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All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

Fan-beam intensity modulated proton therapy

Purpose: This paper presents a concept for a proton therapy system capable of delivering intensity modulated proton therapy using a fan beam of protons. This system would allow present and future gantry-based facilities to deliver state-of-the-art proton therapy with the greater normal tissue sparing made possible by intensity modulation techniques.Methods: A method for producing a divergent fan beam of protons using a pair of electromagnetic quadrupoles is described and particle transport through the quadrupole doublet is simulated using a commercially available software package. To manipulate the fan beam of protons, a modulation device is developed. This modulator inserts or retracts acrylic leaves of varying thickness from subsections of the fan beam. Each subsection, or beam channel, creates what effectively becomes a beam spot within the fan area. Each channel is able to provide 0–255 mm of range shift for its associated beam spot, or stop the beam and act as an intensity modulator. Results of particle transport simulations through the quadrupole system are incorporated into the MCNPX Monte Carlo transport code along with a model of the range and intensity modulation device. Several design parameters were investigated and optimized, culminating in the ability to create topotherapy treatment plans using distal-edge tracking on both phantom and patient datasets.Results: Beam transport calculations show that a pair of electromagnetic quadrupoles can be used to create a divergent fan beam of 200 MeV protons over a distance of 2.1 m. The quadrupole lengths were 30 and 48 cm, respectively, with transverse field gradients less than 20 T/m, which is within the range of water-cooled magnets for the quadrupole radii used. MCNPX simulations of topotherapy treatment plans suggest that, when using the distal edge tracking delivery method, many delivery angles are more important than insisting on narrow beam channel widths in order to obtain conformal target coverage. Overall, the sharp distal falloff of a proton depth-dose distribution was found to provide sufficient control over the dose distribution to meet objectives, even with coarse lateral resolution and channel widths as large as 2 cm. Treatment plans on both phantom and patient data show that dose conformity suffers when treatments are delivered from less than approximately ten angles. Treatment time for a sample prostate delivery is estimated to be on the order of 10 min, and neutron production is estimated to be comparable to that found for existing collimated systems.Conclusions: Fan beam proton therapy is a method of delivering intensity modulated proton therapy which may be employed as an alternative to magnetic scanning systems. A fan beam of protons can be created by a set of quadrupole magnets and modified by a dual-purpose range and intensity modulator. This can be used to deliver inversely planned treatments, with spot intensities optimized to meet user defined dose objectives. Additionally, the ability of a fan beam delivery system to effectively treat multiple beam spots simultaneously may provide advantages as compared to spot scanning deliveries.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2012

Monte Carlo study of in-field and out-of-field dose distributions from a linear accelerator operating with and without a flattening-filter

Purpose: To compare dosimetric characteristics of 6 MV photon fields originating from a linear accelerator operating with (FF) and without (FFF) a flattening-filter. The main objective is to establish a FFF model that results in similar depth-dose and build-up profiles as the original FF model, and subsequently estimate and compare out-of-field dose distributions. Methods: The EGSnrc Monte Carlo user codes BEAMnrc and DOSXYZnrc are used for photon beam simulations of an Elekta linear accelerator and dose calculations in a water phantom, respectively. Three beam models were analyzed: (1) the conventional linear accelerator with the flattening-filter in place and incident electron energy 6.45 MeV (FF 6.45 MeV), (2) similar flattening-filter-free model (FFF 6.45 MeV), and (3) as (2) but with increased electron energy (FFF 8.0 MeV). The field size 5 Multiplication-Sign 5 cm{sup 2} was used for characterization of dose output, depth dose profiles, and photon spectrum. The field size 40 Multiplication-Sign 40 cm{sup 2} was used for characterization of cross-field photon energy, photon fluence, and dose distributions. Out-of-field dose distributions were analyzed in both in-plane and cross-plane directions for 5 Multiplication-Sign 5 cm{sup 2} and 10 Multiplication-Sign 10 cm{sup 2} fields. Results: Comparable depth dose distributions, including the build-up region, for FF and FFF fields were achieved by increasing the electron energy from 6.45 MeV to 8.0 MeV for the FFF beam. The FFF beams result in reduced out-of-field dose compared to the FF beam: the reduction was most apparent in the cross-plane direction and more pronounced by the FFF 8.0 MeV beam compared to the FFF 6.45 MeV beam. Differences in out-of-field dose due to direction (in-plane vs cross-plane) were up to 40% for the FF beam; this effect was significantly reduced for the FFF beams. As the flattening-filter is a major source of contaminating electrons, superficial out-of-field dose was expected, and was found to be, reduced for FFF beams. Conclusions: The build-up and depth-dose characteristics of a conventional '6 MV' beam can be maintained when changing to a flattening-filter-free modality by increasing the incident electron energy from 6.45 MeV to 8.0 MeV. This will at the same time reduce the out-of-field dose for regions up to 20 cm from the central axis by 10%-30% compared to the original FF situation.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2015

Breast dosimetry in transverse and longitudinal field MRI-Linac radiotherapy systems

Purpose: In the framework of developing the integration of a MRI-Linac system, configurations of MRI-Linac units were simulated in order to improve the dose distribution in tangential breast radiotherapy using transverse and longitudinal magnetic field geometries of Lorentz force for both medial and lateral tangential fields. Methods: In this work, the GEANT4 Monte Carlo (MC) code was utilized to compare dose distributions in breast radiotherapy for Linac-MR systems in the transverse and longitudinal geometries within humanoid phantoms across a range of magnetic field strengths of 0.5 and 1.5 T. The dose increment due to scattering from the coils was investigated for both geometries as well. Computed tomography images of two patients were used for MC simulations. One patient had intact breast while the other was mastectomized. In the simulations, planning and methods of chest wall irradiation were similar to the actual clinical planning. Results: In a longitudinal geometry, the magnetic field is shown to restrict the lateral spread of secondary electrons to the lung, heart, and contralateral organs, which reduced the mean dose of the ipsilateral lung and heart by means of 17.2% and 6% at 1.5 T, respectively. The transverse configuration exhibits a significant increase in tissue interface effects, which increased dose buildup in the entrance regions of the lateral and medial tangent beams to the planning target volume (PTV) and improved dose homogeneity within the PTV. The improved relative average homogeneity index for two patients to the PTV at magnetic field strength of 1.5 T with respect to no magnetic field case evaluated was 11.79% and 34.45% in the LRBP and TRBP geometries, respectively. In both geometries, the simulations show significant mean dose reductions in the contralateral breast and chest wall skin, respectively, by a mean of 16.6% and 24.9% at 0.5 T and 17.2% and 28.1% at 1.5 T in the transverse geometry, and 10.56% and 14.6% at 0.5 T and 11.3% and 16.3% at 1.5 T in the longitudinal geometry. Considering the scattered photons which reflected from the coils, the average relative dose of each voxel is slightly increased by 0.53% and 0.32% in the LRBP and TRBP geometries, respectively. Conclusions: Orienting the B0 magnetic field parallel to the photon beam axis, LRBP geometry, tends to restrict the radial spread of secondary electrons which resulted in dose reduction to the lung. Dosimetry issues observed in both Linac-MR geometries, such as changes to the lateral dose distribution, significantly exhibited dose reduction in the contralateral organs on a representative breast plan. Further, the results show sharper edge dose volume histogram curves at 1.5 T for both geometries, especially in the LRBP configuration.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2022

Verification of Upcoming MCNP Features For Estimating Nuclear Data Sensitivities in Fixed Source Simulations [Abstract]

Predictive simulation codes, like the Monte Carlo N-Particle (MCNP) transport code, are used throughout the nuclear community. These simulations are based on nuclear data. Maximizing the accuracy and precision of nuclear data maximizes the accuracy and precision of the overall simulation. This is imperative to applications that rely on simulations. For example, improving nuclear data for special nuclear material improves simulation accuracy in stockpile stewardship applications, which results in larger safety margins and decreased operational costs. The improvement and validation of nuclear data is completed through integral benchmark experiments. Past benchmarks have primarily been limited to focus on the effective multiplication factor ($\kappa$ eff ); broadening the purview of benchmarks beyond $\kappa$ eff -dependent nuclear data addresses nuclear data deficiencies. Different response types depend on different areas of nuclear data. This dependence is quantified as nuclear data sensitivity: the change in response due to perturbation of a contributing parameter. The larger the nuclear data sensitivity of a response, the more the experiment is influenced by the uncertainties of the nuclear data. The optimization of nuclear data sensitivities in future benchmarks would result in more detailed validation of lesser studied areas of nuclear data. Currently, direct sensitivity capabilities are not easily found for all experiment types and parameters. An MCNP tool to directly estimate the cross section sensitivities of tallied values is under development. Additionally, updates have been made to the perturbation feature of MCNP, which can be used in a less direct approach to estimating sensitivities. This work verifies these features to estimate nuclear data sensitivities in fixed source simulations of a 4.5-kg sphere of alpha- phase weapons-grade plutonium surrounded by differing amounts of copper and polyethylene. Integrated estimates made using MCNP’s tools were found to statistically agree with integrated estimates made from manual perturbation of nuclear data proving the validity of the MCNP tools.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

Methodology and Software for Gross Defect Detection of Spent Nuclear Fuel at the Atucha-I Reactor [Novel Methodology and Software for Spent Fuel Gross Defect Detection at the Atucha-I Reactor]

Here, fuel assemblies in the spent fuel pool are stored by suspending them in two vertically stacked layers at the Atucha Unit 1 nuclear power plant (Atucha-I). This introduces the unique problem of verifying the presence of fuel in either layer without physically moving the fuel assemblies. Given that the facility uses both natural uranium and slightly enriched uranium at 0.85 wt% 235 U and has been in operation since 1974, a wide range of burnups and cooling times can exist in any given pool. A gross defect detection tool, the spent fuel neutron counter (SFNC), has been used at the site to verify the presence of fuel up to burnups of 8000 MWd/t. At higher discharge burnups, the existing signal processing software of the tool was found to fail due to nonlinearity of the source term with burnup.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Strategies for thorium fuel cycle transition in the SD-TMSR

Liquid-fueled Molten Salt Reactor (MSR) systems represent advances in safety, economics, and sustainability. The MSR has been designed to operate with a Th/ 233 U fuel cycle with 233 U used as startup fissile material. Since 233 U does not exist in nature, we must examine other available fissile materials to start up these reactor concepts. This work investigates the fuel cycle and neutronics performance of the Single-fluid Double-zone Thorium-based Molten Salt Reactor (SD-TMSR) with different fissile material loadings at startup: High Assay Low Enriched Uranium (HALEU) (19.79%), Pu mixed with HALEU (19.79%), reactor-grade Pu (a mixture of Pu isotopes chemically extracted from Pressurized Water Reactor (PWR) spent nuclear fuel (SNF) with 33 GW d/tHM burnup), transuranic elements (TRU) from Light Water Reactor (LWR) SNF, and 233 U. The MSR burnup routine provided by SERPENT-2 is used to simulate the online reprocessing and refueling in the SD-TMSR. The effective multiplication factor, fuel salt composition evolution, and net production of 233 U are studied in the present work. Additionally, the neutron spectrum shift during the reactor operation is calculated. The results show that the continuous flow of reactorgrade Pu helps transition to the thorium fuel cycle within a relatively short time (≈ 4.5 years) compared to 26 years for 233 U startup fuel. Finally, using TRU as the initial fuel materials offers the possibility of operating the SD-TMSR for an extended period of time (≈ 40 years) without any external feed of 233 U.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Photonuclear production of 47 Ca for 47 Ca/ 47 Sc generator from natural CaCO 3 targets

Here this work investigated the indirect production of 47 Sc from natural Ca targets via 48 Ca(γ,n) 47 Ca → 47 Sc +β - +$\bar{\nu}_e$ with incident electron energies of 30, 35, and 40 MeV. The 47 Ca production yields were simulated using the PHITS Monte Carlo simulation code and compared to experimental data. The simulated production rates for all three irradiations are in good agreement with experimental data within uncertainties. As a demonstration of the 47 Ca/ 47 Sc generator system, one of the irradiated CaCO 3 targets was dissolved in nitric acid, and 47 Sc was isolated from the target material using commercially available Eichrom DGA resin. The 47 Sc was allowed to grow in, and the purification process was repeated with promising 47 Sc and Ca recovery yields.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2011

ENDF/B-VII.1 Neutron Cross Section Data Testing with Critical Assembly Benchmarks and Reactor Experiments

The ENDF/B-VII.1 library is the latest revision to the United States Evaluated Nuclear Data File (ENDF). The ENDF library is currently in its seventh generation, with ENDF/B-VII.0 being released in 2006. This revision expands upon that library, including the addition of new evaluated files (was 393 neutron files previously, now 423 including replacement of elemental vanadium and zinc evaluations with isotopic evaluations) and extension or updating of many existing neutron data files. Complete details are provided in the companion paper [M. B. Chadwick et al., 'ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data,' Nuclear Data Sheets, 112, 2887 (2011)]. This paper focuses on how accurately application libraries may be expected to perform in criticality calculations with these data. Continuous energy cross section libraries, suitable for use with the MCNP Monte Carlo transport code, have been generated and applied to a suite of nearly one thousand critical benchmark assemblies defined in the International Criticality Safety Benchmark Evaluation Project's International Handbook of Evaluated Criticality Safety Benchmark Experiments. This suite covers uranium and plutonium fuel systems in a variety of forms such as metallic, oxide or solution, and under a variety of spectral conditions, including unmoderated (i.e., bare), metal reflected and water or other light element reflected. Assembly eigenvalues that were accurately predicted with ENDF/B-VII.0 cross sections such as unmoderated and uranium reflected {sup 235}U and {sup 239}Pu assemblies, HEU solution systems and LEU oxide lattice systems that mimic commercial PWR configurations continue to be accurately calculated with ENDF/B-VII.1 cross sections, and deficiencies in predicted eigenvalues for assemblies containing selected materials, including titanium, manganese, cadmium and tungsten are greatly reduced. Improvements are also confirmed for selected actinide reaction rates such as {sup 236}U, {sup 238,242}Pu and {sup 241,243}Am capture in fast systems. Other deficiencies, such as the overprediction of Pu solution system critical eigenvalues and a decreasing trend in calculated eigenvalue for {sup 233}U fueled systems as a function of Above-Thermal Fission Fraction remain. The comprehensive nature of this critical benchmark suite and the generally accurate calculated eigenvalues obtained with ENDF/B-VII.1 neutron cross sections support the conclusion that this is the most accurate general purpose ENDF/B cross section library yet released to the technical community.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2015

Inventory Uncertainty Quantification using TENDL Covariance Data in Fispact-II

The new inventory code Fispact-II provides predictions of inventory, radiological quantities and their uncertainties using nuclear data covariance information. Central to the method is a novel fast pathways search algorithm using directed graphs. The pathways output provides (1) an aid to identifying important reactions, (2) fast estimates of uncertainties, (3) reduced models that retain important nuclides and reactions for use in the code's Monte Carlo sensitivity analysis module. Described are the methods that are being implemented for improving uncertainty predictions, quantification and propagation using the covariance data that the recent nuclear data libraries contain. In the TENDL library, above the upper energy of the resolved resonance range, a Monte Carlo method in which the covariance data come from uncertainties of the nuclear model calculations is used. The nuclear data files are read directly by FISPACT-II without any further intermediate processing. Variance and covariance data are processed and used by FISPACT-II to compute uncertainties in collapsed cross sections, and these are in turn used to predict uncertainties in inventories and all derived radiological data.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2014

Measurements and simulations of the cosmic-ray-induced neutron background

The cosmic-ray-induced neutron background at ground level has been measured and simulated in conjunction with EJ-309 organic liquid scintillators with an approximate deposited energy range of 0.5–6 MeV. Specifically, the pulse height distributions, net neutron count rates, and angular dependences were obtained. The simulations were carried out using the Monte Carlo transport code MCNPX-PoliMi combined with the (Cosmic-Ray Shower Generator) CRY source subroutine that returns secondary particles produced by cosmic rays. A scaling formula from literature was also implemented in the simulation. The angular dependence of the neutron count rate was measured by collimating the liquid scintillator with polyethylene to attain 18° angular resolution from 0° downwards to 72° horizontally. The neutron count rate was measured to be 23.10±1.69 h -1 sr -1 at 0°, and 7.20±0.78 h -1 sr -1 at 72°. The simulations and measurements compare well and show similar cosine anisotropy for the angular distribution. Finally, the study thus shows that the neutron background response in detector systems can be efficiently and accurately simulated using the procedures described.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2021

SANDD: A directional antineutrino detector with segmented 6 Li-doped pulse-shape-sensitive plastic scintillator

We present a characterization of a small (9-liter) and mobile 0.1% 6 Li-doped pulse-shape-sensitive plastic scintillator antineutrino detector called SANDD (Segmented AntiNeutrino Directional Detector), constructed for the purpose of near-field reactor monitoring with sensitivity to antineutrino direction. SANDD comprises three different types of module. A detailed Monte Carlo simulation code was developed to match and validate the performance of each of the three modules. The combined model was then used to produce a prediction of the performance of the entire detector. Analysis cuts were established to isolate antineutrino inverse beta decay events while rejecting large fraction of backgrounds. The neutron and positron detection efficiencies are estimated to be 34.8% and 80.2%, respectively, while the coincidence detection efficiency is estimated to be 71.7%, resulting in inverse beta decay detection efficiency of 20.0% ± 0.2%(stat.) ± 2.1%(syst.). Finally, the predicted directional sensitivity of SANDD produces an uncertainty of 20° in the azimuthal direction per 100 detected antineutrino events.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2022

Thick target neutron yields from Beryllium, Carbon, Tungsten, and Lead targets irradiated by 26.7 MeV/nucleon 4 H e ions

The angular and energy distributions of secondary neutrons produced from 26.7 MeV/nucleon 4 He ions, stopping in thick Be, C, W and Pb targets are measured by the time of flight method. The GEANT4, PHITS, and FLUKA Monte Carlo simulation codes with different physics models are employed to simulate the neutron yields and the results are compared with the experimental data. For the thick Be and C targets, GEANT4 with INCL++ reproduces the experimental data reasonably well. Whereas PHITS with JQMD-2.0 overestimates the neutron yields at intermediate energies and FLUKA with PEANUT underestimated the neutron yields at intermediate energies. For W and Pb targets, FLUKA with PEANUT physics model agree with experimental data best. PHITS with JQMD-2.0 overestimates the neutron yield at intermediate energies at 0°, and GEANT4 with INCL++ underestimates the neutron yield at intermediate energies at three angles.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2024

Shielding development for the spallation neutron source VENUS instrument

VENUS is a neutron imaging instrument that will use a broad range of neutron wavelengths, from epithermal to cold energies, and will include enhanced contrast mechanisms. It will offer novel energy-selective imaging techniques directly connecting complex engineering materials and systems’ structures, properties, and functions to reveal practical and fundamental answers about their real-world performance. Here, the instrument will be built at SNS beamline 10, facing the decoupled poisoned hydrogen moderator. The driving cost for the instrument is the beamline and instrument cave shielding. Final analyses were performed to evaluate the thickness and composition of shielding materials for the instrument cave and beamline to meet radiation safety criteria for the instrument to start up in 2024 after completing the SNS proton power upgrade.

43 PARTICLE ACCELERATORS↗
DOE OSTI2015

Low energy electrons and swift ion track structure in PADC

The current work aims at providing an accurate description of the ion track-structure in poly-allyl dyglycol carbonate (PADC) by using an up-to-date Monte-Carlo code-called TILDA-V (a French acronym for Transport d’Ions Lourds Dans l’Aqua & Vivo). In this simulation the ion track-structure in PADC is mainly described in terms of ejected electrons with a particular attention done to the Low Energy Electrons (LEEs). After a brief reminder of the most important channels through which LEEs are prone to break a chemical bond, we will report on the simulated energetic distributions of LEEs along an ion track in PADC for particular incident energies located on both sides of the Bragg-peak position. Lastly, based on the rare data dealing with LEEs interaction with polymers or organic molecules, we will emphasise the role played by the LEEs in the formation of a latent track in PADC, and more particularly the one played by the sub-ionization electrons.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2023

Feasibility of a multigroup Boltzmann–Fokker–Planck solution for electron beam dose calculations

Legacy nuclear-reactor Boltzmann solvers start clinical deployment as an alternative to Monte Carlo (MC) codes and Fermi–Eyges semiemprical models in radiation oncology treatment planning. Today’s certified clinical solvers are limited to photon beams. In this paper, ELECTR, a state-of-the-art multigroup electron cross sections generation module in NJOY is presented and validated against Lockwood’s calorimetric measurements, EGS-nrc and GEANT-4 for 1–20 MeV unidirectional electron beams. The nuclear-reactor DRAGON-5 solver is upgraded to access the library and solve the Boltzmann–Fokker–Planck (BFP) equation. A variety of heterogeneous radiotherapy and radiosurgery phantom configurations were used for validation purpose. Case studies include a thorax benchmark, that of a typical breast Intra-Operative Radiotherapy and a high-heterogeneity patient-like benchmark. For all beams, 100% of the water voxels satisfied the American Association of Physicists in Medicine accuracy criterion for a BFP-MC dose error below 2%. At least, 97.0% of adipose, muscle, bone, lung, tumor and breast voxels satisfied the 2% criterion. The average BFP-MC relative error was about 0.56% for all voxels, beams and materials combined. By irradiating homogeneous slabs from Z = 1 (hydrogen) to Z = 99 (einsteinium), we reported performance and defects of the CEPXS mode [US. Sandia National Lab., SAND-89-1685] in ELECTR for the entire periodic table. For all Lockwood’s benchmarks, NJOY-DRAGON dose predictions are within the experimental data precision for 98% of voxels.

42 ENGINEERING↗
DOE OSTI2011

Characteristics of the Neutron Irradiation Facilities of the PSI Calibration Laboratory

The neutron radiation fields of the Calibration Laboratory at Paul Scherrer Institute (PSI) are traceable to the national standards of the Physikalisch-Technische Bundesanstalt (PTB) in Germany. A Berthold LB6411 neutron dose rate meter for neutron radiation is used as a secondary standard. Recently, a thorough characterization of the neutron irradiation fields of the {sup 241}Am-Be and {sup 252}Cf sources by means of reference measurements and a detailed MCNPX simulation of the irradiation facility has been initiated. In this work, the characteristics of the neutron radiation fields are summarized and presented together with model equations and an uncertainty analysis. MCNPX results are shown for the {sup 241}Am-Be source. A comparison of measured and simulated data shows an excellent agreement. From the simulation, valuable information about the neutron fields like the contribution of scattered neutrons in the fields and the energy spectra could be obtained.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2013

Heavy reflector experiments in the IPEN/MB-01 reactor: Stainless steel, carbon steel and nickel

New experiments devoted to the measurements of physical parameters of a light water core surrounded by a heavy reflector were performed in the IPEN/MB-01 research reactor facility. These experiments comprise three sets of heavy reflector (SS-304, Carbon Steel, and Nickel) in a form of laminates around 3 mm thick. Each set was introduced individually in the west face of the core of the IPEN/MB-01 reactor. The aim here is to provide high quality experimental data for the interpretation and validation of the SS-304 heavy reflector calculation methods. The experiments of Carbon Steel, which is composed mainly of iron, and Nickel were performed to provide a consistent and an interpretative check for the SS-304 reflector experiment. The experimental results comprise critical control bank positions, temperatures and reactivities as a function of the number of the plates. Particularly to the case of Nickel, the experimental data are unique of its kind. The theoretical analysis was performed by MCNP-5 with the nuclear data library ENDF/B-VII.0. It was shown that this nuclear data library has a very good performance up to thirteen plates and overestimates the reactivity for higher number of plates independently of the type of the reflector.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Reactivity worth measurements at the IPEN/MB-01 nuclear reactor

Researches that aim to improve the performance of neutron transport codes and quality of nuclear cross section databases are very important to increase the accuracy of simulations and the quality of the analysis and prediction of phenomena in the nuclear field. In this context, relevant experimental data such as reactivity worth measurements are needed. The objective of this work was to perform a series of experiments of reactivity worth measurements, using a digital reactivity meter developed at IPEN. The experiments employed small metallic and ceramic samples inserted in the central region of the core of the experimental IPEN/MB-01 reactor. The theoretical analysis was performed by the MCNP-5 reactor physics code, developed and maintained by Los Alamos National Laboratory, and the ENDF/B-VII.0 nuclear data library.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗