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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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All sourcesNASA NTRSDOE OSTI
NASA NTRS2009

Comparison of Radiation Transport Codes, HZETRN, HETC and FLUKA, Using the 1956 Webber SPE Spectrum

Protection of astronauts and instrumentation from galactic cosmic rays (GCR) and solar particle events (SPE) in the harsh environment of space is of prime importance in the design of personal shielding, spacec raft, and mission planning. Early entry of radiation constraints into the design process enables optimal shielding strategies, but demands efficient and accurate tools that can be used by design engineers in every phase of an evolving space project. The radiation transport code , HZETRN, is an efficient tool for analyzing the shielding effectiveness of materials exposed to space radiation. In this paper, HZETRN is compared to the Monte Carlo codes HETC-HEDS and FLUKA, for a shield/target configuration comprised of a 20 g/sq cm Aluminum slab in front of a 30 g/cm^2 slab of water exposed to the February 1956 SPE, as mode led by the Webber spectrum. Neutron and proton fluence spectra, as well as dose and dose equivalent values, are compared at various depths in the water target. This study shows that there are many regions where HZETRN agrees with both HETC-HEDS and FLUKA for this shield/target configuration and the SPE environment. However, there are also regions where there are appreciable differences between the three computer c odes.

Heinbockel, John H.↗
DOE OSTI2015

A study on the criticality safety for the direct disposal of used nuclear fuel in Japan: Application of burnup credit to the criticality safety evaluation for the disposal canister - 14049

Japan has recently started the technical development of the direct disposal of used nuclear fuel (UNF) to prepare various disposal options. Unlike the disposal of radioactive waste from UNF reprocessing, criticality safety is important here because of the presence of a certain amount of fissile nuclides in UNF. This paper provides an outline of the research to be addressed in this field and the relevant studies in Japan. It especially includes the first result of a criticality safety evaluation of a disposal canister model adopting burnup credit. In this evaluation, it was assumed that a steel disposal canister loaded with fuel assemblies was surrounded by a clay buffer material, and the fuel assemblies and the disposal canister were intact. Conditions representative of discharged pressurized-water reactor (PWR) fuel were applied in this study - initial U-235 enrichment of 4.5 wt% and burnup of 45 GWd/t. The isotopic composition of the burnt fuel was evaluated using SWAT3.1 code. In the depletion calculation, conservative parameters were applied within the typical PWR operation range. The effective neutron multiplication factor (k{sub eff}) was then calculated using a continuous-energy Monte Carlo code MVP-2.0, adopting the JENDL-4.0 nuclear data library. Uncertainty in k{sub eff} due to depletion calculation errors was evaluated using the correction factors of isotopic composition, which were obtained via multiple comparisons between the calculated and measured isotopic concentrations of irradiated fuels. In addition, the effects of axial burnup profile and horizontal burnup gradient on k{sub eff} were evaluated using the open burnup distribution database. The k{sub eff} including the uncertainties and conservatism was found to be below 0.95 for both actinide-only and actinide and fission product considerations when the fuel assemblies and the disposal canister were assumed to be intact. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Software Quality Metrics for Geant4: An Initial Assessment - Paper 122

In the context of critical applications, such as shielding and radiation protection, ensuring the quality of simulation software they depend on is of utmost importance. The assessment of simulation software quality is crucial both to determine its adoption in experimental applications and to guarantee reproducibility of outcome over time. Geant4 is a simulation system used in a wide range of experimental applications, some of which, such as radiation protection and biomedical studies, may involve critical use cases and would especially profit from an objective assessment of the software quality. In this study, we present the first results from an ongoing analysis of Geant4 code based on established software metrics from the perspective of software engineering. The analysis evaluates the current status of the code to quantify its characteristics with respect to documented quality standards; further appraisals concern evolutions over a series of release distributions. The quantitative assessment of the software is initially focused on a subset of Geant4 packages, which play a key role in scientific applications and are representative of different software development processes. We describe the chosen software metrics that map the entities of a software system to numeric values through mathematical definitions, quantifying software attributes ranging from functionality to maintainability. To measure such metrics, we exploited a set of software metrics tools, i.e. programs that implement a set of software metrics definitions. They allow software engineers to assess a software system by extracting the required attributes from the software and providing the corresponding metric values. To correlate software metrics values in a well-defined way, we used software quality standard models, such as the ISO/IEC 9126 standard, which aggregate numerical values to aid quality analysis and assessment. With this work we aim to create a baseline for evaluating correlations between software quality embedded in the Geant4 development process and simulation observables produced by Geant4-based applications. Furthermore, we intend to highlight what metrics are most effective at evaluating radiation transport software quality. We are confident that the result will provide constructive guidance both to improve key software tools, such as Geant4, and to estimate their contribution in risk analysis concerning sensitive applications. The remainder of this paper is structured as follows. We first discuss the approach adopted to fulfil our assessment. Then, we present the setup of our experiments and the way we collected data. We provide an initial interpretation of a subset of data resulting from measurements made on the selected Geant4 packages. As the analysis is still ongoing at the time of submitting this abstract, some results will be documented in the final paper and in the conference presentation. With this work we aim to create a proper Geant4 data set for evaluating its software characteristics and identifying both strengths and weaknesses. In this initial study, we considered the geometry and processes packages, which play a major role in most simulation applications. The data analysis is in progress at the time of writing; its complexity is related to the differences among software metrics tools, to the large amount of data to be evaluated and to the pioneering character of this project. We are confident that the results will provide valuable indications for the Geant4 user community and constructive guidance to improve key Geant4 packages. Other Monte Carlo codes could profit from the methodology developed in this pilot project for similar assessments of their software quality. At the end of this study, we intend to create a baseline for evaluating correlations between software quality embedded in the Geant4 development process and simulation observables produced by Geant4-based applications. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2015

Back-and-forth approximation as a simple neutron transport model - 14607

The back-and-forth approximation is a very simple neutron transport model that helps get a number of exact solutions which qualitatively describe the actual characteristics of planar, cylindrical and spherical systems with fissile material. The model allows its absolutely accurate implementation in Monte Carlo codes and can be used to test the algorithms which do not treat geometry or interaction data. (author)

42 ENGINEERING↗
DOE OSTI2013

Monte Carlo electron-photon transport using GPUs as an accelerator: Results for a water-aluminum-water phantom

An electron-photon coupled Monte Carlo code ARCHER - Accelerated Radiation-transport Computations in Heterogeneous Environments - is being developed at Rensselaer Polytechnic Institute as a software test bed for emerging heterogeneous high performance computers that utilize accelerators such as GPUs. In this paper, the preliminary results of code development and testing are presented. The electron transport in media was modeled using the class-II condensed history method. The electron energy considered ranges from a few hundred keV to 30 MeV. Moller scattering and bremsstrahlung processes above a preset energy were explicitly modeled. Energy loss below that threshold was accounted for using the Continuously Slowing Down Approximation (CSDA). Photon transport was dealt with using the delta tracking method. Photoelectric effect, Compton scattering and pair production were modeled. Voxelised geometry was supported. A serial ARHCHER-CPU was first written in C++. The code was then ported to the GPU platform using CUDA C. The hardware involved a desktop PC with an Intel Xeon X5660 CPU and six NVIDIA Tesla M2090 GPUs. ARHCHER was tested for a case of 20 MeV electron beam incident perpendicularly on a water-aluminum-water phantom. The depth and lateral dose profiles were found to agree with results obtained from well tested MC codes. Using six GPU cards, 6x10{sup 6} histories of electrons were simulated within 2 seconds. In comparison, the same case running the EGSnrc and MCNPX codes required 1645 seconds and 9213 seconds, respectively, on a CPU with a single core used. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2013

Advances in Sensitivity Analysis with MCNP [Slides]

Monte Carlo codes such as MCNP are used extensively in preventing inadvertent nuclear criticalities in processes involving fissionable material. Their ability to do this relies on the accuracy of the simulation tools, the underlying nuclear data libraries, and their validation of relevant experiments. Recent development of an inline method for importance or adjoint weighting with continuous-energy Monte Carlo in MCNP has allowed for the calculation of sensitivity coefficients without the need of meshes or multigroup collapse, unlike previously available tools. These capabilities allow for quantitative comparisons of processes to benchmark experiments and uncertainty quantification for the calculation of biases and safety margins. The current state-of-the-art is discussed, as well as future research avenues.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Implementation and testing of the on-the-fly thermal scattering Monte Carlo sampling method for graphite and light water in MCNP6

Here, a proper treatment of thermal neutron scattering requires accounting for chemical binding through a scattering law S(α,β,T). Monte Carlo codes sample the secondary neutron energy and angle after a thermal scattering event from probability tables generated from S(α,β,T) tables at discrete temperatures, requiring a large amount of data for multiscale and multiphysics problems with detailed temperature gradients. We have previously developed a method to handle this temperature dependence on-the-fly during the Monte Carlo random walk using polynomial expansions in 1/T to directly sample the secondary energy and angle. In this paper, the on-the-fly method is implemented into MCNP6 and tested in both graphite-moderated and light water-moderated systems. The on-the-fly method is compared with the thermal ACE libraries that come standard with MCNP6, yielding good agreement with integral reactor quantities like k-eigenvalue and differential quantities like single-scatter secondary energy and angle distributions. The simulation runtimes are comparable between the two methods (on the order of 5–15% difference for the problems tested) and the on-the-fly fit coefficients only require 5–15 MB of total data storage.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

SU-E-T-146: Beam Energy Spread Estimate Based On Bragg Peak Measurement

Purpose: ProNova is installing and commissioning a two room proton therapy system in Knoxville, TN. Beam energy out of the 230MeV cyclotron was measured on Jan 24, 2015. Cyclotron beam was delivered into a Zebra multi layered IC detector calibrated in terms of penetration range in water. The analysis of the measured Bragg peak determines penetration range in water which can be subsequently converted into proton beam energy. We extended this analysis to obtain an estimate of the beam energy spread out of the cyclotron. Methods: Using Monte Carlo simulations we established the correlation between Bragg peak shape parameters (width at 50% and 80% dose levels, distal falloff) and penetration range for a monoenergetic proton beam. For large uniform field impinging on a small area detector, we observed linear dependence of each Bragg peak parameter on beam penetration range as shown in Figure A. Then we studied how this correlation changes when the shape of Bragg peak is distorted by the beam focusing conditions. As shown in Figure B, small field size or diverging beam cause Bragg peak deformation predominantly in the proximal region. The distal shape of the renormalized Bragg peaks stays nearly constant. This excludes usage of Bragg peak width parameters for energy spread estimates. Results: The measured Bragg peaks had an average distal falloff of 4.86mm, which corresponds to an effective range of 35.5cm for a monoenergetic beam. The 32.7cm measured penetration range is 2.8cm less. Passage of a 230MeV proton beam through a 2.8cm thick slab of water results in a ±0.56MeV energy spread. As a final check, we confirmed agreement between shapes of the measured Bragg peak and one generated by Monte-Carlo code for proton beam with 0.56 MeV energy spread. Conclusion: Proton beam energy spread can be estimated using Bragg peak analysis.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2014

Benchmark Experiment of Dose Rate Distributions Around the Gamma Knife Medical Apparatus

Dose rate measurements around a gamma knife apparatus were performed by using an ionization chamber. Analyses have been performed by using the Monte Carlo code MCNP-5. The nuclear library used for the dose rate distribution of {sup 60}Co was MCPLIB04. The calculation model was prepared with a high degree of fidelity, such as the position of each Cobalt source and shielding materials. Comparisons between measured results and calculated ones were performed, and a very good agreement was observed. It is concluded that the Monte Carlo calculation method with its related nuclear data library is very effective for such a complicated radiation oncology apparatus.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2012

Study of the secondary neutral radiation in proton therapy: Toward an indirect in vivo dosimetry

Purpose: Secondary particles produced in the collision of protons with beam modifiers are of concern in proton therapy. Nevertheless, secondary radiation can provide information on the dosimetric parameters through its dependency on the modulating accessories (range shifter and range modulating wheel). Relatively little data have been reported in the literature for low-energy proton beams. The present study aims at characterizing the neutron and photon secondary radiation at the low-energy proton therapy facility of the Centre Antoine Lacassagne (CAL), and studying their correlation to the dosimetric parameters to explore possible practical uses of secondary radiation in the treatment quality for proton therapy. Methods: The Monte Carlo code MCNPX was used to simulate the proton therapy facility at CAL. Neutron and photon fluence, {Phi}, and ambient dose equivalent per proton dose, H*(10)/D, were determined across the horizontal main plane spanning the whole treatment room. H*(10)/D was also calculated at two positions of the treatment room where dosimetric measurements were performed for validation of the Monte Carlo calculations. Calculations and measurements were extended to 100 clinical spread-out Bragg Peaks (SOBPs) covering the whole range of therapeutic dose rates (D/MU) employed at CAL. In addition, the values of D and MU were also calculated for each SOBP and the results analyzed to study the relationship between secondary radiation and dosimetric parameters. Results: The largest production of the secondary particles takes place at the modulating devices and the brass collimators located along the optical bench. Along the beam line and off the beam axis to 2.5 m away, H*(10)/D values ranged from 5.4 {mu}Sv/Gy to 5.3 mSv/Gy for neutrons, and were 1 order of magnitude lower for photons. H*(10)/D varied greatly with the distance and angle to the beam axis. A variation of a factor of 5 was found for the different range of modulations (SOBPs). The ratios between calculations and measurements were 2.3 and 0.5 for neutrons and photons, respectively, and remained constant for all the range of SOBPs studied, which provided validation for the Monte Carlo calculations. H*(10)/D values were found to correlate to the proton dose rate D/MU with a power fit, both for neutrons and photons. This result was exploited to implement a system to obtain D/MU values from the measurement of the integrated photon ambient dose equivalent H*(10) during treatment, which provides a method to control the dosimetric parameters D/MU and D. Conclusions: The treatment room at CAL is moderately polluted by secondary particles. The constant ratio between measurements and calculations for all SOBPs showed that simulations correctly predict the dosimetric parameters and the dependence of the production of secondary particles on the modulation. The correlation between H*(10)/D and D/MU is a useful tool for quality control and is currently used at CAL. This system works as an indirect in vivo dosimetry method, which is so far not feasible in proton therapy. This tool requires very simple instrumentation and can be implemented from the measurement of either photons or neutrons.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2014

MO-G-BRF-09: Investigating Magnetic Field Dose Effects in Mice: A Monte Carlo Study

Purpose: In MRI-linac treatments, radiation dose distributions are affected by magnetic fields, especially at high-density/low-density interfaces. Radiobiological consequences of magnetic field dose effects are presently unknown; therefore, preclinical studies are needed to ensure the safe clinical use of MRI-linacs. This study investigates the optimal combination of beam energy and magnetic field strength needed for preclinical murine studies. Methods: The Monte Carlo code MCNP6 was used to simulate the effects of a magnetic field when irradiating a mouse-sized lung phantom with a 1.0cmx1.0cm photon beam. Magnetic field effects were examined using various beam energies (225kVp, 662keV[Cs-137], and 1.25MeV[Co-60]) and magnetic field strengths (0.75T, 1.5T, and 3T). The resulting dose distributions were compared to Monte Carlo results for humans with various field sizes and patient geometries using a 6MV/1.5T MRI-linac. Results: In human simulations, the addition of a 1.5T magnetic field caused an average dose increase of 49% (range:36%–60%) to lung at the soft tissue-to-lung interface and an average dose decrease of 30% (range:25%–36%) at the lung-to-soft tissue interface. In mouse simulations, the magnetic fields had no effect on the 225kVp dose distribution. The dose increases for the Cs-137 beam were 12%, 33%, and 49% for 0.75T, 1.5T, and 3.0T magnetic fields, respectively while the dose decreases were 7%, 23%, and 33%. For the Co-60 beam, the dose increases were 14%, 45%, and 41%, and the dose decreases were 18%, 35%, and 35%. Conclusion: The magnetic field dose effects observed in mouse phantoms using a Co-60 beam with 1.5T or 3T fields and a Cs-137 beam with a 3T field compare well with those seen in simulated human treatments with an MRI-linac. These irradiator/magnet combinations are suitable for preclinical studies investigating potential biological effects of delivering radiation therapy in the presence of a magnetic field. Partially funded by Elekta.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2020

Development of a predictive capability of short-pulse laser-driven broadband x-ray radiography

High intensity, short-pulse laser interaction with a solid metal target produces broadband hard x-rays potentially for various applications of x-ray radiography. In this work, experimental benchmarking of numerical modelling for short-pulse laser-driven broadband x-ray radiography is presented. Angular dependent x-ray spectra are first calculated with a hybrid particle-in-cell code, Large Scale Plasma (LSP), using fast electron parameters inferred from an analysis of measured bremsstrahlung signals. Subsequently, a calculated x-ray spectrum in the direction of radiography is used in photon transport calculations using a Monte Carlo code, Particle and Heavy Ion Transport code System (PHITS), to simulate a radiographic image including a modelled 3D test object, an x-ray attenuation filter and an image plate detector. Simulated radiographic images are compared with measurements obtained in an experiment using a 50-TW Leopard short-pulse laser at the University of Nevada Reno. Results show that simulations reproduce the experimental images well for three different attenuation filters (plastic, aluminium, and brass), while 1D transmission profiles for the plastic and aluminium filters are quantitatively in good agreement. The modelling approach established in this work could be used as a predictive tool to simulate radiographic images of complex 3D solid objects at any arbitrary angular position or to optimize experimental components such as the source spectrum, x-ray attenuation filters and a detector type depending on a radiographic object without carrying out radiographic experiments.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Lattice physics calculations using the embedded self-shielding method in polaris, Part II: Benchmark assessment

Polaris is a 2D multigroup lattice physics capability in the SCALE code system used for the analysis of light water reactor fuel designs. In this work, verification and validation results of Polaris in SCALE 6.2 are presented. The SCALE/KENO continuous-energy Monte Carlo code is used as a reference to compare the results of more than 2,000 combinations of pressurized water reactor and boiling water reactor lattice designs and state points. Radiochemical assay measurements were also used to assess the accuracy of depletion calculations. Results show that Polaris, with ENDF/B-VII.1 cross section library, produces accurate lattice physics data for core simulators.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

The theoretical aspects of UrQMD & AMPT models

The field of high energy physics is very challenging in carrying out theories and experiments to unlock the secrets of heavy ion collisions and still not cracked and solved completely. There are many theoretical queries; some may be due to the inherent causes like the non-perturbative nature of QCD in the strong coupling limit, also due to the multi-particle production and evolution during the heavy ion collisions which increase the complexity of the phenomena. So for the purpose of understanding the phenomena, variety of theories and ideas are developed which are usually implied in the form of Monte-Carlo codes. The UrQMD model and the AMPT model are discussed here in detail. These methods are useful in modeling the nuclear collisions.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2014

Poster — Thur Eve — 21: Off-axis dose perturbation effects in water in a 5 × 5 cm{sup 2} 18 MV photon beam for the PTW microLion and Exradin A1SL ionization chambers

A PTW microLion liquid ionization chamber and an Exradin A1SL air-filled ionization chamber have been modeled using the egs-chamber user code of the EGSnrc system to determine their perturbation effects in water in a 5 × 5 cm{sup 2} 18 MV photon beam. A model of the Varian CL21EX linear accelerator was constructed using the BEAMnrc Monte Carlo code, and was validated by comparing measured PDDs and profiles from the microLion and A1SL chambers to calculated results that included chamber models. Measured PDDs for a 5 × 5 cm{sup 2} field for the microLion chamber agreed with calculations to within 1.5% beyond a depth of 0.5 cm, and the A1SL PDDs agreed within 1.0% beyond 1.0 cm. Measured and calculated profiles at 10 cm depth agreed within 1.0% for both chambers inside the field, and within 4.0% near the field edge. Local percent differences increased up to 15% at 4 cm outside the field. The ratio of dose to water in the absence of the chamber relative to dose in the chamber's active volume as a function of off-axis distance was calculated using the egs-chamber correlated sampling technique. The dose ratio was nearly constant inside the field and consistent with the stopping power ratios of water to detector material, but varied up to 3.3% near the field edge and 5.2% at 4 cm outside the field. Once these perturbation effects are fully characterized for more field sizes and detectors, they could be applied to clinical water tank measurements for improved dosimetric accuracy.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2018

Application of the Denovo Discrete Ordinates Radiation Transport Code to Large-Scale Fusion Neutronics

Fusion energy systems pose unique challenges to the modeling and simulation community. These challenges must be met to ensure the success of the ITER experimental fusion reactor. ITER’s complex systems require detailed modeling that goes beyond the scale of comparable simulations to date. In this work, the Denovo radiation transport code was used to calculate neutron fluence and kerma for the JET streaming benchmark. This work was performed on the Titan supercomputer at the Oak Ridge Leadership Computing Facility. Denovo is a novel three-dimensional discrete ordinates transport code designed to be highly scalable. Sensitivity studies have been completed to examine the impact of several deterministic parameters. Furthermore, results were compared against experiment as well as the MCNP and Shift Monte Carlo codes.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2012

Comet whole-core solution to a stylized 3-dimensional pressurized water reactor benchmark problem with UO{sub 2}and MOX fuel

A stylized pressurized water reactor (PWR) benchmark problem with UO{sub 2} and MOX fuel was used to test the accuracy and efficiency of the coarse mesh radiation transport (COMET) code. The benchmark problem contains 125 fuel assemblies and 44,000 fuel pins. The COMET code was used to compute the core eigenvalue and assembly and pin power distributions for three core configurations. In these calculations, a set of tensor products of orthogonal polynomials were used to expand the neutron angular phase space distribution on the interfaces between coarse meshes. The COMET calculations were compared with the Monte Carlo code MCNP reference solutions using a recently published an 8-group material cross section library. The comparison showed both the core eigenvalues and assembly and pin power distributions predicated by COMET agree very well with the MCNP reference solution if the orders of the angular flux expansion in the two spatial variables and the polar and azimuth angles on the mesh boundaries are 4, 4, 2 and 2. The mean and maximum differences in the pin fission density distribution ranged from 0.28%-0.44% and 3.0%-5.5%, all within 3-sigma uncertainty of the MCNP solution. These comparisons indicate that COMET can achieve accuracy comparable to Monte Carlo. It was also found that COMET's computational speed is 450 times faster than MCNP. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

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

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 two 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)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗