IAEA Research Contract No R0. Application of the TRANSURANUS-WWER Version Code in the FUMEX-III Project

Transcription

1 IAEA Research Contract No R Application of the TRANSURANUS-WWER Version Code in the FUMEX-III Project Institute for Nuclear Research and Nuclear Energy Sofia, Bulgaria Chief Scientific Investigator - Stefka Boneva 6 October 28 6 July 29 1

2 Table of contents 1. Introduction Description of the experiment Base irradiation in Kola NPP, unit Tests in the MIR reactor Techniques of the re-fabrication and instrumentation Techniques of the LHR determination Tests description TRANSURANUS simulations and results Base irradiation modelling Tests simulation RAMP test - simulation and results FGR-1 test simulation and results FGR-2 test - simulation and results Post Irradiation Examination Puncturing results Fuel microstructure Conclusions References: Introduction The present progress report presents a part of the research work, done in the Research Contract RC 15164/R, at the Institute for Nuclear Research and Nuclear Energy (INRNE) in the frame of the CRP FUMEX-III supported by IAEA. According to the working program of the project INRNE had to analyse the fuel performance of WWER rods included into the CRP FUMEX-III. During the first 9 months of the year the team has worked on MIR ramp tests on Kola3 rods (Kola3- MIR experiment) by using the latest TRANSURANUS-WWER version v1m1j9 [1] on the basis of the IFPE-OECD/ IAEA-NEA database [2]. Fuel rods included in the tests have been operated under normal conditions at Kola NPP up to maximum burnup of about 5 6 MWd/kgU. Nine re-fabricated rods have been cut from selected FA s and carried out under single ramp conditions (RAMP test) and 2 step-by-step power increase tests with instrumented rods. The goal of the study was: to prove the applicability of the TRANSURANUS-WWER version code to fuel performance under transient conditions; 2

3 to study TRANSURANUS code predictions of WWER fuel rods under normal and off-normal conditions (ramp tests); to demonstrate the dependence between the burnup accumulated in the base irradiation prior to the ramp tests and the changes of the fuel rod structure and geometry fuel and cladding; The attention was concentrated on: fuel central temperature during the ramp irradiation (FGR2-test). Comparison between TRANSURANUS assessments and thermocouple records; pin pressure (FGR1-test). In-pile pressure measurements for two rod with different accumulated burnup during base irradiation; distribution. fission gas release and gas mixture (from PIE); microstructure changes of the fuel-central hole closing and porosity The calculations were done for base irradiation as well as for the test irradiation, using restart option of TRANSURANUS code for account rod cutting and refilling. 2. Description of the experiment In , a number of over power transient tests were carried out in the MIR reactor (SSC RIAR, Dimitrovgrad) using high burn-up WWER-44 fuel. Fuel rods used were from FA-198 and FA-222 and had operated under normal conditions at Kola NPP Unit 3 during 4- and 5 fuel cycles up to a maximum burn-up of ~ 5 and 6 MWd/kgU. Details of the design and main operating conditions of the NPP Kola-3 during 4-5 fuel cycles are given in [2].The objectives of the experiments were to determine the power ramps influence on fuel rod behaviour, to evaluate the threshold value of linear power for enhanced fission gas release, and to study the dependence between the linear power, temperature, structure and properties of the fuel [3, 4, 5]. The experiment Kola-3-MIR covers data on: pre-test initial rod state (fuel burnup, pin gas pressure, fuel-cladding gap, rim layer presence, mechanical properties of the cladding); test conditions (linear power values, power increase rate, power ramp magnitude, hold up time at maximum power); in-pile data of the pin pressure during transient test for 2 rods; in-pile data of the fuel central temperature, measured by thermocouples for two rods with different burnup accumulated during the base irradiation post irradiation examination of 7 rods ( FGR and gas mixture). 3

4 fuel microstructure of four rods after FGR1 and RAMP tests Base irradiation in Kola NPP, unit 3 The operating conditions of the FA-198 (Rods 2, 76 and 99) and FA-222 (rods 2, 3, 5, 6, 25 and 46) have been recorded in ASCII data files time, linear heat rate, outer clad temperature and fast neutron flux. The axial distribution of all data is presented for 21 slices (uniform separation). Before re-fabrication, the rods were examined by non-destructive examinations [6] (γ-scanning, outer diameter measurement, eddy-current measurements of the cladding, X-raying) and the results are presented in the table. 2.1 below. Table 2.1 The state of the fuel rods after base irradiation. Parameter ~5 MWd/kgU ~6 MWd/kgU Fuel pellet state Fuel pellets are divided into 6-8 parts, but the initial configuration is preserved Central hole diameter, mm 1.6 Average grain size, µm 6-8 Rim-layer width, µm ~1 ~12 Fuel swelling, % Fission gas release, %.5-1. ~3 Pin gas pressure, MPa Fuel-cladding gap, µm Outer cladding diameter decrease in the central part, µm Oxide film thickness on the outer cladding surface, µm Oxide film thickness on the inner cladding surface, µm The maximum and averaged over the whole length burnups for the 9 fuel rods after base irradiations were defined by Cs 137 and given in the database. The temperature of the fuel column centre during base irradiation does not exceed 11 o C according to code calculations [7], and 15 o C according to destructive PIE [8]. The fuel column maintained its integrity and the pellet fragmentation is insignificant (six to eight fragments). The grain size and the shape do not differ from the initial ones. 4

5 2.2. Tests in the MIR reactor Techniques of the re-fabrication and instrumentation Fuel rod fragments with maximum burnup were cut for further investigation. The re-fabrication process consisted of the following main steps: cutting the fuel rods into fragments; removing the fuel pellets from the fragment ends; assembling the fragment of the fuel rod with the tailpieces; thermovacuum dry out of the fuel rod fragment at temperature of 3 C and vacuum of 1-1 mm Hg for 3 minutes; helium filling at temperature of C; pressurization at MPa. The thermocouple (wolfram-rhenium, diameter of 1.2 mm) was located in the central fuel hole at about 1 mm from the fuel column top. The pressure detectors were located in the upper part of the re-fabricated fuel rod with the mechanicalelectrical pressure transducers. The total error of pressure measurements was ±.4 MPa, and the fuel temperature determination error was ± 2% Techniques of the LHR determination The required level of linear heat rate (LHR) during the tests was reached by the total reactor power increase. The total fuel rod power was monitored by thermal balance taking into account the thermal losses and heating in the construction materials of the loop channel. Dynamics of the fuel rod power change during ramp was varied by: reactor power change; local channel power change with absorbers; combination of the above methods. The error in determination of the maximum LHR was ± 6-8%, and of the fuel rod cladding temperature - ± 6 C. The gamma peak of 14 Ba (E= kev) was used as a reference nuclide (during RAMP- and FGR1-tests) to determine the LHR axial distribution. During the test experiment FGR-2, 95 Zr was used as a reference nuclide. 5

6 2.2.3 Tests description RAMP test Three non-equipped re-fabricated fuel rods (numbers 33, 37 and 38) were tested under power ramp conditions. The re-fabricated fuel rods are of identical height (95mm) and with about the same free volume and are pressurized up to 1.1 MPa. The RAMP test can be divided into 3 stages: Stage 1 - irradiation at initial power level (12-17 kw/m) of duration ~346 hours; Stage 2 - power ramp from initial to maximum level (duration ~23 min); Stage 3 - hold stage after ramp (27-36 kw/m) of duration ~17 hours. The coolant pressure during the RAMP test was held at a level of 12.5 MPa. All data for these 3 re-fabricated rods, namely - the initial parameters (pin geometry, free volume, pin pressure), and the test conditions (LHR, fast neutron flux and cladding temperature) as well as the post irradiation examination results are available in IFPE database. Depressurization of the re-fabricated fuel rods was not observed in the RAMP test FGR-1 test The step by step power increase experiment, FGR-1, included three re-fabricated rods: one (No41) with accumulated burnup Bu~5MWd/kgU, and two rods - No32 and No48 with Bu~6MWd/kgU. Rods 41 and 48 were instrumented with pressure transducers, rod 32 was not instrumented. The total test cycle in the experiment FGR-1 was carried out in two stages. Each stage consisted of several steps. Stage one was divided in steps 1, 2 and 3, while stage two was divided in 4 steps -1A, 2A, 3A and 4A. The irradiation parameters, such as power, coolant pressure, outer cladding temperature and fast neutron flux in 11 sections are included in the database. The maximal power of the test irradiation was reached at the step 4A. The hold-up time at maximum power was 94 hours. After the first stage, the pressure sensor of the rod 48 broke down. The pressure transducers registrations, as well as puncturing results of fission gas release, are available in IFPE database FGR-2 test The FGR-2 experiment was the second of the tests series of WWER fuel investigation at different power levels. Test rig included three re-fabricated fuel rods: 5, 51 and 52 (fuel column high=4 mm and filling gas pressure 1.2 MPa). Rods 5 6

7 and 51 were instrumented with thermocouples, inserted in the pellet central hole (thermocouples coordinate x= 1 mm from the top of the fuel stack). Rod 52 was not instrumented. The FGR-2 test was carried out in two stages. Each stage was performed in four steps. The range of the linear power rate increase is kw/m and the hold-up time at maximum power was 74 hours. The thermocouples recorded data, together with the heat linear rate data evolution and the coolant pressure changes are included in the database. The three re-fabricated rods were leak-tight after test FGR-2. During disassembling of the test rig, rods 5 and 51 were damaged, and puncturing results of fission gas release and gas mixture only for the rod 52 are available. The data for all re-fabricated rods in the three experiments are presented for 11 sections along the fuel stack (uniform separation). The section one is the fuel column bottom, the section 11 is the fuel stack top. Results of metallographic analysis are given in the database, too. 3. TRANSURANUS simulations and results 3.1. Base irradiation modelling Using the available data and our experience from KOLA3 simulation [9], inputs for TRANSURANUS code calculations with 21 equidistance slices of the rods from FA-198 and FA-222, were prepared. This simulation with the whole length of the rods was performed in order to obtain integral results and to compare them with the experimental ones. The power history, the fast neutron flux, the cladding outer temperature and the general system pressure were prescribed in the macro step block of the input model. The model options comprised the standard TU recommendations for simulation of WWER fuel rods. UO 2 material properties were modelled by standard TU models for LWR, including the standard correlation for thermal conductivity of the fuel (ModFuel(6)=21), accounting for the local porosity. Standard WWER cladding material correlations, models and options for the TRANSURANUS-WWER version v1m1j9 were selected. Irradiation growth of Zr1Nb cladding was modelled by correlation 25, depending on the fast neutron flux and texture factors, given in KOLA- 3 database [2]. The fuel densification at the first stage of operation was taken according the recommendation in IFPE database, 1.2%. 7

8 Table 3.1 Comparison of the post base irradiation experimental results with the calculated ones for the whole length rods Parameter ~5 MWd/kgU ~6 MWd/kgU Calculated Measured Calculated Measured Rim-layer width, µm. ~ 15 ~1 ~ 7 ~12 FGR, % ~ ~3 Pin gas pressure, MPa ~.8 1. ~ Fuel-cladding gap, µm ~ ~ Outer cladding diameter decrease in the central part, µm Oxide film thickness on the outer cladding surface, µm ~ The base irradiation simulation results are reasonable in comparison with experimental ones and the same options are used by base irradiation simulation of the truncate pieces of rods before ramp. It should be noticed that results of mother rods modelling by TRANSURANUS code version v1m1j9 are somewhat different than the results in [1] when version v1m1j6 was applied. The RIM zone width (fully developed) is considerable smaller Tests simulation The test rods have different lengths of the fuel column and different fill gas pressures. The PIE after tests determines the axial distribution of LHR, fast neutron flux and cladding outer temperature for 11 fuel nodes. The base irradiation and the ramp test were simulated by applying the same fuel geometries and models. Boundary conditions for two reactors (KOLA and MIR) are including too. The time dependent part of the input files included both - the base irradiation and the test power histories. Using the restart options of TU code the restart file was modified to include the changes in the filling gas, pin pressure and number of cracks in the fuel pellets ( from 6 to 15). TU version v1m1j9 was broadened with a model for enhance release of grain boundary gas inventory under ramp conditions. This model was chosen by option igrbdm=3. In this simulation the new implemented model of linear heat rate radial distribution in WWER fuel pellet was tested. The difference between WWER and LWR type of fuel rod linear heat rate radial distributions is illustrated on Fig. 3.1 by comparison of the calculated fuel central temperatures (WWER and LWR) in the most loaded slice. 8

9 Test FGR-1, Rod Fuel Central Temperature ( o C) LWR WER Radial Position (mm) Fig. 3.1.Radial distribution of the FCT according to two TRANSURANUS models of radial distribution of the linear heat rate. The results of the base irradiation TRANSURANUS simulation with this set of option are summarized in the table below. Table 3.2. The main parameters (experimental and calculated) of the refabricated fuel rods after base irradiation simulation and before gas refilling. RFR number Equip. Burnup, MWd/kgU Outer cladding diameter, mm Pellet-cladding gap (radial), µm Exp. Calc. Exp. Calc. Exp. Calc. RAMP: 33 (198) No (222) No (222) No FGR-1: 41 (198) PT (222) No (222) PT FGR-2: 51 (198) TC* (222) TC* (222) No There is satisfactory agreement between data given by the supplier and the data after base irradiation simulation by the TRANSURANUS code. 9

10 3.2.1 RAMP test - simulation and results The RAMP test consists of 3 stages irradiation in initial power (346 hours); power ramp to maximum power (23 min); hold stage after ramp (17 hours). Three rods with different accumulated burnup during base irradiation, with equal length and He filling pressure are irradiated. Fig. 3.2 and Fig. 3.3 show the linear heat rate of the three rods during the ramp test and the TRANSURANUS prediction of the fuel central temperature and fission gas release. Maximal Linear Heat Rate (kw/m) Rod 33 Rod 37 Rod 38 Rod 33 Rod 37 Rod 38 RAMP Test Time (h) LHR FCT Fuel Centre Temperature ( o C) Fig Linear heat rate for 3 rods in RAMP experiment. On the right ordinate the fuel central temperature is presented. 1

11 RAMP Test Fission Gas Release (%) Rod 33, 5.8 MWd/kgU Rod 37, 6.1 MWd/kgU Rod 38, 6.2 MWd/kgU Time (h) Fig. 3.3.The calculated fission gas release during base and ramp irradiation. These rods were not instrumented and only puncturing results of FGR are available. The TRANSURANUS predictions of FGR for the fuel rods with initial burnup 6MWd/kgU revealed weak sensibility to the ramp in linear heat rates up to the maximum heat rate kw/m. Rod 33 with initial burnup of 5 MWd/kgU (maximum linear heat rate during ramp of >35 kw/m and FCT ~ 14 o C) shows fission gas release increasing, but it is still far from the experiment (31.3%, see Table 3.3). Fission gas release was simulated according URGAS algorithm with diffusion coefficients of Hj.Matzke (thermal) and athermal diffusion coefficient according to data of R.White. For intergranular fission gas release model of burst release of the gas grain boundary inventory due to micro-cracking was included (igrdbdm =3). In this model complete release from grain boundaries is considered when thresholds for local temperature (T loc >15(1-bu loc /8)) and change of LHR ( q >3.5kW/m) in one time-step are fulfilled. TRANSURANUS code takes account of the fission products from the high burnup structure. Details on the physical ground can be found in [11]. The high burnup phenomenon concerning FGR process was included in tests simulations with threshold for the start of enhance release from RIM zone of 85 MWd/kgU. The authors of the Kola3-MIR experiment [4,12,13] established linear heat rate threshold of the gas release enhancement. It is different for the rods with different burnups and is ~ kw/m for the fuel with burnup of 5 MWd/kgU and ~ kw/m for the fuel with burnup of 6 MWd/kgU. 11

12 3.2.2 FGR-1 test simulation and results Two rods with different initial burnup and equal fuel stack length (75mm) were instrumented with pressure transducer and the third was not instrumented. Further on, prior to the test, rod 41 (Bu max = 48.9 MWd/kgU) and rod 48 (Bu max =6.5 MWd/kgU) were pressurized by refilling up to 2 and 1.1 MPa respectively. Rod 32 (Bu ~6 MWd/kgU) with longer fuel stack (95 mm) was pressurized up to 1.1Mpa. During the test FGR-1, the coolant pressure in the reactor was changed and the data is included in the database. Coolant pressure was properly modelled in the input files. In the next Fig. 3.4 the maximum LHR v/s time of the re-fabricated rods 41 and 48 are presented. The in-pile measured internal rod pressures are given on the right figure axis. Test FGR Linear Heat Rate (kw/m) LHR, Rod 41, 48.9 MWd/kgU LHR, Rod 48, 6.5 MWd/kgU Pin pressure, Rod 41 Pin pressure, Rod Pressure increment (MPa) Time (h) Fig Linear heat rate evolution and measured internal rod pressure (right axis) of the rods 41 and 48 during the test FGR-1. The increasing levels of the pin pressure recorded in-pile are due to two processes: gas temperature increase at power increase and additional fission gas release from the fuel [14].The rate of the measured pressure increase with the temperature is the same as the power increase (two first steps in the picture above). When the linear power was increased over 26 kw/m (rod 41) and 22 kw/m (rod 48), the second process started and the pressure in rod 48 reached the pressure of rod 41 (the initial pressure was 1.1 and 2 MPa, respectively). In [4] it is pointed out that the pressure increase is delayed from the power ramp for about 25 h. The linear heat rate in the third step (time interval hours) decreased slightly (2.5 12

13 3.9 kw/m) but the pressure remained the same. The second stage of the FGR-1 test experiment illustrated more clearly the LHR threshold presence. Thus, during the initial three steps of the second stage, the pressure followed the temperature increase and during the hold stage the linear heat rate decreased from 43.5 to 4.8 kw/m while the pin pressure is still increasing (rod 41). The pressure sensor of the rod 48 failed at the end of the first stage. The calculated by TRANSURANUS internal pin pressures of the two rods during the test are compared with the pressure transducers records and the results are shown in Fig Test FGR-1 15 Rod 48 - calculated Rod 41 - calculated Rod 48 - exp. data Rod 41 - exp. data 15 Gap Pressure (MPa) Time (h) Fig Internal rod pressure comparison of the two instrumented rods with measured ones. The measured and calculated values of pin pressure differ significantly. The calculated values followed the linear heat rate changes (temperatures) during the first stage. The calculated values increased at the second stage, when linear heat rate (see Fig. 3.5) exceed 4 kw/m. The internal pin pressures (calculated and measured) continued increasing, although linear heat rate decreased from 43.5 to the 4.8 kw/m FGR-2 test - simulation and results The test rig included three re-fabricated short rods (fuel column length of 4 mm). Two rods were instrumented with thermocouples (1 mm from the top of fuel column) and one was not instrumented. 13

14 Test FGR-2 was carried out in two stages too. Both stages of the test irradiation were fulfilled in 4 steps. The LHR evolution (in thermocouple section) of the instrumented rods 5 and 51 are shown on Fig In the same picture the corresponding fuel central temperatures (thermocouple records) are presented (see the right axis). Test FGR-2 Linear Heat Rate (kw/m) LHR, Rod 5, 58.4 MWd/kgU LHR, Rod 51, 49.5 MWd/kgU FCT, Rod 5 FCT, Rod Fuel Centre Temperature ( o C) Time (h) Fig Linear heat rate and measured fuel temperatures v/s time, rods 5, 51. The thermocouples of the rods 5 and 51 failed before the end of the experiment. The thermocouple of the rod 5 broke down during the first stage after 78 hours in operation. The second thermocouple of rod 51 failed after 189 hours just before the upper level of the second stage. Comparison between the measured and calculated temperatures is presented in the next figures. 14

15 Fuel Centre Temperature ( o C) Burnup = 49.5 MWd/kgU Exp. data Calculated Test FGR-2, Rod Time (h) Fig.3.7 Comparison between measured and calculated fuel central temperatures of rod 51 v/s time Test FGR-2, Rod 5 Burnup = 58.4 MWd/kgU Rod 5 - exp.data Rod 5 - calculated Fuel Centre Temperature ( o C) Time (h) Fig Comparison between the measured and calculated fuel central temperatures of rod 5 v/s time. The calculated fuel central temperature is in good agreement with measured values only for rod 51, which started test irradiation with lower initial burnup, 49.5 MWd/kgU. The thermocouple of the re-fabricated rod 5 broke down at the beginning of the fourth step of the first stage. The data recorded before the failure noticeably overestimated the calculated temperatures. One could suppose that the 15

16 disagreement of FCT for the rod 5 is due to incorrect work of thermocouple before the damage but it should be pointed out that the results of registered fission gas release pressure in FGR-1 test showed the enhanced release in rods with higher burnup after certain linear rate threshold. The dependence of fuel central temperature v/s linear heat rate in the range where the thermocouples worked properly (points) are presented on Fig The calculated rates of temperature increase (solid lines) are shown in this figure too. Test FGR-2 Rod 5 (58.4 MWd/kgU); Rod 51 (49.5 MWd/kgU) 2 Fuel Central Temperature ( o C) Rod 5, exp. Rod 51, exp. Rod 5, exp., fit: y = 62.1x+22.8 Rod 51, exp., fit: y = 32.9x+155 Rod 5, calc., fit: y = 42.x+64.6 Rod 51, calc., fit: y = 39.6x Linear Heat Rate (kw/m) Fig Measured and calculated fuel central temperatures versus linear heat rates. In order to assess the rate of temperature enhancement with linear heat rate increasing a linear approach to the points (experimental and calculated) was done. The measured rates of the temperature enhancement with the power increase differ up to two times. The pin inner gas and pressure, cladding diameters and fuelcladding gaps of two rods 5 and 51 before the test start were similar (see Table 3.1). The two rods differ basically by accumulated burnup and responding fuel structure changes. The calculated fuel temperatures follow the changes in linear heat rates for the two rods. The calculated rate of temperature increase with the power for rod 5 with higher burnup is slightly higher in comparison with rod 51. The same tendency is observed for the measured rates of temperature increase but the rate of rod 5 is almost twice as much as that of rod

17 3.3. Post Irradiation Examination Puncturing results The Kola3-MIR database comprises in-pile data (pressure and fuel central temperature), PIE puncturing results of FGR and gas mixture, and results of metallographic analysis. Measured and calculated with different model options values of the fission gas release are presented in the Table 3.3. In the column of burnups the measured values are given. RFR number/ (FA number) RAMP: Table 3.3 Puncturing results of fission gas release and corresponding calculated values. Exp. Burnup [MWd/kg U] Calc. FGR after base irradiation, [%] igrbdm=1 igrbdm=3 FGR after test irradiation, [%] TU calculation Igrbdm=1 Igrbdm=3 Igrbdm= 1.8*LHR igrbdm=1 1.8*LHR igrbdm= Punct. results 33 (198) (222) (222) FGR-1: 41 (198) (222) (222) FGR-2: 51 (198) (222) (222) The calculated values of FGR differ significantly from the measured one. In order to analyse the impact of systematically modifying selected model parameters the model of intergranular fission gas release was chosen. The simplest concept of describing the intergranular fission gas behaviour is the concept of grain boundary saturation. Specific concentration (number of gas atoms per unit area) allows the release of gas atoms from grain boundaries to the free volume (further assumption concerns gas release from grain boundaries to the free volume). The model option igrdbdm determines the possibility to choose different values for gas concentration threshold for starting gas release from grain boundaries. In the TRANSURANUS code four possibilities are available: 1) no gas on the grain boundary (igrdbdm =); 2) grain boundary saturation concentration is constant and can be chosen 17

18 (igrdbm=1, standard option); 3) grain boundary saturation concentration depends on temperature (~a/t), (igrdbdm =2) and 4) a burst release of the gas grain boundary inventory due to micro-cracking was included (igrdbdm =3). The tetter is the case when so-called ramp conditions are fulfilled (T loc >15*(1-bu loc /8 and q >3.5kW/m), which are controlled by routine RampFlag. The standard set of options for the calculated FGR analysis includes igrdbdm =1 with default value of 1x1-4 mol/m 2 for the saturation coefficient and number of cracks in the fuel pellet changed from 6 to 15 at the beginning of the test irradiation according the observation of the data supplier [13, fig.6]. The second column presents the results with new model of ramp gas release from the grain boundaries. The column with the option igrdbm= in the Table 3.3 presents calculations for the boundary conditions, when all fission gas available at the grain boundaries is released immediately, i.e. the maximum. Normally one should expect that all experimental data should be within the 'bands' simulated by these extreme options. Probably this is not the case for the MIR experiment simulation, it seems that there is not sufficient fission gas arriving at the grain boundaries (simulated by diffusion). The next step in the maximum possible FGR simulation was to use the upper limit of rod linear heat rate. The specified in the database experimental error - ±8%, was taken (option-1.8*lhr). The last column of calculated FGR in Table 3.3 presents the results with higher linear heat rate during test irradiation and maximum release from the grain boundaries. Nevertheless the calculated values of fission gas release are far from puncturing results. Therefore, comparison with the available data for the gas mixtures of the re-fabricated rods was not done. To facilitate the combined analysis of the three tests - RAMP, FGR-1 and FGR- 2, the calculated fuel central temperatures are presented in the next figure Fig The presented temperatures are maximum - in the most loaded section of the rod. They are below the Vitanza threshold (~11 o C for the burnups of 5 6 MWd/kgU) only for the rods 38 from the RAMP test experiment. 18

19 r Fuel Cemtral Temperature ( o C) r33 r38 r37 r41 r32 r48 r52 r RAMP FGR-1 FGR Time (h) Fig Maximum calculated fuel central temperature during test irradiations RAMP, FGR-1 and FGR-2. The FGR up to 5% under transient conditions in reactor MIR was detected. Migration of cesium nuclides along the re-fabricated rods (in the area of the maximum heat release) was experimentally observed in most of the cases [3]. The redistribution of the free fission products proves the increased movement of Xe and Kr and their intensive release. It was established that 137 Cs intensive migration starts at linear power over kw/m for the lower burnup level and kw/m for the fuel with burnup of ~6 MWd/kgU. The data obtained with pressure detectors [4, 12] shows intensive fission gas release when the linear heat rate is over kw/m for fuel with burnup of ~5 MWd/kgU and kw/m for the fuel with burnup of ~6 MWd/kgU. Thus, the linear heat rate threshold for the intensive gas release starting was specified experimentally.. The next two figures are an illustration of the results of TRANSURANUS simulations presented in Table

20 Rod 33 (Kola FA-198 / MIR RAMP) LHR 2 Fission Gas Release (%) ) +4) 4) LHR + 8% 3) max. release from gb 2) ramp release from gb 1) standard LHR (kw/m) 1 FGR Time (h) Fig Calculated FGR during RAMP experiment for rod 33 initially irradiated up to burnup 5.8 Mwd/kgU. On the right axis maximum LHR is plotted. Rod 41 (Kola FA-198 / MIR FGR-1) LHR 3 2 Fission Gas Release (%) ) +4) 4) LHR + 8% 3) max. release from gb 2) ramp release from gb 1) standard LHR (kw/m) FGR Time (h) Fig Comparison of calculated FGR with different model options for rod 41, initially irradiated to burnup 48.9 Mwd/kg. On the right axis maximum LHR is plotted Fuel microstructure An important result of this experiment is the observed relatively small variations in the rod cladding geometry and significant changes in the fuel structure after 2

21 transient irradiation in both cases, i.e. ramp and step by step LHR increase. The tests resulted [4] in outer cladding diameter increase (relatively to the diameter after base irradiation) in the central part of the re-fabricated rods. For the rods with burnup of about 5 MWd/kgU the increase was µm and for Bu ~6 MWd/kgU µm. The measured fuel cladding gap in the post test cold state is 4 65 µm for Bu~5 MWd/kgU and µm for the rods with greater Bu~6 MWd/kgU. The pellet cracking modelling impacts on the size of fuel inner and outer radius and the result is different gap and cladding outer radius. Prior to the test irradiation simulations the number of pellet cracks was changed from 6 to 15 by the restart option of TRANSURANUS code (second row of calculations in the Table 3.4). Table 3.4 The pellet cracking modelling impact on the rod behaviour in the most loaded section. Refabr. fuel rod number Calc. BU Fuel centre temp., oc 6 cracks 15 cracks d inn Pellet hole diam.change, µm 6 cracks 15 cracks Centre grain size, µm 6 cracks 15 cracks Pellet-cladding gap (diam.), µm 6 cracks 15 cracks D out Clad outer diam.change, µm 6 cracks RAMP: cracks FGR-1: FGR-2: The burnup data in the second column are the calculated TRANSURANUS values which differ from the experimentally determined values up to 7.7% (see Table 3.3) The results in Table 3.4 show that in cold state (after test irradiation) the gap remains closed only in the rod 41 (Bu before test ~49 MWd/kgU). The fuel centre temperature calculated with 6 and 15 cracks of the fuel pellet is changed significantly only in this rod too and subsequently the grain size is considerably increased. The simulation with TU code shows a cladding diameter increase in comparison with cladding outer diameter after base irradiation only in the case of 6 (non changed) pellet cracks; see Table 3.4, last column. Only calculations of rod 41 show some diameter increase when 15 cracks are supposed. The next figure presents the 21

22 axial cladding changes for the rod 41. One of the KOLA 3 - MIR experiment results observed was the number of the fuel pellet cracks increasing with the LHR increase. At the same LHR, the number of the cracks in fuel with burnup of 6 MWd/kgU is higher than those in the fuel burned to 5 MWd/kgU [5, fig.8]. The coaxial cracks appeared when LHR is over the 21.3 kw/m Rod41, (Kola FA198/MIR FGR-1) Cladding outer radius (mm) pellet cracks=15 pelet cracks=6 before test 4.52 initial value Axial Position (mm) Fig Axial distribution of the calculated cladding rod radius of rod 41 after FGR-1 test irradiation. Rod fuel performance simulation with different pellet cracks was done with the grain growth model of Ainscough and Olsen (standard option, igrnsz=1) and an interaction layer between fuel and cladding in the gap conductance model (intaxl=1). The calculated FCT for the rods with initial burnup~6mwd/kgu are below 17 o C (TRANSURANUS underestimates the FCT in the case of higher burnup, see Fig. 3.8) and no considerable grain size change was assessed. The grain growth radial distribution of rod 48 (FGR-1 test, Bu~6 MWd/kgU) and rod 51 (FGR-2 test, Bu~5 MWd/kgU) are presented on Fig In the same figure the grain size radial distributions of rod 41 simulated with two pellet cracks assumptions are presented too. 22

23 .7 Grain Size (mm) rod48,t max =1468, ncracks=15 rod51,t max =1844, ncracks=15 rod41,t max =21, ncracks=15 rod41,t max =1778, ncracks=6.2.1 initial value Radial Position (mm) Fig Calculated grain size radial distribution after test irradiation. Post irradiation examination of fuel microstructure revealed [4] that the area of columnar crystals was formed during the test irradiation. The central area of recrystallized grains of 5-7 µm in grain length and 2-3 µm in their width was determined. The calculated grain size of ~ 63 µm is of the same order. PIE of tested rods revealed significant changes in UO 2 structure in both ramp and step by step power increasing test irradiations. The following four zones of the radial fuel pellet restructuring were observed: rim layer of 7-15 µm (porosity and width did not differ for both burnup levels); initial structure zone decreases to one-third of pellet radius; gas swelling area (1.2 2 mm) with increased porosity; central area of grain growth. The grains are stretched mostly in radial direction. The thermocouples of re-fabricated rods (FGR-2 test) revealed the temperature threshold for the start of structural changes and FGR intensification. These temperatures (linear heat rates) are different (111 and 13 o C).for the rods with different burnup levels [14, fig.18]. The process of fuel restructuring is considered in TRANSURANUS code using different models that are selected by code option istzne. Two different calculations were tested. The first - without including the model for fuel restructuring zone determination (istzne=) and the second - with the fuel restructuring zone calculation according to the original Olander model (istzne=2). These two assumptions did not manifested any differences in the results for the fuel porosity and grain size. Table 3.5 below presents the comparison of the experimental data for the fuel structure and porosity with the results after TRANSURANUS modelling by standard options. 23

24 Table 3.5 Fuel microstructure after FGR-1 and RAMP test RFR number Measured/ TU. Calc. RIM-layer Initial Structure area Gas swelling area Columnar grains area Do - Di P, % Do - Di P, % Do - Di P, % Do - Di P, % Dch, rel. unit 41 Measured ± ± ± No data.85 - Cracks Cracks Measured ± ± No data No data.13 - Cracks Cracks Measured ± ± ±.6 No data - Cracks Cracks Measured 1.97 No data No data.66.2 No data No data No data.2 - Cracks Cracks Do - area outer diameter to pellet diameter ratio Di - area inner diameter to pellet diameter ratio Dch - central hole diameter to pellet diameter ratio P - porosity * - for the TU calc. porosity, the maximum values in the interval are given 24

25 Table 3.5 contains data of the fuel pellet hole diameter. The experiment gave the data of hole diameter in rod 33 (RAMP test, Bu~5) the pellet hole is completely closed and in the rod 41 (FGR1-test, Bu~5) a new central hole (~.6 mm) was observed [5] in the range of the maximum linear power (44.1 kw/m). The TRANSURANUS simulation with higher number of cracks shows increasing of fuel centre temperature only in the rod41 and consequently start of fuel microstructure changes, but the hole diameter decreasing is about 4.2%. The TRANSURANUS predictions for hole diameter decreasing varied from (rod 33, RAMP test) to 2.7% for rod 52 (FGR-2 test ). The calculated results of porosity (Table 3.5), grain size and fuel inner radius evolution during test overpower irradiation are only preliminary. This problem needs deeper analysis Conclusions In order to analyse the fuel rod behaviour under transient conditions nine fuel rods were cut from the rods of two fuel assemblies irradiated at Kola NPP Unit 3, reinstrumented and putted to the test irradiation. In the experimental MIR reactor in Dimitrovgrad, Russia three tests were carried out, one under single ramp conditions - RAMP experiment, and two step-by-step power increase experiments -FGR-1 and FGR-2 The data for the rods from in-pile and post irradiation examinations were submitted to the IFPE databank and compared with TRANSURANUS simulations. The TU base irradiation simulation results are in a good agreement with the experimental ones and the same options are used for the simulation of the truncate pieces of the rods. Two rods (rods 5 and 51 FGR-2) with different burnups were equipped with thermocouples. Rod 5 (Bu~6 MWd/kgU) has reached higher temperature than rod 51 (Bu~5 MWd/kgU) although its linear heat rate, during test irradiation, was lower. Comparison between the measured and the calculated temperatures of rod 51 shows very good results. However this is not the case for rod 5. Two re-fabricated rods rod 41 (Bu max ~5 MWd/kgU) and rod 48 (Bu max ~6 MWd/kgU) with different burnups were equipped with pressure sensors (FGR1-test). The rods were fulfilled with He and pressurized up to 2 and 1.1 MPa respectively. The calculated values underestimate the measured ones significantly. The in-pile measurements show that the test rod with higher burnup and lower initial pressure

26 rapidly increased the pressure after an operational time of about 25 hours. The calculated pressures follow the trend of LHR. All tested rods passed the over power irradiation successfully and were leaktight after the tests, but during disassembling of the FGR-2 test rig rods 5 and 51 were damaged. This explains why there is data for fission gas release and gas mixture from PIE only for seven of the rods. The comparison between the measured and calculated results is not satisfactory. The measured fission gas release is underestimated by the code for all available rods, even when combining options for extreme conditions. The calculations with number of cracks 6 and 15 was performed in order to investigate the impact of the initial number of pellet cracks over the fission gas release and over the fuel microstructure. No significant differences in code predictions have been observed for all rods, except for the rod 41 (linear heat rate during hold ramp time was 44 KW/m and fuel central temperature was over 2 o C). In conclusion, the calculated values of FGR differ from the measured ones. For this type of the fuel rod transient loading an additional source of fission gas release as a function of linear heat rate threshold as well as changes in the fuel microstructure should be considered. Kola3-MIR experiment provides data for WWER fuel under transient conditions and it will become a good tool for developing some of the code models responsible for such type of enhancement. The only disadvantage is the insufficient number of instrumented rods and therefore the reliability of the drown conclusions remains questionable. Acknowledgements The work, described in the present report, was done in close collaboration with the Modelling Group of the Institute for Transuranium Elements in Karlsruhe, Germany. We cordially thank P. van Uffelen, A. Schubert, J. van de Laar and C. Györi for the valuable help and the fruitful discussions in performing this work. 26

27 4. References: 1 K. Lassmann, TRANSURANUS: a fuel rod analysis code ready for use, Journal of Nuclear Materials, Vol. 188, pp , The Public Domain Database on Nuclear Fuel Performance Experiments for the Purpose of Code Development and Validation, International Fuel Performance Experiments (IFPE) Database, edition April 27, in: 3 M. Solonin et al., Proc. II Intern. Seminar on Fuel performance, Modelling and Experimental Support, p. 48, April 1997, Sandanski, Bulgaria 4 A. Smirnov, В Kanashov, V Ovchinnikov et al. Study of Behaviour of WWER- 44 Fuel Rods of Higher Fuel Burnup under Transient Conditions, Report HPR- 349/43, Enlarged HPG Meeting on High Burnup Fuel Performance, Safety and Reliability, OECD Halden Reactor Project, Norwey, Lillehammer, 15-2 March,1998, p.1. 5 Smirnov, В Kanashov, G. Lyadov et al. Examination of Fission Gas Release and Fuel Structure of High Burnup WWER-44 Rods under Transient Conditions, Proc. of the Third Int. Seminar WWER fuel Performance, Modeling and Experimental Support, Pamporovo, Bulgaria, 4-8 October 1999, p A.Smirnov, V.Smirnov, A. Petuhov, et al, Peculiarities of the WWER-44 fuel behaviour at high burnups, Proc. of the Second Int. Seminar WWER fuel Performance, Modeling and Experimental Support, Sandanski, Bulgaria, April 1997, p M.Manolova, D.Elenkov, et al, Validation of the TRANSURANUS Code WWER Version by Updated KOLA-3 Data Set, Proc. of the Third Int. Seminar WWER fuel Performance, Modeling and Experimental Support, Pamporovo, Bulgaria, 4-8 October 1999, p A.Smirnov, B.Kanashov, V.Kuzmin, et al., Results of Post Irradiation Examination to Validate WWER-44 and WWER-1 Fuel Efficiency at High Burnups, Proc. of the Fourth Int. Seminar WWER fuel Performance, Modeling and Experimental Support, Varna, Bulgaria, 1-5 Oct., 21., p D.Elenkov, S.Boneva, at al., Verification of the TRANSURANUS-WWER Code Version v1m2j by Sofit and KOLA-3 Data Bases, Proc. of 4-th Int. Proc. of 27

28 the Fourth Int. Seminar WWER fuel Performance, Modeling and Experimental Support, Varna, Bulgaria, 1-5 Oct., 21, p D.Elenkov, Final Report Study on the verification of the TRANSURANUS code for nuclear Fuel used in Bulgaria, SC no F1ED KAR BG 11 K. Lassmann, A. Schubert, J. van de Laar, C. Venix, TCM on Nuclear Fuel Behaviour Modelling at High Burnup, June 2, lake Windermere, UK 12 A. Smirnov, B. Kanashov, et al., FGR from high-burnup WWER-44 fuel under steady-state and transient operation conditions, Int. Seminar Fission Gas Behaviour in Water Reactor Fuels, Cadarache, France,26-29 Sept., 2, p A.Burukin, D.Markov, G.Mayorshina, Results of examination of fission gas release and fuel structure of the VVER fuel rods with a burnup of 5 MWd/kgU and higher after operation under normal conditions and testing in the MIR reactor, 7 th International seminar WWER fuel performance, modeling and experimental support, Albena, Bulgaria, September 27, p A. Smirnov, B. Kanashov, A.Panyushkin et al., FGR from high-burnup WWER-44 fuel under steady-state and transient operation conditions, Int. Seminar Fission Gas Behaviour in Water Reactor Fuels, Cadarache, France,26-29 Sept., 2, p.63 28