Overview of interpretive modelling of fusion performance in JET DTE2 discharges with TRANSP

In the paper we present an overview of interpretive modelling of a database of JET-ILW 2021 D-T discharges using the TRANSP code. The main aim is to assess our capability of computationally reproducing the fusion performance of various D-T plasma scenarios using different external heating and D-T mixtures, and to understand the performance driving mechanisms. We find that interpretive simulations confirm a general power-law relationship between increasing external heating power and fusion output, which is supported by absolutely calibrated neutron yield measurements. A comparison of measured and computed D-T neutron rates shows that the calculations' discrepancy depends on the absolute neutron yield. The calculations are found to agree well with measurements for higher performing discharges with external heating power above ∼20 $\mathrm{MW}$, while low-neutron shots display an average discrepancy of around +40% compared to measured neutron yields. A similar trend is found for the ratio between thermal and beam-target fusion, where larger discrepancies are seen in shots with dominant beam-driven performance. We compare the observations to studies of JET-ILW D discharges, to find that on average the fusion performance is well modelled over a range of heating power, although an increased unsystematic deviation for lower-performing shots is observed. The ratio between thermal and beam-induced D-T fusion is found to be increasing weakly with growing external heating power, with a maximum value of $\gtrsim$1 achieved in a baseline scenario experiment. An evaluation of the fusion power computational uncertainty shows a strong dependence on the plasma scenario type and fusion drive characteristics, varying between ±25% and 35%. D-T fusion alpha simulations show that the ratio between volume-integrated electron and ion heating from alphas is $\lesssim$10 for the majority of analysed discharges. Alphas are computed to contribute between ∼15% and 40% to the total electron heating in the core of highest performing D-T discharges. An alternative workflow to TRANSP was employed to model JET D-T plasmas with the highest fusion yield and dominant non-thermal fusion component because of the use of fundamental radio-frequency heating of a large minority in the scenario, which is calculated to have provided ∼10% to the total fusion power.This work has been carried out within the framework of the EUROfusion Consortium, funded by the European Union via the Euratom Research and Training Programme (Grant Agreement No. 101052200—EUROfusion). Views and opinions expressed are however those of the author(s) only and do not necessarily reflect those of the European Union or the European Commission. Neither the European Union nor the European Commission can be held responsible for them. This work has been part-funded by the EPSRC Energy Programme with grant number EP/W006839/1. The Barcelona Supercomputing Center part of this work has contributed through the Spanish National R&D Project PID2019-110854RB-I00 funded through MCIN/AEI/10.13039/501100011033. In addition BSC are grateful for the support received from the Departament de Recerca i Universitats de la Generalitat de Catalunya via the Research Group Fusion Group with code: 2021 SGR 00908. The Laboratorio Nacional de Fusión contribution was funded in part via the Spanish National R&D Project PID2021-127727OB-I00 funded through MCIN/AEI /10.13039/501100011033.Peer Reviewed"Article signat per 43 autors/es: Ž. Štancar, K.K. Kirov, F. Auriemma, H.-T. Kim, M. Poradziński, R. Sharma, R. Lorenzini, Z. Ghani, M. Gorelenkova, F. Poli, A. Boboc, S. Brezinsek, P. Carvalho, F.J. Casson, C.D. Challis, E. Delabie, D. Van Eester, M. Fitzgerald, J.M. Fontdecaba, D. Gallart, J. Garcia, L. Garzotti, C. Giroud, A. Kappatou, Ye.O. Kazakov, D.B. King, V.G. Kiptily, D. Kos, E. Lerche, E. Litherland-Smith, C.F. Maggi, P. Mantica, M.J. Mantsinen, M. Maslov, S. Menmuir, M. Nocente, H.J.C. Oliver, S.E. Sharapov, P. Sirén, E.R. Solano, H.J. Sun, G. Szepesi and JET Contributors"Postprint (published version

Analiza nevtronskih diagnostičnih sistemov v velikih tokamakih

Modern fusion neutronics studies play a crucial role in the support of the development of fusion devices. Their contribution varies from the design of plasma diagnostics systems, fusion power measurements, tritium breeding studies, evaluation of radiation induced structural embrittlement to radiation protection of personnel. Present-day neutron calculations are almost in their entirety based on advanced stochastic neutron transport codes. One of the foundations of these programs is knowledge about the neutron source, in our case a hot plasma. Since uncertainties in basic simulation parameters are being propagated through the system together with the neutrons, there is an ongoing effort of trying to identify and study major uncertainty sources and improving existing physics models for describing the generation of neutrons in a tokamak. The dissertation focuses on the description of the plasma as a neutron source and begins with a study of the state-of-the-art modelling capabilities of neutron emission in tokamak plasmas. The core of the thesis is the description and application of a novel methodology for generation of realistic plasma neutron sources, called PLANET. The methodology is based on calculations of plasma transport with the TRANSP code and neutron spectra with the DRESS code, coupled to the MCNP stochastic neutron transport code. Diagnostic data and modelling results of two representative JET deuterium fuel discharges are used for neutron generation in computational JET models, analysing basic source parameters - emissivity profile, spectra shape, source anisotropy and synthetic detector response. By comparing the realistic source results with a thermal plasma, it is shown that discrepancies of integral neutron detector response, from which the total neutron rate and hence the fusion power is calculated, of up to several percent are computed, exhibiting relatively low sensitivity to changes in the neutron source. The analysis of neutron spectra shows distinct structural characteristics which arise due to the fact that plasma heating and fusion reaction anisotropy are modelled. Material activation studies show that certain threshold reactions yield results of orders of magnitude difference for different neutron source models. The developed plasma neutron source is shown to be applicable to detailed tokamak neutron source effect studies.Razvoj modernih fuzijskih naprav v veliki meri temelji na nevtronskih raziskavah. Slednje se uporabljajo za razvoj diagnostičnih sistemov, meritve fuzijske moči, napovedi oplojevanja tritija, ocene sevalnih poškodb materialov in obratovanja v skladu z načeli varstva pred sevanji. Novodobni nevtronski preračuni temeljijo predvsem na programih za stohastični transport nevtronov. Eden izmed temeljev stohastičnih kod je opis nevtronskega izvora, v našem primeru vroče plazme. Ker se negotovosti v izvornih simulacijskih parametrih preko transporta nevtronov propagirajo v izračunane fizikalne količine, je del nevtronskih raziskav namenjen identifikaciji in preučevanju večjih izvorov negotovosti in izboljšavi trenutnih fizikalnih modelov, ki opisujejo nastajanje nevtronov v tokamakih. V disertaciji opisujemo plazemske izvore nevtronov, začenši s študijo najmodernejših zmožnosti modeliranja emisije nevtronov iz plazme tokamaka. Glavni del naloge predstavljata opis in uporaba razvite izvirne metodologije PLANET za ustvarjanje realističnih plazemskih izvorov nevtronov. Metodologija temelji na izračunih plazemskega transporta s programom TRANSP in nevtronskih spektrov s programom DRESS, ki so sklopljeni s kodo za stohastični transport nevtronov MCNP. Diagnostične podatke in rezultate modeliranja dveh reprezentativnih devterijevih plazemskih strelov tokamaka JET uporabimo kot osnovo za postopek žreba izvornih nevtronov v Monte Carlo računskih modelih tokamaka JET, nakar preučimo osnovne izvorne parametre - profil izseva, obliko spektrov, anizotropijo izvora in sintetični odziv diagnostike. Primerjave med realističnimi in poenostavljenimi izvori pokažejo nekaj procentne razlike v integralnem odzivu detektorjev in splošno nizko občutljivost na spremembe v izvoru. Zaradi upoštevanja učinkov gretja plazme in anizotropije fuzijskih reakcij v obliki nevtronskih spektrov opazimo značilne fizikalne strukture. Pri študiji aktivacije vzorcev zabeležimo odstopanja izračunanih reakcijskih hitrosti za več redov velikosti, pri uporabi materialov s pragovnimi reakcijami. V disertaciji pokažemo, da so razviti plazemski izvori nevtronov primerni za podrobne analize učinka izvora na odziv detektorjev v tokamakih

Computational design and characterization of a subcritical reactor assembly with TRIGA fuel

The TRIGA fuel of the Philippine Research Reactor-1 (PRR-1) will be used in a subcritical reactor assembly (SRA) to strengthen and advance nuclear science and engineering expertise in the Philippines. SRA offers a versatile and safe training and research facility since it can produce neutrons through nuclear fission reaction without achieving criticality. In this work, we used a geometrically detailed model of the PRR-1 TRIGA fuel to design a subcritical reactor assembly and calculate physical parameters of different fuel configurations. Based on extensive neutron transport simulations an SRA configuration is proposed, comprising 44 TRIGA fuel rods arranged in a 7 × 7 square lattice. This configuration is found to have a maximum keff value of 0.95001±0.00009 at 4 cm pitch. The SRA is characterized by calculating the 3-dimensional neutron flux distribution and neutron spectrum. The effective delayed neutron fraction and mean neutron generation time of the system are calculated to be 748pcm±7pcm and 41μs, respectively. Results obtained from this work will be the basis of the core design for the subcritical reactor facility that will be established in the Philippines. Keywords: Nuclear reactor, Subcritical assembly, TRIGA fuel, MCN

Predicting Ex-core Detector Response in a PWR with Monte Carlo Neutron Transport Methods

An approach for calculating ex-core detector response using Monte Carlo code MCNP was developed. As a first step towards ex-core detector response prediction a detailed MCNP model of the reactor core was made. A script called McCord was developed as a link between deterministic program package CORD-2 and Monte Carlo code MCNP. It automatically generates an MCNP input from the CORD-2 data. A detailed MCNP core model was used to calculate 3D power distributions inside the core. Calculated power distributions were verified by comparison to the CORD-2 calculations, which is currently used for core design calculation verification of the Krško nuclea power plant. For the hot zero power configuration, the deviations are within 3 % for majority of fuel assemblies and slightly higher for fuel assemblies located at the core periphery. The computational model was further verified by comparing the calculated control rod worth to the CORD-2 results. The deviations were within 50 pcm and considered acceptable. The research will in future be supplemented with the in-core and ex-core detector signal calculations and neutron transport outside the reactor core

Predicting Ex-core Detector Response in a PWR with Monte Carlo Neutron Transport Methods

An approach for calculating ex-core detector response using Monte Carlo code MCNP was developed. As a first step towards ex-core detector response prediction a detailed MCNP model of the reactor core was made. A script called McCord was developed as a link between deterministic program package CORD-2 and Monte Carlo code MCNP. It automatically generates an MCNP input from the CORD-2 data. A detailed MCNP core model was used to calculate 3D power distributions inside the core. Calculated power distributions were verified by comparison to the CORD-2 calculations, which is currently used for core design calculation verification of the Krško nuclea power plant. For the hot zero power configuration, the deviations are within 3 % for majority of fuel assemblies and slightly higher for fuel assemblies located at the core periphery. The computational model was further verified by comparing the calculated control rod worth to the CORD-2 results. The deviations were within 50 pcm and considered acceptable. The research will in future be supplemented with the in-core and ex-core detector signal calculations and neutron transport outside the reactor core

VALIDATION OF SERPENT FOR FUSION NEUTRONICS ANALYSIS AT JET

Fusion neutronics analysis before and after experiments at JET is traditionally performed using Monte Carlo particle transport code Monte Carlo N-Particle. For redundancy and diversity reasons there is a need of an additional Monte Carlo code, such as Serpent 2, capable of fusion neutronics analysis. In order to validate the Serpent code for fusion applications a detailed model of JET was used. Neutron fluxes and reaction rates were calculated and compared for positions outside the tokamak vacuum vessel, in the vacuum vessel above the plasma and next to a limiter inside the vacuum vessel. For all detector positions with DD and DT neutron sources the difference between neutron fluxes calculated with both Monte Carlo codes were within 2σ statistical uncertainty and for most positions (more than 90 % of all studied positions) even within 1σ uncertainty. Fusion neutronics analysis in the JET tokamak with Serpent took on average 10 % longer but this can be improved by changing the threshold value for determination of the transport method used. With the work presented in this paper the Serpent Monte Carlo code was validated to be a viable alternative to MCNP for fusion neutronics analysis for the JET tokamak

Representation and modeling of charged particle distributions in tokamaks

Experimental diagnostics, analysis tools and simulations represent particle distributions in various forms and coordinates. Algorithms to manage these data are needed on platforms like the ITER Integrated Modeling & Analysis Suite (IMAS), performing tasks such as archiving, modeling, conversion and visualization. A method that accomplishes some of the required tasks for distributions of charged particles with arbitrarily large magnetic drifts in axisymmetric tokamak geometry is described here. Given a magnetic configuration, we first construct a database of guiding center orbits, which serves as a basis for representing particle distributions. The orbit database contains the geometric information needed to perform conversions between arbitrary coordinates, modeling tasks, and resonance analyses. Using that database, an imported or newly modeled distribution is mapped to an exact equilibrium, where the dimensionality is reduced to three constants of motion (CoM). The orbit weight is uniquely given when the input is a true distribution: one that measures the number of physical particles per unit of phase space volume. Less ideal inputs, such as distributions estimated without drifts, or models of particle sources, can also be processed. As an application example, we reconstruct the drift-induced features of a distribution of fusion-born alpha particles in a large tokamak, given only a birth profile, which is not a function of the alpha’s CoM. Repeated back-and-forth transformations between CoM space and energy-pitch-cylinder coordinates are performed for verification and as a proof of principle for IMAS

VALIDATION OF SERPENT FOR FUSION NEUTRONICS ANALYSIS AT JET

Fusion neutronics analysis before and after experiments at JET is traditionally performed using Monte Carlo particle transport code Monte Carlo N-Particle. For redundancy and diversity reasons there is a need of an additional Monte Carlo code, such as Serpent 2, capable of fusion neutronics analysis. In order to validate the Serpent code for fusion applications a detailed model of JET was used. Neutron fluxes and reaction rates were calculated and compared for positions outside the tokamak vacuum vessel, in the vacuum vessel above the plasma and next to a limiter inside the vacuum vessel. For all detector positions with DD and DT neutron sources the difference between neutron fluxes calculated with both Monte Carlo codes were within 2σ statistical uncertainty and for most positions (more than 90 % of all studied positions) even within 1σ uncertainty. Fusion neutronics analysis in the JET tokamak with Serpent took on average 10 % longer but this can be improved by changing the threshold value for determination of the transport method used. With the work presented in this paper the Serpent Monte Carlo code was validated to be a viable alternative to MCNP for fusion neutronics analysis for the JET tokamak

Reaction Rate Benchmark Experiments with Miniature Fission Chambers at the Slovenian TRIGA Mark II Reactor

A series of fission rate profile measurements with miniature fission chambers, developed by the Commisariat á l’énergie atomique et auxénergies alternatives, were performed at the Jožef Stefan Institute’s TRIGA research reactor. Two types of fission chambers with different fissionable coating (235U and 238U) were used to perform axial fission rate profile measurements at various radial positions and several control rod configurations. The experimental campaign was supported by an extensive set of computations, based on a validated Monte Carlo computational model of the TRIGA reactor. The computing effort included neutron transport calculations to support the planning and design of the experiments as well as calculations to aid the evaluation of experimental and computational uncertainties and major biases. The evaluation of uncertainties was performed by employing various types of sensitivity analyses such as experimental parameter perturbation and core reaction rate gradient calculations. It has been found that the experimental uncertainty of the measurements is sufficiently low, i.e. the total relative fission rate uncertainty being approximately 5 %, in order for the experiments to serve as benchmark experiments for validation of fission rate profiles. The effect of the neutron flux redistribution due to the control rod movement was studied by performing measurements and calculations of fission rates and fission chamber responses in different axial and radial positions at different control rod configurations. It was confirmed that the control rod movement affects the position of the maximum in the axial fission rate distribution, as well as the height of the local maxima. The optimal detector position, in which the redistributions would have minimum effect on its signal, was determined