Mixed-oxide (MOX) fuel performance benchmarks

Transcription

1 Mixed-oxide (MOX) fuel performance benchmarks L. J. Ott a,*, Terje Tverberg b, Enrico Sartori c a Oak Ridge National Laboratory, Oak Ridge, Tennessee, U.S.A. b OECD Halden Reactor Project, Halden, Norway c OECD/NEA Data Bank, Paris, France Abstract Within the framework of the OECD/NEA Expert Group on Reactor-based Plutonium disposition (TFRPD), fuel modeling code benchmarks for MOX fuel were initiated. This paper summarizes the calculation results provided by the contributors for the first two fuel performance benchmark problems. A limited sensitivity study of the effect of the rod power uncertainty on code predictions of fuel centerline temperature and fuel pin pressure also was performed and is included in the paper. 1. Introduction The OECD/NEA Nuclear Science Committee has established an Expert Group that deals with the status and trends of reactor physics, nuclear fuel performance, and fuel cycle issues related to the disposition of weapons-grade plutonium as MOX fuel. The objectives of the group are to provide NEA member countries with up-to-date information on, and develop consensus regarding, core and fuel cycle issues associated with weapons-grade plutonium disposition in thermal water reactors (PWR, BWR, VVER-1, and CANDU) and fast reactors (BN-6). These issues concern core physics, fuel performance and reliability, and the capability and flexibility of thermal water reactors and fast reactors to dispose of weapons-grade plutonium in standard fuel cycles. The activities of the NEA Expert Group on Reactor-based Plutonium Disposition are carried out in close cooperation with the NEA Working Party on Scientific Issues in Reactor Systems (WPRS), and in most cases jointly. An major part of these activities includes benchmark studies. MOX fuel behavior benchmarks finalized or in progress are as follows: Halden Reactor Project (HRP) MOX fuel irradiation experiment benchmark (completed); Belgonucléaire (BN) and SCK CEN PRIMO ramped MOX fuel rod performance benchmark (nearly finalized); U.S. Department of Energy weapons-grade MOX fuel irradiation experiment irradiated at the advanced test reactor (ATR) benchmark (started); MOX fuel rod behavior in fast power pulse conditions (started). The following benchmarks relative to the reactor physics activities of the Expert Group are completed or in progress: VENUS-2 MOX core benchmarks, carried out jointly with the WPRS (completed); * Corresponding author, ottlj@ornl.gov, Tel: (865) ; Fax: (865)

2 VVER-1 LEU and MOX benchmark (completed); KRITZ-2 benchmarks, carried out jointly with the WPRS (completed); benchmark using dosimetry data from the VENUS-2, MOX core experiments (completed); VVER-1 in-core self-powered neutron detector calculational benchmark (started); VENUS-7 weapons-grade MOX core benchmark (started). This paper describes the results of the first two benchmarks relative to MOX fuel behavior. The corresponding experimental data have been released, compiled, and reviewed for the International Fuel Performance Experiments (IFPE) Database. 2. OECD Halden Reactor (HRP) MOX Fuel Benchmark (Tverberg, 27) 2.1. Irradiation data The blind benchmark exercise was performed on a data set provided by the OECD Halden Reactor Project (HRP) of two short MOX rods (one hollow and one solid). The rods were instrumented with fuel thermocouples (TF) and internal rod pressure transducers (PF) and irradiated in the Halden Boiling Water Reactor (HBWR). Details of the fuel fabrication data as well as information about the provided irradiation data (for a period of 626 irradiation days) are provided by Tverberg (27). The rod average linear heat generation rate (ALHGR), including rod power uncertainty of 5 1%, is illustrated in Figs. 1 and 2 for Rod 1 (solid pellets) and Rod 2 (hollow pellets), respectively. Rod 1 had an active fuel length of 224 mm and consisted of 17 solid fuel pellets and 4 annular pellets (that allowed the insertion of a fuel centerline thermocouple); the fuel stack in Rod 2 was 22 mm in length with all 21 pellets being annular in design (again allowing a fuel TF). Of special note in the power histories is the short power spike at ~17 irradiation days followed by a ~5 day period with slightly higher power than before, during which notable gas release was observed in the rods. Also notable is the power increase towards the later stages of irradiation (at about ~525 days), which caused large gas release in both rods. The internal pressure, measured in the rods during rod puncturing after irradiation, was in good agreement with the end-of-life (EOL) pressure from the in-pile measurements shown in these figures. The ALHGRs and the combined uncertainties resulting from the 5% experimental uncertainty of an in-pile power calibration and the sensitivity of neutronics calculations to changes in local surroundings are plotted for Rod 1 and Rod 2 in Figs. 1 and 2, respectively. The estimated error increases with burnup from the initial 5% (power calibration uncertainty) to approximately 1% at the end of the data set. The results of a limited sensitivity study of the effects of a ±5% power uncertainty on the computed fuel pin centerline temperature and fuel pin internal pressure are presented in Section 2.4. Fig. 1. ALHGR for Rod 1 (solid) including power uncertainty. 2

3 Fig. 2. ALHGR for Rod 2 (hollow) including power uncertainty Contributing organizations and codes The following countries and organizations, and the respective fuel modeling codes that they used, have provided contributions to the benchmark exercise. Organization Nexia Solutions (UK) KAERI (Korea) Kurchatov Institute (RF) ORNL (USA) SCK CEN (Belgium) VNIINM-Bochvar (RF) 2.3. Calculational results Fuel modeling code ENIGMA FRAPCON-3 TRANSURANUS FEMAXI-V START-3 For this paper, only the data and calculational results for Rod 2 will be illustrated. Tverberg (27) summarizes the major modeling differences (fuel thermal conductivity and fission gas release models) in the fuel performance codes (Section 2.2) applied to this benchmark exercise. The predicted fuel temperatures for Rod 1 are generally less than the measured temperature prior to the high power period starting at ~525 irradiation days. During the high power period, all predictions (except for the FEMAXI-V results) are higher than the measured temperatures. At the beginning of life (BOL), the code predictions bracket the measured temperature for Rod 1 with a spread of ~14 C; this range quickly (by ~5 days) decreases to less than 6 C and remains less than 1 C spread through the period prior to the power increase at 525 days. After the power increase, the calculations bracket the measurements with a range of ~15 C at 545 days to a range of ~2 C at 62 days. For Rod 2, see Fig. 3, the calculated temperatures are more closely bunched throughout the irradiation. Again, through ~23 days, the predictions are below the measured temperatures, with the spread ranging from ~11 C at BOL to ~75 C at 215 days. After the power reduction at ~23 days and through ~525 days, all code predictions very closely match the measured response with spreads of less than 6 C. After the power increase at 525 days, the calculations again closely match the measured response with a spread of less than 12 C (excluding Frapcon-3, the predictions are very close to the measurements). The predicted and measured fuel pin pressures are illustrated for Rod 2 in Fig. 4. Initially (at BOL), the code predictions can be split (roughly) into two groups with a difference of bar in the predicted pressure response. The bottom grouping consists of ENIGMA,, and TRANSURANUS, and the top grouping is, FRAPCON-3 and FEMAXI-V. The top group does reflect the pressure increase due to the power spike at 17 days for Rod 1 (but not for Rod 2). In general (except for the predictions of and FRAPCON-3), the codes do not properly predict the fuel pin pressure increase after the power increase at 525 days. 3

4 Rod 2 Fuel Centerline Temperature Measured, Rod 2 BNFL KAERI Kurchatov SCK-CEN ORNL-FRAPCON-3 (v1.3) ORNL-TRANSURANUS Temperature ( C) Fig. 3. Rod 2 (hollow) experimental and calculated fuel centerline temperature. Rod 2 Fuel Pin Pressure 3 25 Measured, Rod 2 BNFL KAERI Kurchatov SCK-CEN ORNL-FRAPCON-3 (v1.3) ORNL-TRANSURANUS 2 Pressure (bar) Fig. 4. Rod 2 (hollow) experimental and calculated fuel pin pressure Sensitivity results In order to determine the effect of the rod power uncertainty on the calculated fuel centerline temperature and fuel pin pressure, a sensitivity study was performed at the ORNL. The basis for the sensitivity study is the uncertainties in the ALHGRs, which show an uncertainty varying from 5% initially to 1% at EOL. In the ORNL calculations, however, a 5% uncertainty is assumed throughout the irradiation. The code results presented here are the predictions from the FRAPCON-3 code. The predictions in the referenced figures are for nominal power and nominal power ±5%, and these predictions are plotted along with the measured responses. Figures 5, 6, and 7, respectively, illustrate the calculated results for the fuel centerline temperature, fuel pin internal pressure, and the fuel fission gas release [along with the EOL post-irradiation examination (PIE) rod puncture results] for Rod 2. General conclusions on power uncertainty effects are as follows: Without the FGR feedback, an uncertainty of ±5% in the power generation results in an uncertainty of <±5% in the calculated fuel centerline temperature. 4

5 Rod 2 Centerline Temperature Temperature ( C) Experimental Data Frapcon-3 (nominal LHGR) Frapcon-3 (.95*LHGR) Frapcon-3 (1.5*LHGR) Fig. 5. Rod 2 (hollow) experimental and calculated fuel centerline temperature. Rod 2 Internal Pressure 3 Experimental Data 25 Frapcon-3 (nominal LHGR) Frapcon-3 (.95*LHGR) Frapcon-3 (1.5*LHGR) Pressure (bars) Fig. 6. Rod 2 (hollow) experimental and calculated fuel pin pressure. Rod 2 Calculated Fission Gas Release Frapcon-3 (nominal LHGR) Frapcon-3 (.95*LHGR) Frapcon-3 (1.5*LHGR) Experiment 12 FGR (%) Fig. 7. Rod 2 (hollow) experimental and calculated fuel fission gas release. 5

6 Realistically, the temperature differences (produced by ±5% power uncertainty) significantly affect the FGR models and FGR, which in turn affects the pellet-to-clad gap conductance. This can result in: 1 to +15% range on the computed fuel centerline temperature and 3 to +4% range on the computed fuel rod pressure. 3. Belgonucléaire (BN) and SCK CEN PRIMO MOX Fuel Benchmark (Ott, 28) 3.1. Irradiation data The program PWR Reference Irradiation of MOX Fuels (PRIMO) was started in October It was jointly organized by SCK CEN (Studiecentrum voor Kernenergie Centre d Etudes de L énergie Nucléaire) and BN (Belgonucléaire) and was co-sponsored by ten participants including fuel vendors, utilities, nuclear centers, and national authorities. The PRIMO program was an investigation on MOX fuel, with the following major objectives: irradiation of MOX fuel rods to different burnup stages, following power histories representative of those of PWR power plants, to determine their behavior as far as their mechanical, thermal, and neutronic properties are concerned; execution of a ramp test program, to find out the failure threshold of MOX fuel rods and to obtain mechanical and thermal data under ramp conditions; and fast power transients on a MOX rod to simulate a class II incident in a PWR. BN and SCK CEN provided the PRIMO data (fabrication and irradiation) for rod BD8 for use by the Expert Group as a MOX fuel performance benchmark. Rod BD8 was base irradiated in the BR3 reactor of SCK CEN (cycles 4D1and 4D2 for a total of 746 days at power). The average rod burnup reached 3.1 GWd/tM, corresponding to a peak pellet burnup of 38. GWd/tM. After base irradiation in the BR3 reactor, the BD8 rod was transported to CEA-Saclay for irradiation in the OSIRIS reactor. The irradiation device used for the PRIMO ramps was the ISABELLE 1 loop. Rod BD8 was preconditioned in the ISABELLE loop at a peak power level of 189 W/cm with a hold time of 27 hours. The subsequent power excursion rate amounted to 77 W/(cm min), reaching a terminal peak power level of 395 W/cm that lasted for 2 hours. Details of the fuel fabrication data as well as information about the provided irradiation data are given by Ott (28) Contributing organizations and codes The following countries and organizations, and the respective fuel modeling codes that they used, have provided contributions to the benchmark exercise. Organization Fuel modeling code KAERI (Korea) Kurchatov Institute (RF) ORNL (USA) FRAPCON-3 (v. 1.3 and 2.) TRANSURANUS PSI (Switzerland) FALCON SCK CEN (Belgium) FEMAXI-V VNIINM-Bochvar (RF) START Calculational results Rod BD8 was not instrumented; therefore, only calculated results of the benchmark participants are presented in this paper. The PIE of BD8 (and a sister rod which did not undergo the power ramp) does provide fission gas release data. A comparison of the calculated average burnup for Rod BD8 (from the submittals) and the fuel burnup computed by SCK CEN using CONDOR and the PIE determination of burnup is essentially the same curve (no illustration needed). Basically, the calculated rod average burnups indicate that all benchmark participants are modeling the rod geometry and heating histories (including the axial power distribution) correctly. The calculated fuel midplane centerline temperatures (prior to the ramp) are illustrated in Fig. 8. Through ~253 days of irradiation, all code predictions are within an ~125 C band; after 253 days, the Russian Federation (RF) codes ( 6

7 Calculated Rod Midplane Centerline Temperature FRAPCON-3 v1.3 Halden Criteria FRAPCON-3 v2. TRANSURANUS (v1m3j4) Start-3 FEMAXI-V FALCON Temperature ( C) Fig. 8. PRIMO Rod BD8 calculated fuel midplane centerline temperature, prior to ramp. and START-3) yield results that are higher and diverge from the remaining codes (whose results are within an ~1 C band). Prior to the ramp at the end of cycle 4D2, the RF codes predict fuel temperatures of 113 C () and 1175 C (START-3), while all other code predictions are within C. The peak fuel temperatures during the ramp phase of the test are given in Fig. 9; except for the START-3 prediction, all code predictions are within an ~25 C band. The calculated fuel pin internal pressures (preramp) are given in Fig. 1. As in the first benchmark, there are two distinct groupings with ~1 MPa difference in the predictions. Figure 11 shows the calculated and experimental FGR for Rod BD8. For the base irradiation in BR3, the experimental FGR value is.5% (from a sister rod of BD8 which did not undergo the ramp phase). The RF code predictions for the FGR during the base irradiation in BR3 are all greater than 2%, and the FALCON results are ~1.5%; all other code simulations are close to the experimental value. The PIE of BD8 (after the ramp) yielded a FGR of 11.2%. The START-3 and FALCON (second simulation) predict the highest values of 15.4 to 16.2%; the lowest prediction is that of TRANSURANUS at 4.7%. All other code predictions are within the range of 9.1 to 13.7%. Calculated Midplane Rod Centerline Temperature Temperature ( C) Halden Criteria FRAPCON-3 v1.3 FRAPCON-3 v2. TRANSURANUS (v1m3j4) Start-3 FEMAXI-V FALCON FALCON Fig. 9. PRIMO Rod BD8 calculated fuel midplane centerline temperature during ramp. 7

8 Calculated PRIMO Rod BD8 Internal Pressure 18 Rod Pressure (MPa) FRAPCON-3 (Version 1.3) Cold TRANSURANUS (v1m3j4) FRAPCON-3 (Version 2.) FEMAXI-V (model 1) FEMAXI-V (model 2) FALCON FALCON Fig. 1. PRIMO Rod BD8 calculated fuel pin internal pressure. Calculated PRIMO Rod BD8 Fission Gas Release 18 FGR (%) PRIMO Estimated (base) PRIMO Estimated (ramp) FRAPCON-3 v1.3 FRAPCON-3 v2. TRANSURANUS (v1m3j4) Start-3 FEMAXI-V (model 1) FEMAXI-V (model 2) FALCON FALCON Fig. 11. PRIMO Rod BD8 experimental and calculated fuel fission gas release. 4. Summary The OECD/NEA Nuclear Science Committee has established an Expert Group that deals with the status and trends of reactor physics, nuclear fuel performance, and fuel cycle issues related to the disposition of weapons-grade plutonium as MOX fuel. The activities of the NEA Expert Group on Reactor-based Plutonium Disposition are carried out in close cooperation with the NEA Working Party on Scientific Issues in Reactor Systems (WPRS). An eminent part of these activities include benchmark studies. MOX fuel behavior benchmarks that are finalized and summarized in this paper: Halden Reactor Project (HRP) MOX fuel irradiation experiment benchmark (completed); Belgonucléaire (BN) and SCK CEN PRIMO ramped MOX fuel rod performance benchmark (nearly finalized). This paper describes the results of these first two benchmarks relative to MOX fuel behavior. The corresponding experimental data have been released, compiled, and reviewed for the International Fuel Performance Experiments (IFPE) Database. 8

9 References Ott, L., 28. Mixed-oxide (MOX) Fuel Performance Benchmark: Summary of the Results for the PRIMO MOX Rod BD8, in draft form to be NEA/NSC/DOC report, to be published in late 28. Tverberg, T., 27. Mixed-oxide (MOX) Fuel Performance Benchmark: Summary of the Results for the Halden Reactor Project MOX Rods, NEA/NSC/DOC(27)6. 9