MONTE CARLO CALCULATIONS ON THE FIRST CRITICALITY OF THE MULTIPURPOSE REACTOR G.A. SIWABESSY. Liem Peng Hong Center for Multipurpose Reactor - BATAN

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1 MONTE CARLO CALCULATIONS ON THE FIRST CRITICALITY OF THE MULTIPURPOSE REACTOR G.A. SIWASSY Liem Peng Hong Center for Multipurpose Reactor - BATAN ABSTCT MONTE CARLO CALCULATIONS ON THE FIRST CRITICALITY OF THE MULTIPURPOSE REACTOR G.A. SIWASSY. Monte Carlo first criticality calculations have been conducted on the first core of the Indonesian 30 MWth Multipurpose Reactor G.A. Siwabessy. A vectorized, continuous energy Monte Carlo code developed by JAERI was used for the whole analyses and calculations, while the results were then compared to the experiment data collected during the first core reactor physics experiments. Monte Carlo calculations produced k eff values with excellent agreement with experiment data. INTRODUCTION The construction of the Indonesian multipurpose research reactor, Reaktor Serba Guna G.A. Siwabessy (RSG GAS) [1], has been completed and the first criticality of the reactor was achieved in July 29, RSG GAS has five transition cores (smaller cores) before a full core configuration (40 standard fuel elements and 8 control fuel elements) can be achieved in the sixth core. As a part of the commissioning work, a series of reactor physics experiments has been done on the transition cores especially for the first core. These efforts have been aimed to check and fulfill the safety and operational requirements. However, sufficient calculation efforts using sophisticated method such as Monte Carlo method have not been conducted. In this report, Monte Carlo calculation results for the first core of RSG GAS will be presented and compared with the experiment data. The first core was chosen because of, first, the completeness of the experiment data, and secondly, the fuel elements, absorber and beryllium reflectors were still fresh so that uncertainties raised from the fuel burnup and absorber depletion calculations, and from the calculation of lithium poisoning in the beryllium reflectors can be eliminated. The vectorized, continuous energy Monte Carlo code, MVP [2], developed by the Japan Atomic Energy Research Institute (JAERI), was used for the entire calculations. In the present work, as the nuclear data libraries for the code, the third version Japanese Evaluated Nuclear Data Library, JENDL-3.1 [3] and JENDL-3.2 [4] were used to check their accuracy by comparing to the experiment data. The Monte Carlo method has been selected for the present work since it is the most accurate method for neutron criticality calculations. The exact geometry, dimension and composition of the reactor details can be modeled in 3-D Monte Carlo method, and the neutron transport can be simulated 51

2 without any approximation. However, in the computer code implementation of Monte Carlo method one has to consider the type of neutron cross section data which will be used in the simulation. In the early application of Monte Carlo code for criticality calculations the many group cross section type was commonly used since the computer memory and computational capacity were then limited. In this case the continuous neutron energy is approximated by many discrete energy groups. With the fast development of computer technology the limitation can be relaxed significantly. The most advanced Monte Carlo code, such as MVP code that was used in the present work, used the energy-continuous cross section data which removes the approximation applied in the neutron energy variable. On top of that, in order to save computational time, the code has been vectorized by taking the advantage of the parallel processing technology available recently. The organization of the paper is as follows. In the next section, the criticality experiment for establishing the first core of RSG GAS is presented in detail which consists mainly of first criticality experiment, excess reactivity loading experiments and control rod calibrations. The third section discussed the modeling adopted and benchmark Monte Carlo calculations for simulating the experiments. In the fourth section, the Monte Carlo calculation results are discussed and compared with the experiment results. The last section gives the concluding remarks of the present work. CRITICALITY EXPERIMENTS Firstly, the RSG GAS is briefly over-viewed here. RSG GAS is a 30 MWth open pool type, light water moderated and cooled, beryllium reflected reactor. Reactor main data are shown in Table 1. Fuel elements used for RSG GAS are of the material testing reactor (MTR), i.e. plate-type fuel elements, and one fuel element consists of 21 fuel plates assembled by two side plates. The cross-sectional view of RSG GAS standard fuel element is shown in Fig. 1. One fuel plate consists of % enriched uranium oxide meat (with uranium density of 2.96 gu/cc) embedded in aluminum matrix, and aluminum cladding. The active length of the fuel element (or the meat height) is 60 cm. The nominal 235 U loading per standard fuel element is 250 g. Control fuel elements, shown in Fig. 2, with identical outer dimension consist of 15 fuel plates, that is, three fuel plates at both outer sides of the fuel elements are removed to provide space for absorber blades, and therefore the nominal 235 U loading for a control fuel element reduces to g. At both sides of the control fuel element, two absorber guide plates (aluminum) are installed. A fork type control rod (0.38 cm thick Ag-In-Cd absorber meat with SS-321 cladding) can be inserted into or withdrawn out of the control fuel element. 52

3 The full configuration of the first core of RSG GAS consists of 12 fresh standard and 6 control fuel elements while the configuration for the first criticality needs only fresh 9 standard and 6 control fuel elements. The first criticality configuration of the first core is shown in Fig. 3. The experiment procedure and results to achieve the first criticality are briefly described. First, estimation of the number of standard and control fuel elements required for the first criticality was conducted with IAFUEL code [5], an incore fuel management code to date used for RSG GAS in-core fuel management. The code was provided by Interatom. The code adopted 2-D, X-Y reactor geometry 4 group (direct coupled) diffusion theory. From the calculation results, it was expected that the first criticality would be achieved with 6 control fuel elements, 8± 1 standard fuel elements while all control rods were withdrawn. Secondly, six control fuel elements were loaded in the core with the absorber plates were all fully inserted. 252 Cf neutron source was inserted into the E-7 grid position to initiate fission chain reactions, and then loading of standard fuel elements was done step by step. The fuel loading and first criticality were planned to be achieved at the most critical condition, that is, the water coolant temperature in the primary circuit was kept low (25 o C), all neutron beam tubes were flooded (filled) by water, all fuel elements were free of xenon and samarium by operating the reactor at very low power. At each loading step, the inverse multiplication rates were plotted for four control rod configurations to estimate the critical loading. The four control rod configurations were (A) all rods inserted, (B) regulating rod (grid position C-8) withdrawn and shim bank rods inserted, (C) shim bank rods withdrawn and regulating rod inserted, and (D) all rods withdrawn. At critical loading, the establishment of the control rod configuration C were closely monitored by source and start-up range channels. Thirdly, the first criticality was not achieved at the 8-th loading of standard fuel elements and the 9-th fuel element was inserted (RI-20 into grid position F-7). The first criticality was achieved in this loading step when the control rod configuration was changed from configuration C to D, i.e. the regulating rod was pulled out slowly (reactor periods were greater than 50 s) while the other 5 shim rods were fully withdrawn. Regulating rod position of 475 mm (inserted length) gave the first criticality condition for the first core of RSG GAS. This configuration was the most appropriate for benchmark calculations such as by the Monte Carlo method. Fourthly, after first criticality, loading of fuel elements and reflector elements were conducted to achieve a full core configuration with sufficient excess reactivity for one core cycle. At each loading step, reactivity gains were measured by calibrating the difference of the regulating rod position with a reactivity meter (compensation method with the other 5 shim rods in 53

4 bank configuration). The accumulated excess reactivity measured by this method should be considered as an uncorrected excess reactivity value. The excess reactivity value was then corrected by a method described in Ref. [6] using 2-D multigroup neutron diffusion code, Batan-2DIFF [7] in XY reactor geometry. In the calculations the group constant sets were obtained from the IAFUEL code. Throughout the excess reactivity loading, the shutdown reactivity margin for one stuck-rod condition must be equal or greater than 1 % k / k (xenon and samarium free, cold condition). After the full configuration of the first core was achieved the measured uncorrected excess reactivity value was 8.14 % k / k (10.63 $) while the calculated value by IAFUEL code was 9.2 % k / k. After correction on the measured excess reactivity was done the corrected excess reactivity of the first core was found to be 8.46 % k / k ( $). This excess reactivity was converted to effective multiplication factor of The full configuration of the first core is shown in Fig. 4. Following the excess reactivity loading, control rod calibrations were conducted for the six control rods with various methods. During the calibrations one can find many combinations of control rod positions which gave a critical core condition suitable for our Monte Carlo calculations. In the present work, the combinations of control rod positions occurred during control rod calibration with bank compensation method will be checked with Monte Carlo calculations. It should be noted here that the critical conditions were achieved when the primary cooling pumps worked producing coolant flow of 2500 m 3 /hour which increased slightly the core temperature (Joule effect) isothermally. The reactivity effect of the Joule effect from the primary cooling system was measured to correct the experiment data. MONTE CARLO CALCULATIONS The active part (7.71 x 8.1 x 60 cm 3 ) of both standard and control fuel elements were modeled as their exact geometry and dimensions while the top and end-fitting of the elements were modeled in an approximate manner since their geometry are very complicated, that is, the structure materials were homogenized with water by volume weighting. Exact modeling approach was also taken for the active parts of the beryllium reflector elements, beryllium block elements and irradiation positions. Considering their complicated geometry, the core grid and bottom support were also treated approximately as for the top or end-fitting of fuel elements. This approximation did not deteriorate the accuracy of the Monte Carlo calculation results since it was applied in the non active parts of the core. The movable control rods (absorber blades) were modeled as their exact geometry and dimensions. Consequently, 60 cm water layer above the core had to be included in the calculation to provide enough space for the 54

5 absorber blades when a control rod was fully withdrawn. Approximately 30 cm water layers were included below the core bottom support, and around the beryllium block and element reflectors. Vacuum boundary conditions were imposed on the outer boundary of the reactor system. All Monte Carlo calculations in the present work were conducted with libraries from JENDL-3.1 and JENDL-3.2 for temperature of 300 K. The measured critical effective multiplication factors were corrected when the core isothermal temperature was not identical with 300 K. The total number of particles simulated was for all cases considered. RESULTS AND DISCUSSIONS Table 2 shows the comparison between experiment data and Monte Carlo calculation results for first criticality and excess reactivity of RSG GAS first core. The first criticality predictions by Monte Carlo method were very close to the experiment data, especially the one with JENDL-3.1 library. JENDL-3.2 library produced a slightly higher effective multiplication factor although the difference with the experiment data was still below 1 %. For the full configuration of the RSG GAS first core, the (corrected) excess reactivity data and Monte Carlo calculation results also showed a good agreement. However, again JENDL-3.2 library gave a slightly higher effective multiplication factor. It should be noted here that the measured excess reactivity must be converted from $ to absolute value to get the (absolute unit) effective multiplication factor, and therefore, the accuracy depended on the accuracy of the calculated β eff (= ). No measured data were available for β eff so that further assessment on the accuracy of the calculated β eff was not possible. This factor contributed also in the difference between the experiment data and Monte Carlo results. The total control rod worth shown in Table 2 was obtained by a simple arithmatic summation of single control rod worth, while the single control rod worth was measured by a reactivity meter with shim rod bank compensation method. It is well known that the interference between control rods to some extent may deteriorate the accuracy of the total control rod worth obtained by the summation of single control rod worth. Combined with the uncertainty of the calculated β eff value, the Monte Carlo results can be judged to be well agreed with the experiment data. Table 3 shows the comparison between Monte Carlo results with several critical conditions of the RSG GAS first core occurred during control rod calibrations. These critical conditions were achieved when the calibrated rod was fully inserted and the other rods were in a certain bank position. The measured critical conditions shown in the table were already corrected for the Joule effect (which turned out to be negligible) of the primary cooling 55

6 pumps. It can be observed from the table that Monte Carlo results using JENDL-3.1 library gave excellent agreement with experiment data. JENDL-3.2 library results systematically produced slightly higher effective multiplication factors but still below 1 % differences. The slightly higher effective multiplication factors calculated with JENDL-3.1 library are consistent with the evaluation reported by Nakagawa et al. (1995). It is attributed to the smaller 235 U capture cross sections in the unresolved region (below 30 kev) in JENDL-3.2 than the one in JENDL-3.1. CONCLUDING REMARKS Monte Carlo calculations have been conducted on the first core of the Indonesian 30 MWth Multipurpose Reactor, RSG GAS. Vectorized, continuous energy Monte Carlo code, MVP, developed by JAERI combined with nuclear data derived from JENDL-3.1 and -3.2 libraries was used for the whole benchmark calculations and the results were compared with experiment data. Monte Carlo calculations with both JENDL-3.1 and -3.2 libraries produced k eff values with excellent agreement to experiment data. Systematically, JENDL-3.2 library showed a slightly higher k eff values than JENDL-3.1 library. ACKNOWLEDGMENTS Firstly, the author expresses his gratitude to Mr. Yasuhiro Nakano, Department of Research Reactor, Tokai Research Establishment, JAERI, for his valuable advisory and admission in performing the work during the author stay in JAERI. Secondly, the hard work of the commissioning group of RSG GAS for collecting and providing the criticality experiment data is greatly acknowledged. 56

7 REFERENCES 1. Batan, Multipurpose Reactor G.A. Siwabessy Safety Analysis Report, Rev. 7, Jakarta (1989) 2. MORI, T. and NAKAGAWA, M., JAERI-Data/Code , Japan (1994) 3. SHIBATA, K. et al., JAERI 1319 (1990) 4. NAKAGAWA, T. et al., Jour. Nucl. Sci. Tech., 32(12), (1995) 5. WICKERT, M., "Concept and Methods of the Program MAIN: Controlling the IAFUEL Program Cycle for Neutronic Calculations Regarding Research Reactor", Interatom Bericht/Report Ident-No: (1986) 6. NABBI, R. and WICKERT, M., Internal Report, Batan, Jakarta (1987) 7. Liem, P.H., Atom Indonesia, 20, 1 (1994) 57

8 General Table 1. Reactor Main Design Data of RSG GAS. Reactor Type Pool Type Fuel Element Type LEU Oxide MTR Cooling System Forced Convection Down Flow Moderator/Coolant H 2 0 Reflector Be & H 2 0 Maximum Power (MWt) 30 Core Characteristics (full core configuration) No. of Fuel Elements 40 No. of Control Elements 8 No. of Fork Type Absorber (pairs) 8 Nominal Cycle Length (fpd) 25 Ave. Burn-up at BOC (% loss of U-235) 23.3 Ave. Burn-up at EOC (% loss of U-235) 31.3 Ave. Discharge Burn-up at EOC (% loss of U-235) 53.7 Fuel/Control Elements Fuel/Control Element Dimension (mm) 77.1x81x60 Fuel Plate Thickness (mm) 1.3 Coolant Channel Width (mm) 2.55 No. of Plate per Fuel Element 21 No. of Plate per Control Element 15 Fuel Plate Clad Material AlMg 2 Fuel Plate Clad Thickness (mm) 0.38 Fuel Meat Dimension (mm) 0.54x62.75x600 Fuel Meat Material U 3 O 8 Al U-235 Enrichment (w/o) Uranium Density in Meat (g/cc) 2.96 U-235 Loading per Fuel Element (g) 250 U-235 Loading per Control Element (g) Absorber Meat Material Ag-In-Cd Absorber Thickness (mm) 3.38 Absorber Clad Material SS-321 Absorber Clad Thickness (mm)

9 Table 2. Comparison of Monte Carlo calculation results with experiment data for first criticality and excess reactivity of RSG GAS first core (300 K). Core configuration Experiment Monte Carlo calculations data JENDL-3.1 MVP code JENDL-3.2 MVP code First criticality (9 FEs, 6 k eff ± ± CEs, RR=475 mm) C/E Full core (9 FEs, 6 CEs, k eff a) ± ± CRs all up) C/E Full core (9 FEs, 6 CEs, k eff n.c. b) ± ± CRs all down) C/E - - Control rods worth ρ (%) c) C/E Note : a) β = , excess reactivity was already corrected (see text) b) n.c. = not conducted c) Shim rod bank compensation method, measured by reactivity meter, summation of single control rod worth 59

10 Table 3. Comparison of Monte Carlo calculation results with several critical conditions of RSG GAS first core occurred during control rod calibrations (300 K). Calibrated rod / grid position Experiment Monte Carlo calculations ( calibrated rod position / other rod bank position ) data JENDL-3.1 MVP code JENDL-3.2 MVP code JDA06 / C-8 k eff a) ± ± (600 mm / 290 mm) C/E JDA01 / E-9 k eff ± ± (600 mm / 284 mm) C/E JDA03 / F-8 k eff ± ± (600 mm / 293 mm) C/E JDA05 / C-5 k eff ± ± (600 mm / 288 mm) C/E JDA04 / F-5 k eff ± ± (600 mm / 290 mm) C/E JDA07 / D-4 k eff ± ± (600 mm / 282 mm) C/E Note : a) β = , excess reactivity was already corrected (see text) 60

11 Figure 1. RSG GAS standard fuel element (unit mm). 61

12 Figure 2. RSG GAS control fuel element with absorber blades inserted (unit mm). 62

13 K J H G BF3 F C-2 F-20 C-4 PN E C-5 F-13 DE * F-19 HY D BS+ F-18 F-15 F-14 DE F-12 C-6 HY C DE C-1 F-16 F-17 C-3 HY B BS- DE HY A BF Beryllium Block Reflector F : Fuel Element BS- : Be Refl. El. without Stopper C : Control Element BS+ : Be Refl. El. with Stopper : Be Reflector Element DE : Dummy Element PN : Pneu. Rabbit System HY : Hydraulic Rabbit System BF 3 : BF 3 Neutron Detector * : Neutron Source (Cf-252) Figure 3. First criticality core configuration of the RSG GAS first core. 63

14 K J H G BF3 BS+ 4 BS+ 5 BS F C-2 F-20 F-22 C-4 DE PN E C-5 F-13 DE F-19 F-21 F-23 HY * D BS+ F-18 F-15 F-14 DE F-12 C-6 10 HY C DE C-1 F-16 F-17 C HY B BS- DE HY A BF Beryllium Block Reflector F : Fuel Element BS- : Be Refl. El. without Stopper C : Control Element BS+ : Be Refl. El. with Stopper : Be Reflector Element DE : Dummy Element PN : Pneu. Rabbit System HY : Hydraulic Rabbit System BF 3 : Neutron Detector * : Neutron Source (Cf-252) Figure 4. Full core configuration of the RSG GAS first core. 64