Continuous-energy Monte Carlo Transport Research Articles

Studies of intra-pin power distributions in operated BWR fuel assemblies using MCNP with a cycle check-up methodology

At PSI, the reference approach for core safety analyses of all the Swiss operating reactors is based on a deterministic 2-D lattice depletion/3-D two-group nodal diffusion code platform. As a complement to this, a cycle check-up (CHUP) methodology is being established where, for instance, the continuous energy Monte Carlo transport solver MCNP can be plugged-in at any selected operating point of a given reactor/cycle to conduct in-depth examinations of the 3-D flux/power distributions. After briefly outlining the concepts of the CHUP methodology, its application with MCNP to obtain additional 3-D pin power information than provided by current state-of-the-art nodal core simulators is illustrated. More specifically, the pin power and local intra-pin power distributions in operated assemblies, representative of modern BWR fuel designs, are examined. Focus is given to fuel rods characterised by strongly heterogeneous spectral environments. The CHUP/MCNP methodology is applied to evaluate and compare pin and intra-pin power behaviour between different cycles and/or burnup steps, for normal conditions as well as when assuming local perturbations of the coolant and/or mechanical structures. Although only pin-average nuclide compositions are at this stage used, it is shown that already in this case, strong to very-strong azimuthal gradients take place both radially as well as axially and that these can become substantially more pronounced by very small local perturbations, in particular from geometrical changes related to assembly bowing.

Monte Carlo Sensitivity and Uncertainty Analysis with Continuous-Energy Covariance Data

The continuous-energy Monte Carlo (MC) sensitivity and uncertainty (S/U) analysis conducted using the multigroup covariance matrices has a theoretical pitfall in that it is inconsistent with the principle of continuous-energy MC neutronics calculations because the use of the multigroup covariance matrices means treating covariance data as multigroup variables rather than continuous-energy variables. As a way to get around this deficiency and perform the MC S/U analysis on the theoretically consistent principle, this paper presents a new continuous-energy MC S/U formulation which directly utilizes the continuous-energy covariance data in the evaluated nuclear data libraries instead of the multigroup covariance matrices produced by nuclear data processing codes. The validity of the new MC S/U formulation is examined in terms of the input-nuclear-data-induced k uncertainty of the Godiva critical assembly and the TMI-1 pin cell problem by inputting the continuous-energy covariance data of nuclides involved directly into the continuous-energy MC transport calculations by a Seoul National University MC code, McCARD.

Measurement and calculation of neutron leakage spectra from slab samples of beryllium, gallium and tungsten irradiated with 14.8 MeV neutrons

In order to make benchmark validation of the nuclear data for gallium (Ga), tungsten (W) and beryllium (Be) in existing modern evaluated nuclear data files, neutron leakage spectra in the range from 0.8 to 15 MeV from slab samples were measured by time-of-flight technique with a BC501 scintillation detector. The measurements were performed at China Institute of Atomic Energy (CIAE) using a D-T neutron source. The thicknesses of the slabs were 0.5 to 2.5 mean free path for 14.8 MeV neutrons, and the measured angles were chosen to be 60∘ and 120∘ . The measured spectra were compared with those calculated by the continuous energy Monte-Carlo transport code MCNP, using the data from the CENDL-3.1, ENDF/B-VII.1 and JENDL-4.0 nuclear data files, the comparison between the experimental and calculated results show that: The results from all three libraries significantly underestimate the cross section in energy range of 10–13 MeV for Ga; For W, the calculated spectra using data from CENDL-3.1 and JENDL-4.0 libraries show larger discrepancies with the measured ones, especially around 8.5–13.5 MeV; and for Be, all the libraries led to underestimation below 3 MeV at 120∘ .

ICONE19-44144 Gamma Spectra Measurement and Calculations Using Different Data Libraries on Nickel and Iron Benchmark Assemblies

The leakage gamma spectra measurements have been studied on benchmark spherical assemblies-nickel and iron spheres with diameter of 50 cm. The Cf-252 neutron sources with different emissions were placed into the centre of sphere. The fine structure of gamma spectrum was measured by HPGe spectrometer. The gamma energy range of spectrometer was from 0.1 to 10 MeV. The experimental data were compared to results of transport calculations based on different evaluated nuclear data libraries (ENDF/B-VII.0, ENDF/B-VI.3, JENDL 3.3, JENDL 4.0, JEFF-3.1.1, TENDL-2008, TENDL-2009 and CENDL 3.1). The continuous energy Monte Carlo transport calculation code MCNP-4C was employed for the calculations. Main observed differences between experiments and calculations are discussed.

Proposal of Direct Calculation of Kinetic Parameters βeff and Based on Continuous Energy Monte Carlo Method

Direct calculation methods of kinetic parameters are proposed based on the continuous energy Monte Carlo method. In the proposed methods, the effective delayed neutron fraction βeff and the neutron generation time ∧ are estimated using eigenvalue calculations. The expected number of fission neutrons in the next generation is newly applied to the proposed methods instead of the adjoint flux that has been conventionally used. The algorithms to estimate the kinetic parameters are established and incorporated into the continuous energy Monte Carlo transport calculation code MCNP-4C, which is versatile for eigenvalue calculations of nuclear reactor cores with various types of neutron energy spectrum and geometry. The proposed methods were validated since the calculated values agreed with the experimental data of βeff and βeff/∧ for the critical cores within accuracies of 4.5% and 10%, respectively.

Proposal of Direct Calculation of Kinetic Parameters .BETA.eff and .LAMBDA. Based on Continuous Energy Monte Carlo Method

Direct calculation methods of kinetic parameters are proposed based on the continuous energy Monte Carlo method. In the proposed methods, the effective delayed neutron fraction βeff and the neutron generation time ∧ are estimated using eigenvalue calculations. The expected number of fission neutrons in the next generation is newly applied to the proposed methods instead of the adjoint flux that has been conventionally used. The algorithms to estimate the kinetic parameters are established and incorporated into the continuous energy Monte Carlo transport calculation code MCNP-4C, which is versatile for eigenvalue calculations of nuclear reactor cores with various types of neutron energy spectrum and geometry. The proposed methods were validated since the calculated values agreed with the experimental data of βeff and βeff/∧ for the critical cores within accuracies of 4.5% and 10%, respectively.

Measurements and analyses of angular neutron flux spectra on graphite and lithium-oxide slabs irradiated with 14.8MeV neutrons.

Angular neutron fluxes leaking from the surface of lithium-oxide and graphite slab assemblies have been measured with irradiation of D-T neutrons. The spectrum measurement was performed using the time-of-flight technique with an NE213 scintillation detector. The thicknesses of the slabs were 0.6 to 5 mean free path for 14.8 MeV neutrons, and the measured leaking angles of the angular fluxes were 0.0°, 12.2°, 24.9°, 41.8° and 66.8°. The experimental results have been compared with the results calculated by the continuous energy Monte Carlo transport code MCNP, using the data in the JENDL-3PR1, −3PR2, and ENDF/B-V nuclear data files. The comparisons between the experimental and calculated results show that the data of 7Li in JENDL-3PR2 is improved for the secondary emission spectra of the 4.63 MeV level and (n, 2n) reactions; the angular distributions of 3rd-and 4th-level inelastic reactions of C in the JENDLs are questionable. The thickness dependences for high energy neutrons also suggest that the total cross section of 7Li and the elastic cross sections of C are slightly inadequate.

Experiments and calculations of 14MeV neutron streaming through multi-layered slit assembly.

A streaming experiment using a D-T neutron source was carried out to verify the calculational technique for neutron transport in a shield assembly with multi-layered slits. Reaction rate distributions of a small spherical NE213 scintillation detector to fast neutrons were measured in the slits made of 304SS and in the mortar surrounding the slits. The energy spectrum of fast neutrons in the slit was also measured with the same detector. These measurement were compared with calculations using the continuous energy Monte Carlo transport code MCNP. The calculated reaction rates in the slits agreed with the measured ones within experimental and calculational errors. Besides, it is suggested that the attenuation of fast neutrons in the mortar is significantly different from that in the slits and the behavior is nearly traced by the calculation with the MCNP code. The measured and calculated spectra at a position close to the exit inside the lower slit agreed within the both errors.

Measurement and analysis of neutron leakage spectrum from nickel sphere for 14 MeV neutron source.

A neutron leakage spectrum from a nickel sphere surrounding a 14-MeV neutron source is measured and analyzed in order to verify the accuracy of nickel cross sections. Measurement is done by means of a time-of-flight technique in the range of 2–15 MeV using an NE213 scintillator, and compared with calculations carried out with MCNP, a continuous energy Monte Carlo transport code, using JENDL-3PR1 and ENDFIB-IV neutron libraries. In spite of an overestimation of neutron flux near 13 MeV, the calculated result employing JENDL-3PR1 shows generally better agreement with the measured spectrum. In ENDF/B-IV usage, there is disagreement between measured and calculated spectra between 5 and 12 MeV. Problems in evaluated nuclear data for nickel are also described.

A Comparative Multidimensional Nuclear Analysis of Candidate Blanket Designs for Tokamak and Tandem Mirror Reactor Concepts

Three-dimensional, tritium breeding, and energy multiplication analyses were performed for nine tandem mirror reactor (TMR) designs and seven tokamak reactor designs that had evolved as the ''rank-1'' concept from the latest Blanket Comparison and Selection Study. These designs include a variety of combinations of candidate breeder, coolant, and structural materials that seem most promising for the respective TMR/tokamak concept application. The analysis was performed using a single consistent computational technique, i.e., a continuous energy Monte Carlo transport method based on MCNP with the associated cross section libraries based on the latest ENDF/B-V data. Through a comparative study of those 16 candidate blanket designs, an attempt is made to rank them from a standpoint of the neutronic blanket performance.