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KAMINI reactor benchmark analysis
Available online at www.sciencedirect.com annals of
NUCLEAR ENERGY Annals of Nuclear Energy 35 (2008) 570–575 www.elsevier.com/locate/anucene
KAMINI reactor benchmark analysis C. Sunil Sunny b
a,*
, D.K. Mohapatra b, P. Mohanakrishnan b, K.V. Subbaiah
a
a Safety Research Institute, Atomic Energy Regulatory Board, Kalpakkam, Tamil Nadu 603102, India Reactor Physics Division, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu 603102, India
Received 5 April 2007; received in revised form 22 August 2007; accepted 22 August 2007 Available online 17 October 2007
Abstract Monte Carlo modeling of the Kalpakkam Mini reactor (KAMINI) had been made earlier by us for the first time using Monte Carlo code (MCNP4A) and ENDF/B-VI.2 data in the year 2004. KAMINI has been accepted as a part of the International Criticality Safety Benchmark Evaluation Project (ICSBEP) recently and its detailed configuration is available for analysis in the International Handbook of Evaluated Criticality Safety Benchmark Experiments, 2006. New neutron cross section data is also now available for nuclides of thorium–uranium fuel cycle through IAEA coordinated research and these are based on ENDF/B-VII. Examination of the new cross section data for nuclides 233U and 27Al, the major constituents in the fuel alloy of KAMINI, shows the addition of more resolved resonance structure in them. Hence, a review of our earlier work is made using new cross section data and by using the improved MCNP model of KAMINI supplied by ICSBEP evaluators. It is observed that by using the new data keff value changed from 0.9890 to 0.99357 and is closer to the experimental value. 2007 Elsevier Ltd. All rights reserved.
1. Introduction The Kalpakkam Mini reactor (KAMINI) is the newest built research reactor in India, which had attained criticality on 29th October 1996. It is a 30 kW tank-in-pool-type reactor, operated with an alloy of 233U and aluminum as fuel, light water as moderator as well as coolant and beryllium oxide (BeO) as reflector. Located at the Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, KAMINI has the unique distinction of being the only operating pool-type reactor in the world at present that employs 233 U as fuel. The reactor is used regularly for neutron radiography of both radioactive and non-radioactive materials, activation analysis and radiation physics studies. There are three horizontal neutron beam tubes, two thimble irradiation locations and one pneumatic fast transfer facility available in this reactor for research purposes. The three neutron beam tubes emerge out from the core reflector junction and are situated on the west, south and north sides
*
Corresponding author. Tel.: +91 44 27480164; fax: +91 44 27480165. E-mail address: [email protected] (C. Sunil Sunny).
0306-4549/$ - see front matter 2007 Elsevier Ltd. All rights reserved. doi:10.1016/j.anucene.2007.08.016
of the reactor. While the west and south neutron beam tubes are situated at the core mid-height, the north neutron beam tube is situated near the core top. The thimble irradiation locations are located outside the BeO reflectors on the west side of the reactor core. It is possible to irradiate large samples (100 ml), both in solid and liquid form for longer duration at these locations. Since its criticality, several theoretical and experimental studies have been carried out in KAMINI. The criticality calculations and prediction of reactor safety parameters have earlier been carried out in detail by neutron diffusion and transport theory based codes (Mohapatra and Mohanakrishnan, 2000, 2002; Mohapatra et al., 2004a). Similarly, many reactor physics experiments have also been carried out to measure the neutron flux and spectra at various locations of the reactor, reactivity coefficients as well as worth of control plates (Mohapatra et al., 2004a). Since this reactor is very compact and has much heterogeneity with three air filled neutron beam tubes adjacent to its core, a detailed three-dimensional continuous energy Monte Carlo model of KAMINI was developed for the first time using the MCNP4A code (Briesmeister, 1993; Mohapatra et al., 2004b).
C. Sunil Sunny et al. / Annals of Nuclear Energy 35 (2008) 570–575
Recently KAMINI has been accepted as a part of the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and its detailed configuration is available for analysis in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (NEA, 2006). The evaluators have carried out a detailed Monte Carlo modeling of KAMINI by using the MCNP code and have reproduced the MCNP input listings along with the benchmark report (Radha et al., 2006). Continuous energy neutron cross section data based on ENDF/B-VI.8 has been used by the evaluators for calculations. The keff as predicted by the evaluators is 0.9890 ± 0.0003 as against our earlier value of 1.017 ± 0.00092 (Mohapatra et al., 2004b). In order to figure out the reason for this discrepancy, the benchmark analysis of KAMINI is once again carried out, using MCNP input listing from ICSBEP handbook and different neutron cross-section data, the details of which are presented in this paper. 2. Brief design description of KAMINI system The design description of KAMINI has been discussed in many earlier works (Pasupathy et al., 1993; Mohapatra and Mohanakrishnan, 2000, 2002;Mohapatra et al., 2004a,b; Usha et al., 2006). But, a very detailed configuration with minute details of the reactor is very diligently described in the recently published ICSBEP benchmark handbook. However for the clarity of understanding a schematic view of KAMINI reactor core is shown in Fig. 1. KAMINI has a small core volume of nearly 10 l with dimension 207 · 207 · 276 mm. The core of KAMINI reactor comprises nine rectangular fuel subassemblies,
571
arranged in a 3 · 3 array. Because of the highly efficient reflector material BeO, it has a very low fuel inventory of 612 g. The entire core and reflector assembly is immersed in a stainless steel tank containing ordinary demineralised light water. Fuel assembly design is similar to other MTR reactor fuel assemblies. Each fuel subassembly consists of eight fuel plates. The fuel subassembly specifications are mentioned in Table 1. The reactor core is, surrounded by BeO reflectors on all the six sides and their thickness is typically 200 mm each. The system is designed to accommodate three numbers of adjustable reflector blocks (ARB) on the east side, and they can be installed or removed in order to increase or decrease the core reactivity. At present the reactor is being operated with only one ARB (ARB1), which has a dimension of 204 · 104 · 530 mm. There is a water gap between the fuel subassemblies and also between the core and BeO reflector blocks. A spacer type aluminum core cage maintains the uniformity of the water gaps. The reactor power is controlled; by two safety control plates (SCPs), each 300 mm in length, located on the east side of the core. They consist of cadmium plates clad in aluminum moving in a waterfilled zircaloy box. The dimensions of Cd, Al and zircaloy regions are given in Table 2. The west and south neutron beam tubes have their centers located near the core center, which is nearly at an elevation of about 355 mm from the base plate and the north beam tube is above the core top at a height of about 500 mm from the base plate. The beam tubes consist of an inner extension (extending inside the reactor tank) made of 1.3 mm thick nuclear grade zircaloy-2 and an outer extension (extending outside the reactor tank) fabricated
North Beam Tube
North Thimble Location
West Beam Tube
Beo Reflector Block
Core Adjustable Reflector Block
South Thimble Location South Beam Tube
Fig. 1. Schematic view of KAMINI reactor core and thimble locations.
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C. Sunil Sunny et al. / Annals of Nuclear Energy 35 (2008) 570–575
Table 1 KAMINI fuel subassembly (SA) specifications Fuel material Fuel meat dimension (mm) Al clad Fuel plate dimensions (mm) Nominal water gap between plates (mm) Number of fuel plates/SA Fuel subassembly dimensions (mm) Nominal weight of uranium per plate (g) Nominal weight of uranium per SA (g) Number of fuel SAs in core
Al 20 wt.% U alloy 1 · 55 · 250 0.5 mm on each side 2 · 62 · 260 6 8 66 · 66 · 275 8.5 68 9
Table 2 Safety control plates (SCP) dimensions (mm) Zr box outer dimensions Zr box thickness Outer dimensions of Al clad SCPs Dimensions of Cd meat Cd meat length
10.6 · 205.2 1.3 (side of guide channel between SCP and core) 5 · 70 · 505 1.5 · 40 · 300 300
of stainless steel. The zircaloy portions of the beam tubes have a total length of 890 mm out of which nearly 290 mm is of 53 mm outer diameter while the remaining portion has an outer diameter of approximately 110 mm. The stainless steel extensions consist of three tubes having diameters 128, 203 and 260 mm and lengths 108, 420 and 430 mm, respectively, which pass through suitable cavities in the concrete biological shield surrounding the reactor. Each portion of the beam tube is suitably coupled to the next one so that their centerlines coincide. Each beam tube is provided with cadmium collimator to ensure parallel orientation of the neutron beam. Finally each beam tube is provided with a beam shutter made of 250 mm thick lead at its end. This helps in reducing the radiation dose from the core center with the reactor in shut down condition. The beam shutters are moved in or out of position with the help of a motorized drive mechanism. Two hollow aluminum tubes having an inside diameter of 50 mm are provided symmetrically outside the BeO reflector blocks on both sides of the west beam tube, through which the samples can be lowered from the reactor top in an aluminum container for irradiation. The two tubes are about 88 mm away from the reflector surface and about 75 mm away from the west beam tube. These two irradiation locations mentioned in the introduction are called north and south thimble irradiation locations. 3. MCNP modeling of KAMINI: comparison of earlier work with ICSBEP Our earlier modeling (Mohapatra et al., 2004b), of KAMINI was mostly based on the safety assessment report (Pasupathy et al., 1993), which was used as a
standard for all calculations. Thus all the dimensions of various core components, isotopic composition of materials and their geometries used in our MCNP modeling were obtained from the above document. However, in the recent ICSBEP evaluation, the evaluators have attempted to furnish as much technical details as possible, furnishing the minute details on geometries and compositions of materials involved in the reactor system. Detailed evaluations were carried out by performing fresh measurements wherever required to confirm the dimensions, densities and isotopic compositions of some of the important materials present in the core. We delineate below some of the important differences between the MCNP modeling carried out in our earlier work and the recent ICSBEP evaluation. (1) In our earlier work, the core was modeled as a lattice filled with fuel subassemblies, water gaps and core cage. Each fuel plate was modeled as per the dimensions quoted in KAMINI safety assessment report (Pasupathy et al., 1993). The repeated structure capability of MCNP was utilized for filling the lattice with all the fuel subassemblies by which all the fuel plates are treated identically. In case of the ICSBEP evaluation, the evaluators have furnished the information on the weight and isotopic composition on each fuel plate separately as obtained from the fuel fabricator. An average plate mass for each subassembly is obtained by averaging the masses of the individual plates in that subassembly. Variation of the uranium density within the plate or within the subassembly is not considered. Thus in the ICSBEP evaluation the repeated structure capability has been avoided and each fuel subassembly is modeled independently. (2) The density of BeO reflector used in our earlier modeling was taken as 2.76 g/cm3 while in the ICSBEP evaluation it is determined as 2.9318 g/cm3. The densities of water and safety control plate (Cd) and other materials are almost the same in both the calculations. (3) The material impurities present in the fuel alloy and Al clad was considered as quoted in an earlier publication on the KAMINI fuel fabrication experience (Ganguly et al., 1991). In the current evaluation the same reference is used for taking into account the impurities in the fuel and clad. However, in the current evaluation the impurities present in the BeO reflector and zircaloy used as the canning material for the BeO reflectors are also taken into account, which was absent in our earlier modeling. It is realized from the above observations that the recent ICSBEP modeling is an improved modeling as compared to our earlier one with regard to geometry and atomic number densities. Hence, as a first step an investigation is made by computing the keff of the KAMINI core by
C. Sunil Sunny et al. / Annals of Nuclear Energy 35 (2008) 570–575
573
4. Criticality calculation with different cross section sets
Fig. 2. MCNP modeling view of one fuel subassembly.
employing the MCNP model input listing given in ICSBEP handbook but by using the neutron cross section data which was used in our earlier work. This is necessary to find out whether neutron cross section data used in our earlier MCNP model was responsible for the discrepancy in keff values as discussed in Section 1. Fig. 2 depicts the 2D plot of one fuel subassembly along with the water gaps as modeled by MCNP. An overall modeling view of the whole core, west and south beam tubes and thimble locations are shown in Fig. 3.
The criticality calculation or in other words, the calculation of the effective multiplication factor (keff) was performed with the ‘KCODE’ option of the MCNP code. The calculations were performed for 2500 cycles of iteration with 5000 neutrons per cycle. The initial source distribution for the keff calculations were given at the center of the core and first 50 cycles were ignored for the keff calculation. The thermal scattering law data S(a,b) were used for hydrogen in water and for beryllium in BeO to account for the molecular binding effects at low neutron energies. Since the geometry and material data remain the same as that of the ICSBEP evaluation, more stress was imparted to verify the effects of using various cross-section sets. In KAMINI, criticality is achieved by fully raising the SCP1 out of the core and partially raising SCP2. In the present calculation, SCP1 and SCP2 were raised up by 286.8 mm (fully out) and 187.8 mm from the bottom of the core, respectively. The KAMINI core criticality calculations are performed for the following cases and the corresponding keff values obtained are listed in Table 3. Case-1: ICSBEP model but neutron cross-section data (ENDF/B-VI.2) for all materials same as that used in our earlier work, Case-2: same as Case-1 but 233U data of IAEA Th-U cycle research project (Trkov, 2006) based on ENDF/B-VII, Case-3: same as Case-1 but 233U, 234U, data of IAEA ThU cycle research project, Case-4: same as Case-1 but 233U, 234U, 232U, data of IAEA Th-U cycle research project, Case-5: same as Case-4 but 27Al data based on ENDF/BVII; this data is also found to be same as that of IAEA ADS nuclear data library based on JEFF 3.1 (Aldama and Trkov, 2005).
Surrounding Water SBT
Table 3 KAMINI reactor keff values obtained for different neutron cross-section data Cases
BeO Case-1
NT
ST
Case-2 Case-3
WBT
Case-4 Case-5
Fig. 3. MCNP modeling of the whole core and portions of beam tubes inside the tank.
Cross-sections ICSBEP Reported value (ENDF/B-VI.8) ICSBEP MCNP input listing but earlier work CXS data (ENDF/B-VI.2) Same as Case-1 but 233U – (ENDF/B-VII) Same as Case-1 but 233U, 234 U – (ENDF/B-VII) Same as Case-1 but 232U, 233 U, 234U – (ENDF/B-VII) Same as Case-4 but 27Al from IAEA ADS Library (based on JEFF 3.1 which is also same as ENDF/B-VII)
keff(1r interval)
q (mk)
0.9890 ± 0.0003
11.1223
0.98899 ± 0.00025
11.1326
0.99171 ± 0.00025
8.3593
0.99194 ± 0.00024
8.1255
0.99176 ± 0.00025
8.3085
0.99357 ± 0.00025
6.4716
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C. Sunil Sunny et al. / Annals of Nuclear Energy 35 (2008) 570–575
Table 4 Comparison of cross-section (barns) data of energy, 0.025 eV
233
U and
27
Al at neutron
Neutron reaction type
ENDF/B-VI.8
ENDF/B-VII
27
0.23441 534.04 46.2324
0.23583 536.74 45.692
Al Capture 233 U Fission 233 U Capture
Data from ENDF/B-VI.8 and ENDF/B-VII for neutron capture and fission cross-sections of U233 and capture cross-section of Al27 at neutron energy of 0.025 eV are compared in Table 4. 5. Results and discussion The keff value computed for KAMINI reactor by the evaluators using MCNP code and presented in the ICSBEP handbook is 0.9890 ± 0.0003 using ENDF/B-VI.8 crosssection data and it is identical with the keff value obtained
Fig. 4. Comparison of U233 fission cross-section of (a) ENDF/B-VI (b) ENDF/B-VII.
for Case-1 (i.e., using ENDF/B-VI.2 cross-section data in ICSBEP model of KAMINI). This shows that the crosssection data used in our earlier work is similar to that used by the evaluators in ICSBEP benchmark computational model. Hence, it should be believed that the improved details of geometry and atomic densities given in ICSBEP model of KAMINI is the reason for the difference in our earlier prediction of keff = 1.017 compared to the ICSBEP evaluators’ prediction of 0.9890. In the next analysis (i.e., Case-2), the 233U neutron cross-section data alone is changed from ENDF/B-VI.8 to ENDF/B-VII (IAEA Th-U cycle) in the input model by which the keff computed changed from 0.98899 to 0.99171, which is more closer to the experimental value of 1.000 by about 2.7733 mk (see Table 3). It can be inferred from Table 3 that the change of crosssection data from ENDF/B-VI.8 to ENDF/B-VII made for 232 U and 234U (Cases 3 and 4) in the input model did not alter the keff value substantially. This may be because these
Fig. 5. Comparison of Al27 capture cross-section of (a) ENDF/B-VI (b) ENDF/B-VII.
C. Sunil Sunny et al. / Annals of Nuclear Energy 35 (2008) 570–575
isotopes are present only in trace quantities in KAMINI reactor fuel. Since aluminum is a major constituent in the KAMINI reactor fuel alloy, the effect of changing cross-section data of 27Al from ENDF/B-VI.8 to ENDF/B-VII is studied next (Case-5), the results of which showed the keff moving closer to the experimental value of 1.000 by another 1.8369 mk (see Table 3). Hence by changing the cross-section data of 233U, 232U, 234U and 27Al from ENDF/B-VI.8 to ENDF/B-VII the keff value improved from 0.9890 to 0.99357 which is equivalent to a reactivity change of 4.6507 mk. The changes in neutron cross data between ENDF/BVI.8 and ENDF/B-VII of two major constituents of KAMINI reactor fuel, i.e. 233U and 27Al was closely examined especially with respect to the fission cross section of 233 U and capture cross-section of 27Al which are compared in Figs. 4 and 5. It can be seen from Fig. 4 that more resolved resonance structure has been introduced into the 233 U fission cross-section data of ENDF/B-VII beyond neutron energy of 100 eV. From Fig. 5, it can be seen that there is change in the capture cross-section resonance data of 27Al beyond neutron energy of 0.1 MeV. These changes in the resonance data for 233U and 27Al lead to the reactivity changes mentioned above. 6. Conclusions
(a) The MCNP modeling of KAMINI given in ICSBEP benchmark differs from our earlier model in the atomic densities of fuel plates, which lead to a reactivity difference of about 27.8 mk between both the models. By using ENDF/B-VI.2 cross section data we get the same keff as ICSBEP model result quoted in the benchmark report with ENDF/B-VI.8 data. Thus, it is confirmed that there is negligible difference between the use of ENDF/B-VI.8 and ENDF/B-VI.2 data. (b) By employing the new cross section data from ENDF/B-VII for 233U alone in the criticality calculations a reactivity change of 2.7733 mk is obtained and is more closer to the experimental value. Similarly, using 27Al cross-section data of ENDF/B-VII showed a reactivity change of 1.8369 mk, which also tends closer to the experimental value. It is inferred that by changing the cross-section data of 233U, 232U, 234 U and 27Al from ENDF/B-VI.8 to ENDF/B-VII the keff value improved from 0.9890 to 0.99357. (c) It is evident from Table 4 that thermal cross sections are not contributing to the improved results. Examination of the new cross-section data shows the more
575
resolved resonance data in the fission cross-section of 233 U beyond neutron energy of 100 eV as well as change in the capture cross-section resonance data of 27Al beyond neutron energy of 0.1 MeV. These changes improved the computed keff value of KAMINI reactor benchmark.
Acknowledgement Authors would like to acknowledge Dr. S. Ganesan of RPDD, BARC (India) for giving insight to start this work and for fruitful discussions regarding the IAEA coordinated research project on thorium–uranium fuel cycle cross-section data. References Aldama, D.L., Trkov, A., 2005. ADS-Lib/V1.0: A Test Library for Accelerator Driven Systems, INDC (NDS)-0474, IAEA. Available from:
NUCLEAR ENERGY Annals of Nuclear Energy 35 (2008) 570–575 www.elsevier.com/locate/anucene
KAMINI reactor benchmark analysis C. Sunil Sunny b
a,*
, D.K. Mohapatra b, P. Mohanakrishnan b, K.V. Subbaiah
a
a Safety Research Institute, Atomic Energy Regulatory Board, Kalpakkam, Tamil Nadu 603102, India Reactor Physics Division, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu 603102, India
Received 5 April 2007; received in revised form 22 August 2007; accepted 22 August 2007 Available online 17 October 2007
Abstract Monte Carlo modeling of the Kalpakkam Mini reactor (KAMINI) had been made earlier by us for the first time using Monte Carlo code (MCNP4A) and ENDF/B-VI.2 data in the year 2004. KAMINI has been accepted as a part of the International Criticality Safety Benchmark Evaluation Project (ICSBEP) recently and its detailed configuration is available for analysis in the International Handbook of Evaluated Criticality Safety Benchmark Experiments, 2006. New neutron cross section data is also now available for nuclides of thorium–uranium fuel cycle through IAEA coordinated research and these are based on ENDF/B-VII. Examination of the new cross section data for nuclides 233U and 27Al, the major constituents in the fuel alloy of KAMINI, shows the addition of more resolved resonance structure in them. Hence, a review of our earlier work is made using new cross section data and by using the improved MCNP model of KAMINI supplied by ICSBEP evaluators. It is observed that by using the new data keff value changed from 0.9890 to 0.99357 and is closer to the experimental value. 2007 Elsevier Ltd. All rights reserved.
1. Introduction The Kalpakkam Mini reactor (KAMINI) is the newest built research reactor in India, which had attained criticality on 29th October 1996. It is a 30 kW tank-in-pool-type reactor, operated with an alloy of 233U and aluminum as fuel, light water as moderator as well as coolant and beryllium oxide (BeO) as reflector. Located at the Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, KAMINI has the unique distinction of being the only operating pool-type reactor in the world at present that employs 233 U as fuel. The reactor is used regularly for neutron radiography of both radioactive and non-radioactive materials, activation analysis and radiation physics studies. There are three horizontal neutron beam tubes, two thimble irradiation locations and one pneumatic fast transfer facility available in this reactor for research purposes. The three neutron beam tubes emerge out from the core reflector junction and are situated on the west, south and north sides
*
Corresponding author. Tel.: +91 44 27480164; fax: +91 44 27480165. E-mail address: [email protected] (C. Sunil Sunny).
0306-4549/$ - see front matter 2007 Elsevier Ltd. All rights reserved. doi:10.1016/j.anucene.2007.08.016
of the reactor. While the west and south neutron beam tubes are situated at the core mid-height, the north neutron beam tube is situated near the core top. The thimble irradiation locations are located outside the BeO reflectors on the west side of the reactor core. It is possible to irradiate large samples (100 ml), both in solid and liquid form for longer duration at these locations. Since its criticality, several theoretical and experimental studies have been carried out in KAMINI. The criticality calculations and prediction of reactor safety parameters have earlier been carried out in detail by neutron diffusion and transport theory based codes (Mohapatra and Mohanakrishnan, 2000, 2002; Mohapatra et al., 2004a). Similarly, many reactor physics experiments have also been carried out to measure the neutron flux and spectra at various locations of the reactor, reactivity coefficients as well as worth of control plates (Mohapatra et al., 2004a). Since this reactor is very compact and has much heterogeneity with three air filled neutron beam tubes adjacent to its core, a detailed three-dimensional continuous energy Monte Carlo model of KAMINI was developed for the first time using the MCNP4A code (Briesmeister, 1993; Mohapatra et al., 2004b).
C. Sunil Sunny et al. / Annals of Nuclear Energy 35 (2008) 570–575
Recently KAMINI has been accepted as a part of the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and its detailed configuration is available for analysis in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (NEA, 2006). The evaluators have carried out a detailed Monte Carlo modeling of KAMINI by using the MCNP code and have reproduced the MCNP input listings along with the benchmark report (Radha et al., 2006). Continuous energy neutron cross section data based on ENDF/B-VI.8 has been used by the evaluators for calculations. The keff as predicted by the evaluators is 0.9890 ± 0.0003 as against our earlier value of 1.017 ± 0.00092 (Mohapatra et al., 2004b). In order to figure out the reason for this discrepancy, the benchmark analysis of KAMINI is once again carried out, using MCNP input listing from ICSBEP handbook and different neutron cross-section data, the details of which are presented in this paper. 2. Brief design description of KAMINI system The design description of KAMINI has been discussed in many earlier works (Pasupathy et al., 1993; Mohapatra and Mohanakrishnan, 2000, 2002;Mohapatra et al., 2004a,b; Usha et al., 2006). But, a very detailed configuration with minute details of the reactor is very diligently described in the recently published ICSBEP benchmark handbook. However for the clarity of understanding a schematic view of KAMINI reactor core is shown in Fig. 1. KAMINI has a small core volume of nearly 10 l with dimension 207 · 207 · 276 mm. The core of KAMINI reactor comprises nine rectangular fuel subassemblies,
571
arranged in a 3 · 3 array. Because of the highly efficient reflector material BeO, it has a very low fuel inventory of 612 g. The entire core and reflector assembly is immersed in a stainless steel tank containing ordinary demineralised light water. Fuel assembly design is similar to other MTR reactor fuel assemblies. Each fuel subassembly consists of eight fuel plates. The fuel subassembly specifications are mentioned in Table 1. The reactor core is, surrounded by BeO reflectors on all the six sides and their thickness is typically 200 mm each. The system is designed to accommodate three numbers of adjustable reflector blocks (ARB) on the east side, and they can be installed or removed in order to increase or decrease the core reactivity. At present the reactor is being operated with only one ARB (ARB1), which has a dimension of 204 · 104 · 530 mm. There is a water gap between the fuel subassemblies and also between the core and BeO reflector blocks. A spacer type aluminum core cage maintains the uniformity of the water gaps. The reactor power is controlled; by two safety control plates (SCPs), each 300 mm in length, located on the east side of the core. They consist of cadmium plates clad in aluminum moving in a waterfilled zircaloy box. The dimensions of Cd, Al and zircaloy regions are given in Table 2. The west and south neutron beam tubes have their centers located near the core center, which is nearly at an elevation of about 355 mm from the base plate and the north beam tube is above the core top at a height of about 500 mm from the base plate. The beam tubes consist of an inner extension (extending inside the reactor tank) made of 1.3 mm thick nuclear grade zircaloy-2 and an outer extension (extending outside the reactor tank) fabricated
North Beam Tube
North Thimble Location
West Beam Tube
Beo Reflector Block
Core Adjustable Reflector Block
South Thimble Location South Beam Tube
Fig. 1. Schematic view of KAMINI reactor core and thimble locations.
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C. Sunil Sunny et al. / Annals of Nuclear Energy 35 (2008) 570–575
Table 1 KAMINI fuel subassembly (SA) specifications Fuel material Fuel meat dimension (mm) Al clad Fuel plate dimensions (mm) Nominal water gap between plates (mm) Number of fuel plates/SA Fuel subassembly dimensions (mm) Nominal weight of uranium per plate (g) Nominal weight of uranium per SA (g) Number of fuel SAs in core
Al 20 wt.% U alloy 1 · 55 · 250 0.5 mm on each side 2 · 62 · 260 6 8 66 · 66 · 275 8.5 68 9
Table 2 Safety control plates (SCP) dimensions (mm) Zr box outer dimensions Zr box thickness Outer dimensions of Al clad SCPs Dimensions of Cd meat Cd meat length
10.6 · 205.2 1.3 (side of guide channel between SCP and core) 5 · 70 · 505 1.5 · 40 · 300 300
of stainless steel. The zircaloy portions of the beam tubes have a total length of 890 mm out of which nearly 290 mm is of 53 mm outer diameter while the remaining portion has an outer diameter of approximately 110 mm. The stainless steel extensions consist of three tubes having diameters 128, 203 and 260 mm and lengths 108, 420 and 430 mm, respectively, which pass through suitable cavities in the concrete biological shield surrounding the reactor. Each portion of the beam tube is suitably coupled to the next one so that their centerlines coincide. Each beam tube is provided with cadmium collimator to ensure parallel orientation of the neutron beam. Finally each beam tube is provided with a beam shutter made of 250 mm thick lead at its end. This helps in reducing the radiation dose from the core center with the reactor in shut down condition. The beam shutters are moved in or out of position with the help of a motorized drive mechanism. Two hollow aluminum tubes having an inside diameter of 50 mm are provided symmetrically outside the BeO reflector blocks on both sides of the west beam tube, through which the samples can be lowered from the reactor top in an aluminum container for irradiation. The two tubes are about 88 mm away from the reflector surface and about 75 mm away from the west beam tube. These two irradiation locations mentioned in the introduction are called north and south thimble irradiation locations. 3. MCNP modeling of KAMINI: comparison of earlier work with ICSBEP Our earlier modeling (Mohapatra et al., 2004b), of KAMINI was mostly based on the safety assessment report (Pasupathy et al., 1993), which was used as a
standard for all calculations. Thus all the dimensions of various core components, isotopic composition of materials and their geometries used in our MCNP modeling were obtained from the above document. However, in the recent ICSBEP evaluation, the evaluators have attempted to furnish as much technical details as possible, furnishing the minute details on geometries and compositions of materials involved in the reactor system. Detailed evaluations were carried out by performing fresh measurements wherever required to confirm the dimensions, densities and isotopic compositions of some of the important materials present in the core. We delineate below some of the important differences between the MCNP modeling carried out in our earlier work and the recent ICSBEP evaluation. (1) In our earlier work, the core was modeled as a lattice filled with fuel subassemblies, water gaps and core cage. Each fuel plate was modeled as per the dimensions quoted in KAMINI safety assessment report (Pasupathy et al., 1993). The repeated structure capability of MCNP was utilized for filling the lattice with all the fuel subassemblies by which all the fuel plates are treated identically. In case of the ICSBEP evaluation, the evaluators have furnished the information on the weight and isotopic composition on each fuel plate separately as obtained from the fuel fabricator. An average plate mass for each subassembly is obtained by averaging the masses of the individual plates in that subassembly. Variation of the uranium density within the plate or within the subassembly is not considered. Thus in the ICSBEP evaluation the repeated structure capability has been avoided and each fuel subassembly is modeled independently. (2) The density of BeO reflector used in our earlier modeling was taken as 2.76 g/cm3 while in the ICSBEP evaluation it is determined as 2.9318 g/cm3. The densities of water and safety control plate (Cd) and other materials are almost the same in both the calculations. (3) The material impurities present in the fuel alloy and Al clad was considered as quoted in an earlier publication on the KAMINI fuel fabrication experience (Ganguly et al., 1991). In the current evaluation the same reference is used for taking into account the impurities in the fuel and clad. However, in the current evaluation the impurities present in the BeO reflector and zircaloy used as the canning material for the BeO reflectors are also taken into account, which was absent in our earlier modeling. It is realized from the above observations that the recent ICSBEP modeling is an improved modeling as compared to our earlier one with regard to geometry and atomic number densities. Hence, as a first step an investigation is made by computing the keff of the KAMINI core by
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4. Criticality calculation with different cross section sets
Fig. 2. MCNP modeling view of one fuel subassembly.
employing the MCNP model input listing given in ICSBEP handbook but by using the neutron cross section data which was used in our earlier work. This is necessary to find out whether neutron cross section data used in our earlier MCNP model was responsible for the discrepancy in keff values as discussed in Section 1. Fig. 2 depicts the 2D plot of one fuel subassembly along with the water gaps as modeled by MCNP. An overall modeling view of the whole core, west and south beam tubes and thimble locations are shown in Fig. 3.
The criticality calculation or in other words, the calculation of the effective multiplication factor (keff) was performed with the ‘KCODE’ option of the MCNP code. The calculations were performed for 2500 cycles of iteration with 5000 neutrons per cycle. The initial source distribution for the keff calculations were given at the center of the core and first 50 cycles were ignored for the keff calculation. The thermal scattering law data S(a,b) were used for hydrogen in water and for beryllium in BeO to account for the molecular binding effects at low neutron energies. Since the geometry and material data remain the same as that of the ICSBEP evaluation, more stress was imparted to verify the effects of using various cross-section sets. In KAMINI, criticality is achieved by fully raising the SCP1 out of the core and partially raising SCP2. In the present calculation, SCP1 and SCP2 were raised up by 286.8 mm (fully out) and 187.8 mm from the bottom of the core, respectively. The KAMINI core criticality calculations are performed for the following cases and the corresponding keff values obtained are listed in Table 3. Case-1: ICSBEP model but neutron cross-section data (ENDF/B-VI.2) for all materials same as that used in our earlier work, Case-2: same as Case-1 but 233U data of IAEA Th-U cycle research project (Trkov, 2006) based on ENDF/B-VII, Case-3: same as Case-1 but 233U, 234U, data of IAEA ThU cycle research project, Case-4: same as Case-1 but 233U, 234U, 232U, data of IAEA Th-U cycle research project, Case-5: same as Case-4 but 27Al data based on ENDF/BVII; this data is also found to be same as that of IAEA ADS nuclear data library based on JEFF 3.1 (Aldama and Trkov, 2005).
Surrounding Water SBT
Table 3 KAMINI reactor keff values obtained for different neutron cross-section data Cases
BeO Case-1
NT
ST
Case-2 Case-3
WBT
Case-4 Case-5
Fig. 3. MCNP modeling of the whole core and portions of beam tubes inside the tank.
Cross-sections ICSBEP Reported value (ENDF/B-VI.8) ICSBEP MCNP input listing but earlier work CXS data (ENDF/B-VI.2) Same as Case-1 but 233U – (ENDF/B-VII) Same as Case-1 but 233U, 234 U – (ENDF/B-VII) Same as Case-1 but 232U, 233 U, 234U – (ENDF/B-VII) Same as Case-4 but 27Al from IAEA ADS Library (based on JEFF 3.1 which is also same as ENDF/B-VII)
keff(1r interval)
q (mk)
0.9890 ± 0.0003
11.1223
0.98899 ± 0.00025
11.1326
0.99171 ± 0.00025
8.3593
0.99194 ± 0.00024
8.1255
0.99176 ± 0.00025
8.3085
0.99357 ± 0.00025
6.4716
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Table 4 Comparison of cross-section (barns) data of energy, 0.025 eV
233
U and
27
Al at neutron
Neutron reaction type
ENDF/B-VI.8
ENDF/B-VII
27
0.23441 534.04 46.2324
0.23583 536.74 45.692
Al Capture 233 U Fission 233 U Capture
Data from ENDF/B-VI.8 and ENDF/B-VII for neutron capture and fission cross-sections of U233 and capture cross-section of Al27 at neutron energy of 0.025 eV are compared in Table 4. 5. Results and discussion The keff value computed for KAMINI reactor by the evaluators using MCNP code and presented in the ICSBEP handbook is 0.9890 ± 0.0003 using ENDF/B-VI.8 crosssection data and it is identical with the keff value obtained
Fig. 4. Comparison of U233 fission cross-section of (a) ENDF/B-VI (b) ENDF/B-VII.
for Case-1 (i.e., using ENDF/B-VI.2 cross-section data in ICSBEP model of KAMINI). This shows that the crosssection data used in our earlier work is similar to that used by the evaluators in ICSBEP benchmark computational model. Hence, it should be believed that the improved details of geometry and atomic densities given in ICSBEP model of KAMINI is the reason for the difference in our earlier prediction of keff = 1.017 compared to the ICSBEP evaluators’ prediction of 0.9890. In the next analysis (i.e., Case-2), the 233U neutron cross-section data alone is changed from ENDF/B-VI.8 to ENDF/B-VII (IAEA Th-U cycle) in the input model by which the keff computed changed from 0.98899 to 0.99171, which is more closer to the experimental value of 1.000 by about 2.7733 mk (see Table 3). It can be inferred from Table 3 that the change of crosssection data from ENDF/B-VI.8 to ENDF/B-VII made for 232 U and 234U (Cases 3 and 4) in the input model did not alter the keff value substantially. This may be because these
Fig. 5. Comparison of Al27 capture cross-section of (a) ENDF/B-VI (b) ENDF/B-VII.
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isotopes are present only in trace quantities in KAMINI reactor fuel. Since aluminum is a major constituent in the KAMINI reactor fuel alloy, the effect of changing cross-section data of 27Al from ENDF/B-VI.8 to ENDF/B-VII is studied next (Case-5), the results of which showed the keff moving closer to the experimental value of 1.000 by another 1.8369 mk (see Table 3). Hence by changing the cross-section data of 233U, 232U, 234U and 27Al from ENDF/B-VI.8 to ENDF/B-VII the keff value improved from 0.9890 to 0.99357 which is equivalent to a reactivity change of 4.6507 mk. The changes in neutron cross data between ENDF/BVI.8 and ENDF/B-VII of two major constituents of KAMINI reactor fuel, i.e. 233U and 27Al was closely examined especially with respect to the fission cross section of 233 U and capture cross-section of 27Al which are compared in Figs. 4 and 5. It can be seen from Fig. 4 that more resolved resonance structure has been introduced into the 233 U fission cross-section data of ENDF/B-VII beyond neutron energy of 100 eV. From Fig. 5, it can be seen that there is change in the capture cross-section resonance data of 27Al beyond neutron energy of 0.1 MeV. These changes in the resonance data for 233U and 27Al lead to the reactivity changes mentioned above. 6. Conclusions
(a) The MCNP modeling of KAMINI given in ICSBEP benchmark differs from our earlier model in the atomic densities of fuel plates, which lead to a reactivity difference of about 27.8 mk between both the models. By using ENDF/B-VI.2 cross section data we get the same keff as ICSBEP model result quoted in the benchmark report with ENDF/B-VI.8 data. Thus, it is confirmed that there is negligible difference between the use of ENDF/B-VI.8 and ENDF/B-VI.2 data. (b) By employing the new cross section data from ENDF/B-VII for 233U alone in the criticality calculations a reactivity change of 2.7733 mk is obtained and is more closer to the experimental value. Similarly, using 27Al cross-section data of ENDF/B-VII showed a reactivity change of 1.8369 mk, which also tends closer to the experimental value. It is inferred that by changing the cross-section data of 233U, 232U, 234 U and 27Al from ENDF/B-VI.8 to ENDF/B-VII the keff value improved from 0.9890 to 0.99357. (c) It is evident from Table 4 that thermal cross sections are not contributing to the improved results. Examination of the new cross-section data shows the more
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resolved resonance data in the fission cross-section of 233 U beyond neutron energy of 100 eV as well as change in the capture cross-section resonance data of 27Al beyond neutron energy of 0.1 MeV. These changes improved the computed keff value of KAMINI reactor benchmark.
Acknowledgement Authors would like to acknowledge Dr. S. Ganesan of RPDD, BARC (India) for giving insight to start this work and for fruitful discussions regarding the IAEA coordinated research project on thorium–uranium fuel cycle cross-section data. References Aldama, D.L., Trkov, A., 2005. ADS-Lib/V1.0: A Test Library for Accelerator Driven Systems, INDC (NDS)-0474, IAEA. Available from: