Nuclear Code Research Articles

Excitation functions of (n, p) reactions for the production of iodine isotopes used in medical applications

The cross sections of 124Xe (n, p)124I, 125Xe (n, p)125I, 126Xe (n, p)126I, 128Xe (n, p)128I, 129Xe (n, p)129I, 130Xe (n, p)130I, 131Xe (n, p)131I, 132Xe (n, p)132I, 133Xe (n, p)133I, 134Xe (n, p)134I and 135Xe (n, p)135I for production of iodine isotopes were calculated at neutron energy ranges between 1 and 20 MeV with the help of three computer nuclear codes (EMPIRE 3.2.3, ALICE-ASH and TALYS 1.95). Furthermore, the cross sections were calculated by the empirical formula derived in some previous studies at 14.5 MeV. The previous calculated data from the evaluated data from the Evaluated Nuclear Data File (ENDF), such as ENDF/B-VIII.0; JEFF-3.3; ROSFOND-2010; CENDL-3.2; JENDL-4.0; JENDL-5.0; and TENDL-2021, as well as the experimental nuclear reaction cross-section data (EXFOR) about the above reactions, have been collected. In general, the calculated results by computer codes are in good agreement when compared with earlier empirical formulas, the measured values in the EXFOR library, and the evaluated data (ENDF/B-VIII.0; JEFF-3.3; JENDL-4.0; JENDL-5.0; ROSFOND-2010; CENDL-3.2; and TENDL-2021) with some exceptional.

Predicting the production of 51Mn as a useable radionuclide for PET imaging via the 52Cr(p,2n)51Mn reaction using GEANT4, MCNPX, TALYS, and EMPIRE codes

51Mn is an ideal radioisotope for positron emission tomography (PET) in pancreatic imaging dues to its positron emission characteristics (Eβ+ = 963.7 keV, Iβ+ = 97%) and its advantageous short half-life (T1/2 = 46.2 min). In this study, the production of 51Mn via the 52Cr(p,2n)51Mn reaction was predicted by calculating cross-section and production yield parameters using GEANT4-10.7, MCNPX, TALYS-1.96, SRIM, and EMPIRE-3.2.3 nuclear codes. The GEANT4 and SRIM codes were applied to investigate range of protons in 52Cr target. The cross-section of the 52Cr(p,2n)51Mn reaction was determined by GEANT4, EMPIRE, and four methods of TALYS code. The thick yield of the reaction was evaluated via two different ways of GEANT4 toolkit, and also TALYS and MCNPX codes. At the end, the simulated values were compared to previous reported experimental data, which were in good compatibility to each other.

Thermo-neutronic integrated coupling effects on nuclear reactor core calculations

Small Modular Reactors (SMRs) have gained significant attention as the next generation of nuclear reactors due to their unique features, such as improved economic viability and safety. This study focuses on the SMART reactor, the first certified design of an SMR, and investigates various Thermo-Neutronics aspects using an integrated coupling scheme of nuclear codes. The Neutronics cell and core calculations are performed using the DRAGON/DONJON codes, while the COBRA code is employed for thermal-hydraulic calculations. The analysis considers parameters including axial and radial power peaking factors (PPFs), critical heat flux (CHF), minimum departure from nucleate boiling ratio (MDNBR), axial average coolant temperature in hot channel and core, and maximum fuel temperature. These parameters are evaluated using the developed coupling system. Code-to-code verification is conducted to ensure the accuracy of the results, demonstrating good agreement. The coupling approach reveals significant effects compared to stand-alone code modeling, emphasizing the importance of considering the coupling between Neutronics and thermal-hydraulics, especially during the cycle length. This study contributes to the understanding of SMRs, particularly the SMART reactor, by providing insights into key parameters and their interdependencies. The integrated coupling scheme enhances the accuracy of the analysis and highlights the significance of considering coupling effects in the modeling process during the steady state and the cycle length.

Activation cross section for 85Rb(n,2n)84mRb and 85Rb(n,p)85mKr reactions with uncertainty propagation and covariance analysis

The cross section of 85Rb(n,2n)84mRb and 85Rb(n,p)85mKr reactions were measured at neutron energy range 12–20 MeV using activation analysis followed by off-line γ-ray spectroscopic technique. The quasi mono-energetic neutrons were produced through 7Li(p,n)7Be reaction. The measurements were done relative to 27Al(n,α)24Na reference monitor reaction cross section. The detailed uncertainty propagation from the attributes present in measurements was performed using covariance analysis. The γ-ray self-attenuation and background low neutron energy corrections were performed in the measurement studies. Theoretical calculations were performed by TALYS-1.96 nuclear code. The comparison of measured values with the available data in EXFOR database and evaluated data from JENDL-5 and EAF-2010 shows the measured cross section values of 85Rb(n,p)85mKr reaction were slightly higher than the published values, evaluated data and theoretical data while for 85Rb(n,2n)84mRb, the measured values were consistent with the published data.

Overview of T and D–T results in JET with ITER-like wall

Abstract In 2021 JET exploited its unique capabilities to operate with T and D–T fuel with an ITER-like Be/W wall (JET-ILW). This second major JET D–T campaign (DTE2), after DTE1 in 1997, represented the culmination of a series of JET enhancements—new fusion diagnostics, new T injection capabilities, refurbishment of the T plant, increased auxiliary heating, in-vessel calibration of 14 MeV neutron yield monitors—as well as significant advances in plasma theory and modelling in the fusion community. DTE2 was complemented by a sequence of isotope physics campaigns encompassing operation in pure tritium at high T-NBI power. Carefully conducted for safe operation with tritium, the new T and D–T experiments used 1 kg of T (vs 100 g in DTE1), yielding the most fusion reactor relevant D–T plasmas to date and expanding our understanding of isotopes and D–T mixture physics. Furthermore, since the JET T and DTE2 campaigns occurred almost 25 years after the last major D–T tokamak experiment, it was also a strategic goal of the European fusion programme to refresh operational experience of a nuclear tokamak to prepare staff for ITER operation. The key physics results of the JET T and DTE2 experiments, carried out within the EUROfusion JET1 work package, are reported in this paper. Progress in the technological exploitation of JET D–T operations, development and validation of nuclear codes, neutronic tools and techniques for ITER operations carried out by EUROfusion (started within the Horizon 2020 Framework Programme and continuing under the Horizon Europe FP) are reported in (Litaudon et al Nucl. Fusion accepted), while JET experience on T and D–T operations is presented in (King et al Nucl. Fusion submitted).

The combination effect of optical potential and nuclear level density for alpha induced reaction cross sections on 92,94,95Mo isotopes

The theoretical studies of charged particle induced reactions are important to improve and study the radioisotope production, especially in reactions where the experimental data are incomplete or insufficient. In this study, cross section of alpha induced of 92,94,95Mo isotopes are calculated and analyzed by using different alpha optical model potentials and nuclear level density models with TALYS code. The effects of alpha optical model potentials and nuclear level density are analyzed separately and together for each reaction. To determine the best combination of models, chi-squared values are calculated for all combination cases. The commonly used program is TALYS to calculate the reaction cross section. For the calculations, the latest version of TALYS nuclear code version 1.96 is used. The experimental data are taken carefully for all reactions from experimental nuclear reaction data base (EXFOR). The obtained results are compared with these data and discussed. A good agreement between the calculated and experimental data is clearly presented for all analyzed reactions. It is seen from the results that alpha optical model potentials and nuclear level density models have an effective role on cross section calculations of alpha induced reactions.

Neutronic analysis of a tokamak hybrid blanket cooled by thorium-molten salt fuel mixture

In this study, a novel approach has been investigated by using a mixture of thorium and molten salts as a dual-purpose coolant and medium for the production of fissile fuel in a Fusion Fission Hybrid Reactor (FFHR) for the reference geometry of ITER. The study highlighted the broader benefits of thorium fuel cycling, safety features, and reduced radioactive minor actinides generation. The use of a thorium-melted salt coolant for fissile fuel production in a fusion-fission hybrid reactor represented a promising path towards efficient and sustainable nuclear energy, with potential benefits in terms of safety features and reduced generation of radioactive minor actinides. In this study, SS 316 LN-IG was selected as the first wall material for the reactor, and a molten salt fuel mixture of LiF–ThF4 was used as the coolant, taking into account the eutectic points of the material, the nominal fusion power in the FFHR for the Tokamak design concept is considered to be 500 MW. The nuclear code MCNP6 was used with the nuclear data libraries ENDF/B-VIII and CLAW-IV for the neutron calculations. The time evolution of the isotopes in the reactor was calculated with the interface code MCNPAS. The study results are evaluated in terms of tritium breeding ratio, energy multiplication factor, radiation damage, fissile fuel production and fuel burn-up value.The 4-year operation history of total TBR value is calculated and always above 1.05 and increases with time.Th initially decreased from 631.3 tonnes to 587.2 tonnes, while 233U production during this period was 9.1 tonnes. According to these results, the first wall replacement period was calculated as 3.9 years.

Use of rational approximations for phonon density of states in evaluation of neutron thermal scattering law

We propose and explore a new analytical formulation for evaluation of Thermal Scattering Law (TSL) using continuous phonon spectra for neutron inelastic scattering. It involves approximating Phonon Density of States (PDOS) in terms of rational approximations in energy transfer. Debye Waller coefficient calculated with these forms are in reasonable agreement with LEAPR calculated values. TSL can be expanded as a power series, each term of which can be integrated analytically. The method leads to TSL being a series of exponential cosine integrals which again have closed rational forms in energy transfer. Their shapes are analogous to asymmetric resonance line shapes and show maxima for certain values of energy transfer which could correspond to phonon emission or absorption. The series is shown to be convergent and easy to compute as it does not involve any numerical integration. The approach is applied to Beryllium metal and Graphite which have vastly different PDOS shapes, and is shown to be capable of giving most of the trends shown by TSL and cross section results from NJOY nuclear code processing system, even with smaller number of power series terms. Since TSL can be calculated for both negative and positive energy transfer with equal ease, one can easily check for the validity of the approximation obeying Principle of Detailed Balance (PDB). The results overall have shown promise of our formulation in making a case for rapid calculations of TSL and also for using the rational forms in energy transfer for analysis of experimental/calculated PDOS data.

Coupling between LOTUS and CTF with DYN3D within a multiscale and multiphysics software development

Coupling of nuclear codes can be performed at several scale levels and is necessary to improve their reliability and sustainability. Mainly, the coupling of nuclear codes that use nodal expansion neutronics with channel thermal hydraulics has been carried out at the fuel assembly level while, only recently, the coupling of nuclear codes that use advanced neutronics, thermal hydraulics, and thermo-mechanics has been carried out at either the fuel pin or materials level. Meanwhile, in the UK, a multiscale and multi-physics software development between NURESIM and CASL is being developed, which includes a coupling software environment that enables the coupling of nuclear codes at several scale levels. Full coupled reactor physics at either the fuel pin or materials level can be obtained by coupling the transport code LOTUS, the subchannel code CTF, and the nodal code DYN3D. In this journal article, a multi ways coupling between LOTUS and CTF at either the fuel pin or materials levels with DYN3D at the fuel assembly level is compared to a multi ways coupling between DYN3D and CTF at the fuel pin level with DYN3D at the fuel assembly level and a multi ways coupling between Open MC and CTF at the materials level with DYN3D at the fuel assembly level. These comparisons have been carried out to present the coupled reactor physics verifications at either the fuel pin or materials levels. These show that the multi ways coupling between LOTUS and CTF at either the fuel pin or materials levels with DYN3D at the fuel assembly level outperforms the multi ways coupling between DYN3D and CTF at the fuel pin level with DYN3D at the fuel assembly level due to the application in the former of full neutron transport. Also, these show that the multi ways coupling between LOTUS and CTF at the materials level with DYN3D at the fuel assembly level agrees with the multi ways coupling between Open MC and CTF at the materials level with DYN3D at the fuel assembly level due to the application in both of full neutron transport.

Implementation of dynamic start-up test experimental data as a main part of the nuclear code validation procedure: Developed RELAP5 model for VVER-1000

The main purposes of start-up tests in nuclear power plants (NPPs) are to ensure safe and reliable operation, verify system functionality, comply with regulatory requirements, optimize performance, and establish a foundation for ongoing plant operation and maintenance. However, the start-up tests of NPPs also could be used as a main part of the nuclear code validation procedure for several reasons including: realistic simulation, comprehensive evaluation, detection of code limitations, validation of safety margins and confidence in code predictions. The main purpose of the current study is to define and assess the validation procedure based on actual start-up test data. In this regard, the developed RELAP5 model has been validated against the actual data of VVER-1000 plant during a dynamic start-up test. The results of this full-scale validation show a good agreement between the developed RELAP5 model results and actual plant data. Finally, by defining a step by step validation procedure, it has been recommended to use the start-up phase test data as a more robust validation process which allow for full-scale validation of the nuclear code by comparing its predictions with actual plant measurements and also other advantages which have been demonstrated in the current study.

Comparison of PM-HIP to forged SA508 pressure vessel steel under high-dose neutron irradiation

Powder metallurgy with hot isostatic pressing (PM-HIP) is an advanced manufacturing process that is envisioned to replace forging for heavy nuclear components, including the reactor pressure vessel (RPV). But PM-HIP products must at least demonstrate comparable irradiation tolerance than forgings in order to be qualified for nuclear applications. The objective of this study is to directly compare PM-HIP to forged SA508 Grade 3 Class 1 low-alloy RPV steel at two neutron irradiation conditions: ∼0.5–1.0 displacements per atom (dpa) at ∼270 °C and ∼370 °C. PM-HIP SA508 experiences greater irradiation hardening and embrittlement (total elongation) than forged SA508. However, uniform elongation and approximate toughness are comparable across all irradiated materials, suggesting irradiated PM-HIP SA508 exhibits superior ductility at maximum load-bearing capacity. The irradiation hardening mechanism is linked to composition rather than fabrication method. Since PM-HIP SA508 has higher Mn and Ni concentration, it is more susceptible to irradiation-induced nucleation of Mn-Ni-Si-P (MNSP) nanoprecipitates and dislocation loops, which both contribute to hardening. Conversely, the forged material nucleates fewer MNSPs, causing dislocation loops to control irradiation hardening. These results show promise for the irradiation performance of PM-HIP SA508 and can motivate future nuclear code qualification of PM-HIP fabrication for RPVs.

Excitation functions of alpha-particle induced nuclear reactions on nat Sn

Abstract Excitation functions of alpha-particle induced nuclear reactions on nat Sn have been presented in the 11–40 MeV energy range. In the present study, the stacked-foil activation technique followed by offline gamma-ray spectrometry was used to measure the production of 119Te, 121Te, 122Sb and 126Sb from alpha-particle induced reactions on nat Sn. The TALYS nuclear code was used to calculate the theoretical predictions of the excitation functions of nat Sn(α,x) nuclear reactions. The measured data of the above-mentioned nuclear reactions were compared with the theoretical predictions and the experimental results available from EXFOR. In this study, covariance analysis was performed to calculate the uncertainty propagation in the measured cross sections.

NEA GitLab: A centralised platform for development and deployment of nuclear codes and data

The NEA Data Bank is an international reference centre for computer codes, nuclear and thermochemical data which has traditionally used simple file servers, and even DVDs, to deliver valuable content to end users across the globe. With the recent implementation of a self hosted GitLab system at the NEA it has enabled the Data Bank to streamline delivery, automate processes and testing, while empowering code owners and developers with a secure platform to collaborate and develop codes. In this paper we present the NEA Git- Lab system with some concrete examples of codes such as Kraken, PHITS and FISPACT-II taking advantage of many of the services and functionalities provided by GitLab. We illustrate methodologies on how to work effectively with third party software in a position of being a custodian of code, rather that a code owner, providing DevSecOps as a service.

Proton, neutron and deuteron induced nuclear reactions in medical application

Radioisotopes production is important in medical science as they are used for different purposes. The studies of the nuclear reactions cross-section for the production of 64Cu were performed for the nuclear reactions of 64Ni(p, n)64Cu, 65Cu(p, x)64Cu, 67Zn(p, a)64Cu, 65Cu(n, 2n)64Cu, 64Zn(n, p)64Cu, 64Ni(d, 2n)64Cu, natNi(d,x)64Cu and 66Zn(d, a)64Cu. The calculations were performed using TALYS1.6 nuclear code. The results of excitation functions were compared with experimental data from letratures. It was seen that the results are important to produce copper-64 using different reaction mechanism and thus experimental results and calculations are in good agreement for interesed reactions types.

Study of trends in cross-section values of primary reactions of most abundant isotope of elements with Z= 6 to 116 induced by 14 MeV neutrons using TALYS code and the EXFOR database

Study was carried out to understand the trends of variation of cross-section values of all possible primary reactions induced by 14 MeV neutrons incident on the most abundant isotope of elements from atomic number 6 to 116. The available nuclear reaction channels and cross-section values of all possible primary reactions were obtained using the theoretical values calculated using TALYS nuclear code version 1.96. The trends were explained on the basis of nuclear binding energy curve, the pairing effect and shell effect. Most of the trends show a regular variation except at some points, which give interesting information for the nucleus where trend is irregular. The variation of the available experimental cross-section values obtained from the EXFOR database, if any, was also obtained and compared with the trends obtained from the theoretical values. The discrepancies in the trends of theoretical and experimental cross-section values of various possible reactions were explained on the basis of the possible errors in experiments and the assumptions of the theoretical model of the TALYS nuclear code.

Is the Paleomammalian Brain of Leaders an Immutable Barrier to Lasting World Harmony?

This exploratory paper hypothesizes that sustained universal peace is inexorably limited by the human genetic drive to survive. National and corporate leaders need to be particularly aware of their powerful, persistent, reptilian brains, especially when making decisions in conflict situations with the nuclear button readily accessible. Brain structure can explain the complexity. On the one hand, the slower, problem-solving, rational, neo cerebral brain is in the frontal cortex. On the other hand, the paleomammalian, subjective, primeval brain is located in the limbic system. The paleomammalian brain is fast, self-centered, and protective. Its quickness dominates. The unfortunate corrosive sides of humanity emerge from the limbic system when survival is, or perceived to be, threatened. War, cruelty, hypocrisy and impossible fantasies manifest from this brain structure. Given that some leaders—Trump, Putin, and Kim Jong Un, have the nuclear code, the possibility of destroying humanity is real.

Measurement of Isfahan heavy water zero-power reactor kinetic parameters using advanced pulsed neutron source method in a near critical state

The pulsed neutron source (PNS) technique was used to determine the prompt neutron decay constant for two different lattice pitches in the HWZPR heavy water zero power reactor. The results were compared to the variance-to-mean ratio (VTM) method. The neutron mean generation time was also calculated for both pitches, and the results were compared to previous Monte Carlo calculations. The findings of this research can be used as a benchmark nuclear codes to validate kinetic parameters.

Design studies for A 50 MWth molten salt fast reactor

In this paper, design studies for a 50 MWth Molten Salt Reactor (MSR) have been carried out for the eutectic point of the molten salt mixture. Neutronic calculations are performed with the 19.75% enriched uranium and 100% 7Li isotope contained. The MCNP6 nuclear code was used with the ENDF/B-VIII nuclear data library to determine geometry dimensions and criticality. The time-evolution of Pu and other heavy isotopes in the reactor are calculated with the interface code MCNPAS. Four models are investigated with different reactor vessel materials: Model 1: Ni alloy (NiCrW-Hastelloy steel), Model 2: Beryllium, Model 3: Graphite and Model 4: Silicon Carbide (SiC). For the respective models, time-dependent criticality calculations are performed with a startup criticality value of keff = 1.0262, 1.0298, 1.0404, and 1.0223. The 235U consumed for the corresponding models over the 10 years of reactor operation are 257.9 kg, 258.8 kg, 259.4 kg and 257.9 kg, respectively. At the same time, the 238U consumptions are 400.6 kg, 380.0 kg, 379.8 kg, and 400.9 kg, respectively. The amount of the higher quality new fuel (239Pu) produced for 10 years is calculated as 129.10 kg, 127.41 kg, 122.95 kg and 128.66 kg, for the respected models.

Neural network predictions of [formula omitted] reaction cross sections at 18.5±3 MeV using the Levenberg-Marquardt algorithm

In recent developments, artificial neural networks (ANNs) have demonstrated their capability to predict reaction cross-sections based on experimental data. Specifically, for predicting (α,n) reaction cross-sections, we meticulously fine-tuned the neural network’s performance by optimizing its parameters through the Levenberg-Marquardt algorithm. The effectiveness of this approach is corroborated by notable correlation coefficients; an R-value of 0.90928 for overall correlation, 0.98194 for validation, 0.99981 for testing, and 0.94116 for the comprehensive network prediction. We conducted a rigorous comparison between the results and theoretical computations derived from the TALYS 1.95 nuclear code to validate the predictive accuracy. The mean square error value for artificial neural network results is 7620.92, whereas for TALYS 1.95 calculations, it has been found to be 50,312.74. This comprehensive evaluation process validates the reliability of the ANN based on the Levenberg-Marquardt algorithm in approximating the reaction sections, thus demonstrating its potential for comprehensive investigations. These recent developments confirm the feasibility of using ANN models to gain insight into (α,n) reaction cross-sections.

Embrittlement of WCLL blanket and its fracture mechanical assessment

In the European fusion programme, the Water Cooled Lithium Lead breeding blanket (WCLL BB) uses EUROFER as a structural material cooled with water at temperatures between 295 °C–328 °C and a pressure of 155 bar. The WCLL BB will be significantly irradiated (>2 dpa), while some parts will not receive significant heat loads, e.g. the sidewalls or the back-supporting structures. The irradiation, together with the irradiation temperature of EUROFER below 350 °C, produces a shift of the ductile-to-brittle-transition temperature (DBTT) to levels above room temperature at neutron doses, causing material damage as low as 2–3 dpa. Even though the DBTT does not reach the operating temperature level, brittle/non-ductile fracture is a concern during in-vessel maintenance when the BB temperature is below the DBTT. Two loading scenarios were identified as severe in this respect: (i) re-pressurization of the WCLL BB cooling loop after in-vessel maintenance, and (ii) dead weight loads during lifting of the BB segment. The embrittlement of the WCLL BB was investigated by quantifying the local DBTT shift in its parts based on current knowledge of the embrittlement behaviour of EUROFER under neutron irradiation. Therefore, a suitable, not overly conservative procedure was derived considering dpa damage and transmuted helium effects. The results demonstrate the ability to identify the 3D spread of the severely embrittled zones in the structure whose impact on the structural integrity was assessed considering the risk of brittle/non-ductile fracture. Thereby, the fracture mechanics approach established in nuclear codes was applied assuming its applicability to EUROFER. The embrittled zones in the first wall (FW) and its sidewalls pass the criteria when assessing the relatively low stresses resulting from the coolant pressure. The assessment was then continued considering stresses appearing in the FW during maintenance, in particular, when lifting the BB segment and transporting it out of the vacuum vessel. In this context, the maximum tolerable flaw sizes were determined in a parameter study considering designs of the FW with different cooling channel wall thicknesses.