In this study, analysis of the complete operational history of the “Joˇzef Stefan” Institute (JSI) TRIGA reactor was performed. Reactor power changes, core configurations and weekly excess reactivity measurements were analysed to obtain the needed data for fuel burnup calculations.
Progress in Nuclear Energy 130 (2020) 103536 Contents lists available at ScienceDirect Progress in Nuclear Energy journal homepage: http://www.elsevier.com/locate/pnucene Computational burnup analysis of the TRIGA Mark II research reactor fuel ˇ ˇc a, Luka Snoj a, b, * Anˇze Pungerˇciˇc a, b, Duˇsan Cali a b Reactor Physics Department, Joˇzef Stefan Institute, Jamova cesta 39, 1000, Ljubljana, Slovenia Faculty of Mathematics and Physics, University of Ljubljana, Jadranska ulica 19, 1000, Ljubljana, Slovenia A R T I C L E I N F O A B S T R A C T Keywords: TRIGA research reactor and Fuel Burnup Operational history analysis Serpent-2 TRIGLAV Excess reactivity measurements In this study, analysis of the complete operational history of the “Joˇzef Stefan” Institute (JSI) TRIGA reactor was performed Reactor power changes, core configurations and weekly excess reactivity measurements were ana lysed to obtain the needed data for fuel burnup calculations More than 50 years of reactor operation was simulated using deterministic code TRIGLAV and stochastic code Serpent-2 The calculated core reactivities are in good agreement compared with the excess reactivity measurements Code-to-code comparison is presented Clear agreement is observed when comparing changes in core excess reactivity, and discrepancies are observed in the comparison of individual fuel element burnup and its isotopic composition The Serpent-2 results are in better agreement with the measurements compared to the TRIGLAV code; nevertheless, a conclusion can be made that the TRIGLAV code is viable for TRIGA fuel management and burnup analysis A three-dimensional (3D) burnup study was conducted, where individual fuel elements were further divided into multiple angular and axial depletion zones Notable burnup effects were observed, and an explanation using surrounding water distance is presented Introduction Determination of fuel element burnup in research reactors is an important activity, as it is related to fuel utilisation and management, characterisation of radiation fields in the reactor, reactor safety pa rameters and safety analyses, as well as spent fuel management The “Joˇzef Stefan” Institute (JSI) has been operating a TRIGA Mark II reactor (Douglas et al., 2003) since 1966 Over this time, multiple burnup cal culations and measurements have been performed (Ravnik et al., 1999; ˇ Zagar and Ravnik, 2000; Perˇsiˇc et al., 2000; Jeraj et al., 2002) Mea surements were performed using the fuel element reactivity worth method (Ravnik et al., 1987) Burnup calculations for the JSI TRIGA reactor were performed using only part of the operational history with simplified operational data; this was done because the operational his tory analysis was not available Burnup was calculated with determin istic codes, such as the in-house developed TRIGLAV code (Jeraj et al., 2002; Perˇsiˇc et al., 2017) TRIGLAV is a two-dimensional (2D) diffusion code used for calculation of fuel element burnup, excess reactivity and power peaking factors (Snoj and Ravnik, 2008) As the TRIGA core is highly heterogeneous and asymmetric, three-dimensional (3D) Monte Carlo codes are superior to diffusion codes for burnup calculations Hence, we decided to repeat the analysis (reactivity worth of important isotopes) using modern tools, which enable 3D Monte Carlo burnup calculations using detailed operational data Our goal was to improve the TRIGA burnup analysis using modern burnup codes and validate the results with multiple excess reactivity measurements The JSI TRIGA reactor had well-recorded individual reactor operation from 1966 to the present We decided to obtain the operational data and simulate the complete history using reactor physics and burn-up code Serpent-2 (Leppă anen et al., 2015; Maria, 2016) Only a few TRIGA reactors have this much information available regarding its operation (Chiesa et al., 2016; Idris Lyric et al., 2013; Khan et al., 2010; El Bakkari et al., 2013) In addition, two experiments (i.e criticality and fission rate profile) performed at the JSI TRIGA reactor have been published in the ICSBEP and IRPhEP handbooks (Jeraj and Ravnik, 2010) Several other exper iments that can be used for validation of computer codes and models have been performed at the JSI TRIGA reactor (Raduloviˇc et al., 2014; ˇ Goriˇcanec et al., 2015; Stancar et al., 2018; Ambrozic et al., 2020) Our primary goal was to record and digitalise the complete opera tional history, different core configurations and individual reactor op erations for the purpose of experimental validation of computer codes and models As for the need for burnup experiments, we also decided to analyse weekly excess reactivity measurements The analysed opera tional history is presented in the first part of the paper * Corresponding author Reactor Physics Department, Joˇzef Stefan Institute, Jamova cesta 39, 1000, Ljubljana, Slovenia E-mail addresses: anze.pungercic@ijs.si (A Pungerˇciˇc), luka.snoj@ijs.si (L Snoj) https://doi.org/10.1016/j.pnucene.2020.103536 Received 22 June 2020; Received in revised form 29 September 2020; Accepted October 2020 Available online 23 October 2020 0149-1970/© 2020 The Authors Published by Elsevier Ltd This is an open (http://creativecommons.org/licenses/by-nc-nd/4.0/) access article under the CC BY-NC-ND license A Pungerˇciˇc et al Progress in Nuclear Energy 130 (2020) 103536 Due to the significant amount of fuel shuffling throughout the history and diverse reactor operation, we developed an automated script called STRIGA (Pungerˇciˇc et al., 2019) This script creates a Serpent input of the TRIGA research reactor (Calic et al., 2016) based on the loading scheme, fuel element composition and burnup parameters, and it pre pares necessary inputs for Monte Carlo burnup calculations This methodology, together with that employed in the TRIGLAV code, is presented in the second part of this paper The last part focuses on the burnup calculations, comparison between both codes, and most impor tantly, the validation performed using the excess reactivity measurements Fig Schematic view of the “Joˇzef Stefan” Institute (JSI) TRIGA fuel elements with its dimensions Two different types of fuel elements based on cladding are presented: Stainless steel SS-304 (top) and aluminium AL (bottom) samarium disk above and below In addition, the fuel elements are divided into three groups The first two groups have 8.5 wt % or 12 wt % of 19.9 % low enriched uranium (LEU) in the U–Zr–H mixture, while the third one is 8.5 wt % of 70 % high enriched uranium (HEU) in the U–Zr–H–Er The latter is a so-called FLIP fuel element, and it contains erbium as a burnable absorber In this study the fresh fuel element composition was determined by considering recorded masses of 235 U and 238 U However, in 1999, all fuel elements except SS 12 wt % were shipped back to the United States, meaning that burnup calculation is the only way to determine their burnup and isotopic composition During the reconstruction of the JSI TRIGA reactor in 1991, the control rods were replaced, and also another different one (transient) was introduced Since then four control rods have been in use: P = transient, S = safety, R = regulating and C = shim The latter three feature a fueled follower SS 12% type fuel, which is thinner and has rfollower = 1.66687 cm instead of rSS12 = 1.82245 cm, while for the transient control rod, only air is left in its place when extracted The older control rods did not feature a fueled follower and were used in different core position Aluminium tube was left in their place when extracted In addition, the absorber for regulating control rod was thinner, where rreg = 1.1 cm instead of rshim,safe = 1.6 cm A schematic view of the control rods before and after the reconstruction in 1991 is presented in Fig 2 JSI TRIGA operational history analysis The first part of our detailed burnup analysis is the JSI TRIGA Mark II operational history analysis, which consists of several important parts that are presented in detail in this section These are as follows: • Reactor description: description of the JSI TRIGA Mark II reactor • Reactor power changes: analysis of individual reactor operations to accurately calculate released energy • Fuel shuffling: analysis of all of 240 core configurations used in the complete operational history • Measurements of excess reactivity: analysis of weekly measure ments used for validating burnup calculations • Measurements uncertainty: evaluation of uncertainties in the reactor operation parameters 2.1 Reactor description Analysis of reactor operation for the purpose of detailed burnup determination was performed on a 250 kW TRIGA Mark II research reactor at the “Joˇzef Stefan” Institute Only a brief description of the reactor is given here, comprising information that is important to un derstand the presented work; a more detailed description can be found in descriptions of JSI TRIGA benchmark experiments (Raduloviˇc et al., ˇ 2014; Stancar et al., 2018; Mele et al., 1994; Jeraj et al., 1997; Jeraj and Ravnik, 2010; Ravnik and Jeraj, 2003) The JSI TRIGA reached first criticality on 31st May 1966 Since then, 300 different fuel elements have been in use Information regarding all different types of fuel element is presented in Table In general, two types of fuel elements were used as follows: stainless steel (SS) with a zirconium rod in the middle and aluminium (AL) without the middle rod These setups are presented in Fig The SS fuel element features a molybdenum disk below the fuel region, while the AL fuel element has a 2.2 Reactor operation from 1966 to 2019 One of the most important parameters in burnup determination is the energy released during reactor operation Therefore, each operation written in the reactor logbooks was analysed In total, 50 logbooks or approximately 20 000 pages were analysed (depicted in Fig 3) Our goal was to digitalise all the needed parameters for future burnup calcula tions The energy released can be calculated from the reactor power level, date and time of both reactor start-up and shut-down or power change A part of this information in a computer readable format is presented in Table Using reactor operation data, the annual released energy in the reactor core is obtained (depicted in Fig 4) It can be seen that, after 1991, the energy released was reduced since the TRIGA reactor was no longer used for isotope production In recent years, the annual average reactor power decreased, mostly due to the higher number of reactor operations at lower power for detector testing, benchmark experiments and similar activities (Goriˇcanec et al., 2015; Raduloviˇc et al., 2014; ˇ ˇ Henry et al., 2015; Filliatre et al., 2015; Stancar and Snoj, 2017; Stancar Table Material composition of four types of fuel elements that were in use in the JSI ˇ TRIGA Mark II research reactor (Zagar and Ravnik, 2000) For each fuel element type, its years of utilisation are depicted Fuel element name AL 8.5 % SS 8.5 % SS 12 % FLIP U–ZrH SS-304 12 U–ZrH–Er SS-304 8.5 277 19.9 (LEU) 192 70 (HEU) a Fresh fuel element material composition Fuel type U–ZrH U–ZrH Cladding Al SS-304 Uranium content [wt 8.5 8.5 %] Uranium mass [g] 185 190 Uranium enrichment 19.9 (LEU) 19.9 (LEU) [wt %] 235 37 38 U [g] Burnable absorber Absorber content [wt %] Years of usage 55 134 / / / / / / Er 1.5 1966–1983 1970–1996 1991–2018 1973–1991 Fig Schematic representation of control rods used before (bottom) and after (top) the reconstruction of the reactor in 1991 Two different types of old control rods were in use, differing in the absorber radius Thicker (rabs = 1.6 cm) rod types were used for Shim (position C-03) and Safety (C-11) control rods, while a thinner (rabs = 1.1 cm) rod type was used for the Regulating control rod (E− 13) a Composition of individual fresh fuel elements can vary but no more than % from the depicted values A Pungerˇciˇc et al Progress in Nuclear Energy 130 (2020) 103536 Fig Photograph of all logbooks written in the history of the JSI TRIGA Mark II research reactor operation (left) Example of logbook input, where reactor startup on 150 W, reactor power change to 250 kW and measurements of excess reactivity are depicted Table Example of the JSI TRIGA reactor operation logbook digitalisation from first criticality in 1966 until 2019 For each operation core configuration, individual com mentary and reactor power, together with the starting and ending date-time [year-month-day hour:minute] is written The table is regularly updated Operation No [#] Start-up [y-m-d h:m] Preactor [kW] End [y-m-d h:m] Δt [h] Erel [MWh] Etot [MWh] Commentary Core conf No ⋮ 1966-5-31 14:15 1966-6-6 8:40 1966-6-6 13:50 0.005 0.02 14 ⋮ 1966-5-31 14:25 1966-6-6 9:50 1966-6-6 14:05 0.17 1.17 0.25 0 0.0035 0 0.0035 ⋮ First criticality 1 ⋮ 27165 2019-2-5 13:15 0.15 2019-2-5 13:20 0.08 0.00 24007.04 2019-2-5 13:20 250 2019-2-5 15:22 2.03 0.51 24007.55 ρexcess measure 240 27166 240 pulse duration ( ̃ 10 ms (Vavtar et al., 2020)); therefore it was not considered in the burnup calculations 2.3 Core configurations Another important part of burnup calculation is the fuel shuffling history Throughout the history, the fuel element position in the core has been changed several times To acquire the positions, all core configu rations (in total 240) were analysed and digitalised so their loading patterns could be used in the burnup calculations or for other activities An example of two recent core configuration changes is presented in Fig During the analysis of the reactor logbooks and the fuel shuffling, it was found that some fuel elements of the SS 8.5 wt % type were already previously used in another TRIGA reactor in Frankfurt am Main, Ger many The problem was that the burnup of those fuel elements is un known Therefore, additional analysis, as presented in 4.1.1, was performed to understand and evaluate the effect of not knowing the burnup of these fuel elements Another important discovery was that, after the reconstruction in 1991 new fresh fuel elements of the SS 12 wt % type, which are still in use today, were mixed together with the old, Fig Analysis of the JSI TRIGA Mark II reactor operation by each year from its start in 1966 to the end of 2018 Annual released energy and average reactor power are depicted for each year Average reactor power was calculated as ∑ Pi ti Pavg = top , where i presents individual operation and top represents total time reactor was operational in year et al., 2018) In addition the pulse mode operation started after the reconstruction was analysed (Pungerˇciˇc and Snoj, 2018); however, the energy released during the individual pulse is negligible due to the short Fig Examples of three different core configurations of the JSI TRIGA reactor Each fuel element is labelled with a four-digit identification number All three core configurations were established in 2018 A Pungerˇciˇc et al Progress in Nuclear Energy 130 (2020) 103536 Fig Measurements of excess reactivity performed at the JSI TRIGA reactor in the complete operational history (left) and selected period of months (right), where changes due to core shuffling and burnup are visible Four core configuration changes were employed in this period Energy released during operation on core configurations 30 and 32 is depicted The last part of the operational history analysis was to record all the excess reactivity measurements Excess reactivity has been determined regularly every Monday since the start of the operation in 1966 Usually, the JSI TRIGA reactor does not operate during the weekend, which means that measurements are without xenon contribution Changes in excess reactivity are directly connected to fuel burnup or fuel shuffling, as presented in Fig As these changes can be used to validate burnup calculations, we have decided to analyse all 2000 measurements per formed in the complete operational history reduce the error Recently, this uncertainty was reduced to %–5 % ˇ with an improved thermal power calibration method (Stancar and Snoj, 2017) Other negligible uncertainty related to released energy is also in the reactor startup or power change as only time on power is written; therefore, the energy released during the transient is not considered Uncertainties in excess reactivity measurements are more difficult to evaluate since they depend on the changes in the reactor core For the comparison of absolute reactivity measurements, the 1σ uncertainty can be up to 500 pcm due to the control rod-worth measurements (Trkov et al., 1987; Merljak et al., 2018), reactor physical parameters (Filliatre et al., 2015; Snoj et al., 2010; Henry et al., 2015) and power redistri bution due to control rod insertion (Goriˇcanec et al., 2015) In the analysis of relative changes in excess reactivity, the assumed uncertainty is much smaller as the same control rod-worth measurements are being used, and the changes in reactor physical parameters and flux redistri bution are negligible if the core configuration is the same 2.5 Major sources of measurement uncertainty Methods of burnup calculations A major source of uncertainty in the fuel burnup determination re lates to uncertainty in reactor power measurements In small research reactors, the power is normally calibrated with respect to a single neutron detector Its response is proportional to the flux at its position Local flux is proportional to the total flux (power) of the reactor only if its radial and axial distributions not change However, this is not the case in the research reactors where operational reactivity changes (burnup, power defect, xenon effect) are compensated for by moving the control rods Redistribution effects on neutron flux distribution due to control rod insertion/withdrawal detected by a single detector may be in the order of 20 % yielding the same error in reactor power readings ˇ (Goriˇcanec et al., 2015; Zerovnik et al., 2014, 2015) Using two or more detectors strategically located at different locations around the core can Until now, all burnup calculations for the JSI TRIGA Mark II, as presented in (Jeraj et al., 2002), have been performed either by using the TRIGLAV (Perˇsiˇc et al., 2017) fuel management deterministic code or other unit-cell based burnup calculations Usually the isotopic compo sition was obtained with a standalone burnup code (e.g WIMS-D (Kulikowska, 1996)) and then used in Monte Carlo code, MCNP For the experiments with burned core configurations, higher discrepancies between the reactivity measurements and those calculated were ˇ observed (Zagar and Ravnik, 2000) For this purpose, we have decided (in addition of using the TRIGLAV code) to simulate the complete operational history using the Monte Carlo neutron transport and burnup ănen et al., 2015), which is known for its burnup code Serpent-2 (Leppa capability (Maria, 2016) Nuclear data libraries for both codes are based on the ENDF/B-VII.1 (Chadwick et al., 2011) nuclear data Comparison of continuous (Serpent) and 69-group cross-section (WIMS-D) energy dependence for total neutron incident on 235 U is presented in Fig coupling the operational history before and after the reconstruction This makes the burnup analysis more complex, as the complete history from 1966 must be considered All the fuel shufflings are presented in JSI TRIGA fuel shuffling animation (Reference to the online animation) 2.4 Measurements of excess reactivity 3.1 The TRIGLAV code The TRIGLAV fuel management and burnup code was developed at the Reactor Physics Department of the “Jozef Stefan” Institute A general description of the code is given in (Perˇsiˇc et al., 2017), and a detailed description of the burnup calculation is provided in (Ravnik et al., 1999) The code is based on a four-group diffusion equation for r − ϕ geometry, solved by the finite difference method The TRIGLAV pro gram package consists of a four group 2D diffusion module and the WIMSD-5B code (Kulikowska, 1996), which is linked automatically to the diffusion module to calculate unit-cell-averaged effective group constants All 91 positions in the reactor core are treated separately in the unit-cell approximation The TRIGLAV geometry model, presented in Fig 8, represents the full TRIGA cylindrical core and graphite Fig Energy dependence for total 235 U microscopic cross-section Continuous (Serpent-2) and 69-group (WIMS) cross-sections are presented The latter is acquired from the WIMS Library Update Project (Coordinated Research), and it is based on the ENDF/B-VIII.1 nuclear data library (Chadwick et al., 2011) A Pungerˇciˇc et al Progress in Nuclear Energy 130 (2020) 103536 Fig TRIGA Mark II reactor geometry in the TRIGLAV model with homogeneous unit-cells (left) TRIGLAV code methodology schematic flow-chart (right) Fig Schematic representation of the STRIGA methodology in which TRIGA reactor parameters are used to create Serpent-2 input for steady-state or burnup calculations Each of the four important inputs is described in the schematic reflector On the reflector outer boundary, the zero flux boundary con dition is imposed The unit-cell-averaged cross-sections were calculated using 69-group WIMS library based on the ENDF/B-VII.1 nuclear data library Fuel element burnup (BUel ) within the TRIGLAV code is calculated using the WIMSD-5B code from which the fuel isotopic configuration is obtained, as seen in Fig Unit-cell homogenised cross-sections at the defined burnup are extracted and collapsed into neutron flux weighted four-group constants that are used in 2D diffusion approximation With diffusion solution, a core power distribution is obtained and the fuel element burnup increment can be calculated by knowing energy released data for the defined cycle Using the described procedure, complete operational history can be simulated where different core configurations and reactor operations can be considered In the current TRIGLAV methodology, the isotopic composition of individual fuel el ements is not transferred from cycle to cycle as only the burnup incre ment for individual fuel elements on each cycle is calculated Such procedure means that the mentioned BUel carries the information regarding the complete operational history (power and loading pattern changes) Fig 10 Graphical representation of the TRIGA Mark II reactor diverse power operation treatment in the simulation of the complete operational history A Pungerˇciˇc et al Fig 11 Differences in calculated nU− 250 kW was used as a reference Progress in Nuclear Energy 130 (2020) 103536 235 (left) and nPu− 239 as a function of fuel element burnup for different reactor powers defined in the burnup simulation Pmax = 3.2 STRIGA tool between the Serpent-2 and the TRIGLAV code Criticality benchmark core 132 (Jeraj et al., 1997; Mele et al., 1994) was selected • Burnup on fuel unit-cell: Burnup simulation on fuel element unitcell (fuel pin surrounded with water) to study the physics of fuel composition changes through burnup The STRIGA tool is a simple data manipulation script, written in standard FORTRAN-77 programming language that reads the TRIGLAV inputs and creates inputs for 3D Monte Carlo calculations Core configuration inputs in TRIGLAV were already prepared for the com plete operational history Detailed description of the STRIGA tool is ˇ ˇc et al., 2017); only a brief presented in (Pungerˇciˇc et al., 2019; Cali description is presented in this paper The STRIGA tool is based on the validated steady-state Serpent-2 model (Calic et al., 2016), that is based on the criticality benchmark (Mele et al., 1994) and MCNP models used ˇ in (Raduloviˇc et al., 2014; Stancar et al., 2018; Goriˇcanec et al., 2015) Steady-state SERPENT-2 calculations are completely consistent with the MCNP ones, indicating that the geometry and material composition employed in the model are well defined for reactor cores with fresh fuel (Calic et al., 2016) STRIGA requires two kinds of input data The first represents reactor component dimension and its material composition, while the second represents the reactor operation data, as depicted in Fig To simulate the complete history from 1966 to 2019, additional types of fuel ele ments and control rod positions have been added to the existing STRIGA ˇ ˇc et al., 2017) Another advantage of the STRIGA tool is that it script (Cali reads the TRIGLAV core configuration input file, which is created using ˇ the graphical user interface (GUI) Triglav-W (Zagar et al., 2006) Through this interface, one can specify the reactor operation (power and time) and select the core loading patter via an user-friendly fuel shuf fling interface All fuel and non-fuel elements can be moved from one location to another by a simple click and point procedure In addition, cool-down of the reactor core was added into the STRIGA tool, enabling detailed calculations of diverse reactor operations For a given fuel cycle calculations using burned fuel the most important information is the fuel isotopic composition that is taken from the previous cycle Within the STRIGA tool, the user can select which isotopes are tracked within the burnup calculation In the analysis, 269 isotopes were tracked After each burnup calculation the fuel isotopic library is created or updated Such principle enables the extraction of individual fuel element isotopic composition at different times in the reactor operational history 4.1 Simulation of complete operational history The complete operational history consists of more than 25 000 reactor power changes and more than 240 core configurations To simulate each individual operation, high computer power and memory would be needed, especially for calculations with Monte Carlo code Serpent-2 Therefore, additional simplification was used to divide the complete operational history into individual long operations on same loading pattern or core configuration, as presented in Fig 10 It has been approximated that the reactor operated on maximum power Pmax = ∫t 250 kW for a period in which the energy released Preactor (t)dt was the same At the end of the operation, fuel cooldown was approximated as the sum of the total time the reactor was not operational This principle was used to create 240 different inputs for both Serpent-2 and the TRIGLAV code In this way, individual fuel element burnup was tracked from its first insertion in the reactor core until today The approximation that the reactor operated on maximum power Pmax = 250 kW was tested by repeating burnup simulation on a hypo thetical core configuration using the Serpent-2 code Calculations were repeated for 100 kW and kW reactor powers To keep fuel burnup the same at different reactor powers, the irradiation time was increased accordingly The differences at the point of maximum TRIGA fuel burnup were less than 0.1 % and % for differences in the calculated nU− 235 and nPu− 239 , respectively We also repeated the simulation for Preactor = 100 W and found similar discrepancy for nU− 235 and higher 10 % discrepancy for nPu− 239 Reason behind higher relative differences in nPu− 239 is that the amount of 239 P in TRIGA fuel elements is low, due to the smaller content of 238 U compared with traditional light-water re actors Relative differences in nU− 235 and nPu− 239 are presented in Fig 11 Despite higher relative differences observed, this effect is negligible when comparing final absolute values in material composi tion Similar conclusion can be made for fuel and moderator tempera ture We have investigated the effects of Tfuel and Twater by repeating the calculations with Tfuel = 600 K and Twater = 350 K It has to be noted that Twater represents the temperature of the water surrounding the fuel and it is only one of the contributions to the neutron moderation in a TRIGA reactor The density of the water surrounding the fuel was changed from 0.9975 cmg to 0.9737 cmg Information for the analysed cases are presented in Table The other contribution is in the fuel itself on hydrogen in the U–Zr–H fuel mixture Regarding the effects of fuel temperature, the differences at the point of maximum TRIGA fuel burnup were less than 0.05 % and 10% for differences in the calculated Burned fuel material composition calculations Fuel element burnup analysis for the TRIGA Mark II reactor is per formed on three different cases, which are as follows • Complete operational history: Analysed data regarding the energy released and fuel shuffling was used to create a burnup simulation of the actual reactor operation performed from 1966 to 2019 • Burnup on benchmark core No 132: Burnup simulation on selected core configurations was performed to study the differences A Pungerˇciˇc et al Progress in Nuclear Energy 130 (2020) 103536 Fig 12 Differences in calculated nU− 235 (left) and nPu− 239 as a function of fuel element burnup for higher fuel temperature Tfuel = 600 K and core water temperature Twater = 350 K Tfuel = Twater = 300 K was used as a reference and for all burnup calculations presented in this paper Both calculations were performed on maximum steady state power Pmax = 250 kW nU− 235 and nPu− 239 , respectively Resuls are presented in Fig 12 The noticeable effect on plutonium production is constant at 10 % This is due to the Doppler broadening of (n, γ) reaction resonances on 238 U There were no noticeable effects of surrounding water temperature on nPu− 239 and for nPu− 239 the differences were below 0.1 % A conclusion can be made that for absolute determination of 239 Pu, fuel temperature has to be taken into account, however for our analysis, mainly consisting of comparing relative change due to burnup, negligible discrepancy is introduced However further more detailed analysis with detailed thermodynamical model should be performed to fully analyse the tem perature effect on burnup Analysing the effect of different reactor powers and fuel tempera tures on depletion of 235 U and production of 239 P showed that there is negligible effect on core excess reactivity due to long-lived isotopes if approximation presented in Fig 10 is taken into account In addition we observed that 239 P production depends slightly more on different reactor power than fuel temperature, especially in smaller burnup below MWd kg(HM) However the amount of 239 P produced in TRIGA fuel is low and Fig 14 Relative difference in calculated fuel burnup between Serpent and TRIGLAV as a function of burnup for all 300 fuel elements used in JSI TRIGA such effects could be neglected in some cases However effect on core excess reactivity is present due to short-lived isotopes such as xenon and samarium Effect of such isotopes is analysed in Sec 6.1 From this a conclusion can be made that if xenon and samarium are of interest, operation Relative difference is defined as detailed operational history for the last couple of days or weeks should be taken into account Burnup of all 300 fuel elements accumulated in different periods of the reactor operation history from 1966 to 2019 was calculated with both the TRIGLAV and the Serpent-2 code Complete operational history simulation with TRIGLAV takes approximately h on one standard PC core, while Serpent simulation takes approximately three weeks on 56 cores (Intel Xeon Processor 25M Cache, 2.80 GHz) Accuracy of the calculated burnup depends mainly on the experimental power calibra ˇ tion accuracy (Stancar and Snoj, 2017) and the precision of the opera tional records As discussed in section 2.5, the 1σ uncertainty in reactor power level is 10 %–15 % However, we assumed only % uncertainty in final burnup as the uncertainty for longer operations averages out The final fuel burnup for randomly selected 16 fuel elements, of each type, (each fuel element is represented with a four-digit fuel identifi cation number) is presented in Fig 13 It should be noted that these fuel elements were not part of the same core configuration, but they were used in different parts of the operational history First, we can observe that FLIP fuel elements have higher burnup (up MWd to 70 kg(HM) ), while the burnup for LEU fuel elements is around Fig 13 Final fuel burnup for 16 out of 300 fuel elements used in the history of the JSI TRIGA reactor operation and calculated by simulating the complete operational history For each fuel type, elements were chosen, which are represented with a four-digit identification number, with exception of fuel followers for regulation (REGU) and safety (SAFE) control rods Table Material information defined in the Serpent-2 model of the JSI TRIGA reactor for the analysis of fuel and water temperature effects on the fuel element burnup Analysed cases Tfuel [K] Reference calculation Fuel temperature test Water temperature test 300 600 300 ρfuel [ g ] cm3 6.04498E+00 6.04498E+00 6.04498E+00 Twater [K] 300 300 350 ρwater (BUSerpent − BUTRIGLAV ) BUSerpent [ g ] cm3 9.97245E-01 9.97245E-01 9.73742E-01 Standard unit for burnup is energy released in MWd per mass of heavy materials (HM) in kg Heavy materials are nuclei with more than 92 protons A Pungerˇciˇc et al Progress in Nuclear Energy 130 (2020) 103536 Fig 15 Individual fuel element burnup calculated by considering the complete operational history Out of 240 core configurations, three were chosen (from left to right: 80, 130, 240), schematically presented at the top For each configuration fuel burnup calculated with Serpent at the end of the cycle is presented together with relative differences between both codes used, defined as 30 MWd kg(HM), (BUSerpent − BUTRIGLAV ) BUSerpent most important from the standpoint of the current reactor operation, is within the ±7.5 % range We can conclude that the difference between both codes for most of the fuel elements is satisfactory, within %, accounting for all the differences between the two codes Code-to-code comparison for all 300 fuel elements is presented in Fig 14 To compare the differences between fuel elements of the same type, core configurations and their position in the reactor core becomes important For this purpose, the following three core configurations were chosen to further analyse the calculated burnup and the differences between TRIGLAV and Serpent-2: due to the higher local power of the FLIP fuel The difference in calc burnup between TRIGLAV and Serpent is due to the higher thermal neutron flux (Snoj and Ravnik, 2008) in the FLIP fuel elements Furthermore, the discrepancies between the codes are analysed Comparing the absolute differences, we can observe higher discrep ancies for FLIP-type fuel elements and the control rod fueled followers, depicted as REGU for regulating and SAFE for safety control rod For FLIP-type fuel elements, the discrepancies can be explained similarly to the difference in calculated burnup Higher localised neutron flux results in higher neutron flux gradients, which are not handled well by the 2D diffusion approximation employed in the TRIGLAV code Nevertheless, the differences in fuel burnup are within 10 % For fueled followers, the relative difference is up to 15 % and this is because in the TRIGLAV code we assume that the fueled follower is the same as a normal SS 12 % fuel element, while in reality, it is smaller; this is considered in the Serpent-2 model, as previously presented in Fig For all 300 fuel elements, relative difference in calculated final burnup with both codes was evaluated The highest discrepancies were observed for control rod fuel followers and for the fuel element type SS 8.5 % with unknown initial burnup (obtained from TRIGA in Frankfurt am Main, Germany) The difference for the SS 12 % type of fuel, which is • Core No 80: due to mixture of three different types of fuel elements (SS, AL 8.5% and FLIP) • Core No 130: as it is the last core that was in operation before the reconstruction in 1991 (maximum burnup in some fuel elements) • Core No 240: as being in operation when the study was performed (February 2019) Fuel element burnup calculated with Serpent-2 and the relative dif ference in comparison with the TRIGLAV code was analysed For core MWd MWd no 80, burnup was between 10 kg(HM) and 30 kg(HM) for all types of fuel A Pungerˇciˇc et al Progress in Nuclear Energy 130 (2020) 103536 Fig 16 Final 239 Pu (top) and 137 Cs (bottom) number densities for 16 fuel elements calculated by simulating complete operational history It should be noted that aluminium fuel element No 288 has only a quarter of fuel material compared with other AL 8.5 % elements elements For core no 130, three clear sets of fuel elements are visible: MWd The first is FLIP, with the highest burnup up to 70 kg(HM) ; the second is SS (Jeraj et al., 2002), only selected isotopes were chosen In Fig 16, the calculated number densities for two important isotopes in burnup analysis 239 P and 137 Cs, are presented The former is the product of neutron absorption in 238 U, and as expected, the calculated amount is lowest in the HEU FLIP fuel elements The latter is the product of nuclear fission; therefore, similar behaviour compared with fuel burnup can be observed For fuel element with ID no 288, the lowest number densities for both isotopes can be observed The reason is that, in this experiment, this fuel element contained only a quarter of the fuel compared with the others Higher discrepancies are observed between the two codes with increased burnup For 235 U, 137 Cs and 239 Pu analysis of the calculated number density differences between both codes was performed for core No 240 The results are presented in Fig 17 For 235 U and 137 Cs, similar behaviour to fuel burnup can be observed, with the highest discrep ancies observed for control rod fuel followers The radial neutron flux distribution, calculated with both codes is highly visible when analysing 239 Pu as for the inner rings (B, C and D) higher number density up to 10 % is calculated with Serpent-2 and lower up to 15 % for outer rings (E and F) compared with the TRIGLAV code The calculated differences for most fuel elements were within %, MWd MWd 8.5 % with burnups between 10 kg(HM) and 20 kg(HM) ; and the last is SS 8.5 %, which were brought from the Frankfurt TRIGA reactor and had MWd recorded burnups below 10 kg(HM) The effect of burnup is clearly visible for core no 240, where burnups in the middle part of the core are higher, MWd MWd , compared with those on the periphery (up to 15 kg(HM) ) at up to 20 kg(HM) Fuel element burnup results are presented in top part of Fig 15, while in the bottom part, relative differences between Serpent-2 and TRIGLAV for the same core configurations are presented A slight shift of calcu lated burnup can be observed when analysing differences between both codes for the complete core, as the burnup calculated with TRIGLAV is higher compared with Serpent-2 on the right side of the core, which can be seen in Fig 15 This difference may occur due to the control rodinduced neutron flux redistribution (Goriˇcanec et al., 2015) or the treatment of burnup in TRIGLAV To understand the mentioned shift, further analysis on just one core configuration was performed; this is presented in section 4.2 In addition to burnup, isotopic composition of individual fuel ele ments was analysed Based on an analysis of isotopic effect on reactivity Fig 17 Comparison of calculated isotopic composition between Serpent and TRIGLAV for last core configuration established in 2019 Relative difference is defined as NSerpent − NTRIGLAV NSerpent Three isotopes are presented; from left to right, these are: 235 U, 239 Pu and 137 Cs A Pungerˇciˇc et al Progress in Nuclear Energy 130 (2020) 103536 while higher discrepancies were due to difference in control rod treat ment in both codes The difference in methodology for calculating neutron flux distribution for both codes comes into effect when comparing distribution of different isotopes, especially 239 Pu These differences are investigated further in the next chapters We can conclude that the difference in the calculated burnup between both codes is relatively low Hence, both methodologies could be used for TRIGA fuel elements’ burnup calculations However, the user should be aware that these differences increase for individual isotope density comparison Fig 18 Relative change in burnup (y-axis) of initially fresh fuel elements (blue dots) due to uncertainty in initial burnup of already irradiated fuel elements, expressed as a relative change from the reference value (x-axis) (For inter pretation of the references to colour in this figure legend, the reader is referred to the Web version of this article.) 4.1.1 Uncertainty propagation of initial fuel element burnup In 1980, already irradiated fuel elements were obtained from the Frankfurt TRIGA reactor As the initial burnup of these fuel elements was not accurately known, additional uncertainty was introduced in the final burnup calculations The effect of this uncertainty was studied, and the results are presented in this section The STRIGA tool was used to study the propagation throughout different core configurations In 1991, soon after the reconstruction, these irradiated fuel elements of type SS 8.5 % were used in mixed-core configurations from core No 138 to core No 146 (reference to the online animation) with the fresh SS 12% Their effect is studied by analysing the burnup of the fresh fuel that is present in the mixed cores In total, fuel was mixed in multiple MWd different cores, which were studied in our case We have chosen 20 kg(HM) as a reference value of the unknown fuel burnup This burnup was later changed from − 100 % to +30 % and effects on accumulated burnup on the fresh fuel elements were analysed The effect on final burnup after core No 146 due to the mentioned burnup changes in a mixed core configuration with fresh fuel is less than % This effect is reduced for further operation and it is practically negligible for fuel burnups at core No 240 The effect on core No 146 is presented in Fig 18 4.2 Burnup on benchmark core no 132 The complete operational history includes a large number of different parameters that could impact the calculated burnup, such as fuel shuffling or mixing of different types of fuel elements To analyse the burnup in higher detail and compare the two methodologies used we decided to perform burnup calculations on the benchmark core config uration with fresh fuel no 132 (Jeraj et al., 1997; Mele et al., 1994) The core schematic is presented in Fig 19 An average core burnup of 50 MWd kg(HM) was assumed and simulated by both Serpent-2 and TRIGLAV This Fig 19 Schematic representation of core configuration No 132, used as a fresh fuel criticality benchmark and available in the ICSBEP handbook (Jeraj et al., 1997; Mele et al., 1994) outermost ring This difference is almost negligible at lower burnups of MWd and becomes evident at larger burnups where the difference 10 kg(HM) between fuel element in ring A and E is 20 approximately represents the energy released in the complete opera tional history, which is an overestimation for only one type of fuel MWd element, while the core burnup of 20 kg(HM) represents the energy MWd kg(HM) Similar behaviour can be observed when analysing the production of isotope 239 Pu Differences in fuel burnup and consequently individual isotopes between both codes were analysed Serpent results were taken as a reference value Highest discrepancies, similar to operational history analysis, are observed for control rod fueled followers and the central released after the reconstruction in 1991 and can be used as a realistic example The validation of the computational model with fresh fuel was already performed (Calic et al., 2016) and taken as an initial condition in this analysis Excess reactivity as a function of core burnup was studied MWd Both codes predict a similar change of ≈ − 6000 pcm at 20 kg(HM) and ≈ − 13000 pcm at 50 MWd kg(HM) Both codes are in good agreement as the discrepancy gradually increases to a difference of only 784 ± 20 pcm at MWd 50 kg(HM) Higher discrepancies are observed for the first couple of steps because of the xenon equilibrium treatment employed in the TRIGLAV code Comparison of reactivity changes with both codes is presented in Fig 20 The individual fuel element burnup and its isotopic composition was studied The results are presented for three different core burnups, MWd MWd MWd which are as follows: 10 kg(HM) , 20 kg(HM) and 50 kg(HM) Serpentcalculated burnup parameters and discrepancies in comparison with the TRIGLAV code are presented in Figs 21 and 22 As expected, the fuel burnup is higher in the inner rings and becomes lower towards the Fig 20 Excess reactivity (top) and difference between Serpent and TRIGLAV criticality calculation (bottom) as a function of TRIGA core No 132 burnup 10 A Pungerˇciˇc et al Progress in Nuclear Energy 130 (2020) 103536 Fig 21 Distribution of calculated fuel burnup and isotopic composition at three different average core burnups (from left to right) From top to bottom, three parameters are presented: individual fuel burnup calculated by Serpent-2, the absolute difference compared with the TRIGLAV code and the number density of 239 Pu calculated by Serpent Control rods are denoted by letters: P = transient, S = safety, C = shim, R = regulating fuel element In addition to fuel burnup, differences in the calculated MWd number densities were analysed at an average core burnup of 20 kg(HM) The results are presented in Fig 22 For 235 discrepancies is comparable to burnup calculations, at between − %and % Similar distribution but higher discrepancies can be observed for 137 Cs, while the distribution for 239 Pu is more uniform, as the TRIGLAV code calculates higher number density (up to 20 %) in U, the distribution of Fig 22 Distribution of calculated differences in number density between Serpent and TRIGLAV for three different isotopes (left to right): 235 U, 239 Pu, 137 Cs Relative difference is defined as NSerpent − NTRIGLAV NSerpent 11 A Pungerˇciˇc et al Progress in Nuclear Energy 130 (2020) 103536 Fig 23 Changes of excess reactivity due to unit-cell burnup for four different types of TRIGA fuel Serpent-2 results are compared with WIMSD-5B inner rings and lower (up to 30 %) in the outermost ring As the pro duction of 239 Pu is a secondary order process the treatment of water surrounding the fuel elements becomes important As can be observed, the TRIGLAV code underestimates number density in the outermost ring surrounded by water, especially for the two elements on the border of E ring Another observation can be made concerning spatial distribution Core configuration no 132 is not symmetrical and the centre is shifted to the left as presented in the core schematic in Fig 19 Analysing the differences between the two codes, we can observe a similar shift This is due to the assumption performed in the first step of the TRIGLAV burnup calculations, where neutron flux or power distribution is assumed The assumption is that the distribution is symetrical and centred on the middle of the core (A1 position) Concerning changes in core reactivity, it can be concluded that the differences between both codes are negli gible The differences in methodology become evident when comparing the burnup of individual fuel element: For the same average burnup, the comparison showed differences in individual rings of the reactor core, which were most visible in the middle three rings Such differences were magnified when analysing isotopic composition We observed that the fuel element surrounding becomes important in the analysis of 239 Pu content It can be concluded that the TRIGLAV code is satisfactory for core excess reactivity analysis; in contrast, isotope composition, higher discrepancies were observed 4.3 Burnup on fuel unit-cell For each type of fuel element, the unit-cell was assumed as a fuel pin surrounded with water and with reflective boundary conditions The same unit cell geometry is used in the TRIGLAV code, where homoge nised cross sections are generated at the unit-cell level with WIMSD-5B (Kulikowska, 1996), where annular geometry with surrounding water radius rwater = 2.31317 cm is defined This is an average water amount in the JSI TRIGA reactor defined in the TRIGLAV code (Persic et al., 2017) ănen et al., 2015) code square geometry has to be In the Serpent-2 (Leppa defined if reflective boundary conditions are to be used For this purpose we defined a square geometry with half-width of 2.05 cm so that water area of the Serpent-2 unit-cell matches the one in the WIMSD-5B code Serpent-2 and WIMSD-5B were used to calculate changes in unit-cell MWd reactivity and isotopic composition for burnups up to 30 kg(HM) The MW assumed power density for all fuel elements was 30 kg(HM) The results of reactivity changes, together with dimensions of individual fuel pins, are presented in Fig 23 The changes in reactivity at the unit-cell level show similar behav iour compared with full core calculations There is a visible difference between SS 8.5 % and AL 8.5 %, meaning that the inner zirconium pin has a notable effect on burnup changes All three LEU fuel elements show similar behaviour, with a slight difference of faster reactivity decrease for SS 8.5% Noticeable differences can be observed for HEU MWd FLIP fuel with burnable absorbers; after average burnups of kg(HM) , reactivity increases linearly with burnup, signifying positive reactivity feedback on core burnup It should be noted that all the changes pre sented here vary in relation to the specific power of each fuel pin, and that FLIP fuel elements were always used in mixed core configurations, resulting in a negative reactivity feedback on core burnup A comparison in changes of excess reactivity due to burnup between WIMSD-5B and Serpent-2 was also performed, where good agreement (between − 200 and + 200 pcm) for all fuel types was observed In addition, we investigated which isotopes made the highest contribution to the discrepancies We found that the effect was proportional to the influence of a single isotope on reactivity changes, meaning, isotopes 235 U and 239 Pu made the highest contribution to the discrepancy From this, we can conclude that burnup on the unit-cell calculation performed with WIMSD-5B in the TRIGLAV code is not the source of observed discrepancies when comparing full core burnup calculations with both Fig 24 Calculated axial distribution of 239 Pu (top) and 235 U (middle) number MWd densities calculated with Serpent-2 for a burnup of 50 kg(HM) A hypothetical core configuration with divided fuel elements on 100 depletion zones was employed Results for all four different types of fuel elements with its sche matics are presented In addition, thermal neutron flux (bottom) in the fuel and graphite region is depicted 12 A Pungerˇciˇc et al Progress in Nuclear Energy 130 (2020) 103536 Fig 25 Changes of core excess reactivity as a function of TRIGA core burnup calculated with Serpent-2 using one and 15 axial depletion zones (left) In addition, manual division and automatic division using the STRIGA tool is compared The reactivity rate of change with burnup is presented, and linear coefficients are calculated (right) Fig 26 Differences of final fuel element burnup due to C-shim and R-regulating control rod insertion All rods out case is taken as a reference for three cases of insertion; from left to right 25 % in, 50 % in and 75 % in codes was subdivided into sections, while the others were left undivided due to the computational limit Four different cases were studied based on the type (AL 8.5 %, SS 8.5 %, FLIP, SS 12 %) of the divided fuel element Schematic of the core is presented on the right side of Fig 27 It was assumed that transient (P) and safety (S) control rods were completely extracted, while regulation (R) and shim (C) were half-inserted This is important only in cases where multiple axial depletion zones are considered (Goriˇcanec et al., 2015) TRIGA fuel burnup spatial effects All the calculations performed with Serpent were conducted using a single depletion zone for the fuel element, without axial, radial or angular division As the division of fuel elements into multiple depletion zones is computationally intensive, it was decided to perform the anal ysis of burnup spatial effects on a hypothetical core configuration con sisting of fuel elements divided into multiple axial and annular depletion zones The core example had only fuel elements of type SS 12 wt %, which were filled into rings from B to E In each ring, one fuel element n235 (burned) Fig 28 Angular distribution of fuel burnup in units of n235U (fresh) for the fuel U element inserted in the core periphery, as depicted in Fig 27 The colour scale represents the difference in burnup in each depletion zone compared with the average fuel burnup (For interpretation of the references to colour in this figure legend, the reader is referred to the Web version of this article.) Fig 27 Computational model of the hypothetical core configuration used in the Serpent-2 code to study the angular distribution of isotopes in burned TRIGA fuel Four depletion zones where the 640-group neutron spectrum was calculated are depicted 13 A Pungerˇciˇc et al Progress in Nuclear Energy 130 (2020) 103536 Fig 29 Definition of angular distribution for angular burnup comparison with surrounding water thickness presented in 30 Fig 30 Burnup in units of depleted 235 U and amount of water as a function of fuel element angle Calculations for four fuel elements one in each ring (B,C,D,E) are depicted Diffusion constant for water DH2O ≈ cm is depicted Graphite reflector and control rod are depicted with shaded area 5.1 Axial distribution (average core burnup of 50 MWd kg(HM)) shows that, as expected, the burnup of all three fuel elements was highest in the middle of the fuel element, where the thermal neutron flux is higher In addition, an increase of burnup on the edges of the fuel meat is observed due to the increase of thermal neutron flux in the graphite region of the fuel element This is not observed for fuel element type AL 8.5 % which has samarium ab sorbers on both sides that neglect this effect Another observation is that the isotope 239 Pu distribution is inverse compared to 235 U, which was an expected result The highest differences can be observed comparing LEU (AL, SS 8.5 % and SS 12 %) and HEU (FLIP) elements Due to the higher enrichment, For all types of fuel elements, it can be observed (in Fig 24) that the distribution was not symmetrical; rather, it was slightly shifted down due to the partially inserted control rods Comparing different types of fuel elements, higher differences can be observed Axial distribution for fresh fuel of neutron thermal flux and isotopic distribution of 235 U and 239 Pu were analysed, and the results—together with differences in fuel dimensions—are presented in Fig 24 The calculated statistical uncer tainty in thermal neutron flux in each of the 100 axial zones was less than % Analysis of 235 U number density at the maximum burnup 14 A Pungerˇciˇc et al Progress in Nuclear Energy 130 (2020) 103536 the past would have to be analysed and individual simulation cycle further divided 5.3 Angular distribution Angular spatial burnup effects were studied by dividing the fuel element by angle into 20 equal (18 ◦ ) depletion zones to evaluate the effect of fuel surrounding on angular burnup distribution One fuel element in the core periphery (depicted in Fig 27) was chosen for this analysis Angular distribution for fuel type SS 12 % was studied at two different burnups The results are presented in Fig 28 It can be observed that the burnup is higher on the side of the fuel element facing towards empty outermost ring filled with water The differences in highest and lowest burnup zones are almost negligible (up to 3%) for realistic MWd average core burnup of 20 kg(HM) and become evident (up to %) at Fig 31 Neutron lethargy spectrum with 640 energy group structure, defined Φ as Ψg = ( g ), for four depletion zones, depicted in Fig 27, in fresh fuel log Eupper Elower element type SS 12 % a lower percentage of the initial number density of 235 U can be observed for FLIP fuel elements As previously mentioned, the AL and SS 8.5 % fuel elements differ in their cladding material The analysis showed no differences in isotope distributions, and it can be concluded that the type of cladding does not have noticeable effect on the burnup of TRIGA fuel elements The analysis shows that using 15 axial depletion zones is adequate to describe the axial effects We compared the excess reactivity changes with the original (without axial distribution) results The difference for fuel burnup of the JSI TRIGA reactor is 400 pcm, meaning that our original results underestimate the reactivity decay The comparison is presented in Fig 25 We can estimate that the complete simulation time using 15 axial depletion zones would increase to three months while preserving the accuracy of the simulations Adding 20 angular and radial divisions would increase the simulations to several years This evaluation was conducted for node with 56 cores (Intel Xeon Processor 25M Cache, 2.80 GHz) higher burnups of 50 MWd kg(HM) As the TRIGA core has an annular configuration that is not periodic, the amount of water surrounding each fuel element depends on the fuel element location Moreover, the amount of water around fuel element also varies significantly with the angle as depicted in Fig 29 To analyse this effect, the neutron spectrum was studied in four different regions, as presented in Fig 27, where region is the one facing the water region It was observed that the thermal peak in region was higher compared with the other regions which directly resulted in higher burnups The 640-group neutron spectrum results are presented in Fig 31 We used a simplse ray-tracing algorithm, described in appendix A, to determine the amount of water around fuel elements We chose four fuel elements, each representing individual ring of the JSI TRIGA reactor (B (inner), C, D, E (periphery), as presented in Fig 29 The amount of water variation was compared with fuel burnup where clear agreement in the shape of the variation was observed, expect for regions with nearby control rods and graphite reflector In the region of the fuel element facing towards the control rod, lower burnup is observed The results are presented in Fig 30 From this, we can conclude that there is a direct connection between the amount of water and fuel burnup, which can be determined with simple and fast ray-tracing algorithms; the longer burnup simula tions with angular division would not be needed in some cases From the axial and angular distribution analysis, we can conclude that the axial effects are more important in the TRIGA reactors, as the differences between the highest and lowest burnup part of the fuel are MWd more than 25 % for realistic (20 kg(HM) ) and even more than 50 % for 5.2 Effect of control rod insertion on fuel burnup Another assumption in our simulation of complete operational his tory was that control rods were extracted from the core This was done for the purpose of comparing excess reactivity measurements with cal culations and because control rod position changes with reactor opera tion and fuel shuffling When the reactor is in operation, safety (S) and transient (P) are extracted, while regulating (R) and shim (C) control rods are inserted between 25 % and 50 % To test this assumption, we have simulated full JSI TRIGA burnup on the last core in operation today and checked the differences in calculated fuel element burnups Three cases were assumed, where R-regulating and C-shim control rods were inserted at 25 %, 50 % and 75 % The results show (Fig 26) a decrease in burnup on fuel elements surrounding the control rods and increase on the rest This decrease is below % for the first case (25 % inserted) and below 15 % for the second (50 % inserted) The results are presented in Fig 26 From this analysis, we can conclude that, for more detailed burnup calculation, control rod positions should be taken into consid eration However, this is extremely difficult to incorporate into complete operational history simulation, since all of the control rods positions in overestimated (50 MWd kg(HM)) burnups It should be noted that this effect could further increase if control rod position is taken into consideration, especially for fuel elements in their close proximity For future burnup calculations, where control rod position will be considered, axial divi sion of depletion zones is necessary The same cannot be stated for angular effects, as even for the extreme cases with highest burnups of MWd 50 kg(HM) and highly heterogeneous fuel element position, the differ ences between the highest and lowest burnups were less than % The axial effect could be considered by using the recorded control rod po sitions in the detailed model created by the STRIGA tool Considering operational history, axial position of the fuel elements is well defined; however, their angular position and rotations were not recorded, and therefore, they cannot be determined This can be treated as an uncer tainty It can be concluded that, in the future, even their angular position should be reported when fresh fuel elements are inserted Validation of burnup calculations The last part of the burnup analysis presented in this paper consists of validating burnup calculations with parameters obtained from the operational history analysis For this purpose, weekly measurements of excess reactivity were analysed and used for comparison with Serpent-2 and TRIGLAV calculations of the complete operational history Fig 32 Effect of individual isotope on TRIGA reactor core reactivity as a function of its average burnup Operation at the maximum steady state Pmax = 250 kW was simulated using the Serpent Monte Carlo code Statistical uncer tainty of a single calculations is σstat (ΔρExcess ) = 12 pcm 15 A Pungerˇciˇc et al Progress in Nuclear Energy 130 (2020) 103536 Fig 33 Excess reactivity at the beginning of each cycle or core configuration for the JSI TRIGA reactor (top graph) Measurements and results of complete operational history simulation using Serpent and TRIGLAV code are presented Excess reactivity for benchmark core configuration No 132 (not included in the graph) is in within the 1σ agreement with both codes The bottom graph depicts the difference between calculated and measured excess reactivity at each core configuration together with 1σ and 2σ of the measurements Fig 34 TRIGA core excess reactivity as a function of average core burnup for three different core configurations (from left to right: 69, 129, 216–232) For each core, measurements are compared with Serpent-2 and TRIGLAV burnup simulations It should be noted that the presented graphs represent only out of 46 analysed cycle excess reactivity changes Core configurations 216–232 represent multiple identical loading patterns between which changes were made that not affect the fuel burnup 6.1 Effect of isotopes on core reactivity reactivity changes due to burnup, we have analysed the effect of indi vidual isotopes on core reactivity of the JSI TRIGA Mark II reactor Similar analysis (Jeraj et al., 2002) was already performed in the past for a TRIGA SS 12 % fuel unit-cell using the WIMS (Kulikowska, 1996) code In the process of fuel burnup number density of more than 1400 different isotopes is constantly changing For the purpose of analysing 16 A Pungerˇciˇc et al Progress in Nuclear Energy 130 (2020) 103536 Fig 35 Top graph shows the burnup reactivity coefficient, defined in Eq (2), for 46 different core configurations Each measurement is compared with the Serpent-2 and TRIGLAV burnup calculations Bottom graph presents the relative difference between measurements and calculations One sigma uncertainty, defined as σdiff = √̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅̅ σ2calculations + σ2measurements , is depicted In that study, 40 different isotopes were studied at three different fuel burnups, at: %, 10 % and 20 % of the initial 235 U burned Analysis was repeated for the 14 isotopes that have the highest effect on reactivity according to the cited analysis using the Serpent-2 neutron transport and burnup code Excess reactivity of the benchmark core configuration No 132 (presented in Fig 19) was calculated at different burnups, each time excluding the analysed isotope The effect of individual isotope i on the core was determined as Eqn where ρall presents core reactivity with all the isotopes and ρall− excluding the one The calculated results were compared with the results presented in (Jeraj et al., 2002) Results from both analyses are included in Table and show extremely good agreement for most isotopes The results agree with our analysis of complete operational history, where the highest discrepancies were observed for 239 Pu It can also be observed that the difference between both codes is increasing with burnup However, very good agreement can be observed for 149 Sm and 135 Xe, especially for higher burnups We can conclude that the repeated (1) Δρi = ρall − ρall− i , Table Effect of individual isotopes on core reactivity Results were obtained by Monte Carlo simulations on benchmark core No 132 Obtained results were compared to WIMS unit-cell calculations, presented in (Jeraj et al., 2002) Calculations were performed with normalisation to maximum reactor power Pmax = 250 kW Units of (ninitial − nburned ) burnup are defined as BU [%] = , where n presents 235 U number density ninitial Effect on core excess reactivity [pcm] Fuel el burnup % Fuel el burnup 10 % Fuel el burnup 20 % Isotope WIMS Serpent Δ WIMS Serpent Δ WIMS Serpent Δ 135 Xe − 850 − 930 80 − 899 − 925 26 − 973 − 958 − 15 149 Sm − 620 − 710 90 − 638 − 690 52 − 645 − 657 12 151 Sm − 101 − 117 16 − 222 − 218 − 284 289 239 Pu ỵ295 ỵ106 189 þ357 þ432 ¡75 þ840 þ984 ¡144 143 Nd − 51 − 54 − 178 − 188 10 − 384 − 444 60 240 Pu − − 10 − 57 − 51 − − 216 − 260 44 236 U − 25 − 44 19 − 84 − 109 25 − 168 − 195 27 147 Pm − 24 − 45 21 − 65 − 61 − − 102 − 114 12 103 Rh − 20 − 36 16 − 82 − 84 − 179 − 181 131 Xe − 16 − 26 10 − 56 − 49 − − 118 − 146 28 133 Cs − 14 − 27 13 − 50 − 68 18 − 105 − 122 17 − 11 − 26 15 − 38 − 33 − 79 − 98 19 99 Tc 145 Nd − − 19 12 − 26 − 26 − 55 − 70 15 155 Eu − − 15 − 14 − 14 − 20 − 27 Δ = ΔρWIMS − ΔρSerpent Serpent statistical uncertainty: 1σ Serpent = pcm 17 A Pungerˇciˇc et al Progress in Nuclear Energy 130 (2020) 103536 calculated by TRIGLAV shows similar behaviour In comparison with the Serpent-2 code, higher discrepancies can be seen for the core configu rations before core No 85 One of the possible reasons for those dis crepancies can relate to the fuel elements, type AL 8.5 %, which were used in those mixed core configurations, as they represent the only variable that changes in comparison after core No 85 Another possible explanation is that the leakage term in TRIGLAV is described with buckling parameter, defined so that the calculated reactivity of bench mark core No 130 matches the measured one Since different mixed core configurations were used before the reconstruction buckling parameter was different, and therefore, the absolute comparison is biased As the source of the discrepancy is still unknown, it should be further analysed in the future Due to the simplicity of the diffusion approximation within the TRIGLAV code, we can conclude that the re sults are in good agreement compared with the Monte Carlo results We can conclude that the TRIGLAV code can be used for simple predictions of excess reactivity changes in TRIGA reactors Table Statistical analysis of differences between the measured and calculated linear burnup reactivity coefficient The total number of coefficients is 46, which represents 46 different core configurations and 2000 measurements of excess reactivity in total Analysis of 46 differences between calculations and measurements Average Median Within 1σ Serpent TRIGLAV − 1.8 % 19.3 % Within 2σ − 1.0 % 24 (52.2 %) 40 (86.9 %) 22.9 % 14 (30.4 %) 28 (60.9 %) analysis, using a different methodology and model, produces similar results to those obtained before At lower burnups, the highest contribution to core reactivity is due to isotopes 135 Xe (constant ≈ 900 pcm) and 149 Sm (constant ≈ 700 pcm), while at higher burnups, the contribution from 143 Nd, 240 Pu, 236 U, 103 Rh and 239 Pu becomes important Additional SERPENT-2 analysis was performed with various burn ups The results are presented in Fig 32 and show that at burnup of MWd 20 kg(HM) , which represents total burnup after the reconstruction in 6.3 Analysis of relative changes due to burnup So far, only the measurements at the beginning of reactor operation on each of the 240 cycles were used We decided to further analyse the relative changes of excess reactivity only due to fuel burnup on indi vidual core configurations For this analysis, we have chosen the core configurations where the burnup increment is substantial and where the measurements were performed regularly without having a poisoned reactor In total, 46 core configurations were chosen resulting in 2000 measurements used for the analysis 26 core configurations were used before the reconstruction in 1991 and consisted mostly of mixtures be tween Al, SS 8.5 % and FLIP-type fuel elements The rest were after the reconstruction and had SS 12 % type fuel elements, except for the already mentioned mixed core configurations 138–146 Fuel composi tion at the beginning of each cycle was taken from the operational his tory analysis Results for three core configurations with their schematics are presented in Fig 34 Results obtained from the burnup calculations were compared using the measured linear coefficient of reactivity change, defined as ] [ ΔρExcess pcm kg(HM) (2) Burnup reactivity coefficient = MWd ΔBurnup 1991, the effect of 239 Pu is ≈ 500 pcm and increases up to ≈ 1700 pcm when energy released in complete operational history is simulated on one core This subsection together with Sec 4.1 provides a clear overview of the reactivity effects in the TRIGA Mark II reactor and our treatment of its diverse operation in the burnup calculations We showed that reac tivity effect of long-lived or stable isotopes (235 U239 Pu, 143 Nd) does not change if different operating power is taken into account, however it is known that reactivity effect 135 Xe, 149 Sm and 151 Sm highly depend on the operating power due to the different saturation level From this a conclusion can be made that our methodology presented in Sec 4.1 is sufficient for the study of relative changes in excess reactivity, due to uranium depletion and plutonium production, presented in the next sections, however for absolute values of core excess reactivity detailed operational history should be taken into consideration to accurately predict the effect of xenon and samarium 6.2 Comparison with measurements Coefficients were determined using linear regression across the measurements and calculations Fig 35 presents the results for all 46 analysed core configurations The measurements predict the reduction of the coefficient for starting core configurations, which is due to the gradual insertion of FLIP-type fuel elements that contain burnable absorber erbium, which reduces the negative reactivity change due to burnup Similar observations can be made when comparing core con figurations before and after reconstruction, as the coefficients are higher after core No 138 (The first core conf after reconstruction was No 132) All the observed changes were well predicted by both the TRI GLAV and Serpent-2 codes, as clear change after the reconstruction was observed For the comparison of individual coefficients, the relative difference between calculations and measurements was analysed The results are presented in the bottom graph of Fig 35 It can be observed that the relative differences are smaller and mostly within 20 % after the reconstruction and around 40 % before the reconstruction, which is due to the improvement in quality of the measured data It can be observed that the coefficient calculated from the TRIGLAV results is almost constantly smaller than the measurements, which again shows a sys tematic discrepancy For better understanding of differences further statistical analysis was performed The results are presented in Table The average value of relative differences between Serpent and the measurements is − 1.8 % and the median is − 1.0 % which means that no systematic discrepancy The main analysis, which encompasses both the acquisition of operational data and the developed methodology of operational history simulation, is the comparison of measurements of excess reactivity performed at the beginning of each of the cycle with calculations per formed by TRIGLAV and Serpent-2 For the measurements, a 1σ uncer tainty of 500 pcm is assumed, as the same uncertainty was determined for the benchmark core configuration This uncertainty is under estimated for individual absolute measurements but overestimated for relative changes on the same core configuration Thus, we have focused our results on the shape of the burnup curve rather than absolute results The comparison between simulations and measurements is presented in Fig 33, where clear agreement between Serpent and the measurements is observed for the first 80 core configurations Almost all values are within the 1σ uncertainty The differences after core No 80 can be explained according to the introduction of older fuel elements with unknown burnup After the reconstruction in 1991, a constant discrep ancy of 1000 pcm − 1500 pcm is observed, which is in agreement with discrepancy for benchmark core No 190, as discussed in the previous chapter Since the calculated absolute values using Serpent-2 are not in perfect agreement with the reactivity measurements, the curve of the calculated Serpent-2 results follow the curve of the measurements throughout the operational history, and therefore, the conclusion is that the simulation methodology presented in this paper is satisfactory Excess reactivity throughout the complete operational history 18 A Pungerˇciˇc et al Progress in Nuclear Energy 130 (2020) 103536 can be observed as the distribution between positive and negative dif ference is equal The systematic discrepancy is clearly visible for the TRIGLAV results as the average value is +19.3 % and the median 22.9 % That the average and the median match for both codes shows that the data are statistically relevant The agreement between Serpent and measurements is good, as 52.2 % of the calculated coefficients are within 1σ, and 86.9 % are within 2σ of the measurements For TRIGLAV, the agreement is not as good because only 30.4 % of coefficients are within 1σ and 60.9 % within 2σ The comparison of excess reactivity measurements and calculations shows that, with the Serpent code and the methodology of simulating complete operational history we can predict the changes of excess reactivity due to burnup relatively well In this case limitations of the TRIGLAV code can be observed as it very well predicts the changes in coefficient due to different fuel elements but fails to accurately describe each operation It should be noted that the number of coefficients was 46 and with further TRIGA reactor operation more will be available and the analysis will become even more statistically relevant In the future, further improvement to the methodology of the operational history calculations can be performed and the coefficients used for further validation addition the position of the control rods should be considered, which could explain the presented discrepancies in comparing calculated and measured excess reactivity A similar conclusion can be made for angular distribution; it was shown that, for realistic burnups, the effect is negligible since the difference between maximum and minimum burnup was below % However, it could be substantial for individual isotopes, due to the spectrum change presented in the paper The main study presented in this paper is the analysis of core reac tivity with burned fuel elements It was shown that the effects of reactor poisons 135 Xe and 149 Sm on TRIGA core reactivity are − 930 pcm and − 690 pcm, respectively The results were compared to similar findings obtained with unit-cell analysis with WIMSD-5B (Jeraj et al., 2002) and it can be observed that highest discrepancies are for 239 Pu Nevertheless, the results are in good agreement, and it can be concluded that the physics of the fuel burnup changes can be sufficiently described with using only unit-cell calculations In addition, weekly measurements of excess reactivity were used to validate the complete operational history simulations, and it can be concluded that the agreement between both absolute and relative excess reactivity data and Serpent-2 simulation is good The relative changes are also well described by the TRIGLAV code One of the main motivations for this work was to investigate the discrepancies of measured and calculated keff for core No 190 By simulating the complete history, detailed fuel burnup information for each core configuration was obtained and used in the validated MCNP model We were able to explain 4000 ± 100 pcm out of a 5500 ± 500 pcm discrepancy Further analysis is needed to understand the remainder The overall conclusion to the presented work is that, by having both full Monte Carlo burnup calculations and detailed opera tional history data, we were able to obtain accurate fuel burnup data that can be used in the future for experimental campaign support, reactor decommissioning, fuel management, and most importantly, validation of new methodologies for calculating burnup (Roskoff and Haghighat, 2018) Discussion and conclusion The JSI TRIGA research reactor offered a unique opportunity to perform a detailed analysis of its fuel element burnup as the complete operational history is well documented With the performed operational history analysis, data were obtained that enable simulation of the complete operational history As each individual operation was recor ded, in the future all of the 27 000 reactor power changes could be simulated In addition, excess reactivity measurements were recorded, enabling the validation of burnup methodologies and the study of un certainty propagation through burnup and multiple sensitivity studies (ParkDong et al., 2018; García-Herranz et al., 2008) The final fuel element burnup and its isotopic composition were calculated with both Monte Carlo Serpent and deterministic TRIGLAV code For most fuel elements the differences between both codes are within 10%, which increases to 30% when comparing isotopes, such as 239 Pu and 137 Cs Two methodologies were also compared on criticality benchmark core No 132 It can be concluded that the highest difference is for fuel elements with higher amounts of surrounding water, meaning that the unit cell employed in the TRIGLAV code could be improved in the future Nevertheless, it is shown that despite the simplicity of the TRIGLAV code the obtained results show great promise in using the code for TRIGA reactor fuel management and burnup analysis The analysis of spatial burnup effects showed that, in the future, axial and angular depletion zone division should be implied as the differences between maximum and minimum burnup in a fuel element could be more than 25 %, resulting in high sensitivity of control rod insertion In Declaration of competing interest The authors declare that they have no known competing financial interests or personal relationships that could have appeared to influence the work reported in this paper Acknowledgements The authors acknowledge the project (Young researcher project Anˇze Pungerˇciˇc, 52060) was financially supported by the Slovenian Research Agency The authors acknowledge the financial support from the Slovenian Research Agency (research core funding No P2-0073) Appendix A Supplementary data Supplementary data to this article can be found online at https://doi.org/10.1016/j.pnucene.2020.103536 A Ray-tracing algorithm for water thickness determination In the analysis of fuel burnup spatial effects, it was determined that fuel surrounding is important from the standpoint of the fuel burnup and isotope distribution Water thickness around fuel elements was determined with a simple ray-tracing algorithm described here For a defined point in Cartesian coordinate system P a set of rays is defined, varying in direction D so that the whole 360∘ angle is covered In a TRIGA research reactor fuel elements, control rods and graphite reflector can all be defined with a circle at position C and radius R > Considering a ⃒ ⃒ parameterized ray X(t) = P + tD and a implicitly defined circle ⃒X − C|2 = R2 , a substitution of the ray equation into the circle equation can be made, by defining Δ = P − C, to obtain quadratic equation in t Eqn 3: ⃒ ⃒ ⃒ 22 ⃒ ⃒D| t + 2(D ⋅ Δ)t + ⃒Δ| − R2 = 0, 19 (3) A Pungerˇciˇc et al Progress in Nuclear Energy 130 (2020) 103536 where the formal roots of the equation are Eqn √̅̅̅ − D⋅Δ ± δ ⃒ t= , ⃒D| (4) ⃒ ⃒ ⃒ ⃒ and the determinant is defined as δ = (D ⋅Δ) − ⃒D|2 (⃒Δ|2 − R2 ) If δ < 0, the line does not intersect the circle If δ = 0, the line is tangent to the circle (one point of intersection) If δ > 0, the line intersects the circle in two points, and the one closer is chosen in our case The algorithm calculates the distance between the source point of the ray and the closest intersection with the circle This is done for all defined rays and circles (core elements) Number of rays defines the angular resolution A python script was created that reads the TRIGA reactor core configuration and defines the needed fuel elements, control rods and graphite reflector Two examples of the described ray-tracing algorithm for a TRIGA reactor are presented in Fig 36 In the last step the distance is averaged over multiple rays to obtain the water thickness at desired angles Fig 36 Screenshot of the ray-tracing algorithm for determination of water thickness around fuel elements in a TRIGA reactor References Jeraj, R., Ravnik, M., 2010 Triga mark ii reactor: U (20)-zirconium hydride fuel rods in water with graphite reactor, ieu-comp-therm-003 In: INternational Handbook of Evaluated Criticality Safety Benchmark Experiments NEA/NSC/DOC, vol 3, 95 Jeraj, Robert, Glumac, Bogdan, Mauˇcec, Marko, 1997 Monte Carlo simulation of the triga mark ii benchmark experiment Nucl Technol 120 (3), 179–187 ˇ Jeraj, Robert, Zagar, Tomaˇz, Ravnik, Matjaˇz, 2002 Monte Carlo simulation of the triga mark ii benchmark experiment with burned fuel Nucl Technol 137 (3), 169–180 Khan, R., Karimzadeh, S., Bă ock, H., 2010 Triga fuel 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in the reactor operation parameters 2.1 Reactor description Analysis of reactor operation for the purpose of detailed burnup determination was performed on a 250 kW TRIGA Mark II research reactor. .. description of the JSI TRIGA Mark II reactor • Reactor power changes: analysis of individual reactor operations to accurately calculate released energy • Fuel shuffling: analysis of all of 240 core