Effect of burnup on the neutronic parameters of ITU TRIGA Mark II research reactor

Progress in Nuclear Energy 83 (2015) 26e34

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Effect of burnup on the neutronic parameters of ITU TRIGA Mark II research reactor Mehmet Türkmen a, *, Üner Çolak b, S¸ule Ergün a a b

Department of Nuclear Engineering, Hacettepe University, Beytepe Campus, Ankara, Turkey a Campus, Maslak, Istanbul, Turkey Energy Institute, Istanbul Technical University, Ayazag

a r t i c l e i n f o

a b s t r a c t

Article history: Received 7 November 2014 Received in revised form 18 February 2015 Accepted 28 February 2015 Available online

This study aims to estimate burnup of the fuel elements for the Istanbul Technical University TRIGA Mark II Research and Training Reactor using a Monte Carlo-based burnup-depletion code. Effect of burnup on the core neutronic parameters, effective core multiplication factor, fast/epithermal/thermal neutron fluxes, and core-average neutron spectrum, and incoming neutron spectrum of the piercing beam port (PBP), is investigated at the Beginning of Life (BOL) and End of Life (EOL). Operational data peculiar to a selected operation sequence, which contains positions of CRs, power level of the reactor, material temperatures and latest core map, are used to determine the current fuel burnup of fuel elements at the time under consideration. A specific operation sequence is selected for the analysis. Furthermore, all control rods are considered fully withdrawn to assess the excess reactivity. Results are obtained using MONTEBURNS2 with ENDFB/V-II.1 neutron/photon library for a full power of 250 kW. Neutron crosssection libraries at the full-power operating temperatures are generated using NJOY. From the results, the calculated burnup values of the core at the sequence considered and EOL are found to be 420 MWh and 560 MWh, respectively. Remaining excess reactivity is calculated to be less than 0.3 $. It is observed that core average thermal neutron flux reduces by 1 % while the fast and epithermal neutron fluxes remain almost unchanged. © 2015 Elsevier Ltd. All rights reserved.

Keywords: ITU TRIGA Mark II Monte Carlo method Fuel burnup MONTEBURNS2 Piercing beam port Neutron spectrum

1. Introduction Accurate burnup calculation for the fuel elements plays important role in the fuel (re)shuffling strategy and precise operation planning of the reactor as well as scheduling for the fuel supply. On the other hand, for the research reactors, burnup calculation of fuel elements is very complicated due to varying operational conditions, power (history), and heterogeneous core configuration. The most common method used to determine the fuel burnup is to utilize total amount of produced energy (time integral of the power history) recorded throughout the reactor operation. Nonetheless, according to Dalle and Veloso (2006), log-books inescapably have an uncertainty of 15% because of thermal calibrations and power oscillations. In addition to this, when the power history is not registered in detail, it causes ambiguity, it prohibits calculating the burnup of the fuel elements accurately.

* Corresponding author. Tel.: þ90 312 297 73 00; fax: þ90 312 299 21 22. E-mail address: [email protected] (M. Türkmen). http://dx.doi.org/10.1016/j.pnucene.2015.02.012 0149-1970/© 2015 Elsevier Ltd. All rights reserved.

At the end of cycle, reactor core requires fresh fuels to be used to replace the most-burned fuels. Similarly, an alternative solution is to (re)shuffle the fuels by employing an in-core fuel management with a standard out-to-in scheme; but, it has been learned from the previous studies that the (re)shuffling does not give any guarantee for maximum utilization of the fuels. For an optimum burnup, varying fuel loading schemes need to be searched (Lyric et al., 2013). In addition, genetic algorithms can be used to find out the optimal loading patterns among the various fuel loading schemes (Do and Nguyen, 2007). In the literature, there are a number of research articles on the burnup calculation of fuel elements for a particular research reactor; however, all of them use the common method mentioned above. From this respect, the method employed in this study is unique. On the other hand, several studies directly related with this work are outlined below. Computer codes are used in burnup calculation of the fuel elements for the TRIGA Mark II type research reactors. MONTEBURNS2 (with ORIGEN2) code was used to compare its results with the experimental data measured in TRIGA Mark II at Josef Stefan

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Institute, Slovenia (Dalle and Jeraj, 2002; Jeraj et al., 2001). For the TRIGA IPR-R1 reactor, the results of calculations for the excess reactivity, control rods (CRs) reactivity worth, neutron fluxes and power distribution were presented (Dalle and Veloso, 2006). In addition to MONTEBURNS2, TRIGLAV linked with WIMS is the other most-frequently used computer code. Burned fuel isotopic composition was obtained by using TRIGLAV, WIMSD4 and ORIGEN2 for the TRIGA reactor (Ravnik et al., 1999). In that study, burnup of several HEU/LEU fuel elements for the mixed core (HEULEU) was measured with the reactivity method that uses the relationship between burnup and reactivity based on the reference values of well-defined burnup versus reactivity curves of several particular fuel elements. The measured values were compared with the calculated results using TRIGLAV and TRIGAP codes. Reactivity defects of fuel elements were used in Istanbul Technical University (ITU) TRIGA Mark II reactor to calculate the burnup (Büke and Yavuz, 2002). For the same reactor, TRIGLAV5 results were compared with the MONTEBURNS2 results by taking some reactor parameters such as excess reactivity and fuel element burnup into account (Türkmen and Çolak, 2014a). Results of TRIGAP computer code were compared with the results of MVP-BURN and MCNP4CORIGEN2.1 codes for the TRIGA Mark II at AEAE (Huda et al., 2008). In-core fuel management was discussed from the viewpoint of criticality, peaking factors, neutron flux, and burnup change with time in this study. Change of the excess reactivity, peaking factors, positions of the CRs and thermal neutron flux distribution with burnup for the Moroccan TRIGA research reactor were carried out using MCNP and BUCAL1 codes (El Bakkari et al., 2013a; El Bakkari et al., 2013b). MVP-BURN code was used to investigate optimum loading strategy in the TRIGA Mark II reactor of BAEC (Lyric et al., 2013). In addition to the simulations, experimental results of reactivity measurements, fuel element reactivity worth and fuel burnup measurements for partly burned reactor core were provided (Persi c et al., 1998). Experimental burnup measurements were presented by measuring 137Cs fission product (FP) activity as a burnup indicator with HPGe detector (Khan et al., 2010). This study is performed as part of an ongoing extensive neutronic research to verify and validate the computer model of the ITU TRIGA Mark II Research and Training Reactor. Validation of the reactor model has been successfully shown in earlier studies (Türkmen and Çolak, 2014b, 2013). In this study, fuel burnup is estimated by using positions of the CRs for the most recent operation registered in the official log-book of ITU TRIGA Mark II reactor. Moreover, effect of the burnup on the reactor parameters such as multiplication factor and neutron flux/ spectrum is investigated. Further, discharge burnup at the End of Life (EOL) and the remaining excess reactivity of the considered operation sequence are obtained while all the Control Rods (CRs) are withdrawn. Up-to-date and EOL burnup values of the fuel elements are calculated using MONTEBURNS2 with ENDFB/V-II.1 neutron/photon library. Neutron cross-section libraries at the fullpower operating temperatures are generated using NJOY. In the study, BOL and EOL (discharge) refer to the beginning of life and end of life burnup calculations for the fully withdrawn positions of CRs. 2. The codes used, Monte Carlo model and the methods The basic method employed for the burnup calculation is to use the total amount of energy generated during the operation (recorded into the log-books of the reactors); but, in this study, it is assumed that a detailed power history of the reactor has not been recorded to the log-book as the reactor operates. Now, a different problem arises: if the log-books are not useful anymore, so how can

27

Fig. 1. MCNP representation of the reactor core.

an up-to-date burnup calculation be achieved? The solution comes from the position of the CRs inside the core. Criticality during a particular reactor operation is ensured by moving the CRs upward or downward inside the core. However, the positions of the CRs inside the core are very peculiar to the reactor operation sequence. This means that burnup values of the fuel elements can be easily calculated using a Monte Carlo (MC) based burnup-depletion code if the reactor is modeled at the conditions similar to the considered operation sequence. Operational data contain the information about not only the positions of the CRs, but also material temperatures, reactor power history and the latest core map. According to the method, any operational data (preferably the most recent) are literally put into the MC model of the reactor. Subsequently, fuels are burned using a burnup code until the excess reactivity1 of the core becomes zero. Once the excess reactivity falls down below zero, the reactor theoretically could not maintain its criticality and, finally shut itself down. On the other side, due to the fact that all the CRs need to be positioned outside the core at the EOL, a relatively easy calculation for the discharge burnup is possible by completely withdrawing the CRs from the core in the MC simulation. The most-recent burnup calculation is carried out using the reactor operation sequence of 1599 (# 1599) recorded into the ITU TRIGA Mark II reactor log-book in March 2013. During the operation, transient and safety rods were set to 96.89% and 94.89% of their full positions, respectively. Position of the regulating rod inside core was not fixed during the operation in order to maintain criticality; however, an average value of 83.83% is used in the MC calculation. An MCNP representation of the reactor including reactor core, Beam Ports (BPs), thermal column and, water and graphite blocks is presented in Fig. 1. Core loading map used in the burnup calculations is shown in Fig. 2.

1 The amount of surplus reactivity that is needed to compensate for decrease of fissile content and accumulation of the FPs.

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Fig. 2. Core loading map used in the burnup calculation.

Discharge burnup at the End of Life (EOL) and the remaining excess reactivity of the considered operation sequence are obtained while all the Control Rods (CRs) are withdrawn. Feedback effects caused by a change in the density and crosssection of materials due to a change in material temperatures are taken into account; however, they are assumed to be invariant. For this purpose, average material temperatures are set to 380 K for the clad, 450 K for the fuel, 540 K for pure zirconium rod, 310 K for water inside the core and 300 K for the rest of structures, respectively. As the zirconium rod is located at the center of the fuel region, average rod temperature nearly equals to maximum temperature of the fuel (to inner surface temperature) and thus, naturally would be higher than that of the fuel. NJOY99 with an update 396 (MacFarlane and Kahler, 2010; MacFarlane and Muir, 1994) is used to generate the neutron cross-sections at the specific material temperatures as well as photo-atomic and photoenuclear reaction libraries. Thermal scattering libraries are also produced for the water and graphite, and the zirconium hydride fuel. A flow diagram of NJOY is presented in Fig. 3. As a nuclear data processing code, NJOY is made up of a set of modules such as MODER, RECORN, and BROADR. Each module (job details of the modules are given in Appendix A) performs a separate task to produce final form of the cross-sections. As the required input parameters for each module are almost automatized by the code developers, default values of the parameters, which are appropriate for most problems, are used. Yet, the code still needs somewhat user knowledge to open the optional cards (not deadly ones). Cross-sections can be generated for any desired temperature/reaction/particle of a selected isotope(s). In brief, after a great deal of straightforward calculation, the code produces ACE

2 The criticality test of the ACE libraries is carried out using the ‘International Handbook of Evaluated Criticality Safety Benchmark Experiments (OECD-NEA ICSBEP)’.

formatted files for the MCNP code and several control plots to check the ACE files.2 Evaluated Neutron Data Files (ENDFs) primarily come from the updated ENDF/B-VII.1 library (Chadwick et al., 2011); however, JEFF3.2 (JEFF Team, 2014), CENDL3.1 (Ge et al., 2010), JENDL4.0 (Shibata et al., 2011) and TENDL2013 (Koning et al., 2013) are used for the isotopes that are not in the library used. MC model incorporates neutron transport calculation with photon transport and photo-nuclear reaction calculations. Photon interactions are turned on/off using the PHYS card. However, it should be noted that photon calculation is not a necessity for a burnup calculation. In burnup calculations, KCODE criticality calculations use 10 000 neutrons per cycle with 50 passive and 500 active cycles. The provided neutron history yields a relative error of less than 0.0007 for the core multiplication factor. Typically, sensitivity of the multiplication factor for an accurate precision should be less than 0.1%. For the other calculations, one million neutrons per cycle with 100 passive and 1500 active cycles are used. F2 surface flux tally (particles/cm2 s) and FMESH averaged over a mesh cell (particles/cm2) are used to obtain the group fluxes of neutrons. Tally results are multiplied with a tally multiplier to convert into the flux:

C¼

v  P  106 ðW=MWÞ 1:602  1013 ðJ=MeVÞ  keff  Qave

where average number of neutrons produced per fission, v, is 2.438; reactor power of P is 250 kW; effective multiplication factor, keff, changes with the burnup, and average fission energy produced from all fissionable material, Qave, is about 200 MeV. The group energy boundaries are 0e0.625 eV for thermal, 0.625 eVe0.1 MeV for epithermal and 0.1 eVe20 MeV for fast neutrons. For the neutron spectrum calculations, number of energy group is defined as 1195 for the particles. MONTEBURNS2 (Poston and Trellue, 1999) [MCNP5 (X-5 Monte Carlo Team, 2003) linked with ORIGEN2.1], a fully-automated

M. Türkmen et al. / Progress in Nuclear Energy 83 (2015) 26e34

29

Fig. 3. NJOY flow diagram.

burnup depletion code, is used to calculate the burnup of fuel elements. The code requires four separate input files, MCNP, MONTEBURNS, feed and cross-section input files. MCNP input file comes from the previous reactor model (Türkmen and Çolak, 2014b). MONTEBURNS input file is created; input parameters are set to 9 for outer burn steps (outer iteration for convergence), 100 for internal burn steps (inner iteration for convergence), no intermediate keff calculation (initial guess for subsequent steps) due to no feed, and zero fractional importance to take all of the FPs into account. Simply, no feed indicates that no fuel is loaded into the core during the burnup calculation. The justification for that is that the reactor records show, thus far, no change in the current reactor core configuration. Furthermore, if the fractional importance is set to zero in the input file, it means that contribution of each FP calculated in the ORIGEN2, even negligible ones, is to be calculated by the code. For the contribution of only major FPs, the fractional importance should be one. On the one hand, an initial ORIGEN2 library, named TRIGA, is produced from a previous test run as an initial library for a TRIGA Mark II has not been provided by Radiation Safety Information Computational Center (RSICC). For tally calculations, a total of 250 critical isotopes, including FPs and actinides, are chosen. A feed input file is also prepared; the time steps are selected to be 3 days (from user experience) for the FPs equilibrium and 20 days for the rest. More, no feed or discharge from the reactor core during the irradiation is considered. The crosssection input file contains a list of transmutation cross-sections for a total of 533 isotopes. Since the power history is not represented accurately in the log book, the reactor is considered to be operated at a constant power of 250 kW. However, constant power approach limits accuracy of the fuel burnup calculation as the accumulation and depletion of the FPs directly depends on power level of the reactor. The most important impact of this simplification is that the user is obligated to make an assumption of FP equilibrium in burnup calculation. A computational flow diagram of burnup calculation is shown in Fig. 4. The output of the code gives burnup values as a function of core multiplication factor (or in terms of reactivity). For an accurate burnup calculation, Non-Linear Reactivity Model (NLRM) (Driscoll et al., 1990) is employed. Therefore, a second order curve fitting for the Eq. (1) is performed assuming a single-batch core. Using the equation, burnup can be easily calculated by setting the fitted equation to zero [r(B) ¼ 0].

rðBÞ ¼ A2  B2 þ A1  B þ A0

(1)

where A2 [1/(MWh)2], A1 [1/MWh], A0 are the fitting constants, and B [MWh] is the specific burnup.

3. Results For the cases of fully withdrawn CRs and the position of CRs in the operation sequence # 1599, the change of the core excess reactivity ($) as a function of cumulative burnup (MWh) (and the irradiation time in the upper x-axis [Full Power Days e FPDs]) is illustrated in Fig. 5 with error bars. The results are fitted to a second order polynomial equation (excluding initial reactivity) with

Fig. 4. Computational flow diagram.

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Fig. 5. Variation of the reactivity ($) with the cumulative burnup (MWh).

Table 1 Coefficients of best fit lines with goodness values. A2 # 1599 Withdrawn

A1 7

6.285  10 3.361  105

3

2.463  10 1.737  102

A0

R2

0.920 1.332

0.55 0.66

sufficient accuracy to obtain the burnup value of each fuel element. Fit results are depicted in the figure as dashed and dotted lines. Coefficients of best fit lines with goodness of fit (R-squared values) are tabulated in Table 1. Excess reactivity decreases smoothly (as quasi-linear) as the burnup goes up after a sharp decrease due to the accumulation of FPs. Initial reactivity of the fresh core at 250 kW is about 3.35 ± 0.1 $. Burnup at sequence # 1599 is obtained to be 420 (17.5 MWd) ± 12 MWh with a total irradiation time of 70 ± 2 days. When all the CRs are withdrawn, the EOL burnup is calculated to be 560 ± 16 MWh at the end of 95 ± 3 days. Further, the core excess reactivity in March 2013 (from the difference between # 1599 and withdrawn position of the CRs) is calculated to be about 0.3 ± 0.1 $. Results show that current fuels can burn as much as 140 MWh for 25 days. In other words, the reactor may practically shut down within one year after operating two hours per day at the full power. After that, either the fuels will have to be replaced with the fresh ones or the (re)shuffling will have to be performed. From March 2013 on, the reactor has been used for the research studies such as Boron Neutron Capture Therapy (BNCT) application and operated several hours per day at full power. Thus, it appears that the reactor has not, as of today, much excess reactivity (with a good estimation, it must be less than 0.3 $). This conclusion is obtained from the simulation results; therefore, it really needs an experimental verification.

3 Although the burnup unit is expressed for the oxide fuels as MWd/kgU, this unit is not appropriate when the hydride fuels are considered. Instead, the literature on hydride fuels, most commonly, expresses the burnup unit in terms of fissions per initial metal atom (FIMA). However, for unfamiliar readers, it will be very helpful to use a conversion factor of 3.678  103 for 20% enriched 8.5 U wt. % hydride fuel to convert from FIMA to MWd/kgU.

For the EOL and sequence # 1599, burnup values of the fuel elements in terms of consumed 235U isotope (%)3 are illustrated in Fig. 6. Core average EOL burnup is about 1.03% while the average burnup of sequence # 1599 is 0.80%. Maximum EOL burnup of 1.51% is observed in fuel rod B2 (marked as dashed) while the minimum is 0.63% in the F27 (dotted). Also, it is calculated from the simulation that thermal neutrons are responsible from 93.8% of the total fission while contributions from the epithermal and fast neutrons are 5.4 and 0.8%, respectively. It remains almost constant as the burnup changes. Variation of the neutron flux with radial distance is shown in Fig. 7 at the BOL and the EOL. The results are obtained using core map given in Fig. 2. The radial distance is divided into 1000 equally sized pieces. Zero radial distance means core center; negative and positive values illustrate right and left hand sides of the core map. Besides reactor core, neutron flux calculation is extended through the graphite and water blocks. As seen in Fig. 7, thermal/fast flux4 sharply decreases with the increasing distance from the core center. On the contrary, epithermal flux decreases smoothly. Inside the graphite block, thermal and epithermal neutron fluxes level off while the fast flux rapidly decreases. Moreover, although it seems that the fluxes are symmetrical in shape and value, in fact, neutron flux at the right section is higher than the other side. This is because the right side of the reactor opens to a thermal column (referring to Fig. 1) while the other loop opens to piercing BP (PBP). This spatial effect clearly leads to change in neutron distribution inside the core (due to leakage from the BP). The fast flux falls down where the thermal flux leaps up due to slowing down process of the neutrons. Neutron fluxes at EOL (dashed lines) in the core decrease as the burnup of the fuel increases. The highest thermal flux is calculated in the core center (Central Thimble) while the fast/epithermal flux is in the ring B. To increase the clarity of Fig. 7 more, fractional change of the neutron fluxes from the BOL with respect to radial distance in units of percentage is plotted in Fig. 8. According to the figure, thermal flux reduces by a maximum value of 3% near the core center. This effect outstands at the locations (e.g., ring B)

4

Maximum relative error is 0.4%.

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Fig. 6. Calculated burnup values (% consumed

where more fissile material is consumed. On the contrary, change in the fast and epithermal fluxes is not as high as the thermal flux; however, all of them show similar trend. It appears that compared with BOL values, neutron fluxes reduce to some degree in the mostinner rings and increase in the most-outer rings. Furthermore, there is a net increase of about 1% inside the graphite block, in particular for the left side of the reactor. Power production shifts from inner rings to outer rings. Since the reactor has relatively low

31

235

U isotope) of fuel elements at EOL and # 1599.

discharge burnup (only 1%), only a small decrease is observed in the fluxes. The change of average neutron spectrum in the core at the BOL and EOL is plotted in Fig. 9. The EOL results are also provided in terms of fractional change from the BOL (EOL/BOL) in units of percentage. For tally purposes, the mesh geometry is limited to the fuel element length (including top/bottom graphite reflectors) in axial direction and to inner surface of the graphite reflector in radial

Fig. 7. Change of the neutron fluxes with radial distance at BOL and EOL.

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Fig. 8. Fractional change of the neutron fluxes from BOL with respect to radial distance.

direction. As can be observed, there is a certain decrease in thermal energies up to 2% even though the obtained spectra5 are nearly the same. This is due to the transient FP content of spent fuel such as 149 Pm, 149Sm, and 135I having huge thermal capture cross-sections. Furthermore, burnup has a little effect on the neutron energies ranging from 0.005 eV to 0.2 keV. On the other hand, more than 2% changes observed in lower/higher neutron energies6 arise from high computational error of the tallies. Also, it is apparent that increase in burnup does not result in spectrum shifting. It has been decided to use the PBP in future studies for BNCT applications. For this purpose, effect of the burnup on the spectrum of the neutrons arriving to the BP has been investigated. The energy distribution of incoming neutrons into the PBP at the BOL and EOL is plotted in Fig. 10. From the results, it is observed that the spectrum7 is significantly affected from the increase in burnup. Change of neutron distribution in epithermal energies with almost 5% is higher than thermal energies. Impact of resonance overlap becomes important since the nuclides having higher resonance capture cross-section accumulate in the spent fuel with burnup. For instance, 151Sm isotope has large cross-section at the 6.16e7.52 eV. 238 U has also a large resonance peak in this energy range. Therefore, neutron spectrum plotted in Fig. 10 shows significant fluctuation. It is also observed that there is no neutron with energies less than 104 eV.

4. Discussion Calculations show that reactor possibly may experience lack of excess reactivity within one year provided that it operates several hours per day at the full power. Current positions of CRs also confirm the presumption suggested in this study to estimate the most-recent burnup. Thus, the reactor will require a number of fresh fuels in the near future. However, since there is a fuel shortage problem in TRIGA reactors due to dilemma in the fabrication of fuel

elements, (re)shuffling can be a good solution to keep the reactor on operating at least for a short while. Shuffling the fuel elements would not be sufficient alone as there are strict constraints on the core map such as allowed maximum excess reactivity and maximum peaking factor, and of course, the fuel utilization. To solve this complicated problem, a suitable way is to use Genetic Algorithm (GA) method for the new core map generation. The GA can readily find the best solution(s) inside the search space for the specified conditions. For this purpose, reshuffling (whether possible or not) will be investigated using the GA method by the authors of this paper in a follow-up article. For comparison purposes, a standard out-to-in fuel management is planned to perform for the ITU TRIGA Mark II reactor, as well. Moreover, to validate the MC results, burnup measurement experiments must be performed by measuring the activity of the long-lived FPs in the fuel elements. For instance, 137Cs isotope with a half-life of about 30 y is a good indicator of burnup. In the course of burnup analyses, CRs are fixed to a constant position and does not move. However, during normal operating condition, the positions of CRs, in particular regulating rod, are continuously changed to maintain criticality. This situation results in somewhat modeling error near the adjacent fuel rods of CRs. Moreover, the reactor operates at varying power levels; but, in our model, it is assumed that it operates at a constant power for the burnup calculation. This affects accumulation of poisonous FPs, the decay of some short-lived isotopes, and hence, the calculated burnup. Other possible impact of constant power approach is the miscalculation of burnup values of individual fuel elements in spite of the fact that core average burnup is totally consistent (from conservation of energy) with the case that uses varying power history. To sum up, it should be expected to find somewhat higher burnup values; unluckily, somewhat will remain an uncertain quantity till the experimental measurement. In the near future, the PBP will be used for industrial applications such as BNCT, radio-isotope production, and Neutron Radiography tests. For this purpose, the availability and capability of the BP, also radial8 and tangential BPs, need to be studied in the future

5

Computational error of the neutron fluxes is less than 0.3%. Typically energies less than 103 eV and higher than 10 MeV. 7 Maximum relative error does not exceed a value of about 4% in the energy interval of 109e13 MeV. 6

8

Several BNCT experiments have been carried out last two years.

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33

Fig. 9. Core average neutron spectrum at BOL and EOL with fractional change of EOL from BOL.

Fig. 10. Incoming neutron spectrum of the PBP at BOL and EOL with fractional change of EOL from BOL.

works. However, it is understood from the results that a burnupcorrected neutron spectrum should be used in the succeeding studies when BPs are modeled. 5. Conclusion In this study, EOL burnup of the each fuel element is determined by using MONTEBURNS2 when all the CRs are out. In order to estimate the up-to-date burnup of each fuel element, positions of CRs of operation sequence # 1599 are used in MC model. Change of the some important reactor parameters (excess reactivity, neutron spectrum, etc.) with burnup is examined. From the results, it is understood that the reactor has very little excess reactivity remained which is expected to be about 0.3 $. The reactor possibly will be operating at most 140 MWh for 25 FPDs after operation sequence # 1599. Core average burnup for the operation sequence # 1599 is found to be 0.8% while the leastburned and most-burned fuels have burnup values of 0.49% and

1.17%, respectively. There is a decrease in EOL neutron fluxes when compared with BOL. In addition, average neutron spectrum of the core, specifically thermal neutrons, is clearly affected from the burnup. Too much fluctuations due to burnup increase are observed in the incoming neutron spectrum of the PBP. A summary of neutron flux change at the BOL and EOL in the core and PBP is listed in Table 2. During the reactor operation, core average fast and epithermal fluxes remain constant although thermal flux reduces to some extent (nearly 1%). In the case of PBP, even though epithermal flux does not change, thermal flux decreases by 0.45% and fast flux increases by 1%. As shown in table, neutron flux values of the core except fast flux are less than the PBP. This is because the provided values under the ‘Core’ columns are averaged over the core volume (that is, those are the mean values). Those are statistically well-converged values since a good deal of number of neutrons inside the core are simulated. On the other side, only the neutrons crossing the port surface located at the intersection of port and core are tallied, and the records are

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Table 2 Neutron fluxes (1012 n/cm2.s) at the BOL and EOL in the Core and Piercing BP. Neutron flux

Core

Piercing BP

BOL

EOL

EOL/BOL (%)

BOL

EOL

EOL/BOL (%)

Thermal (0e0.625 eV) Epithermal (0.625 eVe0.1 MeV) Fast (0.1e20 MeV)

3.0496 ± 0.0006 2.6119 ± 0.0003 2.9416 ± 0.0004

3.0186 ± 0.0006 2.6118 ± 0.0003 2.9417 ± 0.0004

1.02 0 0

4.9542 ± 0.0109 3.4006 ± 0.0095 2.7304 ± 0.0076

4.9317 ± 0.0109 3.4060 ± 0.0095 2.7586 ± 0.0077

0.45 þ0.16 þ1.03

Total

8.6031 ± 0.0009

8.5721 ± 0.0009

0.36

11.085 ± 0.016

11.096 ± 0.017

þ0.10

averaged over the surface area. Therefore, when compared with the core volume, number of neutrons tallied in the entry surface of the beam port are significantly less. This clearly illustrates that burnup is less influential on the average values and more influential on the local fluxes. Appendix A. Detailed description of NJOY Nuclear Data Processing System, shortly called NJOY, produces point-wise and multi-group cross-sections from Evaluated Nuclear Data for a variety of particle transport/diffusion codes including MCNP. NJOY consists of a number of modules, each of which is coded to perform a separate task; however, only the used modules are introduced related with the scope of this study. As shown in Fig. 4, the modules used to produce ACE cross-section files consistent with MCNP are MODER, RECORN, BROADR, THERMR, HEATR, PURR, GASPR and ACER, respectively. In the NJOY, MODER is used to convert binary-to-ASCII or ASCII-to-binary mode; RECORN is used to reconstruct resonance cross-sections from resonance parameters; BROADR is used to generate Doppler-broadened crosssections; THERMR is used to generate point-wise neutron scattering cross-sections in the thermal energy range; HEATR is used to generates point-wise heat production cross-sections and radiation damage energy production for the specified reactions; PURR is used to produce probability tables for MCNP; GASPR is used to add gas production reactions; ACER is used to prepare libraries in ACE format for MCNP. Apart from the mentioned modules, there are a number of distinct modules used to verify the produced neutron cross-sections. ACER is capable of making a series of consistency checks and produces several plot files for which VIEWR module prepares the plot files as Postscript file. Furthermore, PLOTR is very useful when the generated cross-sections are compared with the raw data of ENDF file. References Büke, T., Yavuz, H., 2002. Fuel element burn-up calculation in ITU TRIGA Mark-II Reactor. In: First World TRIGA Users Conference, LENA, Pavia, Italy, June 1620, pp. 96e103. Chadwick, M.B., et al., 2011. ENDF/B-VII.1: nuclear data for science and technology: cross sections, covariances, fission product yields and decay data. Nucl. Data Sheets 112 (12), 2887e2996.

Dalle, H.M., Jeraj, R., 2002. Validation of the MONTEBURNS code for criticality calculations of TRIGA reactors. Res. Dev. Braz. J. 4 (2), 91e94. Dalle, H.M., Veloso, M.A., 2006. Monte Carlo simulation of TRIGA IPR-R1: reactivity worth, burnup, flux and power. In: 3rd World TRIGA Users Conference, Minas Centro-Belo Horizonte, Minas Gerais, Brazil, August 22e25. Do, B.Q., Nguyen, P.L., 2007. Application of a genetic algorithm to the fuel reload optimization for a research reactor. Appl. Math. Comput. 187, 977e988. Driscoll, M.J., Downar, T.J., Pilat, E.E., 1990. The Linear Reactivity Model for Nuclear Fuel Management. American Nuclear Society, La Grange Park, Illinois. El Bakkari, B., et al., 2013a. Fuel burnup analysis for the moroccan TRIGA research reactor. Ann. Nucl. Energy 51, 112e119. El Bakkari, B., et al., 2013b. MCNPX-BUCAL1 code to code verification through burnup analysis. Ann. Nucl. Energy 60, 242e247. Ge, Z.G., et al., 2010. The updated version of Chinese evaluated nuclear data library (CENDL-3.1). In: Proc. International Conference on Nuclear Data for Science and Technology, Jeju Island, Korea, April 26e30. Huda, M.Q., et al., 2008. Burnup analysis and in-core fuel management study of the 3 MW TRIGA Mark II research reactor. Ann. Nucl. Energy 35, 141e147. JEFF Team, 2014. JEFF-3.2: evaluated nuclear data library. https://www.oecd-nea. org/dbforms/data/eva/evatapes/jeff_32. Jeraj, R., et al., 2001. Monte Carlo simulation of the TRIGA Mark II Benchmark Experiment with burned fuel. Nucl. Technol. 137, 169e180. Khan, R., et al., 2010. TRIGA fuel burn-up calculations and its confirmation. Nucl. Eng. Des. 240, 1043e1049. Koning, A.J., et al., 2013. TENDL-2013: TALYS-based evaluated nuclear data library. www.talys.eu/tendl-2013.html. Lyric, Z.I., et al., 2013. Optimum burnup of BAEC TRIGA research reactor. Ann. Nucl. Energy 55, 225e229. MacFarlane, R.E., Kahler, A.C., 2010. Methods for processing ENDF/B-VII with NJOY. Nucl. Data Sheets 111, 2739e2890. MacFarlane, R.E., Muir, D.W., 1994. The NJOY Nuclear Data Processing System, Version 91. Los Alamos National Laboratory report, LA-12740-M. Persi c, A., et al., 1998. Burnup TRIGA Mark II Benchmark Experiment. In: Nuclear Energy in Central Europe'98, Terme Catez, Slovenia, September 7e10. Poston, D.I., Trellue, H.R., 1999. User's Manual, Version 2.0 for MONTEBURNS, Version 5B. LA-UR-99-4999. Ravnik, M., et al., 1999. Fuel element burnup determination in mixed TRIGA core using reactor calculations. Nucl. Technol. 128, 35e45. Shibata, K., et al., 2011. JENDL-4.0: a new library for nuclear science and engineering. J. Nucl. Sci. Technol. 48 (1), 1e30. Türkmen, M., Çolak, Ü., 2013. ITU TRIGA Mark II Research Reactor: a Benchmark analysis with various codes. In: The European Research Reactor Conference: RRFM2013, Saint Petersburg, Russia, April 21e25, pp. 395e400. RRFM2013 Transaction. Türkmen, M., Çolak, Ü., 2014a. Fuel burnup calculation in ITU TRIGA Mark II research reactor by using Monte Carlo method. In: NENE2014: 23rd International Conference Nuclear Energy for New Europe, Portoro z, Slovenia, September 8e11. Türkmen, M., Çolak, Ü., 2014b. Analysis of ITU TRIGA Mark II research reactor using Monte Carlo method. Prog. Nucl. Energy 77, 152e159. X-5 MONTE CARLO TEAM, 2003. MCNP e a General Monte Carlo N-Particle Transport Code, Version 5. LA-CP-03-0245. LANL.