TRIPOLI-4 Code Research Articles

Current Status of TRIPOLI-4® Monte Carlo Radiation Transport Code on Adult and Pediatric Computational Phantoms for Radiation Dosimetry Study

TRIPOLI-4® is a general-purpose Monte Carlo radiation transport code developed by the Service d’Études des Réacteurs et de Mathématiques Appliquées at CEA-Saclay. It uses continuous-energy nuclear data to simulate neutron, photon, electron, and positron transport in fields like radiation shielding, reactor physics, and nuclear criticality safety. To study radiation protection dosimetry in human tissues and organs, male and female adult computational phantoms from the Medical Internal Radiation Dose–Oak Ridge National Laboratory phantoms family and the International Commission on Radiological Protection (ICRP) publication 110 were recently modeled and calculated using the geometry options of TRIPOLI-4. To easily use the ICRP 110 voxel-based phantoms in different exposure scenarios, a newly developed phantom option is available in TRIPOLI-4 and its display tool T4G. This new phantom option is helpful for modeling one or more phantoms and for improving calculation performance in real irradiation environments. The 2020 published pediatric computational reference phantoms are accessible from ICRP publication 143. Male and female pediatric phantoms are also verified with the new T4G tool and TRIPOLI-4 code. This paper reports on recent works using TRIPOLI-4 on adult and pediatric computational phantoms. The modeling methods of stylized and voxel-based phantoms, the graphic displays of modeled phantoms with T4G, and the verification procedures for single-phantom and two-phantom application cases are presented. Validation for external and internal dosimetry calculations has been performed. Calculation results on organ dose S values for nuclear medicine applications are presented for single female and single male voxel phantoms using 131I and 177Lu radiation sources. Effective dose calculations for two-phantom cases using 99mTc and 18F sources are compared with traditional H*(10) calculations from nuclear medicine patient to patient caregiver.

Monte Carlo calculation of the organ equivalent dose and effective dose due to immersion in a 16N beta source in air using the ICRP reference phantoms

This work summarises the results of a comparison organized by EURADOS focused on the usage of the ICRP Reference Computational Phantoms. This activity aimed to provide training for the implementation of voxel phantoms in Monte Carlo radiation transport codes and the calculation of the dose equivalent in organs and the effective dose. This particular case describes a scenario of immersion in a16N beta source distributed in the air of a room with concrete walls where the phantom is located. Seven participants took part in the comparison of results using GEANT4, TRIPOLI-4 and MCNP family codes, and there was detected a general problem when calculating the dose to skeletal tissue and the remainder tissue. After a process of feedback with the participants the errors were corrected and the final results reached an agreement of ±5%.

TRIPOLI-4 simulation of the FNG copper benchmark experiment and the tritium production in the lithium diamond detector for ITER-TBM

Abstract For the future ITER (International Thermonuclear Experimental Reactor) operation needs, several D–T fusion neutron experiments have been conducted at FNG (Frascati Neutron Generator) facility. These FNG experimental results are useful to validate neutron transport codes and nuclear data libraries. The TRIPOLI-4 Monte-Carlo radiation transport code has been widely used on radiation shielding, criticality safety and fission reactor physics fields to support nuclear energy research and industrial applications. With the growing interest in using the TRIPOLI-4 code for ITER applications, a key issue that arises is whether or not the released TRIPOLI-4 code and its associated nuclear data libraries are verified and validated for the 14 MeV fusion neutronics calculations. Copper is a key material used in different components of ITER tokamak machine and the FNG copper benchmark experiment funded by F4E (Fusion for Energy) was recently introduced into the OECD/NEA SINBAD database. In the first part of this work, TRIPOLI-4 calculations were performed to simulate the D–T fusion neutron transport in a pure copper block of FNG-Cu experiments. The reaction rates were evaluated for various activation foils (93Nb(n,2n), 27Al(n,α), 58Ni(n,p), 115In(n,n’), 197Au(n,2n), 186W(n,γ), and 197Au(n,γ)) versus penetration depth in the Cu block. To study the measurement tool of tritium production for the future ITER Test Blanket Module (TBM), the second FNG experiment analyzed in this work arranged a thin layer of 6LiF (LiDia detector) placed inside a polyethylene cylinder and irradiated with FNG 14 MeV neutrons. TRIPOLI-4 simulation models were run to evaluate the tritium production in the LiDia detector. The purpose of this study is to investigate the TRIPOLI-4 calculations against the FNG experimental ones. TRIPOLI-4 calculation methods and results are reported.

Organ Dose Calculations Using ICRP Adult Voxel Phantoms and TRIPOLI-4 Monte Carlo Code

Abstract The ICRP 110 adult male and female voxel phantoms are the official computational models representing the ICRP (International Commission on Radiological Protection) Reference Male and Reference Female. In 2018, the Working Group 6 (WG6) of European Radiation Dosimetry Group (EURADOS) organized a study on the usage of the ICRP voxel reference phantoms. Organ dose calculation tasks with radiation transport codes were proposed in occupational, environmental, and medical dosimetry. The TRIPOLI-4 Monte Carlo radiation transport code has been widely used in radiation shielding, criticality safety, and reactor physics fields for supporting French nuclear energy research and industrial applications. To enhance the application fields of TRIPOLI-4, the 2018 EURADOS-WG6 tasks are being taken into account by using different features of the TRIPOLI-4 code. In this work, the ICRP reference voxel phantoms were first adapted into TRIPOLI-4. More than 14 × 106 voxels were represented in a mixed lattice geometry including 140 organs-tissues and 52 tissue media. Diverse exposure scenarios were then investigated by using 60Co and 241Am gamma-ray sources, 16N beta source, and 10 keV neutron source. The TRIPOLI-4 standard nuclear data library was utilized on these neutron, photon, electron, and positron-coupled transport calculations. Energy deposition estimators for electron, positron, neutron, and photon coupled with mesh tally options were used to calculate the organ absorbed dose DT and the effective dose E. TRIPOLI-4 calculation methods and primary results for the EURADOS-WG6 voxel phantom exercise on organ dose study tasks are reported here.

Comprehensive stochastic sensitivities to resonance parameters

Integral experiments in reactors or critical configurations claim to have very small experimental and technological uncertainties. Therefore these latter can be considered valuable experimental information in nuclear data evaluation. Because in the evaluation process the information is carried by model parameters, to perform a rigorous feedback on a nuclear model parameters p - for instance using a measured reactivity ρ-sensitivities S =∂ρ/ρ⁄∂p/p are needed. In usual integral feedbacks, sensitivity to multi-group cross sections are first obtained with deterministic code using perturbation theory. Then these multi-group cross section sensitivities are “convoluted” with parameter sensitivities in order to provide the sensitivity on nuclear model parameter. Recently stochastic approaches have been elaborated in order to obtain continuous cross-section sensitivities thus avoiding the multi-group discretization. In the present work we used the recent Iterated Fission Probability method of the TRIPOLI4 code [1] in order to obtain directly the sensitivity to nuclear physics parameters. We focus here on the sensitivity on resonance parameters and exemplified the method on the computation of sensitivities for 239Pu and 16O resonance parameters one the ICSBEP benchmark PST001. The underlying nuclear model describing resonant cross sections are based in the R-matrix formalism [2] that provides not only the interaction cross sections but also the angular distribution of the scattered neutrons i.e. differential cross sections. The method has thus been updated in order to compute parameter sensitives that include both contributions: cross section and angular distributions. This extension of the method was tested with exact perturbation of angular distribution and fission spectrum.

14 MeV neutron streaming calculations for JET-like maze entrance using TRIPOLI-4 Monte Carlo code

For the future ITER (International Thermonuclear Experimental Reactor) operation needs, D-T fusion experiments named DTE2 will be conducted in 2020 at JET (Joint European Torus). During the DTE2 operations, 14 MeV fusion neutrons will be generated. One of the aims of the DTE2 neutronics experiments is to investigate the D-T neutron streaming along the penetrations of biological shield of the JET Torus Hall. The TRIPOLI-4 Monte-Carlo radiation transport code has been extensively used on radiation shielding analyses. The purpose of this study is to evaluate the variance reduction (VR) techniques of TRIPOLI-4 on the 14 MeV neutron streaming calculations for a JET-like maze entrance. A JET-like torus hall building of 40 m x 40 m x 4.2 m and a maze entrance of 11.8 m length in 5-section configuration of concrete were first modeled. The standard INIPOND VR method and the new developed VR method AMS (Adaptive Multilevel Splitting) were then presented and discussed. Using the single-step approach combined with AMS techniques, the TRIPOLI-4 calculations results including neutron flux and ambient dose equivalent maps, the VR performance, and the user-friendly AMS input of TRIPOLI-4 code are reported.

Analysis of Dogleg Duct Experiments With 14-MeV Neutron Source Using TRIPOLI-4 Monte Carlo Transport Code

TRIPOLI-4 Monte Carlo transport code, developed by CEA, has been widely used in fission reactor physics, and it can be used in fusion device neutronics. In order to verify the calculation features of TRIPOLI-4 code, a simple dogleg duct model was built to simulate the 14-MeV neutron transport based on a SINBAD fusion benchmark, called dogleg duct streaming experiment. The reaction rates in the bent duct and on the back surface of the iron experimental assembly for 93Nb(n,2n)92mNb, 115In(n,n’)115mIn, and 197Au(n, $\gamma$ )198Au neutron activation dosimeters were calculated with the TRIPOLI-4 code. To improve the calculation efficiency, variance reduction techniques of TRIPOLI-4 were also performed. The calculation results showed that the variance reduction methods of the TRIPOLI-4 code are helpful and obviously decrease the calculated time and increase the convergence efficiency. The calculated reaction rate results of 11 points inside and outside of the dogleg duct assembly were taken into account. Results from the TRIPOLI-4 simulation were compared with the experimental ones obtained from the measurements of Fusion Neutronics Source facility in the Japan Atomic Energy Agency. The benchmark results show that the TRIPOLI-4 code has a good potential to calculate and estimate neutron streaming effects in fusion device design.

Radiation shielding calculations for a 3D ITER benchmark model using TRIPOLI-4® Monte Carlo code

The International Thermonuclear Experimental Reactor (ITER) is currently under construction in France. From the radiation shielding and radiation protection points of view, an intense and large neutron source with energy around 14.1 MeV will be generated from the D-T plasma zone during the ITER operation and diverse gamma-ray sources from neutron activation of the reactor structure materials and coolant should be considered for the reactor operation and maintenance. To decrease the radiation impacts caused by these neutron and gamma sources, iterative designs and nuclear analyses of ITER components are currently performed with three dimensional Monte Carlo radiation transport calculations. Due to the important dimensions of ITER, the thick tokamak blanket modules, and the diagnostic and functional port openings, variance reduction techniques are essential in these Monte Carlo neutron transport calculations. To verify the reactor components design models and to check the radiation transport calculation results, advanced graphic features of the calculation tool are also necessary. With the growing interest in using the TRIPOLI-4® Monte Carlo radiation transport code for ITER applications, the aim of this paper is to study the feasibility to use variance reduction features of TRIPOLI-4 code on a 3D ITER benchmark model which is a 40° toroidal segment including 5796 vol cells. The calculation results reported in this paper include the axial and radial profiles of the inboard TF coil heating and the neutron flux attenuation through the equatorial port plugs and shield. The performance of the TRIPOLI-4 graphic tool under its parallel computing mode was also evaluated.

Re-interpretation of the ERMINE-V experiment validation of fission product integral cross section in the fast energy range

The current knowledge of nuclear data in the fast neutron energy range is not as good as in the thermal range, resulting in larger propagated uncertainties in integral quantities such as critical masses or reactivity effects. This situation makes it difficult to get the full benefit from recent advances in modeling and simulation. Zero power facilities such as the French ZPR MINERVE have already demonstrated that they can contribute to significantly reduce those uncertainties thanks to dedicated experiments. Historically, MINERVE has been mainly dedicated to thermal spectrum studies. However, experiments involving fast-thermal coupled cores were also performed in MINERVE as part of the ERMINE program, in order to improve nuclear data in fast spectra for the two French SFRs: PHENIX and SUPERPHENIX. Some of those experiments have been recently revisited. In particular, a full characterization of ZONA-1 and ZONA-3, two different cores loaded in the ERMINE V campaign, has been done, with much attention paid to possible sources of errors. It includes detailed geometric descriptions, energy profiles of the direct and adjoint fluxes and spectral indices obtained thanks to Monte Carlo calculations and compared to a reference fast core configuration. Sample oscillation experiments of separated fission products such as 103Rh or 99Tc, which were part of the ERMINE V program, have been simulated using recently-developed options in the TRIPOLI-4 code and compared to the experimental values. The present paper describes the corresponding results. The findings motivate in-depth studies for designing optimized coupled-core conditions in ZEPHYR, a new ZPR which will replace MINERVE and will provide integral data to meet the needs of Gen-III and Gen-IV reactors.

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

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.

Analysis of the ITER computational shielding benchmark with the Monte Carlo TRIPOLI-4® neutron gamma coupled calculations

With the growing interest in using the continuous-energy TRIPOLI-4® Monte Carlo radiation transport code for ITER applications, a key issue that arises is whether or not the released TRIPOLI-4 code and its associated nuclear data libraries are verified and validated for the D–T fusion neutronics calculations. Previous published benchmark results of TRIPOLI-4 code on the ITER related activities have concentrated on the first wall loading, the reactor dosimetry, the nuclear heating, and the tritium breeding ratio. To enhance the TRIPOLI-4 verification and validation on neutron-gamma coupled calculations for fusion device application, the computational ITER shielding benchmark of M. E. Sawan was performed in this work by using the 2013 released TRIPOLI-4.9S code and the associated CEA-V5.1.1 data library. First wall, blanket, vacuum vessel and toroidal field magnet of the inboard and outboard components were fully modelled in this 1-D toroidal cylindrical benchmark. The 14.1MeV source neutrons were sampled from a uniform isotropic distribution in the plasma zone. Nuclear responses including neutron and gamma fluxes, nuclear heating, and material damage indicator were benchmarked against previous published results. The capabilities of the TRIPOLI-4 code on the evaluation of above physics parameters were presented. The nuclear data library from the new FENDL-3.0 evaluation was also benchmarked against the CEA-V5.1.1 results for the neutron transport calculations. The results show that both data libraries can be run with TRIPOLI-4 for the fusion neutronics study. This work also demonstrates that the “safety factors” concept is necessary in the nuclear analyses of ITER.

TRIPOLI-4 Simulation of Core Physics Parameters for Two 3600-MW(thermal) Sodium-Cooled Fast Reactors

To support the development of Sodium-cooled Fast Reactors (SFRs) of Generation IV nuclear energy systems and to study the use of the TRIPOLI-4® Monte Carlo code and the JEFF-3.1.1 nuclear data library on the core neutronics of large fast neutron reactors, in this work two recent Organisation for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA) computational benchmarks of two 3600-MW(thermal) SFRs were analyzed with the continuous-energy TRIPOLI-4 code. Both a mixed oxide [(U,Pu)O2] core and a carbide [(U,Pu)C] core were investigated. Under two different fast neutron spectra, the reactor physics parameters—Keff, βeff (effective delayed neutron fraction), sodium void worth, Doppler constant, control rod worth, and core power distribution—were calculated for the beginning of equilibrium cycle condition. Both the pin-by-pin heterogeneous and fuel assembly–level homogeneous calculation models were applied in the whole-core simulation in order to evaluate their impact on the calculation results of SFR reactor physics parameters. The ENDF/B-VII.0 data library from the evaluation was also used with TRIPOLI-4 to study its impact on the SFR core reactivity and the boron carbide control rod worth. Using the mesh tally option, the energy deposition tally, and the upgraded display tool of TRIPOLI-4, radial power distribution and core power maps of the two cores were calculated and compared.

Reactor radiation skyshine calculations with TRIPOLI-4 code for Baikal-1 experiments

*Neutron and photon fluxes generated from the operation of fission reactors and fuel cycle facilities are normally well attenuated with the shielding and ceiling walls. However, skyshine dose rates were observed at 1.5 - 2 km away from the JCO criticality accident site at Tokai-mura. In this study reactor neutron and photon skyshine calculations of Baikal-1 experiments have been performed with the TRIPOLI-4 Monte Carlo transport code. Baikal-1 experiments were carried out with the 300 kW RA research reactor. The 85 cm thick concrete shielding plug in the upper direction of the RA reactor was removed during the experiments. Various distances up to 1500 m from the reactor were considered for skyshine radiation measurements. To study the neutron and photon skyshine dose rates off the reactor with TRIPOLI-4 code, both photon and coupled neutron-photon calculations were performed so as to include primary photons, primary neutrons, and secondary photons contributions. TRIPOLI-4 perturbation calculations on the ground density and on the hydrogen concentration contained in the ground composition were also applied to study the groundshine.

Benchmark test of TRIPOLI-4 code through simple model calculation and analysis of fusion neutronics experiments at JAEA/FNS

In order to examine a basic performance of the TRIPOLI code, two types of analyses were carried out with TRIPOLI-4.4 and MCNP5-1.40; one is a simple model calculation and the other is an analysis of iron fusion neutronics experiments with DT neutrons at the Fusion Neutronics Source (FNS) facility in Japan Atomic Energy Agency (JAEA). In the simple model calculation, we adopted a sphere of 0.5m in radius with a 20MeV neutron source in the center and calculated leakage neutron spectra from the sphere. We also analyzed in situ and Time-of-Flight (TOF) experiments for iron at JAEA/FNS. For the in situ experiment, neutron spectra and reaction rates for dosimetry reactions were calculated for several points inside the assembly. For the TOF experiment, angular neutron leakage spectra from the assembly were calculated. Results with TRIPOLI were comparable to those with MCNP in most calculations, but a difference between TRIPOLI and MCNP calculation results, probably caused by inadequate treatment of inelastic scattering data in TRIPOLI, appears in some calculations.

APOLLO2 and TRIPOLI4 solutions of the OECD VVER-1000 LEU and MOX assembly benchmark

The OECD benchmark for VVER-1000 UGd and MOXGd assemblies was solved with the APOLLO2 and TRIPOLI4 codes. The objective was to verify the TRIPOLI4 Monte-Carlo solution and to assess the APOLLO2 method of characteristics based calculation routes for VVER assemblies. The test problems address important VVER V&V topics such as advanced fuel assemblies, depletion and branch calculations. Solutions with Monte-Carlo and deterministic codes from the OECD benchmark report are available for comparison. The APOLLO2 results obtained with reference and two-level 281/37g MOC calculation schemes are close to the Monte-Carlo reference solutions and the mean of all codes. The higher-order Linear Surface MOC is shown to give accurate and computationally efficient solutions.

Contributions to the neutronic analysis of a gas-cooled fast reactor

In this work the Monte Carlo codes MCNPX and TRIPOLI-4 were used to perform the criticality calculations of the fuel assembly and the core configuration of a gas-cooled fast reactor (GFR) concept, currently in development. The objective is to make contributions to the neutronic analysis of a gas-cooled fast reactor. In this study the fuel assembly is based on a hexagonal lattice of fuel-pins. The materials used are uranium and plutonium carbide as fuel, silicon carbide as cladding, and helium gas as coolant. Criticality calculations were done for a fuel assembly where the axial reflector thickness was varied in order to find the optimal thickness. In order to determine the best material to be used as a reflector, in the reactor core with neutrons of high energy spectrum, criticality calculations were done for three reflector materials: zirconium carbide, silicon carbide and natural uranium. It was found that the zirconium carbide provides the best neutron reflection. Criticality calculations using different active heights were done to determine the optimal height, and the reflector thickness was adjusted. Core criticality calculations were performed with different radius sizes to determine the active radial dimension of the core. A negative temperature coefficient of reactivity was verified for the fuel. The effect on reactivity produced by changes in the coolant density was also evaluated. We present the main neutronic characteristics of a preliminary fuel and core designs for the GFR concept. ENDF-VI cross-sections libraries were used in both the MCNPX and TRIPOLI-4 codes, and we verified that the obtained results are very similar.

Validation of the sodium void reactivity effect prediction using JEFF-3.1 nuclear data

Near 60 Na void experiments performed in the zero power reactors MASURCA (CEA-Cadarache) and ZPPR (Argonne West – Idaho) have been analyzed using JEFF-3.1 nuclear data and the ERANOS-2.1 (deterministic) and TRIPOLI-4 (Monte-Carlo) codes. Some comparative calculations have been performed also using either JEFF-3.1, ENDF/B-VII.0 or JENDL-3.3 nuclear data for 23Na, as these three 23Na evaluations show marked differences. The Na void experiments have been selected to cover spectral conditions ranging from the relatively hard flux in the outer zone of a small fast reactor to the relatively soft flux in the inner zone of a large fast reactor. For in-fuel Na void patterns, there is a good agreement between ERANOS and TRIPOLI computations, while the deterministic calculations significantly underestimate the leakage component for Na void patterns in fertile regions. The agreement between ERANOS-2.1 + JEFF-3.1 predictions and experimental values is excellent for in-fuel Na void patterns in MASURCA experiments, but a significant underestimation of the leakage component occurs for in-fuel Na void patterns in ZPPR. For fertile Na void patterns, there is a clear underestimation of the leakage component, quantitatively different for MASURCA and ZPPR experiments. Variations in 23Na cross-section data also result in significant differences: ENDF/B-VII.0 and JENDL-3.3 nuclear data for 23Na increase noticeably the predicted Na void worth values with respect to JEFF-3.1 data. The three 23Na evaluations differ at high energy (>500 keV, and especially >2 MeV), and this stresses the need for accurate additional measurements in this energy range.

Analysis of the tritium breeding ratio benchmark experiments using the Monte Carlo code TRIPOLI-4

Tritium breeding is an essential element of fusion nuclear technology. A tritium breeding ratio greater than unity is necessary for self-sufficient fueling. To simulate the 14 MeV neutron transport in tritium breeding systems from the D–T fusion reaction, the 3D realistic modeling with Monte Carlo code and the point-wise nuclear data are recommended. Continuous-energy TRIPOLI-4 Monte Carlo transport code has been widely used on the radiation shielding, criticality safety, and fission reactor physics. For supporting the ITER TBM (test blanket module) neutronics study with TRIPOLI-4 code, this paper presents the TRIPOLI-4 simulation of TBR (tritium breeding ratio) for six OKTAVIAN spherical assemblies of Osaka University: Li, Li–C, Pb–Li, Pb–Li–C, Be–Li, and Be–Li–C. It also investigates the impact of nuclear data libraries on TBR calculations from ENDF/B-VI.4, ENDF/B-VII.0, JEFF-3.1, JENDL-3.3, and FENDL-2.1. In general, TRIPOLI-4 produced satisfactory C/E values. Only beryllium of JEFF-3.1 library introduces higher uncertainties.

97/00407 ENDF/B-VI benchmark calculations for the Doppler coefficient of reactivity

Tritium breeding is an essential element of fusion nuclear technology. A tritium breeding ratio greater than unity is necessary for self-sufficient fueling. To simulate the 14 MeV neutron transport in tritium breeding systems from the D–T fusion reaction, the 3D realistic modeling with Monte Carlo code and the point-wise nuclear data are recommended. Continuous-energy TRIPOLI-4 Monte Carlo transport code has been widely used on the radiation shielding, criticality safety, and fission reactor physics. For supporting the ITER TBM (test blanket module) neutronics study with TRIPOLI-4 code, this paper presents the TRIPOLI-4 simulation of TBR (tritium breeding ratio) for six OKTAVIAN spherical assemblies of Osaka University: Li, Li–C, Pb–Li, Pb–Li–C, Be–Li, and Be–Li–C. It also investigates the impact of nuclear data libraries on TBR calculations from ENDF/B-VI.4, ENDF/B-VII.0, JEFF-3.1, JENDL-3.3, and FENDL-2.1. In general, TRIPOLI-4 produced satisfactory C/E values. Only beryllium of JEFF-3.1 library introduces higher uncertainties.