i.e. cell calculation using DRAGON and full core calculation using TRIVAC and DONJON. For verification, the comparation between the calculation results of DRAGON, TRIVAC, and DONJON, and those of other computer codes results found in literature was performed. The effective multiplication factors calculated by DONJON for conditions where all control rods were fully up and fully down, and for ‘ one stuck rod ’ condition were 1.23399, 0.69705, and 0.91340, respectively. The core maximum power and average power were found to be 21166.1 kW and 16891.7 kW, respectively, with the power peaking factor of 1.2530. The maximum and average neutron fluxes in the core were 4.019038E+14 n/cm 2 s and 9.035312E+13 n/cm 2 s, respectively. The results of fuel cell calculation using DRAGON and the full core calculation using TRIVAC and DONJON have the differences of 4% and 1.3%, respectively, compared to the calculation results using other computer codes. This research shows that DRAGON, TRIVAC, and DONJON have capabilities to perform the calculation of neutronics parameters for NPP, specifically PWR type.
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