Neutronic analysis for Tehran Research Reactor mixed-core

Progress in Nuclear Energy 60 (2012) 31e37 Contents lists available at SciVerse ScienceDirect Progress in Nuclear Energy journal homepage: www.elsevier.com/locate/pnucene Neutronic analysis for Tehran Research Reactor mixed-core Ahmad Lashkari a, Hossein Khalafi a, *, S. Mohammad Mirvakili a, b, Shokufe Forughi a a b Nuclear Science and Technology Research Institute (NSTRI), Atomic Energy Organization of Iran (AEOI), North Kargar Ave., Tehran 14399-51113, Iran Department of Nuclear Engineering, School of Mechanical Engineering, Shiraz University, 71348-51154 Shiraz, Iran a r t i c l e i n f o a b s t r a c t Article history: Received 2 October 2011 Received in revised form 17 April 2012 Accepted 23 April 2012 In this research, neutronic calculation of current low enriched uranium control fuel elements replacement with high enriched uranium control fuel elements in the reference core of Tehran Research Reactor (TRR) has been investigated and the results of calculations are compared with the TRR neutronic safety criteria. Results show that all neutronic parameters of the reference and each mixed-core are lower than the safety criteria. Nuclear reactor analysis codes including MTR_PC package and MCNP5 were employed to carry out these calculations. Ó 2012 Elsevier Ltd. All rights reserved. Keywords: Mixed-core Neutronic parameters HEU LEU Safety criteria 1. Introduction Tehran Research Reactor (TRR) became critical using Highly Enriched Uranium (HEU) that was more than 90% enriched in 235U, in 1967. In later years, based on the International Atomic Energy Agency Non Proliferation Treaty (IAEA-NPT), the new fuel with Low Enriched Uranium (LEU), containing less than 20% enrichment in 235 U, was used. According to the history of this reactor, unlike other research reactors, it has not passed the mixed-core period, in which the previous HEU fuel is gradually replaced by new LEU fuel. According to the relatively small consumption of HEU fuel that its maximum burn-up is 20%, designing of a mixed-core for TRR is a perfect solution for optimal use of this kind of fuel. In the present situation, regarding the TRR Control Fuel Elements (CFE) are approaching the range of permissible burn-up, and some of the Shim Safety Rods (SSR) are close to saturation conditions, using old control rods (Oval type) along with Highly Enriched Uraniume Control Fuel Elements (HEU_CFE) can be one of options to continuing TRR activities. In this research, neutronic calculation of current LEU_CFE replacement with HEU_CFE in the reference core has been investigated and the neutronic calculations are carried out in the direction of fuel management calculations to obtain neutronic parameters of the core under operational circumstances and * Corresponding author. Tel.: þ98 21 82064265; fax: þ98 21 88221219. E-mail address: hkhalafi@aeoi.org.ir (H. Khalafi). 0149-1970/$ e see front matter Ó 2012 Elsevier Ltd. All rights reserved. doi:10.1016/j.pnucene.2012.04.006 comparing them with the reactor safe operation criteria (Faghihi and Mirvakili, 2009). Firstly, all neutronic parameters of equilibrium core and then the mixed-core configurations have been studied and the results of each mixed-core were compared to the reference core. Results show that all neutronic parameters of the reference and each mixed-core are lower than the safety criteria. 2. Description of TRR TRR is an open pool type, light water moderated with the thermal power of 5 MW (Zaker, 2003). Its core configuration contains MTR-type fuel elements that are arranged in 9  6 grid plate assembly. In TRR fork-type control rods, i.e. four shim safety rods and one regulating rod are responsible for reactivity safe control in the core (Khalafi et al., 2011). The cross-sectional view of LEU Standard Fuel Elements (SFE) and CFE are given in Fig. 1a and b (AEOI, 1989). Other details of LEU and HEU fuel assemblies and core parameters are given in TRR-Safety Analysis Reports (SAR) (AEOI, 1966, 2001). CFE specifications of TRR for HEU and LEU are compared in Table 1. HEUeCFE is different in composition and dimension respect to LEUeCFE. Main differences are the number of fuel plates, lateral walls and the shape of absorber. Comparison between oval- and fork-type absorbers, follows that considering the same absorber material and fuel enrichment, fork-type control rods are more effective than oval by a ratio ranging from 1.25 to 1.36 (AEOI, 2001). Minimum and maximum average burn-up of HEU_CFE is about 7 and 20 percent. 32 A. Lashkari et al. / Progress in Nuclear Energy 60 (2012) 31e37 Fig. 1. The cross-sectional view of the TRR fuels: (a) SFE, (b) CFE (all dimensions in cm) (T.R.R: AMENDMENT, SEPT/1989). The LEU-core configuration of the reference core and burn-up of the fuel elements (in percent of the initial value of 235U) at beginning of cycle is given in Fig. 2. 3. Methodology 3.1. Simulation methodology The MTR_PC package has been developed by INVAP S.E in order to perform neutronic, thermal hydraulic and shielding calculations of MTR-type reactor for personal computers (PC). In this research, WIMS-D4 (Taubman and Lawrence, 1980), POS_WIMS, HXS and CITVAP (Villarino and Lecot, 1988) neutronic codes of MTR_PC package are used to calculate neutronic-core parameters of TRR and TRR mixed-core. Table 1 Specifications of HEU and LEU control fuel elements. Parameters LEU Meat material Enrichment Number of fuel plates No of outer dummy plates Total fuel plate thickness Meat thickness Cladding thickness Water channel thickness Meat width Total plate width Meat length Side wall thickness Length of side plates Inner distance between side wall CFE dimension FP cladding and side walls material Absorber type Absorber material for shim safety rods Absorber material for fine regulating rod Material in the gap between cladding and absorber Guide plates material Cladding material for absorber plate Uranium per fuel plate Weight Of U-235 per fuel assembly Density of total uranium in meat Total density of meat Density of U-235 in meat UeAl alloy U3O8eAl 20% 93.15% 14 8 None 2 0.15 cm 0.127 0.07 cm 0.0508 0.04 cm 0.0381 0.27 cm 0.310 cm 6 cm 6.1 6.700 cm 6.644 cm 61.5 cm 59.99 0.45 cm 0.476 8.010 cm 7.950 cm 6.7 cm 6.61 8.1  7.7  61.5 cm 8  7.61  87.31 cm Al-6061 Al-6061 Fork Oval AgeIneCd AgeIneCd HEU AISI-316L ss AISI-316L ss He, at 1 atm. None Al-6061 AISI-316L 15.26 g 213.7 g 3.0 g/cc 4.8 g/cc 0.591 g/cc Al-6061 AISI-316L 12.2 g 97.6 0.69951 g/cc 3.16367 g/cc 0.6516 cc WIMSD is employed for macroscopic cross-section generation, which provides nuclear cross-sections in the form of a 69-energy group structure. POS_WIMS is a post-processor program of WIMS code used to condenses and homogenizes macroscopic crosssection from WIMS output. CITVAP code is a new version of the CITATION-II code. It solves 1, 2 or 3-dimensional multi-group diffusion equation in rectangular or cylindrical geometries. HXS program (Handle Cross-section) makes the connection between cell calculation and core calculation. Core calculations are performed with the CITVAP diffusion code, in XeYeZ geometry, using the three group energy structure according to Table 2.This energy structure agrees with the 5-45-69 partition of the 69 groups WIMS library. WIMSD code was run with applying DSN and PERSEUS options to carry out required macroscopic cross-section for different states (cold-clean, hot-zero and hot-full power) in each zone. Due to the proximity of DSN answers to the reference, SEQUENCE1 was used in WIMS calculations. For the accuracy of reliability results, neutronic parameters of some core configuration performed with MCNP code and the results are compared with CITVAP code. The MCNP5 code (X-5 Monte Carlo Team, 2003) and ENDF-VII library (Chadwick et al., 2006) are applied for calculating the neutronic parameters of the new core. 3.1.1. Planar modeling of SFE and CFE for WIMS and CITVAP To model SFE, Homogeneous Model (HOM) was used. Evaluation of macroscopic cross-section for Standard Fuel ElementeHomogeneous Model (SFE_HOM) was performed with the WIMS code in slab geometry. SFE_HOM model is based on the simplified drawing in Fig. 1a. The geometrical model and associated WIMS cell are given in Fig. 3a and b. This model preserves actual meat thickness but aluminum and water thickness was calculated according to the total aluminum and water existing in fuel element. To model CFE, control fuel element was divided into four different material regions as shown in Fig. 4a. Cross-section calculation for each region was performed as follows: - Control Fuel ElementeFuel (CFE_FUE): This region is a homogenization of all the 14 fuel plates in their active width. Crosssection evaluation for this region was performed using WIMS code in slab geometry which WIMS cell geometry is given in Fig. 4b. - Control Fuel ElementeFrame (CFE_FRA): This material is a homogenization of all regions external to the active width of 6 cm Macroscopic cross-sections for this material were 33 A. Lashkari et al. / Progress in Nuclear Energy 60 (2012) 31e37 9 E.B GR GR GR E.B GR 8 SFE 0.93 CFE-RR 18.40 SFE 16.04 CFE-SR2 47.81 SFE 11.78 SFE 5.08 7 SFE 20.14 SFE 31.46 SFE 11.81 SFE 48.62 SFE 39.21 SFE 12.88 6 SFE 7.89 CFE-SR1 59.54 SFE 45.05 E.B CFE-SR3 38.06 SFE 8.48 5 SFE 31.56 SFE 35.99 SFE 49.92 SFE 40.02 SFE 31.40 SFE 5.02 4 SFE 17.25 SFE 22.48 CFE-SR4 54.77 SFE 55.66 SFE 22.15 E.B 3 SFE 3.19 SFE 10.83 SFE 24.89 E.B SFE 0.0 GR 2 GR GR E.B GR GR GR 1 GR GR GR GR GR GR A B C D E F SFE: STANDARD FUEL ELEMENT GR-BOX: GRAPHITE BOX SR: SHIM SAFETY ROD CFE: CONTROL FUEL ELEMENT IR-BOX: IRRADIATION BOX RR: REGULATING ROD Fig. 2. TRR core configuration. calculated by mixing (linear combination) the Al-6061 and WATER, as the clad and coolant adjacent to the fuel plate (Fig. 3b), cross-sections with weighing factor of 0.61 and 0.39 respectively. These weighting factors are the volume fractions of each material in this region. - Control Fuel ElementeShim Safety Rod (CFE_SSR), Control Fuel Element-Fine RegulatingeRod (CFE_FRR), or Control Fuel ElementeAbsorber Out (CFE_AOU): This region is a homogenization of all material zones inside the guide plates, in the active width of 6 cm. - Control Fuel ElementeGuide Plate (CFE_GPL): This zone is a homogenization of all water and aluminum that are within the active width of 6 cm but that is out of to the region that includes exactly the 14 fuel plates with their associated water channels. 3.1.2. Modeling of core in the axial direction For considering reflector role of water, a layer of about 30 cm is considered above the active fuel region also grid plate layer of about 15 cm is added under fuel region. Table 2 Energy groups used for macroscopic cross-section generation by WIMSD (T.R.R: AMENDMENT, SEPT/1989). Energy group Energy range Remarks 1 2 3 10.00e0.821 Mev 0.821e0.625 Mev 0.625e0.000 eV Fast Epithermal Thermal 3.2. Validation of simulation methodology In order to validate the simulation methodology, the first core of TRR was simulated. This core contains 14 SFE, 5 CFE and water as reflector. The core configuration and specification are given in the reference documents (Zaker, 2003). The benchmark refers to the first configuration of TRR that allows reactor operation at maximum power level (5 MW). The neutronic parameters for the first operating core were calculated and compared with the value of SAR parameters which summarized in Table 3. Comparing the results presented in Table 3, shows that, the average percentage error with respect to SAR parameters is below 5%. Table 4 shows the position of control rods in critical point for both experimental and calculated situation in the first operating core. Therefore the validity of the results is confirmed. 3.2.1. The control rods effect The control rods effect on the Peaking Factor (PF) is evaluated and the result is shown in Fig. 5. The results show 12% increase in PF relative to the situation in which all control rods are out. Maximum variation of PF is 15% which is the same as value reported in TRReSAR. In order to perform conservative analysis the value of 15% is applied for the maximum variation of PF. 3.3. Mixed-core calculations 3.3.1. Mixed-core modeling Mixed-core configurations are made only by replacing HEUeCFE with LEUeCFE. These fuel elements are made of 93% enriched 34 A. Lashkari et al. / Progress in Nuclear Energy 60 (2012) 31e37 Fig. 3. Standard fuel element: (a) homogeneous model, (b) unit cell of WIMSD. Fig. 4. (a) Planar model of LEU_CFE, (b) unit cell of WIMSD for fuel region of CFE. Table 3 Neutronic parameters for the first operating core. Parameters CITVAP MCNP SAR %Error Excess reactivitya (pcm) Tempereactivity (5 MW) Xenon equilibrium (5 MW) Total worth of safety rods Worth of RR Shut Down Margin (pcm) Peaking factor Radial Axial Total peaking factor (T.P.F) Reactivity safety factor (R.S.F) Excess reactivity at 5 MW 7460 390 3370 20,990 536 13,380 6393 e e 22,856 450 16,463 6916 370 3150 19,457 550 12,500 7.8 5 2.9 8 2.5 7 1.65 1.3 2.2 2.79 3700 1.7 1.29 2.19 3.47 e 1.66 1.3 2.1 to 2.7 2.81 3319 0 a 0.7 11 The unit of reactivity parameters in all tables is in term of pcm, pcm ¼ (DK/ K)  105. uranium aluminum alloy (AEOI, 1966). The HEUeCFE model is based on the simplified drawing in Fig. 6a. The oval-type absorber whose dimensions are shown in Fig. 6b is fitted in the middle part of CFE. For core calculations, the HEUeCFE was divided into five different regions as shown in Fig. 7. Calculation of each region is exactly similar to the methods that provided in the LEU control fuel element except for dummy plate zone. 3.3.1.1. Dummy Plate (DPL). The latest plate of HEU_CFE does not contain fuel meat. Macroscopic cross-sections for this region are calculated by mixing (linear combination) the Al-6061 and WATER Table 4 Position of control rods in critical point. Critical point SR1 SR2 SR3 SR4 RR Excess reactivity experimental calculated 42 42 55 55 55 55 50 50 42 42 0 166 pcm Total peaking factor fluctuation A. Lashkari et al. / Progress in Nuclear Energy 60 (2012) 31e37 35 of WIMS unit cell for fuel region of HEU_CFE (Fig. 7b) cross-sections with weighing factor of 0.60 and 0.30, respectively. 14 12 10 8 6 4 2 0 -2 0 20 40 60 80 -4 -6 Rods position(% in) Fig. 5. The control rods effect on the peaking factor. 100 3.3.2. Mixed-core configurations Core configuration 51 includes 27 SFE and 4 CFE. The percentage of each fuel element is shown in Fig. 2. At the first mixed-core configuration (Mixed Core-1) one LEUeCFE with maximum burnup (59.54%) replaced with one HEUeCFE 7%. In the next step, two LEUeCFE with 54.77 and 59.54 burn-up are replaced with two 7% burn-up HEUeCFE (Mixed Core-2). If in mixed-core configuration 2 LEUeCFE with maximum burn-up 38.06% replaced with HEUeCFE 20%, Mixed-Core 3 has been made (Mixed Core-3). Finally all LEUeCFE are replaced with HEUeCFE as Mixed Core-4 while the LEUeCFE 18.40% was replaced by HEUeCFE 20% and LEUeCFE 47.81% by HEUeCFE 20%. All neutronic parameters of Fig. 6. The cross-sectional view of: (a) HEUeCFE, (b) HEUeCFE absorber of TRR (all dimensions in cm). Fig. 7. (a) Planar model of HEU_CFE, (b) unit cell of WIMSD for fuel region of CFE (all dimensions in cm). 36 A. Lashkari et al. / Progress in Nuclear Energy 60 (2012) 31e37 Table 5 The results of Mixed Core-1. Parameter Keff SR-OUT SR-IN Excess reactivity Shutdown margin Total worth of safety rod S.R.F Worth of RR Peaking factor Radial Axial Total PPF with rods effect CITVAP MCNP Safety criteria (SAR) Mix.Core Ref.Core Mix.Core Ref.Core 1.02887 0.918695 6176 8850 15,026 2.43 370 1.02755 0.90906 6050 10,000 16,050 2.65 370 1.05429 0.95525 5149 4685 9834 1.91 230 1.05082 0.951882 4836 5055 9891 2.04 310 >1.5 <beff 1.6 1.3 2.39 1.43 1.30 2.14 1.55 1.27 2.27 1.65 1.24 2.38 <3 <6916 (without >3000 135 Xe effect) Table 6 The results of Mixed Core-2. Parameter Keff SR-OUT SR-IN Excess reactivity Shutdown margin Total worth of safety rod S.R.F Worth of RR Peaking factor Radial Axial Total P.F with rods effect CITVAP MCNP Safety criteria (SAR) Mix.Core Ref.Core Mix.Core Ref.Core 1.03059 0.93038 6338 7483.4 13,821 2.18 366 1.02755 0.90906 6050 10,000 16,050 2.65 370 1.052388 0.971015 4978 2985 7963 1.60 160 1.05082 0.951882 4836 5055 9891 2.04 310 >1.5 <beff 1.65 1.30 2.47 1.433 1.30 2.14 1.59 1.27 2.32 1.65 1.24 2.38 <3 these configurations have been calculated and compared with core configuration 51 and neutronic safety criteria. <6916 (without >3000 135 Xe effect) increase in excess reactivity and some changes in peaking factor, other parameters do not change significantly. As a result this mixedcore configuration meets all neutronics safety parameters. 4. Results and discussions 4.2. Mixed-core configuration 2 The neutronic calculations of HEU_CFE substitution for the current Low Enrichment UraniumeControl Fuel Elements (LEU_CFE) in a reference core have been investigated and the results of calculations are compared with the TRR Neutronic Safety Criteria. For reliability of the accuracy of the results, neutronic parameters of two mixed-cores configuration have been calculated with two codes MCNP and CITVAP and the results are shown in one table but in the other cases only CITVAP is used. 4.1. Mixed-core configuration 1 In the first mixed-core configuration, one LEUeCFE with maximum burn-up 59.54% is replaced with one HEUeCFE 7%. All neutronic parameters of this configuration have been calculated and compared with neutronic parameters of reference core in Table 5. CITVAP and MCNP results show that excess reactivity increase due to increasing the amount of 235U and the worth of shim and regulating rods decreased that the result is a reduction in Safety Reactivity Factor (S.R.F) and shutdown margin. But in Power Peaking Factor (PPF) CITVAP only shows increase in radial peaking factor but MCNP shows reduction in radial and increase in axial peaking factor. In addition the results show that the total PPF by considering the control rod effect is less than 3. Notable difference in the results is that, unlike CITVAP total peaking factor, in MCNP less than reference core. Results show that although replacing one CFE causes a small Two LEUeCFE with 54.77 and 59.54% burn-up are replaced with two 7% burn-up HEUeCFE. The results of neutronic parameters for this configuration are showed in Table 6. In this mixed-core, the results are repeated with the difference that the increase or reduction is become more intense. In this case the S.R.F is less than mixed-core 1, because the decreasing of the total worth of safety rods is much more than the increasing of excess reactivity in the Table 7 Mixed Core-3. Parameter Keff SR-OUT SR-IN Excess reactivity Shutdown margin Total worth of safety rod S.R.F Worth of RR Peaking factor Radial Axial Total P.F with rod effect CITVAP Safety criteria (SAR) Mix.Core Ref.Core 1.02700 0.93464 6000 1.02755 0.90906 6050 6993.2 12,993 2.16 366 10000 16,050 2.65 370 >1.5 <beff 1.66 1.30 2.48 1.43 1.30 2.14 <3 <6916 (without >3000 135 Xe effect) A. Lashkari et al. / Progress in Nuclear Energy 60 (2012) 31e37 5. Conclusions Table 8 Mixed Core-4. Parameter Keff SR-OUT SR-IN Excess reactivity (pcm) Shutdown margin Total worth of safety rod (pcm) S.R.F Worth of RR Peaking factor Radial Axial Total P.F with rod effect 37 CITVAP Safety criteria (SAR) Mix.Core Ref. Core 1.02615 0.93521 5918 1.02755 0.90906 6050 6927.1 12,845 10,000 16,050 <6916 (without 135Xe effect) >3000 pcm 2.17 320 2.65 370 >1.5 <beff 1.68 1.29 2.49 1.433 1.307 2.14 <3 core. Also in this configuration, axial peaking factor does not change but radial peaking factor changes a little. This study has two main results: - The calculation methodology for neutronic analysis of TRR mixed-core has been validated by performing neutronic calculations of the first TRR operating core. On the basis of this methodology, results for the first core have a good agreement with SAR and experimental results. - In the main part of this study, the possibility of LEUeCFE replacement with HEUeCFE was studied and the changes in results were compared with the reference core. Accordingly, all mixed-core configurations with one up to all CFE replacement were analyzed. The results in the mixed-core configurations show all neutronic parameters are lower than the Safety Criteria and are far from safety margins. Investigation analysis of the results shows that increasing the number of HEUeCFE, reduce the shutdown margin and worth of RR and on other hand increase radial peaking factor. References 4.3. Mixed-core configuration 3 The results of Table 7 show that substituting of HEU 20% instead of LEU 38.06% does not change the safety neutronic criteria. The amount of 235U in the LEU 38% is equal to the HEU 20% 235U in LEUeCFE. Therefore, the excess reactivity of the new core does not change but the higher efficiency of the oval control rod result in changes in shut down margin. 4.4. Mixed-core configuration 4 The last investigated configuration is the replacement of all LEUeCFE with HEUeCFE. In this case the results of pervious configurations are repeated, but none of the safety criteria is violated. 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