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Assessment of safety aspects of first rod-type fuel irradiation at Tehran research reactor, Part I: Neutronic analysis
Progress in Nuclear Energy 79 (2015) 56e63
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Progress in Nuclear Energy journal homepage: www.elsevier.com/locate/pnucene
Assessment of safety aspects of first rod-type fuel irradiation at Tehran research reactor, Part I: Neutronic analysis S. Safaei Arshi, H. Khalafi*, S.M. Mirvakili Nuclear Science and Technology Research Institute (NSTRI), Tehran 143995-1113, Iran
a r t i c l e i n f o
a b s t r a c t
Article history: Received 5 May 2014 Received in revised form 24 July 2014 Accepted 9 November 2014 Available online
Following domestic fabrication of nuclear fuels in Iran, the need to investigate fuel behavior at different burnup with regard to fission gas release, fuel swelling, cladding material behavior and fuel integrity necessitates in-pile tests irradiation in research reactors. Currently, Tehran research reactor is the sole operating research reactor which can be used for fuel irradiation experiments in the country. This is Part I of the articles describing the details of Tehran research reactor core performance during a fuel irradiation experiment. This first part of the article series is aimed at evaluating the potential capability of Tehran research reactor core to conduct nuclear fuel test during irradiation from the neutronic point of view. A neutronic study on the feasibility of irradiating a rod-type fuel assembly of natural UO2 pellets in the Tehran research reactor core, which itself contains 20% enriched plate-type U3O8eAl fuels, is presented in this paper. The feasibility study is performed by calculating safety related neutronic parameters of the core, i.e., excess reactivity, shutdown margin, reactivity worth of the test fuel, safety reactivity factor and power peaking factors, during irradiation experiment using MCNPX nuclear code. Comparing obtained results to the permissible limits reported in the final safety analysis report of the reactor indicates that this irradiation experiment will not induce any reduction of the reactor safety. © 2014 Elsevier Ltd. All rights reserved.
Keywords: Neutronic study Irradiation experiment Safety Research reactor
1. Introduction Research reactors are valuable means to conduct the fuel irradiation programs for investigating main phenomena affecting the fuel performance during irradiation, namely: fuel thermal behavior, clad corrosion, pelletecladding interaction and fission gas release. A research reactor hosts its own fuel elements which might have different characteristics from the fuel being irradiated to test at the core. The operation of experimental fuel must not induce any significant operational change and any reduction of the reactor safety. The irradiation experiment must comply with safety criteria of the research reactor, enabling the latter to operate properly with its maximum availability. There are many examples of research reactors which have been designed or refurbished to conduct irradiation experiments, i.e., ATR (Advanced Test Reactor) (Marshal, 2005), HALDEN boiling water reactor (HBWR) (IAEA-TECDOC1635, 2009), BR2 (Belgian Reactor 2) (IAEA-TECDOC-1635, 2009), JMTR (Japan Materials Testing Reactor) (Inaba et al., 2011), JHR (Jules Horowitz Reactor) (Bignan, 2011), HANARO (Technical Report * Corresponding author. E-mail address: Hossein_khalafi@yahoo.com (H. Khalafi). http://dx.doi.org/10.1016/j.pnucene.2014.11.005 0149-1970/© 2014 Elsevier Ltd. All rights reserved.
Series No.455, 2007), etc. Tehran Research Reactor (TRR) is a 5 MW pool-type reactor with about 1.0 1014 n/cm2 sec maximum local thermal neutron flux at full power in an irradiation box in the center of the core (Mirvakili et.al., 2012), which can be currently used as an appropriate tool for nuclear fuels and materials irradiation studies in Iran in the absence of any other research reactor with higher neutron flux. A comprehensive feasibility study of irradiating a test fuel in the reactor core as one of the reactor fuel elements necessitates neutronic, thermal hydraulic and accident analysis to ensure the safety of the experiment. This paper deals with the neutronic aspect of this feasibility study, whereas thermal hydraulic and accident aspects will be assessed in a separate paper in the near future. Thus, the main objective of this study is neutronic assessment of making use of the potential capability of TRR core for irradiating newly fabricated fuels as one of the fuel elements of the core to study the behavior of these fuels during irradiation at different burn-up levels. In addition to a comprehensive safety analysis, new facilities for online monitoring of fuel, clad and coolant temperatures, online measurement of fuel dimension changes and clad integrity check to avoid leakage must be installed in the reactor to ensure safety of the irradiation program.
S. Safaei Arshi et al. / Progress in Nuclear Energy 79 (2015) 56e63 Table 1 Main characteristics of TRR core (AEOI, 2009). Parameter
Value
Thermal power Fuel
5 MW 20% enriched U3O8eAl fuel with aluminum cladding 3.1 1013 n/cm2 sec 19 for SFE 14 for CFE SFE: 8.01 7.71 89.7 cm CFE: 8.01 7.71 161.5 cm 290 gr 214 gr Light water 500 m3/h 37.8 C 46 C 0.15 cm 0.07 cm 0.04 cm 0.27 cm 6 cm 61.5 cm Ag: 80% In: 15% Cd: 5% AISI-316L stainless steel
Average thermal neutron flux at 5 MW Number of plates per fuel element Fuel elements dimensions U235 per standard fuel element U235 per control fuel element Moderator and coolant Primary coolant flow rate Coolant inlet temperature in 5 MW Coolant outlet temperature in 5 MW Fuel plate thickness Fuel meat thickness Cladding thickness Water channel thickness Fuel meat width Active height of the fuel plate Safety rods absorber
Regulating rod absorber
TRR is a 5 MW pool-type reactor which can be used for research, training and gaining expertise in nuclear technology. The core lattice is a 9 6 array containing Standard Fuel Elements (SFEs), Control Fuel Elements (CFEs) which host safety and regulating absorber rods, Irradiation boxes and Graphite boxes as reflectors (AEOI, 2009). Main characteristics of TRR core are presented in Table 1. Fuel elements of TRR are plate-type U3O8eAl with 20% enrichment while the test fuel is in the form of natural UO2 ceramic pellets that are about 1.363 cm in diameter and they are stacked in zircaloy tubing to an active length of 0.615 m. In this study, the test fuel is going to be treated as one of the TRR fuels. Thus, it will be cooled by the same cooling mechanism as the one uses for the TRR fuels in which, pool water with total flow rate of 500 m3/hr passes through fuel elements and transfers their heat to the heat exchanger. Fig. 1 shows the flow diagram of the TRR cooling system. Due to the fact that TRR fuel elements and the test fuel differ in geometry and enrichment, a comprehensive investigation of both neutronic and thermal hydraulic aspects is indispensable to ensure safe operation of this mixed core as well as its maximum availability during irradiation experiment. This paper deals with this irradiation experiment from the neutronic point of view. Fig. 2 illustrates the core configuration in which the fuel irradiation experiment will be performed.
2. Methodology In order to investigate the possibility of performing aforementioned fuel irradiation experiment, a comprehensive neutronic analysis is essential to obtain neutronic safety parameters, i.e., excess reactivity, shutdown margin, reactivity worth of the test fuel, safety reactivity factor (SRF) and power peaking factors. All components of the core including SFEs, CFEs, one regulating rod (RR) and four control safety rods (SRs) which are placed within CFEs, rod-type fuel assembly, irradiation and graphite boxes and thermal column are accurately modeled by MCNPX nuclear code (MCNPX, 2008) with ENDF/B-VII library. Fig. 3 presents a schematic of the fuel assembly to be tested in the core. This assembly contains 18 fuel rods and one central tube. Main characteristics of test fuel assembly are presented in Table 2.
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The validity of the model is investigated by two processes. First, benchmarking the obtained first core data against the data reported in the reactor SAR and second, benchmarking the calculated axial distribution of thermal neutron flux of an equilibrium core against experimental data. In the former case, three core states, i.e. Cold and Clean, Hot Zero Power (HZP) and Hot Full Power (HFP) are considered. In the cold and clean state, the core is at room temperature and there is no xenon concentration in the fuel. In HZP and HFP states, the core is at operating temperature. However, in the HFP state, the xenon concentration is in equilibrium at full power, i.e., 5 MW. For HZP and HFP, an MCNP library is constructed for the operating temperature in TRR core using “makxsf” (Brown, 2006) auxiliary processing code. According to SAR, the average temperatures of the fuel and coolant at 5 MW power level for which new cross section library is constructed, are considered to be 65 C and 43 C, respectively. The second benchmarking process is performed by comparing the calculated axial thermal flux distribution in the central irradiation box of an equilibrium core with the data obtained by a flux mapping experiment in that core. The flux mapping experiment is performed by irradiating a copper wire and two gold foils, one cadmium-covered gold foil and one bare gold foil, in the irradiation box located in central position of the core. Natural copper has two isotopes, Cu-63 (69.17%) and Cu65 (30.83%) which produce Cu-64 and Cu-66 when exposed to neutron flux. Cu-64 has a half-life of 12.7 h and is suitable for the purposes of this analysis. Measuring the activation intensities of this activated isotope along the entire wire will result in a data set of neutron flux versus distance from the core bottom. Given the fact that cadmium is a thermal neutron absorber, induced activities of foils are measured by a scintillation counter to obtain the thermal flux. After ensuring the validity of the model, the core configuration, in which irradiation experiment is to be conducted, is modeled to find out whether neutronic parameters will be within safe limits during experiment or not. The effect of irradiating the rod-type fuel assembly in two different locations in the core, i.e., the central and peripheral part of the core, on the safety parameters is also evaluated. In addition, an evaluation of the neutronic parameters in the case of irradiating an assembly of six natural UO2 fuel rods as well as only one rod of this type at the center of the core instead of 18 fuel rods is performed. Achieved burnup in the test fuel assembly during several days is a concern for choosing the best position for irradiating the fuel in the core. Thus, burnup calculation is performed for each case. A comprehensive assessment of the whole parameters of concern provides a basis to find out if this irradiation experiment is possible or not. Core under moderation as an indicator for the inherent safety of the core is evaluated through calculating the variation of core multiplication factor versus moderator temperature. Calculating this parameter involves adjustment of water density and crosssection data over a range of operating temperatures. The crosssection data alters by means of makxsf program (Brown, 2006) to represent temperature changes in the moderator while the moderator density is determined through available steam tables.
2.1. Neutronic analysis In order to analyze the neutronic behavior of the TRR core during fuel irradiation experiment, detailed calculations are conducted to obtain core power and flux distribution, cycle length and operation characteristics, core excess reactivity, shutdown margin and power peaking factors. In addition, the following neutronic safety criteria, which must be met in each core configuration (AEOI, 2009), are investigated:
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Fig. 1. Cooling system flow diagram (AEOI, 2009).
Fig. 2. Core configuration for fuel irradiation experiment.
Fig. 3. A schematic of the fuel assembly to be tested in the core.
S. Safaei Arshi et al. / Progress in Nuclear Energy 79 (2015) 56e63
The maximum allowable power peaking factor for 5 MW operation is equal to 3.0.
Table 2 Main characteristics of test fuel assembly. Parameter Fuel: material Diameter Height Clad: material Outer diameter Inner diameter Height Fuel rod pitch Central tube: material Diameter
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Value UO2 (natural) 11.48 mm 615 mm Zr þ 1% Nb 13.63 mm 11.78 mm 713 mm 16 mm Zr þ 2.5% Nb 15 mm
For any core configuration the shutdown margin in absolute value must be greater than 3000 pcm For any core configuration, reactor must be sub-critical with a shutdown margin of at least 500 pcm, meanwhile the most reactive safety rod is 100% out (stuck rod criteria). Minimum shut down margin in any core configuration is 50% of the reactivity excess i.e. SRF 1.5. SRF is defined as the ratio of integral worth of the control rods to the excess reactivity. Maximum reactivity worth of regulating rod must be lower than the effective delayed neutron fraction (beff).
According to the reactor OLC (Operational Limits and Conditions) (AEOI, 2010), for a total excess reactivity of about 700e1000 pcm at the end of cycle, the core configuration is renewed. Therefore, the amount of excess reactivity of each core configuration must be large enough to compensate fission product poisoning and fuel burnup during the cycle. At first sight, in order to reach the higher burnup within a shorter time, it is desirable to irradiate the test fuel at the core center where the thermal neutron flux is maximum. However, the geometry of the test fuel imposed an unfavorable change in the core configuration to provide enough space for it to be irradiated in the center of the core, i.e., removing two assemblies around the test fuel and placing them in the core peripheral region. The effect of this replacement on the flux distribution as well as neutronic safety parameters is investigated in this study. To avoid the aforementioned change in the core configuration, the test fuel is assumed to be irradiated at position A2 of Fig. 2. Furthermore, irradiating reduced number of natural UO2 fuel rods at the center of the core, i.e., six rods and only one rod, is evaluated to avoid the aforementioned problem and to provide more space for installing some instrumentations on the fuel. In order to investigate the safety aspects pertinent to the first irradiation experiment at TRR, neutronic safety parameters of all aforementioned configurations
Fig. 4. MCNP whole core model for TRR core configuration when the test fuel is located at the core peripheral region including burnup of fuel assemblies.
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Table 3 Results of benchmarking MCNP model for the first TRR core configuration. Core state
Cold & clean HZP HFP
Control rod positions (% extraction)
Reactivity (pcm)
SR1
SR2
SR3
SR4
RR
MCNP results
SAR data
Relative difference (%)
100 0 100 100 100
100 0 100 100 100
100 0 100 100 100
100 0 100 100 100
100 100 0 100 100
6954 12841 6456 6673 3333
6916 12541 6364 6549 3319
0.55 2.39 1.45 1.89 0.42
assemblies from the first operating core to the present core is calculated using MCNPX nuclear code to find out the number densities of nuclides in each fuel assembly before the irradiation experiment begins. In this calculation, the period of time at which the reactor is shut down after its operating cycle is also considered. This period is especially important in the case of taking account the change of fission product poisoning content after shutdown. Excess reactivity is calculated by calculating the reactivity of the core if all the control rods are fully withdrawn. For shutdown margin, the amount of subcriticality when all control rods are fully
Fig. 5. Benchmarking calculated thermal neutron flux distribution against experimental results.
are calculated using MCNPX nuclear code by 10,000 source histories per cycle, 130 active cycle and 30 cycles to be skipped before beginning tally accumulation. These neutronic safety parameters include, excess reactivity, shutdown margin, shutdown margin in stuck rod condition, integral worth of control rods and power peaking factors. Fig. 4 presents MCNP whole core model of TRR core configuration when the test fuel is located at the core peripheral region. Burnup of each fuel assembly at the beginning of cycle for this core configuration is also included in Fig. 4. Burnup of all fuel
inserted is computed. However, for the stuck rod condition, it is assumed that all control rods are fully inserted except for the one with the highest reactivity worth that is assumed to be fully withdrawn. SRF is calculated by dividing the integral worth of control rods by the excess reactivity. Radial power peaking factor is also calculated using MCNPX code in the following way. Fission power is calculated over the volume of the fuel meat in every fuel assembly. The ratio of the maximum integral fission power to the average fission power of all assemblies gives radial power peaking
Fig. 6. Contour of Thermal flux in the core.
S. Safaei Arshi et al. / Progress in Nuclear Energy 79 (2015) 56e63
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Table 4 Neutronic safety parameters of the TRR mixed core. Safety Parameter
Value
Accepted limit
Rod-type fuel at position D6
Excess reactivity (pcm) Shutdown margin (pcm) Shutdown margin in stuck rod condition (pcm) Integral worth of SSRs (pcm) Safety reactivity factor (SRF) Reactivity worth of regulating rod (pcm) Reactivity worth of the rod-type fuel (pcm) Maximum power peaking factor
Rod-type fuel at position A2
18 Rods
6 Rods
1 Rod
1189.68 9724.92 5202.25 10,914.59 9.17 327.18 1142 2.383
4277.82 7199.52 3194.91 11,477.34 2.68 355.92 890 2.106
3823.03 7824.85 3428.66 11,647.88 3.05 342.54 396 2.114
e >3000 >500 e >1.5
3641.39 7862.06 3461.83 11,503.46 3.16 267.55 109 2.100
Table 5 Core multiplication factor and burnup of different numbers of rod-type fuel during 10 days of irradiation in the center of the core. Time (days)
0.0 2.0 4.0 6.0 8.0 10.0
Rod type fuel assembly with 18 rods
Rod type fuel assembly with 6 rods
One rod
keff ± s
Burnup (MWd/T)
keff ± s
Burnup (MWd/T)
keff ± s
0.000Eþ00 1.930Eþ01 3.884Eþ01 5.871Eþ01 7.821Eþ01 9.792Eþ01
1.04306 1.01528 1.01496 1.01321 1.01529 1.01350
0.000Eþ00 2.928Eþ01 5.835Eþ01 8.708Eþ01 1.162Eþ02 1.449Eþ02
1.04001 1.01174 1.01001 1.00958 1.01014 1.00816
1.01337 0.98542 0.98390 0.98331 0.98325 0.98350
± ± ± ± ± ±
0.00071 0.00070 0.00070 0.00069 0.00073 0.00069
± ± ± ± ± ±
factor of each assembly. Thereafter, radial power peaking factor of each fuel plate in the assembly with maximum integral fission power, which is called “hot assembly”, is computed. 3. Results and discussion In this study, the results of a comprehensive neutronic assessment of the first irradiation of a rod-type fuel assembly in the TRR core are presented. The results of benchmarking MCNP model for the first TRR core are presented in Table 3. Extreme compatibility of MCNP results and SAR data for the first core configuration proves the validity of the MCNP model. Comparison between cold & clean state and HZP state shows that constructing a cross-section library for the operating temperature of the core will have an effect of about 281 pcm on the results. Besides, xenon equilibrium will reduce the reactivity by amount of 3340 pcm in the hot state. The results of benchmarking MCNP model for an equilibrium core against the flux mapping experiment results are depicted in Fig. 5. The calculated trend of thermal flux and the position of its maximum are the same as the results of experiment. In this calculation, core multiplication factor (keff) for the critical position of the control rods is about 0.99983 with an estimated standard deviation of 0.00068. After ensuring the validity of the model, flux calculation is performed to find out the thermal neutron flux throughout the core Table 6 Multiplication factor and test fuel burnup during 10 days of irradiating the test fuel in the peripheral region of the core. Time (days)
keff ± s
0.0 2.0 4.0 6.0 8.0 10.0
1.03837 1.01086 1.00953 1.00993 1.00845 1.00725
Burnup (MWd/T) ± ± ± ± ± ±
0.00074 0.00075 0.00063 0.00069 0.00071 0.00068
0.000Eþ00 5.732Eþ00 1.156Eþ01 1.751Eþ01 2.329Eþ01 2.923Eþ01
0.00075 0.00075 0.00077 0.00070 0.00076 0.00063
Burnup (MWd/T) ± ± ± ± ± ±
0.00065 0.00071 0.00086 0.00068 0.00070 0.00069
0.000Eþ00 4.065Eþ01 8.163Eþ01 1.227Eþ02 1.647Eþ02 2.070Eþ02
in which irradiation experiment will be conducted. In this study, neutrons with energies less than 1 eV are considered to be thermal, neutrons with energies within 1 eV to 1 keV range are considered to be epithermal and the neutrons with energies from 1 keV up to 20 MeV are considered to be fast. The contour of thermal flux distribution in the top view of the core and the surrounding coolant before loading the rod-type fuel is depicted in Fig. 6. As expected, maximum thermal flux is observed in the irradiation box located in position D6. The average thermal flux in this position is about 8.78 1013 (n/cm2 s), while the value of the point wise maximum thermal flux is about 1.55 1014 (n/cm2 s) in it. Neutronic safety parameters of the present mixed core for both case of irradiating the rod-type fuel at the core center (position D6 of the core) and at the core periphery (position A2 instead of a graphite box) are calculated and the results are presented in Table 4. In the former case, three different configuration of rod-type fuel, namely, an assembly of 18 fuel rods, an assembly of 6 fuel rods and one single fuel rod are evaluated. As mentioned before, in order to provide enough space to irradiate the rod-type fuel at the place of the maximum flux in the center of the core, fuel elements in C6 and E6 must be transferred elsewhere. In this study, it is decided to replace them with the irradiation boxes in positions A3 and F3, respectively. This substitution causes the excess reactivity to be so small that it cannot compensate fission product poisoning and fuel burnup even during a very short cycle. This fact is illustrated by the results of burnup calculation during 10 days of irradiating the rod-type fuel in the center of the core in Table 5. Calculation show that even loading two fresh SFEs instead of irradiation boxes in E9 and A9 cannot compensate the lack of excess reactivity in this case. In order to avoid the aforementioned problem, it is decided to reduce the number of natural UO2 fuel rods which must be irradiated in the core. Thus, an assembly of six natural UO2 fuel rods as well as a single natural UO2 fuel rod is considered to be irradiated in the center of the core without any necessity to change the position of the fuel elements surrounding it. This causes the higher excess reactivity in comparison to the case of irradiating an assembly of 18
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Fig. 7. Variation in
135
Xe concentration following startup.
fuel rods in the position D6 of the core. The neutronic safety parameters of the core for these two cases are also included in Table 4. Results of burnup calculation for these two cases are presented in Tables 5 and 6. To be able to irradiate the assembly of 18 natural UO2 fuel rods without any change in the core configuration or any reduction in the number of the fuel rods and also to mitigate any probable failure in this fuel, irradiation of the aforementioned fuel at position A2 is investigated. According to flux calculation, thermal flux in position A2 is about 2.47 1013 (n/cm2 s). The variation of multiplication factor and the amount of rod-type fuel burnup during several days of irradiation at position A2 are presented in Table 6. As expected, higher test fuel burnup will be achieved in the case of irradiating fuel at the core center but the small core excess reactivity is problematic in this case. The results of burnup calculation using MCNPX code shows that in this case after about 2 days keff will significantly be lower than the criticality point, which is not the case in irradiating fuel in the peripheral region of the core. The large amount of reactivity decrease during the first 2 days of irradiation experiment reveals the effect of xenon buildup. During steady state operation at a constant neutron flux level, the 135Xe concentration builds up to its equilibrium value in about 40e50 h. However, there is a slight change in reactivity after this period of time. This phenomenon is depicted in Fig. 7. 135Xe equilibrium value (X∞) can be calculated by the following formula Lamarsh (1975):
X∞ ¼
ðgI þ gX ÞSf FT lx þ saX FT
Fig. 9. Radial power peaking factor of fuel plates in the hot assembly.
beff ¼ 1
kp k
(2)
In which, kp is effective multiplication factor for prompt neutrons and k is the total effective multiplication factor. The value of beff for the case of irradiating 18, 6 and 1 UO2 fuel rods in position D6 is about 650, 850 and 819 pcm, respectively. For irradiating an assembly of eighteen UO2 fuel rods in position A2, beff is equal to 548 pcm. Therefore, in all aforementioned cases reactivity worth of regulating rod is lower than beff. According to the reactor SAR, the value of axial power peaking factor for the core with rods fully withdrawn is considered to be
(1)
In which, gI and gX are effective yield (atoms per fission) of 135I P and 135Xe from thermal fission in 235U, respectively. f is thermal fission cross section, FT is thermal flux, saX is absorption cross section of 135Xe and lx is b decay constant of 135Xe. Effective delayed neutron fraction (beff) of each case is calculated using MCNPX code via the following formula (Michalek, 2008):
Fig. 10. Calculated effective multiplication factor as a function of moderator temperature.
Fig. 8. Radial power peaking factor of fuel assemblies when the test fuel is irradiated at position A2.
S. Safaei Arshi et al. / Progress in Nuclear Energy 79 (2015) 56e63
1.315. In addition, it is assumed that for the worst situations with rods partially extracted the radial peaking factors increase about 15%. The multiplication of these two factors by the calculated values of radial power peaking factor for all assemblies in the core results in total power peaking factors. The maximum total power peaking factor associated with each studied case of irradiation experiment is presented in Table 4. Although neutronic safety parameters of all cases satisfy the permissible limits, except for the low excess reactivity of the core when an assembly of eighteen UO2 fuel rods is irradiated in position D6, irradiating the rod-type assembly in position A2 is chosen for the experiment. The radial power peaking factors of all assemblies in this case are presented in Fig. 8. Radial power peaking factors of the hot assembly and the rod-type fuel assembly are indicated in this figure by red and green color (in web version), respectively. Fig. 9 illustrates the layout of the hot assembly, which is an SFE, and the ratio of the power of its 19 fuel plates to the average power in it. As can be seen in this figure, power distribution in the fuel plates of an assembly is almost uniform. The result of calculating effective multiplication factor corresponding to different water temperatures for the case of the core configuration with test fuel at the core peripheral region is illustrated in Fig. 10. It indicates that increasing the water temperature and consequently decreasing water density reduces the effective multiplication factor. The calculated average moderator temperature coefficient of reactivity is 7.93 pcm/K and this negative value guarantees core under-moderation. The inherent uncertainty of the MCNP model is illustrated by the error bar for each calculated effective multiplication factor in this figure. According to available steam tables, under the average operating pressure of the core, namely, about 0.176 MPa, 389 K is the maximum temperature at which the water is in liquid phase. Thus, 389 K is considered as the upper limit of the temperatures considered in evaluating moderator temperature coefficient of reactivity. 4. Conclusion In this study, a neutronic study on the feasibility of irradiating a rod-type fuel assembly of natural UO2 pellets in a research reactor core containing 20% enriched plate-type fuels was conducted. In order to ensure that the irradiation experiment complies with neutronic safety criteria of the research reactor, all components of the core were accurately modeled by MCNPX nuclear code and neutronic safety parameters, i.e., shutdown margin, core excess reactivity, Safety Reactivity Factor and power peaking factor were calculated for all considered cases, namely, irradiating 18, 6 and 1 rod-type fuels at the center of the core and irradiating the assembly of 18 rod-type fuel rods at the core peripheral region. Comparing
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the results with permissible limits reported in reactor SAR shows that in all cases the core safety criteria are satisfied and the fuel irradiation experimental dose not induce any significant operational change. However, small amount of excess reactivity was problematic in the case of irradiating 18 fuel rods in the center of the core. In order to demonstrate the inherent safety of the core during irradiation experiment, variation of core multiplication factor versus the moderator temperature was studied. The result proves the core under-moderation and consequently, negative feedback mechanism during any transient which results in moderator temperature increase. To sum up, from static neutronic assessment point of view, TRR core can be used for fuel irradiation studies in addition to its normal operation and activities to produce radioisotopes. However, to provide a comprehensive feasibility study of this irradiation experiment, thermal hydraulic and accident analysis are indispensable and will be discussed in subsequent papers in the near future. In addition, some refurbishments to develop new online monitoring capabilities like, fuel, clad and coolant temperatures monitoring systems, online measurement of fuel dimension changes and clad integrity check to avoid leakage, must be implemented to provide required conditions for fuel irradiation examination in TRR core.
References AEOI, 2009. Final Safety Analysis Report of Tehran Research Reactor. AEOI, 2010. Operational Limits and Conditions (OLC's) of Tehran Research Reactor. Bignan, Gilles, 2011. The Jules Horowitz reactor: a new high performances European material testing reactor (MTR) as an international center of excellence-update status and focus on the modern safety approach. In: IAEA Conference on Research Reactors eRabat, November 2011. Brown, Forrest B., 2006. The Makxsf Code with Doppler Broadening. Los Alamos National Laboratory. IAEA-TECDOC -1635, 2009. Post-irradiation Examination and In-pile Measurement Techniques for Water Reactor Fuels. International Atomic Energy Agency, Vienna. Inaba, Y., et al., 2011. Present status of refurbishment and irradiation technologies in JMTR. J. Nucl. Mater. 41, 1348e1351. Lamarsh, John R., 1975. Introduction to Nuclear Reactor Theory. Addison-Wesley Pub. Co. Marshal, Frances M., 2005. Advanced test reactor testing experience e past, present, and future. In: Proceedings of the International Symposium on Research Reactor and Neutron Science e in Commemoration of the 10th Anniversary of HANARO- Daejeon, Korea, April 2005. MCNPX User’s Manual, Version 2.6.0, April 2008, Editor: Denise B. Pelowitz. Michalek, Svetozar, 2008. MCNP5 delayed neutron fraction (beff) calculation in training reactor vr-1. J. Electr. Eng. 59, 221e224. Mirvakili, S.M., et al., 2012. Possibility evaluation of eliminating the saturated control fuel element from Tehran research reactor core. Nucl. Eng. Des. 248, 197e205. Technical Report Series No. 455, 2007. Utilization Related Design Features of Research Reactors: a Compendium. International Atomic Energy Agency, Vienna.
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Progress in Nuclear Energy journal homepage: www.elsevier.com/locate/pnucene
Assessment of safety aspects of first rod-type fuel irradiation at Tehran research reactor, Part I: Neutronic analysis S. Safaei Arshi, H. Khalafi*, S.M. Mirvakili Nuclear Science and Technology Research Institute (NSTRI), Tehran 143995-1113, Iran
a r t i c l e i n f o
a b s t r a c t
Article history: Received 5 May 2014 Received in revised form 24 July 2014 Accepted 9 November 2014 Available online
Following domestic fabrication of nuclear fuels in Iran, the need to investigate fuel behavior at different burnup with regard to fission gas release, fuel swelling, cladding material behavior and fuel integrity necessitates in-pile tests irradiation in research reactors. Currently, Tehran research reactor is the sole operating research reactor which can be used for fuel irradiation experiments in the country. This is Part I of the articles describing the details of Tehran research reactor core performance during a fuel irradiation experiment. This first part of the article series is aimed at evaluating the potential capability of Tehran research reactor core to conduct nuclear fuel test during irradiation from the neutronic point of view. A neutronic study on the feasibility of irradiating a rod-type fuel assembly of natural UO2 pellets in the Tehran research reactor core, which itself contains 20% enriched plate-type U3O8eAl fuels, is presented in this paper. The feasibility study is performed by calculating safety related neutronic parameters of the core, i.e., excess reactivity, shutdown margin, reactivity worth of the test fuel, safety reactivity factor and power peaking factors, during irradiation experiment using MCNPX nuclear code. Comparing obtained results to the permissible limits reported in the final safety analysis report of the reactor indicates that this irradiation experiment will not induce any reduction of the reactor safety. © 2014 Elsevier Ltd. All rights reserved.
Keywords: Neutronic study Irradiation experiment Safety Research reactor
1. Introduction Research reactors are valuable means to conduct the fuel irradiation programs for investigating main phenomena affecting the fuel performance during irradiation, namely: fuel thermal behavior, clad corrosion, pelletecladding interaction and fission gas release. A research reactor hosts its own fuel elements which might have different characteristics from the fuel being irradiated to test at the core. The operation of experimental fuel must not induce any significant operational change and any reduction of the reactor safety. The irradiation experiment must comply with safety criteria of the research reactor, enabling the latter to operate properly with its maximum availability. There are many examples of research reactors which have been designed or refurbished to conduct irradiation experiments, i.e., ATR (Advanced Test Reactor) (Marshal, 2005), HALDEN boiling water reactor (HBWR) (IAEA-TECDOC1635, 2009), BR2 (Belgian Reactor 2) (IAEA-TECDOC-1635, 2009), JMTR (Japan Materials Testing Reactor) (Inaba et al., 2011), JHR (Jules Horowitz Reactor) (Bignan, 2011), HANARO (Technical Report * Corresponding author. E-mail address: Hossein_khalafi@yahoo.com (H. Khalafi). http://dx.doi.org/10.1016/j.pnucene.2014.11.005 0149-1970/© 2014 Elsevier Ltd. All rights reserved.
Series No.455, 2007), etc. Tehran Research Reactor (TRR) is a 5 MW pool-type reactor with about 1.0 1014 n/cm2 sec maximum local thermal neutron flux at full power in an irradiation box in the center of the core (Mirvakili et.al., 2012), which can be currently used as an appropriate tool for nuclear fuels and materials irradiation studies in Iran in the absence of any other research reactor with higher neutron flux. A comprehensive feasibility study of irradiating a test fuel in the reactor core as one of the reactor fuel elements necessitates neutronic, thermal hydraulic and accident analysis to ensure the safety of the experiment. This paper deals with the neutronic aspect of this feasibility study, whereas thermal hydraulic and accident aspects will be assessed in a separate paper in the near future. Thus, the main objective of this study is neutronic assessment of making use of the potential capability of TRR core for irradiating newly fabricated fuels as one of the fuel elements of the core to study the behavior of these fuels during irradiation at different burn-up levels. In addition to a comprehensive safety analysis, new facilities for online monitoring of fuel, clad and coolant temperatures, online measurement of fuel dimension changes and clad integrity check to avoid leakage must be installed in the reactor to ensure safety of the irradiation program.
S. Safaei Arshi et al. / Progress in Nuclear Energy 79 (2015) 56e63 Table 1 Main characteristics of TRR core (AEOI, 2009). Parameter
Value
Thermal power Fuel
5 MW 20% enriched U3O8eAl fuel with aluminum cladding 3.1 1013 n/cm2 sec 19 for SFE 14 for CFE SFE: 8.01 7.71 89.7 cm CFE: 8.01 7.71 161.5 cm 290 gr 214 gr Light water 500 m3/h 37.8 C 46 C 0.15 cm 0.07 cm 0.04 cm 0.27 cm 6 cm 61.5 cm Ag: 80% In: 15% Cd: 5% AISI-316L stainless steel
Average thermal neutron flux at 5 MW Number of plates per fuel element Fuel elements dimensions U235 per standard fuel element U235 per control fuel element Moderator and coolant Primary coolant flow rate Coolant inlet temperature in 5 MW Coolant outlet temperature in 5 MW Fuel plate thickness Fuel meat thickness Cladding thickness Water channel thickness Fuel meat width Active height of the fuel plate Safety rods absorber
Regulating rod absorber
TRR is a 5 MW pool-type reactor which can be used for research, training and gaining expertise in nuclear technology. The core lattice is a 9 6 array containing Standard Fuel Elements (SFEs), Control Fuel Elements (CFEs) which host safety and regulating absorber rods, Irradiation boxes and Graphite boxes as reflectors (AEOI, 2009). Main characteristics of TRR core are presented in Table 1. Fuel elements of TRR are plate-type U3O8eAl with 20% enrichment while the test fuel is in the form of natural UO2 ceramic pellets that are about 1.363 cm in diameter and they are stacked in zircaloy tubing to an active length of 0.615 m. In this study, the test fuel is going to be treated as one of the TRR fuels. Thus, it will be cooled by the same cooling mechanism as the one uses for the TRR fuels in which, pool water with total flow rate of 500 m3/hr passes through fuel elements and transfers their heat to the heat exchanger. Fig. 1 shows the flow diagram of the TRR cooling system. Due to the fact that TRR fuel elements and the test fuel differ in geometry and enrichment, a comprehensive investigation of both neutronic and thermal hydraulic aspects is indispensable to ensure safe operation of this mixed core as well as its maximum availability during irradiation experiment. This paper deals with this irradiation experiment from the neutronic point of view. Fig. 2 illustrates the core configuration in which the fuel irradiation experiment will be performed.
2. Methodology In order to investigate the possibility of performing aforementioned fuel irradiation experiment, a comprehensive neutronic analysis is essential to obtain neutronic safety parameters, i.e., excess reactivity, shutdown margin, reactivity worth of the test fuel, safety reactivity factor (SRF) and power peaking factors. All components of the core including SFEs, CFEs, one regulating rod (RR) and four control safety rods (SRs) which are placed within CFEs, rod-type fuel assembly, irradiation and graphite boxes and thermal column are accurately modeled by MCNPX nuclear code (MCNPX, 2008) with ENDF/B-VII library. Fig. 3 presents a schematic of the fuel assembly to be tested in the core. This assembly contains 18 fuel rods and one central tube. Main characteristics of test fuel assembly are presented in Table 2.
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The validity of the model is investigated by two processes. First, benchmarking the obtained first core data against the data reported in the reactor SAR and second, benchmarking the calculated axial distribution of thermal neutron flux of an equilibrium core against experimental data. In the former case, three core states, i.e. Cold and Clean, Hot Zero Power (HZP) and Hot Full Power (HFP) are considered. In the cold and clean state, the core is at room temperature and there is no xenon concentration in the fuel. In HZP and HFP states, the core is at operating temperature. However, in the HFP state, the xenon concentration is in equilibrium at full power, i.e., 5 MW. For HZP and HFP, an MCNP library is constructed for the operating temperature in TRR core using “makxsf” (Brown, 2006) auxiliary processing code. According to SAR, the average temperatures of the fuel and coolant at 5 MW power level for which new cross section library is constructed, are considered to be 65 C and 43 C, respectively. The second benchmarking process is performed by comparing the calculated axial thermal flux distribution in the central irradiation box of an equilibrium core with the data obtained by a flux mapping experiment in that core. The flux mapping experiment is performed by irradiating a copper wire and two gold foils, one cadmium-covered gold foil and one bare gold foil, in the irradiation box located in central position of the core. Natural copper has two isotopes, Cu-63 (69.17%) and Cu65 (30.83%) which produce Cu-64 and Cu-66 when exposed to neutron flux. Cu-64 has a half-life of 12.7 h and is suitable for the purposes of this analysis. Measuring the activation intensities of this activated isotope along the entire wire will result in a data set of neutron flux versus distance from the core bottom. Given the fact that cadmium is a thermal neutron absorber, induced activities of foils are measured by a scintillation counter to obtain the thermal flux. After ensuring the validity of the model, the core configuration, in which irradiation experiment is to be conducted, is modeled to find out whether neutronic parameters will be within safe limits during experiment or not. The effect of irradiating the rod-type fuel assembly in two different locations in the core, i.e., the central and peripheral part of the core, on the safety parameters is also evaluated. In addition, an evaluation of the neutronic parameters in the case of irradiating an assembly of six natural UO2 fuel rods as well as only one rod of this type at the center of the core instead of 18 fuel rods is performed. Achieved burnup in the test fuel assembly during several days is a concern for choosing the best position for irradiating the fuel in the core. Thus, burnup calculation is performed for each case. A comprehensive assessment of the whole parameters of concern provides a basis to find out if this irradiation experiment is possible or not. Core under moderation as an indicator for the inherent safety of the core is evaluated through calculating the variation of core multiplication factor versus moderator temperature. Calculating this parameter involves adjustment of water density and crosssection data over a range of operating temperatures. The crosssection data alters by means of makxsf program (Brown, 2006) to represent temperature changes in the moderator while the moderator density is determined through available steam tables.
2.1. Neutronic analysis In order to analyze the neutronic behavior of the TRR core during fuel irradiation experiment, detailed calculations are conducted to obtain core power and flux distribution, cycle length and operation characteristics, core excess reactivity, shutdown margin and power peaking factors. In addition, the following neutronic safety criteria, which must be met in each core configuration (AEOI, 2009), are investigated:
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Fig. 1. Cooling system flow diagram (AEOI, 2009).
Fig. 2. Core configuration for fuel irradiation experiment.
Fig. 3. A schematic of the fuel assembly to be tested in the core.
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The maximum allowable power peaking factor for 5 MW operation is equal to 3.0.
Table 2 Main characteristics of test fuel assembly. Parameter Fuel: material Diameter Height Clad: material Outer diameter Inner diameter Height Fuel rod pitch Central tube: material Diameter
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Value UO2 (natural) 11.48 mm 615 mm Zr þ 1% Nb 13.63 mm 11.78 mm 713 mm 16 mm Zr þ 2.5% Nb 15 mm
For any core configuration the shutdown margin in absolute value must be greater than 3000 pcm For any core configuration, reactor must be sub-critical with a shutdown margin of at least 500 pcm, meanwhile the most reactive safety rod is 100% out (stuck rod criteria). Minimum shut down margin in any core configuration is 50% of the reactivity excess i.e. SRF 1.5. SRF is defined as the ratio of integral worth of the control rods to the excess reactivity. Maximum reactivity worth of regulating rod must be lower than the effective delayed neutron fraction (beff).
According to the reactor OLC (Operational Limits and Conditions) (AEOI, 2010), for a total excess reactivity of about 700e1000 pcm at the end of cycle, the core configuration is renewed. Therefore, the amount of excess reactivity of each core configuration must be large enough to compensate fission product poisoning and fuel burnup during the cycle. At first sight, in order to reach the higher burnup within a shorter time, it is desirable to irradiate the test fuel at the core center where the thermal neutron flux is maximum. However, the geometry of the test fuel imposed an unfavorable change in the core configuration to provide enough space for it to be irradiated in the center of the core, i.e., removing two assemblies around the test fuel and placing them in the core peripheral region. The effect of this replacement on the flux distribution as well as neutronic safety parameters is investigated in this study. To avoid the aforementioned change in the core configuration, the test fuel is assumed to be irradiated at position A2 of Fig. 2. Furthermore, irradiating reduced number of natural UO2 fuel rods at the center of the core, i.e., six rods and only one rod, is evaluated to avoid the aforementioned problem and to provide more space for installing some instrumentations on the fuel. In order to investigate the safety aspects pertinent to the first irradiation experiment at TRR, neutronic safety parameters of all aforementioned configurations
Fig. 4. MCNP whole core model for TRR core configuration when the test fuel is located at the core peripheral region including burnup of fuel assemblies.
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Table 3 Results of benchmarking MCNP model for the first TRR core configuration. Core state
Cold & clean HZP HFP
Control rod positions (% extraction)
Reactivity (pcm)
SR1
SR2
SR3
SR4
RR
MCNP results
SAR data
Relative difference (%)
100 0 100 100 100
100 0 100 100 100
100 0 100 100 100
100 0 100 100 100
100 100 0 100 100
6954 12841 6456 6673 3333
6916 12541 6364 6549 3319
0.55 2.39 1.45 1.89 0.42
assemblies from the first operating core to the present core is calculated using MCNPX nuclear code to find out the number densities of nuclides in each fuel assembly before the irradiation experiment begins. In this calculation, the period of time at which the reactor is shut down after its operating cycle is also considered. This period is especially important in the case of taking account the change of fission product poisoning content after shutdown. Excess reactivity is calculated by calculating the reactivity of the core if all the control rods are fully withdrawn. For shutdown margin, the amount of subcriticality when all control rods are fully
Fig. 5. Benchmarking calculated thermal neutron flux distribution against experimental results.
are calculated using MCNPX nuclear code by 10,000 source histories per cycle, 130 active cycle and 30 cycles to be skipped before beginning tally accumulation. These neutronic safety parameters include, excess reactivity, shutdown margin, shutdown margin in stuck rod condition, integral worth of control rods and power peaking factors. Fig. 4 presents MCNP whole core model of TRR core configuration when the test fuel is located at the core peripheral region. Burnup of each fuel assembly at the beginning of cycle for this core configuration is also included in Fig. 4. Burnup of all fuel
inserted is computed. However, for the stuck rod condition, it is assumed that all control rods are fully inserted except for the one with the highest reactivity worth that is assumed to be fully withdrawn. SRF is calculated by dividing the integral worth of control rods by the excess reactivity. Radial power peaking factor is also calculated using MCNPX code in the following way. Fission power is calculated over the volume of the fuel meat in every fuel assembly. The ratio of the maximum integral fission power to the average fission power of all assemblies gives radial power peaking
Fig. 6. Contour of Thermal flux in the core.
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Table 4 Neutronic safety parameters of the TRR mixed core. Safety Parameter
Value
Accepted limit
Rod-type fuel at position D6
Excess reactivity (pcm) Shutdown margin (pcm) Shutdown margin in stuck rod condition (pcm) Integral worth of SSRs (pcm) Safety reactivity factor (SRF) Reactivity worth of regulating rod (pcm) Reactivity worth of the rod-type fuel (pcm) Maximum power peaking factor
Rod-type fuel at position A2
18 Rods
6 Rods
1 Rod
1189.68 9724.92 5202.25 10,914.59 9.17 327.18 1142 2.383
4277.82 7199.52 3194.91 11,477.34 2.68 355.92 890 2.106
3823.03 7824.85 3428.66 11,647.88 3.05 342.54 396 2.114
e >3000 >500 e >1.5
3641.39 7862.06 3461.83 11,503.46 3.16 267.55 109 2.100
Table 5 Core multiplication factor and burnup of different numbers of rod-type fuel during 10 days of irradiation in the center of the core. Time (days)
0.0 2.0 4.0 6.0 8.0 10.0
Rod type fuel assembly with 18 rods
Rod type fuel assembly with 6 rods
One rod
keff ± s
Burnup (MWd/T)
keff ± s
Burnup (MWd/T)
keff ± s
0.000Eþ00 1.930Eþ01 3.884Eþ01 5.871Eþ01 7.821Eþ01 9.792Eþ01
1.04306 1.01528 1.01496 1.01321 1.01529 1.01350
0.000Eþ00 2.928Eþ01 5.835Eþ01 8.708Eþ01 1.162Eþ02 1.449Eþ02
1.04001 1.01174 1.01001 1.00958 1.01014 1.00816
1.01337 0.98542 0.98390 0.98331 0.98325 0.98350
± ± ± ± ± ±
0.00071 0.00070 0.00070 0.00069 0.00073 0.00069
± ± ± ± ± ±
factor of each assembly. Thereafter, radial power peaking factor of each fuel plate in the assembly with maximum integral fission power, which is called “hot assembly”, is computed. 3. Results and discussion In this study, the results of a comprehensive neutronic assessment of the first irradiation of a rod-type fuel assembly in the TRR core are presented. The results of benchmarking MCNP model for the first TRR core are presented in Table 3. Extreme compatibility of MCNP results and SAR data for the first core configuration proves the validity of the MCNP model. Comparison between cold & clean state and HZP state shows that constructing a cross-section library for the operating temperature of the core will have an effect of about 281 pcm on the results. Besides, xenon equilibrium will reduce the reactivity by amount of 3340 pcm in the hot state. The results of benchmarking MCNP model for an equilibrium core against the flux mapping experiment results are depicted in Fig. 5. The calculated trend of thermal flux and the position of its maximum are the same as the results of experiment. In this calculation, core multiplication factor (keff) for the critical position of the control rods is about 0.99983 with an estimated standard deviation of 0.00068. After ensuring the validity of the model, flux calculation is performed to find out the thermal neutron flux throughout the core Table 6 Multiplication factor and test fuel burnup during 10 days of irradiating the test fuel in the peripheral region of the core. Time (days)
keff ± s
0.0 2.0 4.0 6.0 8.0 10.0
1.03837 1.01086 1.00953 1.00993 1.00845 1.00725
Burnup (MWd/T) ± ± ± ± ± ±
0.00074 0.00075 0.00063 0.00069 0.00071 0.00068
0.000Eþ00 5.732Eþ00 1.156Eþ01 1.751Eþ01 2.329Eþ01 2.923Eþ01
0.00075 0.00075 0.00077 0.00070 0.00076 0.00063
Burnup (MWd/T) ± ± ± ± ± ±
0.00065 0.00071 0.00086 0.00068 0.00070 0.00069
0.000Eþ00 4.065Eþ01 8.163Eþ01 1.227Eþ02 1.647Eþ02 2.070Eþ02
in which irradiation experiment will be conducted. In this study, neutrons with energies less than 1 eV are considered to be thermal, neutrons with energies within 1 eV to 1 keV range are considered to be epithermal and the neutrons with energies from 1 keV up to 20 MeV are considered to be fast. The contour of thermal flux distribution in the top view of the core and the surrounding coolant before loading the rod-type fuel is depicted in Fig. 6. As expected, maximum thermal flux is observed in the irradiation box located in position D6. The average thermal flux in this position is about 8.78 1013 (n/cm2 s), while the value of the point wise maximum thermal flux is about 1.55 1014 (n/cm2 s) in it. Neutronic safety parameters of the present mixed core for both case of irradiating the rod-type fuel at the core center (position D6 of the core) and at the core periphery (position A2 instead of a graphite box) are calculated and the results are presented in Table 4. In the former case, three different configuration of rod-type fuel, namely, an assembly of 18 fuel rods, an assembly of 6 fuel rods and one single fuel rod are evaluated. As mentioned before, in order to provide enough space to irradiate the rod-type fuel at the place of the maximum flux in the center of the core, fuel elements in C6 and E6 must be transferred elsewhere. In this study, it is decided to replace them with the irradiation boxes in positions A3 and F3, respectively. This substitution causes the excess reactivity to be so small that it cannot compensate fission product poisoning and fuel burnup even during a very short cycle. This fact is illustrated by the results of burnup calculation during 10 days of irradiating the rod-type fuel in the center of the core in Table 5. Calculation show that even loading two fresh SFEs instead of irradiation boxes in E9 and A9 cannot compensate the lack of excess reactivity in this case. In order to avoid the aforementioned problem, it is decided to reduce the number of natural UO2 fuel rods which must be irradiated in the core. Thus, an assembly of six natural UO2 fuel rods as well as a single natural UO2 fuel rod is considered to be irradiated in the center of the core without any necessity to change the position of the fuel elements surrounding it. This causes the higher excess reactivity in comparison to the case of irradiating an assembly of 18
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Fig. 7. Variation in
135
Xe concentration following startup.
fuel rods in the position D6 of the core. The neutronic safety parameters of the core for these two cases are also included in Table 4. Results of burnup calculation for these two cases are presented in Tables 5 and 6. To be able to irradiate the assembly of 18 natural UO2 fuel rods without any change in the core configuration or any reduction in the number of the fuel rods and also to mitigate any probable failure in this fuel, irradiation of the aforementioned fuel at position A2 is investigated. According to flux calculation, thermal flux in position A2 is about 2.47 1013 (n/cm2 s). The variation of multiplication factor and the amount of rod-type fuel burnup during several days of irradiation at position A2 are presented in Table 6. As expected, higher test fuel burnup will be achieved in the case of irradiating fuel at the core center but the small core excess reactivity is problematic in this case. The results of burnup calculation using MCNPX code shows that in this case after about 2 days keff will significantly be lower than the criticality point, which is not the case in irradiating fuel in the peripheral region of the core. The large amount of reactivity decrease during the first 2 days of irradiation experiment reveals the effect of xenon buildup. During steady state operation at a constant neutron flux level, the 135Xe concentration builds up to its equilibrium value in about 40e50 h. However, there is a slight change in reactivity after this period of time. This phenomenon is depicted in Fig. 7. 135Xe equilibrium value (X∞) can be calculated by the following formula Lamarsh (1975):
X∞ ¼
ðgI þ gX ÞSf FT lx þ saX FT
Fig. 9. Radial power peaking factor of fuel plates in the hot assembly.
beff ¼ 1
kp k
(2)
In which, kp is effective multiplication factor for prompt neutrons and k is the total effective multiplication factor. The value of beff for the case of irradiating 18, 6 and 1 UO2 fuel rods in position D6 is about 650, 850 and 819 pcm, respectively. For irradiating an assembly of eighteen UO2 fuel rods in position A2, beff is equal to 548 pcm. Therefore, in all aforementioned cases reactivity worth of regulating rod is lower than beff. According to the reactor SAR, the value of axial power peaking factor for the core with rods fully withdrawn is considered to be
(1)
In which, gI and gX are effective yield (atoms per fission) of 135I P and 135Xe from thermal fission in 235U, respectively. f is thermal fission cross section, FT is thermal flux, saX is absorption cross section of 135Xe and lx is b decay constant of 135Xe. Effective delayed neutron fraction (beff) of each case is calculated using MCNPX code via the following formula (Michalek, 2008):
Fig. 10. Calculated effective multiplication factor as a function of moderator temperature.
Fig. 8. Radial power peaking factor of fuel assemblies when the test fuel is irradiated at position A2.
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1.315. In addition, it is assumed that for the worst situations with rods partially extracted the radial peaking factors increase about 15%. The multiplication of these two factors by the calculated values of radial power peaking factor for all assemblies in the core results in total power peaking factors. The maximum total power peaking factor associated with each studied case of irradiation experiment is presented in Table 4. Although neutronic safety parameters of all cases satisfy the permissible limits, except for the low excess reactivity of the core when an assembly of eighteen UO2 fuel rods is irradiated in position D6, irradiating the rod-type assembly in position A2 is chosen for the experiment. The radial power peaking factors of all assemblies in this case are presented in Fig. 8. Radial power peaking factors of the hot assembly and the rod-type fuel assembly are indicated in this figure by red and green color (in web version), respectively. Fig. 9 illustrates the layout of the hot assembly, which is an SFE, and the ratio of the power of its 19 fuel plates to the average power in it. As can be seen in this figure, power distribution in the fuel plates of an assembly is almost uniform. The result of calculating effective multiplication factor corresponding to different water temperatures for the case of the core configuration with test fuel at the core peripheral region is illustrated in Fig. 10. It indicates that increasing the water temperature and consequently decreasing water density reduces the effective multiplication factor. The calculated average moderator temperature coefficient of reactivity is 7.93 pcm/K and this negative value guarantees core under-moderation. The inherent uncertainty of the MCNP model is illustrated by the error bar for each calculated effective multiplication factor in this figure. According to available steam tables, under the average operating pressure of the core, namely, about 0.176 MPa, 389 K is the maximum temperature at which the water is in liquid phase. Thus, 389 K is considered as the upper limit of the temperatures considered in evaluating moderator temperature coefficient of reactivity. 4. Conclusion In this study, a neutronic study on the feasibility of irradiating a rod-type fuel assembly of natural UO2 pellets in a research reactor core containing 20% enriched plate-type fuels was conducted. In order to ensure that the irradiation experiment complies with neutronic safety criteria of the research reactor, all components of the core were accurately modeled by MCNPX nuclear code and neutronic safety parameters, i.e., shutdown margin, core excess reactivity, Safety Reactivity Factor and power peaking factor were calculated for all considered cases, namely, irradiating 18, 6 and 1 rod-type fuels at the center of the core and irradiating the assembly of 18 rod-type fuel rods at the core peripheral region. Comparing
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the results with permissible limits reported in reactor SAR shows that in all cases the core safety criteria are satisfied and the fuel irradiation experimental dose not induce any significant operational change. However, small amount of excess reactivity was problematic in the case of irradiating 18 fuel rods in the center of the core. In order to demonstrate the inherent safety of the core during irradiation experiment, variation of core multiplication factor versus the moderator temperature was studied. The result proves the core under-moderation and consequently, negative feedback mechanism during any transient which results in moderator temperature increase. To sum up, from static neutronic assessment point of view, TRR core can be used for fuel irradiation studies in addition to its normal operation and activities to produce radioisotopes. However, to provide a comprehensive feasibility study of this irradiation experiment, thermal hydraulic and accident analysis are indispensable and will be discussed in subsequent papers in the near future. In addition, some refurbishments to develop new online monitoring capabilities like, fuel, clad and coolant temperatures monitoring systems, online measurement of fuel dimension changes and clad integrity check to avoid leakage, must be implemented to provide required conditions for fuel irradiation examination in TRR core.
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