DEPLETION ANALYSIS ON THE CONTROL ROD ABSORBER OF RSG GAS OXIDE AND SILICIDE FUEL CORES. Liem Peng Hong'

Transcription

1 DEPLETION ANALYSIS ON THE CONTROL ROD ABSORBER OF RSG GAS OXIDE AND SILICIDE FUEL CORES Liem Peng Hong' ABSTRACT DEPLETION ANALYSIS ON THE CONTROL ROD ABSORBER OF RSG GAS OXIDE AND SILICIDE FUEL CORES. The results of the depletion analysis of the Ag-In-Cd absorber nuclides composing control rod of the Reaktor Serba Guna G.A. Siwabessy (RSG GAS) and whole neutronic core calculations for obtaining the decreasing rate of the control rod worth as function of number of core cycle or accumulated core averaged burnup are reported. The main calculation tool used is JAERI SRAC-95 code system with its nuclear data library derived from JENDL-3.2. The RSG GAS considered in the analysis is the present oxide low enriched uranium (2.96 gu/cm3) Typical Working Core representing the equilibrium core. In accordance with the future, higher loading, silicide fuel conversion program to provide longer core cycles but without modifying the main reactor structures and other irradiation facilities, a similar depletion analysis is conducted on the silicide low enriched uranium (3.58 gu/cm3) core of RSG GAS. The analysis results showed that the decreasing rate of the control rod worth as function of core cycle number were 1.37 and 1.55 Percent mile/giga Watt Day/Ton heavy metal (pcmlgwd/t) for oxide and silicde cores, respectively. INTRODUCTION At present, the forty (40) standard fuel elements (FEs) [1] used in the RSG GAS (RSG GAS) are of the material testing reactor (MTR) type, Le. they consist of 21 fuel plates assembled by two side plates as shown in Figure 1. One fuel plate consists of % enriched uranium oxide meat (with uranium density of 2.96 glee) embedded in aluminum matrix and aluminum cladding. The nominal 235U loading per standard FE is 250 g. In addition to the 40 standard FEs, there are eight (8) control fuel elements (CEs). As shown in Figure 2, each of them consists of only 15 fuel plates and accordingly the 235U loading per CE becomes g. In the outer region at both sides of the CE, two absorber guide plates (aluminum) are installed. A fork type control rod (Ag-In-Cd absorber meat, cladding with SS-321) can be inserted or withdrawn into or out of the CEo The RSG GAS main design data are summirized in Table 1. A relatively high flux intensity due to a higher power density in the core (the nominal core average power density is W/cc) results in a higher consumption rate of the absorber nuclides composing the absorber rod meat. Therefore, an accurate estimation on the effective worth of the absorber rod as a function of number of core cycle have to be provided to assure a safe Center for Multipurpose Reactor (PRSG), National Atomic Energy Agency (Batan) 29

2 operation of RSG GAS. It is worthy noted that RSG GAS has entered its 30-th core cycle. On one hand, with a total 235U loading of kg at the core beginning of cycle (BOC), the reactor can be operated at its nominal power of 30 MWth for 25 days or equivalent to 750 MWD at end of cycle (EOC). The excess reactivity generated by this fissile loading is sufficient for compensating fuel burnup, xenon and samarium poisonings, power defect, and other reactivity loss for irradiation or experiments. On the other hand, in the recent years, research and development on the RSG GAS core conversion program from oxide to higher loading silicide fuel, i.e. with 235Uloading of 300 g per fuel element, is being conducted. Among other advantages of silicide fuel, its higher loading of fissile material is expected to increase the operation cycle; hence, higher reactor availability and utilization can be achieved while the fuel cost can be reduced. Specifically, the main design objective for the new silicide core of RSG GAS is to maximize the core cycle length without violating several important constraints. One of them implies that no major modification to the reactor balance of plant, shielding, core main structural components, and civil buildings may be made. Consequently, there is a strong incentive to use the existing Ag-In-Cd control rod, their driving mechanism etc. In the present oxide core, the eight control rod (with fresh absorber blades) have a total reactivity worth of around 14.5% while preliminary investigation showed that the worth decreased as the neutron spectrum in the silicide core becomes slightly harder. The result also suggested that a good estimation of the absorber life time for the new silicide core must also be made. However, problems involving strong absorber materials are essentially difficult to solve since higher order methods such as transport and Monte Carlo methods are needed. These methods require both extremely long computation time as well as an expensive computer resource. Under the Scientist Exchange Program of Japanese Science and Technology Agency, the super computer version of SRAC-95 code system developed by the Japan Atomic Energy Research Institute (JAERI) became available for the author to be used to solve the problems. Both the present oxide as well as the future silicide cores of RSG GAS will be analyzed as a preparatory step of the RSG GAS silicide core conversion program. PHYSICAL CONSIDERATIONS Typical Working Core Configurations Figure 3 shows the standard fuel element (FE) and control element (CE) arrangement in the core grid for the typical working core (TWC) at BOC which will be used as the calculation object in the present analysis. The FE or CE burnup levels are divided into 7 classes where the burnup 30

3 differences between successive classes are 8 and 9 % for oxide and silicide cores, respectively. At the BOC of each core cycle, seven (7) new FEs and one (1) new CE are loaded into the core and refueling operation is conducted to ensure a relatively flat power distribution and sufficient excess reactivity and safety margins. The loading scheme is so-called 7/1. After the loading scheme is repeated 6 times then the succeeding loading scheme changes to 6/2, i.e. six (6) fresh FEs and two (2) CEs are loaded into the core. Then, the loading scheme switches back to 7/1. This fuel management strategy was recommended by the RSG GAS vendor [1]. As already stated previously, the 235U loading for the oxide and silicide fuel elements are 250 g and 300 g, respectively. Calculation Code, Modes and Procedures The calculation code, models and procedures are outlined in this section. The calculation procedures are mainly divided into two parts, as depicted in Figures 4, 5 (first part) and 6 (second part) which are identical for both oxide and silicide cores. Nuclear Desi~n Code and Library The whole analyses of the present work were conducted using the super computer version of SRAC-95 code system [2,3] developed by JAERI combined with SRAC-95 nuclear data library compiled from the JENDL-3.2 [4]. Generation of Macroscopic Cross Section Sets Four types of cross section set for both oxide and silicide cores must be prepared, namely: I) burnup dependent FE cross section set, 2) burnup dependent CE cross section set, 3) core cycle dependent absorber cross section set, and 4) reactor structural material cross section set. In general the above macroscopic cross section sets are generated using the I-D and followed by 2-D collision probability method (SRAC's PH module) in 107 fine neutron energy group, and resonances are treated with the PEACO or NR option of SRAC (Figures 4 and 5). The transport cross section is evaluated using the B] option with critical buckling. Then, the obtained 107 group macroscopic cross section sets are fed into the 2-D, X-Y geometry SRAC's CITATION diffusion module where the TWC BOC configuration with all control rod withdrawn is chosen (Figure 6). 31

4 The detail of the energy and space dependent neutron spectra obtained are finally used to collapse the 107 energy group cross section sets into few group (10) ones. These few group cross section sets will be used in the following whole reactor neutron transport and diffusion calculations. For the oxide (silicide) TWC configuration at BOC there are 7 burnup class dependent cross section sets for both standard and control fuel elements which must be prepared, namely 0, 8, 16,24,32, 40 and 48 % (0, 9, 18,27, 36, 45 and 54 %) loss of In the actual fuel irradiation in the core, the power density levels for each FE/CE are different. However, in the I-D PH cell burnup calculation, the average core power density ( W/cc) is used. The core cycle dependent absorber compositions are obtained by tracing the history of the nuclide densities of the absorber isotopes (Ag, In and Cd) as the CEs are burnt in the core (from BOC to EOC) with I-D PH cell burnup and PEACO options. Here, one core cycle corresponds to 8 % and 9 % burnup periods for oxide and silicide cores, respectively. For example, in case of the oxide core, I-D PH cell burnup calculations for intervals of 0 to 8, 8 to 16,..., until 48 to 56 % will be conducted. Specifically, up to 56 core cycles are simulated where at the 1, 7, 14,..., 49-th core cycle the fuel meat composition of CE is set to be fresh or zero burnup. This procedure is intended to simulate the burnup process and refueling operation of the CEs. It should be noted here that the average power density levels of the CE/FE adjacent to an absorber plate, in the real situation, will slightly affect the neutron flux level inside the absorber plate and in turn will determine at what rate the absorber isotopes deplete. However, these power density level can not be perfectly modeled into the I-D PH cell burnup calculation for each absorber. On top of that, there are 8 pairs of absorber plate (16 absorber plates) with different adjacent CE/FE power density levels. Therefore, for the absorber material depletion calculation the following approximation is taken. The power density levels of the adjacent CE/FE are assumed to be equal to the average power density of the core ( W/cc). Under this assumption! approximation all the 8 pairs of absorber plates will deplete at the same rate and consequently their compositions will also be identical. Further refinement of the calculation on the depletion rates of the absorber materials can be conducted after whole core neutron diffusion or transport calculation where the power density levels of the surrounding CE/FE are known. Neutron Transport and Diffusion Calculations 2-D, XV geometry, few group neutron transport calculations with SRAC's TWOTRAN module will be used to obtain the control rod reactivity worth (cf. Figure 6). Since the transport calculations are conducted in 2-D XV reactor geometry then the axial leakage data in term of the 32

5 equivalent core height must be supplied. The equivalent core height is determined by forcing the effective multiplication factor of the TWOTRAN calculation result to be equal with the effective multiplication factor of the 3-D XYZ geometry CITATION diffusion calculation for the TWC BOC configuration with all control rod fully withdrawn. Since for this case no strong absorber exists in the core then the accuracy of the 3-D CITATION neutron diffusion calculation can be justified. The total reactivity of the control rod is calculated by the reactivity difference between the fully inserted and fully withdrawn conditions of the control rod obtained from the 2-D, XY geometry Po S6 TWOTRAN transport calculations. Estimation of Core Cycle Dependent Control Rod Worth Attenuation Rate Among the three nuclides composing the absorber materials, Cd with the largest thermal absorption cross section will deplete quickly in the early phase (small number of core cycle). Then the other two nuclides, Ag and In will deplete in a more slower manner since their moderate absoption cross sections. In this period, the core cycle dependent reactivity worth of the total control rod can be expressed by a linear correlation, namely, Y = ax +b, where y and x represent the reactivity worth of the total control rod and number of core cycle (in GWOff), respectively. The decreasing rate of the control rod worth is readily represented by the constant a (in pcm/gwdff). Specifically, the coefficients a and b must be found by linear regression method. RESULTS AND DISCUSSIONS The summary of the calculation results for the bumup dependent few group macroscopic cross section is given in Tables 2 and 3 for oxide and silicide fuels, respectively. It can be observed from the tables that the nominal discharged bumup of 56 % and 63 % for oxide and silicide fuels which corresponded to 95 GWDff and 109 GWDff bumup levels were achieved in 184 and 252 days, respectively. Hence, practically the use of silicide fuel (with higher 235U loading per FE/CE) will extend the core life significantly longer. However, the reactivity bumup swings derived from the infinite multiplication factor increased from % for oxide fuel to % for silicide fuel. Therefore, since the TWC number of fuel batches is 7 then the average increase in term of infinite multiplication factor was from 1.78 % per fuel batch for oxide fuel to 2.66 % for silicide fuel. In conclusion, it can be predicted that a reasonable increase of the excess reactivity of the RSG GAS silicide core at BOC will occur. 33

6 The summary of the absorber depletion calculation results is given in Tables 4 and 5 for oxide and silicide fuels, respectively. From the tables, it can be observed that absorber the isotope 133Cdonly played an important role at the earlier stage of the absorber life. Roughly, after the 14-th core cycle the nuclide density could be neglected. The absorber isotopes Ag and In depleted relatively much slower so that in the long-term use of the absorber these isotopes were of the most importance. Next, the total control rod worth was estimated by subtracting the two values of keffcalculated by 2-D, XY geometry, Po S6 TWOTRAN option, i.e. one corresponded to all control rod withdrawn condition and the other to the inserted condition. The complete calculation results for O-th core cycle (fresh absorber) are summarized in Tables 6 and 7 for oxide and silicide cores, respectively. In the tables, additional results from 2-D, XY geometry CITATION neutron diffusion calculations were also included for future references. Two kinds of 2-D diffusion calculations were conducted, namely, one without or with correction on the absorber boundary conditions where the neutron diffusion theory was hardly valid. For the latter case, the logarithmic derivative boundary conditions were applied on the absorber interface surface through the logarithmic derivative constants but only on the thermal energy groups. Without this correction, it can be observed that significant errors in the control rod worth presumably occurred. The calculated core excess reactivity (9.8 %) of the oxide TWC was close to the design data of the RSG GAS shown in Table 1 (9.2 %). Despite of the fact that the excess reactivity of the silicide core (l1.4 %) is larger than the one of the oxide case (9.8 %) so that consequently a larger control rod worth was needed, the control rod worth for the silicide core was relatively smaller than the one of the oxide core. This was attributed from the relatively harder core spectrum of the silicide core compared to the oxide core since the heavy metal (uranium) to moderator ratio was higher for the silicide core so that the effectiveness of the (thermal) absorber decreased. Therefore, a careful consideration must be paid for designing the future RSG GAS silicide core, i.e. several design options to increase the control rod worth (shutdown margin) must be devised. Similar neutron transport calculations were done for various depletion levels of the absorber materials as function of the number of core cycle. The results were summarized in Tables 8 and 9 for oxide and silicide cores, respectively. The control rod worth and their depletion rates as function of the number of core cycle (in term of accumulated core average burnup, MWDrr) is depicted in Figures 7 and 8, respectively. As already discussed, at the early stage of the absorber life, as the cadmium depleted fast then the control rod worth decreased sharply. However, after that period the decreasing rates of the control rod worth were much slower. For the latter depletion phase, using the linear regression method, the correlation for the 34

7 control rod worth as function of core cycle number (accumulated core averaged bumup) was found to be: for oxide core and, ~p (B) = (pcm/gwdff) xb (GWDrr) , GWDrr ~ B ~ GWDrr (1 core cycle = GWDrr) ~p (B) = (pcm/gwdrr) x B (GWDrr) , (334.2 GWDrr~ B ~ GWDrr) (1 core cycle = GWDrr) for silicide core, respectively. Here, ~p and B represent the control rod worth and accumulated core averaged bumup, respectively. The regression linearities, r, for the above correlations were and for oxide and silicide cores, respectively. For operational use, the accumulated core averaged bumup unit (B) is more preferable since the RSG GAS is not always operated in its nominal core cycle length. It can be observed from the two correlations that the decreasing rate of the control rod worth for the silicide core was slightly higher than one for the oxide core. CONCLUDING REMARKS Control rod absorber depletion analyses for the present RSG GAS oxide and the future silicide cores have been conducted to estimate the control rod worth and their decreasing rates as function of number of core cycle. The RSG GAS typical working core was taken as the calculational model, and the JAERI SRAC-95 code system with its nuclear data library compiled from the JENDL-3.2 was used as the computational tool for the whole analyses. The decreasing rates of the control rod worth, derived by linear regression method, were found to be 1.37 and 1.55 pcm/gwdff for oxide and silicide cores, respectively. The calculated decreasing rates should be considered conservative or over-estimated since the core average power density was used in the absorber depletion analyses while the real control rod power densities were much lower when the absorber plates were inserted. The decreasing rates of the control rod worth are essential for predicting how long the control rod can be used in the reactor. Furthermore, the analysis results showed that the control rod worth of the silicide core (12.2 %) was relatively lower than the one of the oxide core (13.1 %) despite of the fact that the silicide core excess reactivity was higher. Hence, a careful consideration must be paid for designing the future silicide 35

8 core, i.e. several design options to increase the control rod worth or to decrease the excess reactivity must be devised. The present result indicated that a more elaborate 3-D neutron diffusion simulation using the same method with the present work should be done for a real (not TWC) core where the accurately measured control rod worth are available. ACKNOWLEDGMENT The author extends his special gratitude to Mr. Y. Nakano of the Research Reactor Technology Development Division, Department of Research Reactor, JAERl, for his constant advisory help in the present work. Fruitful discussions on various aspects of SRAC code system with Mr. T. Kugo and Mr. Okumura of the Reactor System Engineering Department, JAERl, Mr. K. Soyama, Mr. M. Kaminaga, Mr. A. Tsuruno, Mr. M. Matsubayashi, Mr. T. Amano, Mr. H. Ando and other members of the Research Reactor Department were very helpful for the authors to complete the present work. REFERENCES 1) National Atomic Energy Agency (Batan), Safety Analysis Report of RSG GAS, Rev. 8 (1987) 2) TSUCHIHASHI, K. et al., "SRAC: JAERl Thermal Reactor Standard Code System for Reactor Design and Analysis", JAERl-1285 (1983) 3) TSUCHIHASHI, K. et al., "Revised SRAC Code System", JAERl-1302 (1986) 4) NAKAGAWA, T. et al., "Japanese Evaluated Nuclear Data Library Version 3 Revision-2: JENDL-3.2",J. Nucl. Sci. Technol., 32, 1259 (1995) 36

9 Table 1. Reactor main design data ofrsg GAS (oxide core). 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 Core Characteristics No. of Fuel Elements No. of Control Elements No. of Fork Type Absorber (pairs) Nominal Cycle Length (fpd) Ave. Burn-up at BOC (% loss of 235U) Ave. Burn-up at EOC (% loss of 235U). 235 Ave. Discharge Burn-up at EOC (% loss of U) Core Excess Reactivity (%) Core Ave. Power Density (W/cc) Fuel/Control 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 U30gA Ag-In-Cd 3.38 SS

10 Table 2. Fuel element burnup calculation result for oxide fuel (Xe & Sm o free, 300 K). Time (%-MWDrr) Bumup x1O, Integrated Cony x1O 3 K;njB" x1O, x1O x1O x x1O, x1O, x1O, x10, x1O, x1O, xlO xlO, xlO (2-D k;rifb2 (cm,2s,l) (l-d x1014 X1014 Ratio PH) PH) 10'2 Tabel 3. Fuel element burnup calculation result for silicide fuel (Xe & Sm o free, 300 K). Time (%-MWDrr) Bumup xlO Integrated Cony x1O, x1O 3 KinjB x1O x1O, x1O x1O, x1O x1O, x1O, x x1O x1O,2 (2-D x xlO, k;njb (cm,2s,l) (l-d 172x X1014 PH) Ratio PH) '2 38

11 o Table 4. Absorber depletion calculation result for oxide fuel (300 K). 0 Core Accum. Ave x10' x1Q" xlO'l xlO' x10,J) xlO' xlO' x10' xlO-=T 2.301xlO, xlO'2 Ag-I093) (bamlee 2.063xlO, xlO, x x10, xlO' xlO, xlO' xlO' x10" 3.864xlO" 5.661x10' xlO j Burnup 7.217x10' x10" 147xlO'2 178xlO,2 115xlO-=T In-1I5 Cd-1133) (bamlcc) 188xlO 2 132xlO,2 Core Ag-1073) 162xlO" 180xlO" 4) ) j) I) One core cycle equal to 0-8 % burnup period (=nominal core life time, approximately days). o Table 5. Absorber depletion calculation result for silicide fuel (300 K). 0 Core Accum. Ave xlO' xlO xlO'I xlO' x10-IU xlO xlO' xlO' xlO xlO-:T 2.083xlO xlO, xlO'2 (bamlee 2.1I2xlO 2 Ag-I093) 2.188xlO' xlO xlO xlO' xlO' x10' xlO, xlO' xlO, xlO" 5.728x 3.457xlO' xlO" 3.971xlO ' Burnup 7.067xlO' x10" 6.382xlO" 124xlO-:T In-1I5 Cd-l13 (bamlcc) Core Ag-107 4) 10'3 ) j).') 3) 1) 2) 3) 4) One core cycle equal to 0-9 % burnup period (=nominal core life time, approximately days). Total control rod worth. Nuclide densities in the Ag-In-Cd absorber meat. The depletion calculation tor Cd-113 was not continued since its nuclide density was negligible. 39

12 Table 6. Calculation result for control rod worth of RSG GAS TWC oxide o core (BOC, Xe & Sm free, 300 K, fresh absorber) Codes, Geometry CRs ( ( All-Down keff(p) ( %) L'lp and -(%) ( CRs keff(p) All-Up %) ( %) em) 2 I), XY ( %) (Log. Derivative BC not Used) I) Logarithmic derivative boundary condition for absorber rod was only imposed on thermal energy groups. Table 7. Calculation result for control rod worth of RSG GAS TWC silicide o core (BOC, Xe & Sm free, 300 K, fresh absorber) Codes, Geometry CRs L'lpAll-Down -(%) keff(p) ( ( ( and ( %) CRs ( keff(p) All-Up %) %) em) 2 I) ( %), XY (Log. Derivative BC not Used) I) Logarithmic derivative boundary condition for absorber rod was only imposed on thermal energy groups. 40

13 Table 8. Control rod worth as function of core cycle for TWC oxide core o (BOC, Xe & Sm free, 300 K). 0 Number of Core Accumulated (pcm Control Ave. Of~p (%) /GWDff) Decreasing Bumup Rod~pWorth-Rate Table 9. Control rod worth as function of core cycle for TWC silicide core o (BOC, Xe & Sm free, 300 K). 0 Number of Core Accumulated (pcm Control Ave. Of~p (%) /GWDff) Decreasing Bumup Rod~pWorth-Rate 41

14 1 o 0 u; g Ii, I, 1 1 I I I I I I I :~ 4.5 L,.3 r 11= ~" ~ m~ ~."-nmm1t~~ Figure 1. RSG GAS standardfuel element (unit mm).!~i t----- ~ : ' ---- ~ 1 I " 1 'I ?d Figure 2. RSG GAS control fuel element with absorber blades inserted (unit mm). 42

15 Grid IP-I IP-416 IP-3 CIP27 IP FE-18 FE-24- FE-13 FE-27 FE-29 56FE-12 BE0 FE-JO FE FE-14 FE-16 FE-28 FE CE-4 FE-30 FE-36 FE-37 FE-39 FE-5 FE-6 FE-8 FE-40 FE FE lO CE-5 CE-2 CE-3 CE-6 CE FE-32 CE-8 9FE FE CE-I FE-19 FE FE-20 FE-22 18FE-l Figure 3. RSG GAS typical working core configuration for BOC (2-nd and 3-rd rows represent the FE/CE bumup levels for oxide and silicide cores, respectively; BE: berrylium reflector element; IP: irradiation position; CIP: central irradiation position). 43

16 107 ~ 1-D PIJ, PEACO CELL BURN-UP meat, clad, moderator 107 G, 2-D PIJ, HOMO active fuel region, side plates, water gaps 107 G, 1-D PIJ, PEACO CELL BURN-UP absorber, fuel plates Figure 4. Bumup, core cycle dependent, fine group cross section sets generation for fuel and absorber materials. 107 G, 2-D, XY CITATION Space Group Collapsing TWC, BOC, CRs All-Down Reactor Structural Material Compositions 107 G, 2-D, XY CITATION Space Group Collapsing TWC, BOC, CRs All-Up Figure 5. Final few group cross section sets generation for fuel, absorber and reactor structure materials. 44

17 10 G, 2-D, XY TWOTRAN TWC, BOC, CRs All-Down Figure 6. Determination of total control rod worth by whole reactor neutron transport calculations. -cf!. ~ en -.c-'- "0 () a: '-C Silicide ] 5 10 Accumulated Core Ave. Burnup (105MWOrr) Figure 7. The control rod worths as a function of core average bumup for the present oxide and silicide cores. 45

18 en - 0 ' > (!) t a..l:: 10 ~en ro 56 CD a: c: CD 0) I E (,,) Q) IOxide '42 ;9 7.\ ~ilicide 14~ f Accumulated Core Ave. Burnup (105MWOm Figure 8. The depletion rate of the control rod worths as a function of core average bumup for the present oxide and silicide cores. 46