VALIDATION OF BATAN'S STANDARD 3-D DIFFUSION CODE, BAT AN-3DIFF, ON THE FIRST CORE OF RSG GAS. Liem Peng Hong"

Transcription

1 VALIDATION OF BATAN'S STANDARD 3-D DIFFUSION CODE, BAT AN-3DIFF, ON THE FIRST CORE OF RSG GAS Liem Peng Hong" ABSTRACT VALIDATION OF BATAN'S STANDARD 3-D DIFFUSION CODE, BATAN 3DIFF, ON THE FIRST CORE OF RSG GAS. Batan-3DIFF code validation efforts have been conducted on the experiment data compiled during the commissioning phase of Reaktor Serba GUlla G.A. Siwabessy (RSG GAS) first core. The validation work covers the effective multiplication factor, excess reactivity and the control rod worth. One set of 4-group macroscopic cross section library prepared by WIMS/D4 cell calculation code and another set derived from lafuel library were used in the present analysis. In order to simulate the existence of the control rod absorber material in the core, the extrapolation lengths for each group are calculated based on the dirty blackness coefficients. The ratio of the calculated and experiment (C/E) values for Batan-3DIFF code with WIMS/D4 and IAFUEL libraries for the first criticality condition are and 0.994, respectively, while for the excess reactivity of the full core are and 1.000, respectively. In general, the WIMS/D4 library gives slightly lower effective multiplication factors than the IAFUEL library does. The C/E values for Batan-3DIFF code with WIMS/D4 and lafuel libraries for the total control rod worth are worse, that is, and 0.881, respectively. INTRODUCTION Several Batan's standard neutron diffusion codes [1,2,3] as well as a dedicated in-core fuel management code for research reactors [4] have been developed, verified and validated to meet one objective of the research and development activities conducted by the Center for Multipurpose Reactor, National Atomic Energy Agency (Batan). Since the first version of the codes were launched several important improvements, modifications as well as addition of new options on the codes have been continuously conducted to response to the new requirements, users' comments and feedback. Important options developed and verified are the capability of the codes to calculate the adjoint neutron flux (that is solving the adjoint neutron eigenvalue problems), external neutron source (that is solving the fixed source problems), integral kinetic parameters, reactivity changes based on the perturbation theory etc. The options recently developed, which are closely related to the present work, were the two options for treating the existence of strong absorber materials in the reactor [5]. Particularly, from the point of view of reactor safety these two options are of great importance. The first option for treating strong absorber uses a pair of blackness coefficients which are then converted to effective diffusion parameters (effective diffusion constant and absorption cross section of the absorber materials), while the second one Center for Multipurpose Reactor, National AtomicEnergy Agency 47

2 adopts the logarithmic derivative constant applied as an internal boundary condition for the diffusion calculations. In one hand, with all of these options, the Batan's standard codes have to be systematically verified and validated on the available benchmark problems and experimental data. On the other hand, plentiful sources of experimental data are available from the long eleven years of RSG GAS operation; especially for the ones obtained during the commissioning works which have not been utilized for validation works. In the present work, validation results of Batan-3DIFF code (for 3-D multigroup neutron diffusion calculations) on the first core of RSG GAS experiment data will be reported. 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 organization of the paper is as follows. In the next section, the neutronic experiments of the RSG GAS first core are presented in detail. The third section discusses the neutronic modeling and calculation procedure for simulating the experiments. In the fourth section, the Batan-3DIFF calculation results are discussed and compared with the experiment results. The last section gives the concluding remarks of the present work. FIRST CORE NEUTRONIC EXPERIMENTS In this section, some neutronic experiments closely related to the present validation works are briefly reviewed. These include the criticality experiment and control rod calibration. The establishment of the first full core of RSG GAS (12 standard fuel elements and 6 control fuel elements) consists of two steps. The first step was to achieve a critical condition of the core with the smallest number of fuel elements, i.e. achieving the first criticality of the first core. Following the first criticality, in the second step, a certain number of fuel elements were then added into the core to provide sufficient excess reactivity to the core so that the reactor can be operated for a certain period without refueling, i.e. the excess reactivity loading. The configuration for the first criticality needs only fresh 9 standard and 6 control fuel elements (Fig. 1) while the full configuration of the first core established in the second step consists of 12 fresh standard fuel elements and 6 control fuel elements (Fig. 2). In the second step, at each fuel element loading step, the reactivity gains were measured by calibrating the difference of the regulating rod position with a reactivity meter (single rod vs. bank rods compensation method). 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 48

3 2-D multigroup neutron diffusion code, Batan-2DIFF [2] in XY 2-D reactor geometry. The measured uncorrected excess reactivity value was 8.14 % 11k / k (10.63 $) while after correction on the measured excess reactivity was done the corrected excess reactivity of the first core was found to be 8.46 % 11k / k ( $). This excess reactivity was converted to effective multiplication factor of Following the excess reactivity loading, control rod worth calibrations were conducted for the six control rods with various methods. During the calibrations one can find many combinations of control rods' position which gave a critical core condition suitable for the present validation work. Since the control rods were inserted in the core the validity of the code to treat the strong absorber can be checked. Furthermore, during the control rod calibration experiment the primary cooling pumps worked producing coolant flow of 2500 m3lhour 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. CALCULA TION PROCEDURE AND MODELING The calculation procedure of the validation works is shown in Fig. 3, which consists of two main parts, i.e. the 4 group macroscopic cross sections library generation and the whole core neutron diffusion calculations using Batan's standard neutron diffusion code, Batan-3DIFF in XYZ 3-D reactor geometry. The procedure and modeling used in the cross sections generation for fuel element using WIMS/D4 code have been validated and extensively discussed in Ref. [5], therefore, it will only be briefly reviewed here. For standard fuel elements, the multiplate option is chosen where the unit cell consists of eleven and a half identical fuel regions and one extra region with each fuel region is divided further into fuel meat, aluminum cladding, water moderator regions. The extra region is used to include regions that are not covered by meat-cladding-moderator fuel plate configurations. Those regions are side plates, small parts of aluminum cladding adjacent to side plates and water gaps among fuel elements. Besides the cross section library obtained from the WIMS/D4 code, the cross section library derived from IAFUEL code [6] to date used by the Center for the RSG GAS in-core fuel management will also be used for comparison. In addition to the cross section generation, in order to treat the strong absorber in the core, the extrapolation length of the RSG GAS control rod has to be determined. The procedure to determine the extrapolation length of the RSG GAS control rod follows the one already validated and discussed in Ref. [5]. Batan's standard diffusion codes can treat strong absorbers as nondiffusing regions with an appropriate set of internal boundary conditions that usually derived from the blackness coefficient a. 49

4 (1) where <I> and d are neutron flux and extrapolation distance. In this case, the neutron current and flux at both sides of the absorber surface are assumed to be identical. The equation can be rewritten by using the neutron current expression, (2) where D and J denote the neutron diffusion coefficient and neutron current as function of neutron energy. From the definition of a blackness coefficient, the above boundary condition C=a, that is the surface current-to-flux ratio. For example, if the dirty blackness coefficient, a(db), is used to compute the extrapolation distance then one can write, o d = a (DB) 11+3E.(L.<)] = ),{1_ o 2E3 (La") (3) where Au-, La and "t represent the transport mean free path, absorption cross section and absorber material tickness, respectively. In the codes, this latter expression for the boundary condition is used. The energy dependent d is supplied by code users to simulate the existence of the strong absorber. Use of Eq.(3) in Batan-3DIFF code requires a fast and efficient calculation way of the exponential integral functions, En (n=3,4). The rational approximation of these integral functions as reported by Makino [8] is used where the maximum relative error of the approximation are in the order of The whole core diffusion calculations in 3-D reactor geometry with Batan-3DIFF code can simulate precisely the experiment conditions, especially, the position of the control rods within the core. 50

5 RESULTS AND DISCUSSION In Table 2, the comparison between Batan-3DIFF calculation results with experiment data for the first criticality and excess reactivity of RSG GAS first core is shown. In the table, pure experiment data (no correction) are for the critical condition under first criticality core configuration (9 fuel elements - FEs, 6 control elements - CEs, with the regulation rod at the position of 475 mm). For this critical condition Batan 3DIFF code with WIMS/D4 and IAFUEL libraries gives ratio of calculated to experiment (C/E) values of and 0.994, respectively. The IAFUEL library gives a closer calculation results than the one with WIMS/D4 library; however, both libraries underestimate the critical condition. Under the first criticality core configuration, the regulation rod is only inserted 125 mm from the core top position; however, the reactivity on inserted part of the regulation rod is too small to determine the accuracy of the extrapolation length calculation. As already discussed in the previous subsection, the other two experiment data, that is, the excess reactivity under the full core configuration (12 FEs, 6 CEs, and all six control rods fully withdrawn) and the total control rod worth are not purely experiment data. The measured excess reactivity has been corrected using Batan-2DIFF code and then converted from $ unit to absolute value in order to obtain the effective multiplication factor, keff, and consequently, the accuracy depended on the accuracy of the calculated /3eff (== ). No measured data are available for /3effso that further assessment on the accuracy of the calculated /3effis not possible. For the excess reactivity, the C/E values for Batan-3DIFF with WIMS/D4 and IAFUEL libraries are and 1.000, respectively. These C/E values are slightly improved, and again the WIMS/D4 library gives a slightly lower keff value than the IAFUEL library does (around 1 $). Furthermore, the experiment value of the total control rod worth is obtained simply by an arithmetic summation of the individual control rod worth, or in other words, it is assumed that no significant control rod interaction occurs. For this experiment value, the C/E values of the Batan 3DIFF with WIMS/D4 and IAFUEL libraries are and 0.881, respectively. Considerably large deviations between experiment and calculated values can be observed. This will be discussed later. Table 3 shows the comparison between Batan-3DIFF calculation results with several critical conditions of the RSG GAS first core established 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 have been corrected for the Joule effect (which turned out to be negligible) of the primary cooling pumps. 51

6 This experiment is very important in that large portions of control rods were inserted into the core so that the accuracy of the extrapolation length calculation for simulating the existence of strong absorber in the core can be evaluated accurately. For this experiment, the cm values for Batan-3DIFF with WIMS/D4 and IAFUEL libraries are closer than those for the first criticality condition. Since both conditions were critical so that no correction (neither for excess reactivity calculation nor from the calculated /3elfl are needed then the following conclusion can be drawn. The results strongly suggest that the calculated extrapolation lengths produce slightly underestimated reactivity values for the control rods under consideration. In other words, if the extrapolation lenghts were correct then the cm values for the control rod calibration would produce the same order of, but not better, accuracy with the cm value for the first criticality experiment. This result is consistent with the underestimated calculation results of the total control rod worth shown in Table 2. However, the deviation from the experiment value of total control rod worth is still considered large and may not just because of the error from the extrapolation length estimation. Thus, the assumption and consequently the procedure to obtain the experimental value of the total control rod worth need to be reevaluated. The next discussion will focus on the differences of the nuclear data libraries used for the validation which were calculated by WIMS/D4 code and derived from IAFUEL library. The neutron reaction rates as well as the neutron balance over the reactor are shown in Tables 4 and 5, respectively, for the first criticality condition. In the lowest parts of the tables, the relative differences between the twos are given. From Table 4, it can be observed that, in one hand, the total fission or production rates for WIMS/D4 library are about 0.5 % greater than the ones for IAFUEL library, on the other hand, the absorption rate are about 1.1 % greater than the one for IAFUEL library. This is one reason for the slightly lower keffofwims/d4iibrary. From the neutron balance (Table 5) this can also be observed more clearly. While the neutron production was 100 % (normalized) the neutron absorption were % and 99.8 % for WIMS/D4 and IAFUEL libraries, respectively. Besides neutron absorption, neutron loss may arise from escaping from the reactor system (total leakage) and from control rod absorption. The control rod absorption for the two libraries was essentially the same, but the total leakage for WIMS/D4 library was slightly larger than the one for IAFUEL library. The latter trend made the values of keff for WIMS/D4 library smaller. To elaborate more on these differences, the magnitude of the fuel element macroscopic cross sections for the two libraries have been investigated. It turns out that the eta values (1]) for WIMS/D4 library in the thermal and epithermal energy groups are slightly lower than, while the diffusion coefficients for the thermal energy group are higher than, that for IAFUEL library. The former factor gives relatively lower 52

7 infinity multiplication factor while the latter one enhances the neutron leakage. They combine to produce lower kejffor WIMSlD41ibrary. Fig. 4 shows the fuel element channel factor (FECF) distribution or commonly known as the radial power peaking factor distribution for the two libraries. Essentially there is no significant difference between the FECF distributions of the two libraries. Fig. 5 shows the maximum values of the fuel element axial power distribution, commonly known as the axial power peaking factor, for the two libraries. Again no significant difference between the axial power peaking factors of the two libraries is observed. The last figure, Fig. 6, shows the axial power profiles for the control element at core grid C-8 (where the regulation rod is inserted 12.5 cm) and the fuel element at core grid E-6 under first criticality of the RSG GAS first core calculated by Batan-3DIFF code. No significant difference is observed between the two libraries. CONCLUDING REMARKS Batan-3DIFF code validation efforts have been conducted on the experiment data compiled during the commissioning phase of RSG GAS first core. The validation work covers the effective multiplication factor, excess reactivity and the control rod worth. Two sets of macroscopic cross section library were used, i.e. one set was prepared by WIMSID4 cell calculation code and another set was derived from IAFUEL library. The C/E values for Batan-3DIFF code with WIMSID4 and IAFUEL libraries for the first criticality condition are and 0.994, respectively, while for the excess reactivity of the full core are and 1.000, respectively. In general, the WIMSID4 library gives slightly lower effective multiplication factors than the IAFUEL library does. The C/E values for Batan-3DIFF code with WIMSID4 and IAFUEL libraries for the total control rod worth are worse, that is, and 0.881, respectively. This indicates that the assumption that no significant control rods interference occurs, from which the experiment value of the total control rod worth may be obtained by a simple arithmetic summation of the individual control rod worth, have to be reevaluated. ACKNOWLEDGMENTS The author express their gratitude to the commissioning GAS for providing the criticality experiment data. group of RSG 53

8 REFERENCES 1. P.H. LIEM, "Introduction to Diffusion Code Programming", IAEA Regional Training Course on Calculation and Measurement of Neutron Flux Spectrum for Research Reactors, Lecture Notes, Jakarta (1993) 2. P.H. LIEM, "Development and Verification of Batan's Standard, Two Dimensional Multigroup Neutron Diffusion Code (Batan-2DIFF)", Atom Indonesia 20 (1) (1994) 3. P.H. LIEM, "Pengembangan Program Komputer Standar Batan Difusi Neutron Banyak Kelompok 3-D (Batan-3DIFF)", Risalah Komputasi dalam Sains dan Tekonologi Nuklir V, PPI-Batan, Jakarta (1995) 4. P.H. LIEM, "Batan-FUEL: A General Fuel Management Code", Atom Indonesia 22 (2) (1996) 5. P.H. LIEM and T.M. SEMBIRING, "Validation of Batan's Standard Neutron Diffusion Codes for Control Rod Worth Analysis", Atom Indonesia 23 (2) (1997) 6. M. WICKERT, "Concept and Methods of the Program MAIN: Controlling the IAFUEL Program Cycle for Neutronic Calculations Regarding Research Reactor", Interatom Bericht/Report Ident-No: (1986) 7. T.M. SEMBIRING and P.H. LIEM, "Validation of Batan's Standard Diffusion Codes on IAEA Benchmark Static Calculations", Atom Indonesia 23 (2) (1997) 8. Y. MAKINO, "Some Rational Approximations for the Exponential Integral Functions En (n==i,2,3,4)", Nukleonik 9 (1967). 54

9 Table 1. Reactor main design data of RSG GAS. General Reactor Type Fuel Element Type Cooling System Moderator/Coolant Reflector Nominal Power (MWt) Pool Type LEU Oxide MTR Forced Convection Down Flow H20 Be & H20 30 FueUControl Elements Fuel/Control Element Dimension (mm) Fuel Plate Thickness (mm) Coolant Channel Width (mm) No. of Plate per Fuel Element No. of Plate per Control Element Fuel Plate Clad Material Fuel Plate Clad Thickness (mm) Fuel Meat Dimension (mm) Fuel Meat Material U-235 Enrichment (w/o) Uranium Density in Meat (g/cm3) U-235 Loading per Fuel Element (g) U-235 Loading per Control Element (g) Absorber Meat Material Absorber Thickness (mm) Absorber Clad Material Absorber Clad Thickness 77.1x81x AIMg x62.75x600 U30gAl Ag-In-Cd 3.38 SS

10 VI 0'\ Table 2. Comparison of Batan-3DIFF calculation results with experiment data for first criticality and excess reactivity of RSG GAS first core (300 K). Core a) WIMS/D4 configuration Batan-3DIFF C/E n.c. b) Library IAFUELExperiment C/E0.993 Library diffusion c) i1p(%) calculations Keff Note: a) p = , 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

11 Table 3. Comparison of Batan-3DIFF calculation results with experiment data obtained during control rod calibration experiments ofrsg GAS first core (300 K). JDA06 / C-8 Calibrated WIMS/D4 rod a) / grid Batan-3DIFF C/E Library position IAFUELKetf Ken Keff Library diffusion Experiment data calculations Note: a) p = \J1 -.J

12 Table 4. Neutron reaction rates over the reactor calculated by Batan-3DIFF code for the first criticality of the RSG GAS first core. WIMSID4 Group TOTAL library Production O.OOE+OO Fission Absorption _ Top Leakage Bottom Leakage Left Leakage Right Leakage E Forward Leakage Backward Leakage Total Leakage CRAbsorption IAFUEL Library Production Fission Absorption Top Leakage Bottom Leakage Left Leakage O. Right Leakage Forward Leakage Backward Leakage E Total Leakage CRAbsorption (WIMS-IAFUEL)/IAFUEL*lOO % Production Fission _ Absorption Total Leakage CRAbsorption _

13 TOTAL Table 5. Neutron balances over _ the reactor calculated by Batan-3DIFF Production for the first criticality ofthe RSG GAS first core. Group code 59

14 F C BE PNRA BF3 : Fuel Element : Control Element : Be Reflector Element : Pneu. Rabbit System : BF3 Neutron Detector BS- : Be Refl. El. without Stopper BS+ : Be Refl. El. with Stopper DE : Dummy Element HYRA : Hydraulic Rabbit System * : Neutron Source (Cf-252) Figure 1. First criticality core configuration of the RSG GAS first core. 60

15 F C BE : Fuel Element : Control Element : Be Reflector Element PNRA : Pneu. Rabbit System BFJ : Neutron Detector BS BS+ DE HYRA * : Be Refl. EI. without Stopper : Be Refl. EI. with Stopper : Dummy Element : Hydraulic Rabbit System : Neutron Source (Cf-252) Figure 2. Full core configuration of the RSG GAS first core. 61

16 Absorber geometry & composition FE/CE geometry & composition First core config. First core criticality experiments Experimental keff Figure 3. Batan-3DIFF code validation procedure on the first core of RSG GAS. 62

17 I :m~1$,1~ ti.1f410;k Fuel element channels factors (FECF) under first criticality calculated third FECFs grid 12.5 em marks figures calculated from bythe denote Batan-3DIFF theregulation core withetop). WIMS/D4 ratiorod (the of which first WIMS and and IAFUEL isover second inserted IAFUEL; libraries figures at the the represents while depth shaded the of 8 Figure 5. Maximum values of the fuel element axial power distribution under first criticality calculated by Batan-3DIFF (the first and second figures represents FECFs calculated with WIMS/D4 and IAFUEL l,ibraries while the third figures denote the ratio of WIMS over IAFUEL; the shaded grid marks the regulation rod which is inserted at the depth of 12.5 em from the core top). 63

18 =- 1.1 'c ::J.e 1.1 ~ CI> IE a. e 1.1 J (u/ IjII! I~I) ~ _I -4(10 fa c I) ~ ~ (Ld IjII! /4 I) -tf-c ~ (10 (0 1 I) I 8./ I II 11 \8 II 11 II II I ~ Axial Direction (cm) ---.J Figure 6. The axial power profiles for the control element at core grid C-8 (where the regulation rod is inserted 12.5 cm) and the fuel element at core grid E-6 under first criticality of the RSG GAS first core calculated by Batan-3DIFF code. 64