IMPROVEMENT OF ROFEM AND CAREB FUEL BEHAVIOUR CODES AND UTILISATION OF THESE CODES IN FUMEX 3 CRP

Transcription

1 Institute for Nuclear Research Pitesti, Romania IMPROVEMENT OF ROFEM AND CAREB FUEL BEHAVIOUR CODES AND UTILISATION OF THESE CODES IN FUMEX 3 CRP G. Horhoianu INR Pitesti,Romania Final Report of the IAEA Research Project 14974/RO December 211 1

2 Contract Number: 14974/RO Title of Proiect: IMPROVEMENT OF ROFEM AND CAREB FUEL BEHAVIOUR CODES AND UTILISATION OF THESE CODES IN FUMEX 3 CRP Institute where research was being carried out: Institute For Nuclear Research, Pitesti, Romania Chief Scientific Investigator: Dr. Grigore Horhoianu IAEA Project Officer: Dr. John Killeen Time period covered: to

3 Introduction. The IAEA Vienna organized a Coordinated Research Project (CRP) on improvement the computer codes used for fuel behaviour simulation under the name: FUMEX III [1].The major research objective of this CRP would be to test and develop fuel modeling codes against experimental data and cases provided by IAEA and OECD/NEA [1,2]. Institute for Nuclear Research (INR) Pitesti participated at this CRP with ROFEM and CAREB computer codes [3, 4].The main aim of ROFEM code was to calculate fuel behaviour during steady state operating conditions [5].CAREB code was developed for fuel transients analyses such as LOCA and RIA [6].Recently both codes have been improved with new models in order to extend their capabilities [3 ]. References 1. Killeen,J.; Specific Research Objectives of FUMEX-III CRP, First IAEA Technical Meeting-FUMEX III, Vienna, December 1-12, Sartori,E.; CD with IFPE Data Base Selected for FUMEX-III CRP, First IAEA Technical Meeting-FUMEX III, Vienna, December 1-12, Horhoianu,G. at al.; Improvement of ROFEM and CAREB Fuel Behaviour Codes and Utilization of these Codes in FUMEX III Exercise, Progress Report to IAEA Research Contract No: 14974,IAEA Vienna,24 September Horhoianu,G.at al; Application of ROFEM and CAREB Codes to FUMEX 3 Exercise, IAEA Technical Meeting-FUMEX III, Pisa, Italy,1-4 June,21 5. Moscalu,D.R.; Aspects Regarding Nuclear Fuel Burnup Increase, Ph.D.Thesis, Institute for Nuclear Research,Pitesti, Arimescu,I.; High Temperature Transients Fuel Performance Modelling, Ph.D. Thesis, Institute for Nuclear Research,Pitesti, Objective of the INR Pitesti in FUMEX 3 exercise In order to utilize the versatile nature of the CANDU type reactors, INR-Pitesti has started a development program aiming to introduce extended burnup,recovered Uranium (RU) fuel in the reactors under construction[1,2]. An important aspect that has been taken into account was the need for improvement of the actual fuel element design performance at higher burnup.with a target burnup of about 3 MWd/KgU, which is three times grater than the maximum burnup of the natural uranium fuel,ru fuel design is under development. In order to calculate the performance of the RU fuel at extended burnup, efforts have been focused on the improvement of the existing fuel behavior codes, using information derived from INR experimental date base and from open literature. This activity, which have been started under IAEA research contract 6197/RB is still in progress. In this process, an important stage is the validation of the improved fuel behavior modelling codes in a comparison exercise,like FUMEX 3. 3

4 The design and operating conditions of CANDU PHWR fuel differ from LWR fuels in many respects. The fuel elements in CANDU use a thin wall collapsible sheath. They are short length (.5m) and have no separate plenum. Because of collapsible sheath, the fuel and sheath come in contact right from the beginning of life. The CANDU fuel elements operate at considerably higher linear heat rating compared with LWR fuel elements. Higher heat rating in CANDU fuel elements causes significant fuel restructuring, which promotes new processes of fission gas release which are not present in LWR fuels. Because of closed fuel-sheath gap, the sheath creeps outwards unlike LWR fuels, which initially experience sheath creep down till gap is closed and subsequently creep out at higher burnup after the gap closure. The fuel modeling codes for CANDU fuel have to take into account the above factors. The main objective of ROFEM and CAREB codes participation in FUMEX3 exercise is the need for verifying code prediction by comparison with experimental data, in order to identify the areas in which further improvements are necessary. References 1. Horhoianu G., Nuclear Fuel R&D Program at INR Pitesti for the Period 26-21, Internal Report No.7212/25, INR Pitesti. 2. G.Horhoianu, I.Patrulescu, Technical feasibility of using RU-43 fuel in the CANDU-6 reactors of the Cernavoda NPP, Kerntechnik, volume 73, No1-2, (28) 2. ROFEM Computer Code Main models and code capabilities ROFEM is a FORTRAN-IV computer code to predict the in-pile thermal and mechanical behavior of reactor fuel rods as a function of the reactor operating history. [1,2] The code can not treat the fuel rod behavior during fast transient conditions. The steady-state thermal calculations are used at each time step with an iterative procedure. The code assumes an axisymmetry. The code consists of two major calculation parts, the thermal analysis part and the mechanical analysis part. In the thermal analysis part, the integral behavior of a whole fuel rod is analyzed and the temperature distribution, the dimensional changes of fuel and cladding, the fission gas release, and the associated inner gas pressure are determined. Then, the results of temperature distribution and inner gas pressure are transferred to the mechanical analysis part where the localized mechanical behavior is analyzed by the twodimensional axisymmetric finite element method. The finite element method (FEM) analysis is applied to a small part of fuel rod which is expected to have the most severe mechanical interaction. The results from the mechanical analysis part is regarded as a subcode for the analysis of localized behavior. As a function of irradiation time and axial position, the thermal analysis part calculates the following: The temperature distribution in the fuel and cladding; The radial deformation of the fuel due to thermal expansion, swelling, densification, and relocation; 4

5 The radial deformation of the cladding due to thermoelastic,plastic and creep; The gap or the contact pressure between fuel and cladding; The fission gas release and inner gas pressure. As a function of irradiation time, the FEM mechanical analysis part calculates the following: Stress and strain distributions in the fuel and cladding considering elastoplasticity, creep, thermal expansion, fuel cracking and healing, relocation, hourglassing, densification, swelling, hot pressing, fuel-cladding mechanical interaction, inner gas pressure, and coolant pressure. The basic fuel rod geometry handled in the code consists of fuel in the form of a stack of sintered fuel pellets and a cylindrical Zircaloy cladding tube, closed at each end, with the upper plenum. The fuel consists of uranium deoxide in the form of pellets. The fuel pellets may be solid or annular and may not be dished, or be chamfered. The filling gas is considered as any composition, at any pressure, of the following four gases: helium, nitrogen, krypton, and xenon. The design and characterization data of fuel and cladding such as geometry, density, grain size, filling gas composition and pressure, and cladding type are given as input data. The fuel rod is supossed to be cooled by pressurized water. The coolant temperature and pressure are given as input data as a function of time and are assumed to be constant in axial direction. The power and flux distribution in axial direction is given as input data as a function of time. The major characteristics of ROFEM are the following: ROFEM can analyze the integral behavior of a whole fuel rod throughout its life, as well as the localized mechanical behavior at a small part of a fuel rod; Localized behavior is analyzed in detail by the two-dimensional axisymmetric finite element method; Elasto-plasticity, creep, thermal expansion, fuel cracking and crack healing, relocation, densification, swelling, hot pressing, heat generation distribution, fission gas release, and fuel-cladding mechanical interaction are modelled; A quadratic isoparametric element is used to obtain a more accurate finite element solution with fewer elements than that are required when linear elements are used. Contact problem between fuel and cladding is exactly treated, where the contact condition, is determined by iterative procedure; An implicit algorithm, which necessitate use of iteration, is applied to obtain a accurate and stable solution for non-linear problems; Fuel is assumed as a no-tension material. Crack healing under compression is treated as recovering its stiffness gradually to nominal value. The recovery of fuel stiffness is related to initial relocation; Finite element analysis is applied only to a region of half-pellet height; The code can treat a problem of long irradiation history including power ramps with reasonable running time. 5

6 In the framework of IAEA Research Project the individual models included in the ROFEM code have been reviewed in order to improve these models and the predictive capabilities of the codes.in particular from the ROFEM code the folllowing models have been reviewed: pellet cracking and pellet relocation, pellet densification and swelling, UO2 thermal conductivity degradation with burnup, fission gas release, and pellet and cladding mechanical behavior. In order to improve fission gas release predictive capability the ROFEM code has been coupled with DCHAIN5V code[3,4]. In order to analyse the CANDU fuel behavior in transient/accident conditions,the code ROFEM has been coupled with CAREB code[3]. References [1] D.R.Moscalu,et al,validarea codurilor de analiza a comportarii combustibilului nuclear de tip CANDU existente in institut,raport ICN No. 438,(1994) [2] D.R.Moscalu,Aspecte privind cresterea gradului de ardere in cobustibilul nuclear,ph.d. Thesis (1997) [3] Horhoianu,G. at al, Improvement of ROFEM and CAREB Fuel Behaviour Codes and Utilization of these Codes in FUMEX 3 Exercise,Progress Report to IAEA Research Project 14974/RO,INR Pitesti,June,29 [4] M.J.F.Notley,I.J.Hastings,A Microstructure_Dependent Model for Fission Product Gas Release and Swelling in UO2 Fuel,Report AECL No.5838 (1978) 3. CAREB Computer Code Main Models andcode Capabilities The CAREB computer code was developed to simulate the thermal-mechanical response of a fuel element during rapid, high-temperature transients [1].The model assumed is a single UO 2 /Zircaloy sheath element with axi-symmetric properties. Physical effects considered in the code are: - Expansion, Contraction, cracking and melting of the fuel, - Variation of internal gas pressure, during the transient, - Changes in the fuel/sheath heat transfer, - Thermal, elastic and plastic sheath deformation (anisotropic) - Zr/H 2 O chemical reaction effects - Beryllium assisted crack penetration of the sheath (initiated from Be-brazed appendages) The new release of the code, CAREB. 1B, extends the capability of CAREB to model of sheath failure due to oxygen embrittlement upon rewetting and effect of oxide strengthening on sheath creep [2]. Other new features improved ROFEM-CAREB interface [2]. The equations used in the model to represent these physical effects use transient boundary conditions of coolant temperature, coolant pressure and sheath/coolant heat 6

7 transfer coefficient evaluated by thermal/hydraulic codes. The conditions at the start of the transient are obtained from steady/state fuel performance code ROFEM [2].The CAREB code monitor conditions leading to sheath failure. Several failure mechanisms are explicitly represented in the code: Sheath overstrain (>.15 ) Localized overstrain under oxide cracks Excessive sheath creep rate (>1-1 s -1 ) Low ductility sheath failure (>.4 strain) Beryllium-assisted crack penetration High fuel enthalpy (>838 kj/kg UO 2 ) Oxygen embrittlement Beryllium-assisted crack penetration failure mechanism specific to CANDU fuel elements. Intergranular cracking of the Zircaloy fuel sheath can occur at bearing pad and spacer pad locations of the element brought on by the penetration of a beryllium-braze alloy in the presence of an applied hoop stress[3]. The thermal transient experienced by the sheath represent a complex heat treatment which cause changes in anisotropy, annealing, grain size, phase transformation(α-β) plus other changes in sheath microstructure which effect the plastic creep behaviour of the sheath. An improved transient creep model which accounts for these transient changes in microstructure [4] allow detailed examination of the plastic deformation of the sheath which occur under a variety of postulated LOCA transients. As changes in sheath microstructure are evaluated by CAREB during transient, differences in initial sheath properties can be explicitly represented in output data. CAREB was designed to be used as self-contained code with the minimum amount of input information required for execution. However, the fuel element description at the start of the transient is best obtained directly from the steady-state companion code ROFEM, due to detailed information available. These two codes, ROFEM and CAREB, can be run sequentially to describe any arbitrary pre-transient/transient reactor irradiation history desired. The capability of the CAREB code was extensively verified through the comparison with a large number of in-reactor and out of reactor test [5]. In the framework of IAEA Research Project the individual models included in the CAREB code have been reviewed in order to improve these models and the predictive capabilities of the codes.from the CAREB code the following models have been reviewed:transient creep,clad/steam reaction, beryllium assisted crack penetration.a new model for beryllium assisted crack penetration has been included in CAREB code. References 1. Arimescu, I., High Temperature Transient Fuel Performance Modelling, Ph.D. Thesis, Institute for Atomic Physics, Bucharest, Horhoianu, G. at al, Improvement of ROFEM and CAREB Fuel Behaviour Codes and Utilization of these Codes in FUMEX 3 Exercise,Progress Report to IAEA Research Project 14974/RO,INR Pitesti,June,29 3. Kohn, E. and Sagat, S., Beryllium Assisted Cracking of Zircaloy, 6 th Canadian Fracture Conference, Harrison Hot Springs, B.C., Canada, 1982 June 7

8 4. Sills, H.E. and Holt, R.A., Predicting High Temperature Transient Deformation from Microstructural Models,4 th International Conference on Zirconium in Nuclear Industry, Stratford-upon-Avon (1978) 5. Ion, S., Ionescu, D.V., CAREB Verification and Validation Manual, Internal Report No.484, INR Pitesti, March, The IAEA research contract No /RO In the frame of IAEA research contract No /RO the following activities have been already performed: - Analysis of the basic models of the ROFEM and CAREB codes and adaptation of these models in order to meet FUMEX exercise requirements; - Selection from open literature of various irradiation experiments in order to verify the code response in evaluating the behavior of different types of fuel; - Analysis of the FUMEX cases input data received from NEA for identifying the aspects that have to be clarified or need additional information; - Selection of the FUMEX cases that are very close to INR objective in this exercise; - Preparation of ROFEM and CAREB input files for the preliminary runs; - Preliminary run (with some uncertainties in the input data) on FUMEX-3 case; - Verification of the input data; - Final runs for selected cases; - Analysis of released experimental results; - Comparison between code predictions and experimental results. - Technical specification for LOCA instrumented tests planed to be performed in TRIGA research reactor. 5. Application of ROFEM and CAREB to the FUMEX 3 exercise In accordance with INR objectives for this exercise we have selected for calculations with ROFEM-2 and CAREB codes four cases: Case1: NR bundle and JC bundle,both irradiated in NRU reactor of AECL. Case 2: EC89 and EC51 in TRIGA research reactor of INR. Case 3: FIO 131 and FIO 13 LOCA tests in NRX reactor of AECL. The principal reason for this selection was the CANDU type fuel elements used in these irradiation tests, requirements of the actual version of the codes. 5.1 Case 1: JC bundle and NR bundle, both irradiated in NRU reactor (AECL, Chalk River Labs., Canada) [1]. The irradiation parameters are summarized in Table 1 [12]. 8

9 Table 1. Fuel elements irradiation parameters Fuel Element No. Parameter JC NR Max. element average power (kw/m) Rod average burnup (MWh/KgUO 2 ) Coolant H 2 O, CANDU conditions Coolant pressure (MPa) 9-1 Coolant temperature ( o C) 3 Loop coolant consisted of light water at nominal conditions of 3C (573 K), and pressures between 9.3 and 1. MPa. Nominal fabrication parameters of JC fuel elements are listed in Table 2. Table 2 Nominal Dimensions of Bundle JC Outer Elements Fabrication Parameters For Bundle JC fuel form single dished pellets enrichment %U-235 in U 1.55 pellet i/d mm pellet o/d mm dish depth mm.4 dish void volume (per dish).21 ml pellet land width mm.254 pellet chamfer height mm. pellet chamfer width mm. pellet length mm 2.89 pellet density g/cc 1.65 grain size µm 9 clad material Zircaloy-4 clad inside diameter mm clad outside diameter mm clad inside surface coating none diametral gap mm.1 axial gap between mm 1.67 end of fuel stack and end of sheath fuel length mm (stack length) 9

10 pellets per element 23 fill gas 9% Ar, 1% He pressure 1 atm. Calculated filling gas ml 2.5 volume Nominal fabrication parameters of NR fuel elements are listed in Table 3. Three different pellet-stack-to-end-cap types, as mentioned above, were used in the outer elements of the bundle. Table 3 Nominal Dimensions of Bundle NR Outer Elements No Plenum 8mm Plenum (.35 cc) 12mm Plenum (.58 cc) fuel form single dished single dished single dished pellets pellets pellets enrichment %U-235 in U pellet i/d mm pellet o/d mm dish depth mm dish volume (per dish) ml pellet land width mm.44 +/ / /-.19 pellet chamfer height mm.175 +/ / /-.125 pellet chamfer width mm.44 +/ / /-.19 pellet length mm pellet density g/cc grain size µm clad material Zircaloy-4 Zircaloy-4 Zircaloy-4 clad inside diameter mm clad outside diameter mm clad inside surface coating graphite (CANLUB) graphite (CANLUB) graphite (CANLUB) diametral gap mm axial gap between mm end of fuel stack and end-cap, or plenum insert fuel length mm (stack length) pellets per element fill gas 1% He 1% He 1% He pressure 1 atm. 1 atm. 1 atm. Calculated element voidage ml Note: Pellet chamfer is a bevel edge. 1

11 5.1.1 JC Fuel Bundle Bundle JC was a prototype 37-element fuel bundle for the Bruce-A Ontario Hydro reactors[1].for irradiation in the NRU reactor, the centre fuel element was removed and replaced by a central tie rod for irradiation purposes in the vertical test section. Coolant for the test was pressurized light water under typical PHWR conditions of 9 to 1.5 MPa and 3 C. The fuel elements used 1.55 wt% U235 in U uranium dioxide fuel and were clad with Zircaloy-4 material. The bundles elements were coated with a graphite coating. The fuel is somewhat atypical of 37 element-type fuel since the length to diameter ratio (l/d) is large (1.73) due to the pellets being ground down from a OD of 14.3 mm to mm. The outer element average measured burnup was MWh/kgU on discharge.the standard deviation was 7.9Mwh/kgU,and maximum and minimum values were 654 and 636 MWh/kgU respectively. Burnups were measured at the mid-plane of the four elements. Outer element powers varied between 57 kw/m near the beginning of life and 23 kw/m at discharge.uncertainty in linear power was ±1%. Due to the long irradiation, the bundle experienced 153 short shutdowns, and 129 longer duration shutdowns. No element instrumentation was used during the irradiation. However, the bundle was subjected to extensive post-irradiation examination (PIE) that included dimensional changes, fission gas release, fuel burnup analysis, and metallography that included grain size measurement. The measured FGR and residual sheath strains are within the range that is expected for similar-operated commercial power reactors and reported in open literature [2,3]. Analysis of JC fuel behavior with ROFEM Code The power history are presented in Figure 1. It can be noticed that in the cases JC bundle the power history were of the ramps type with the maximum power at the beginning of the power history. The average linear power represent the heat rate contributed by the fuel element to the coolant over an irradiation interval. It is the integral of the power as a function of time, divided by the time period. The peak linear power is the maximum instantaneous power that occurred during the irradiation interval. The power history,originaly based only on loop calorimetry and reactor physics calculations,have been corrected for measured chemical burnup. 11

12 8 7 6 Linear Power (kw/m) Time (h) Figure 1. Power history of JR case. Figure 2 show the fuel central temperature evolution during irradiation. Figure 3 show the FGR evolution during irradiation. The calculated release show an appreciable increase at the end of irradiation and underestimate the measured data by about 18.5 cm 3 (measurements: cm 3 ) Temperature (C) Time (h) Figure 2. Fuel central temperature vs. time 12

13 8 Fission Gas Release (mm3) Experimental P.I. Data ROFEM Results Time (h) Figure 3. Fission gas release vs. time. The prediction and measured sheath deformations are presented in Figures 4 and 5 and summarized in Table 4. The calculated data show a increase at the power ramp followed by the decrease and again slow increase during irradiation. It can be noticed the calculated values for plastic sheath strain are in good agreement with experimental data in the ridge region and superestimate the measurements in the pellet mid plan region by about.5 % Sheath Hoop Strain at Midplane (%) Experimental P.I Data ROFEM Results Time (h) Figure 4. Sheath hoop strain at the pellet midplane region vs. time. 13

14 2 1.8 Sheath Hoop Strain at Ridge (%) Experimental P.I Data ROFEM Results Time (h) Figure 5. Sheath deformation at the pellet end (ridge) region vs. time. Table 4. Comparison between ROFEM code calculation results and P.I experimental data JC bundle Fission Gas Release Sheath Hoop Strain (%) (cm 3, STP) Pellet Midplane Pellet End (Ridge) calculated measured calculated measured 1 calculated measured Average values on axial direction NR Fuel Bundle Bundle NR was a prototype 37-element fuel bundle for the CANDU 6 reactor[1]. For irradiation in the NRU reactor, the centre fuel element was removed and replaced by a central tie rod for irradiation purposes in the vertical test section. Coolant for the test was pressurized light water under typical PHWR conditions of approximately 9 to 1.5 MPa and 3 C.The fuel elements used 1.41 wt% U235 enriched UO2 fuel pellets and were clad with Zircaloy-4 material. The inner sheath surface was coated with a graphite layer. Three types of pellet-stack-to-end-cap geometries were used for the outer elements: a 35 mm3 plenum insert (six elements), a 58 mm3 plenum insert (six 14

15 elements), and no plenum insert (six elements). Intermediate and inner element rings had no plenum insert.outer element burnups reached average measured burnups of MWh/kgU (the maximum was 238. MWh/kgU and minimum MWh/kgU). Outer element powers were steady during the irradiation and average on time interval ranged between 58 and 62 kw/m during the irradiation. No element instrumentation was used during the irradiation. However, the bundle was subjected to extensive post-irradiation examination (PIE) that included dimensional changes, fission gas release, and fuel burnup analysis. Analysis of NR fuel behavior with ROFEM Code The power history is presented in Figure 6. It can be noticed that in the case NR bundle the power history were of the steady state type with the maximum power at the beginning of the power history Linear Power (kw/m) Time (h) Figure 6. Power history of NR case. 15

16 3 25 Temperature (C) Time (h) Figure 7. Fuel central temperature vs. time. Figure 7 show the fuel central temperature evolution during irradiation. Figure 8 show the FGR evolution during irradiation. The calculated release show an appreciable increase at the end of irradiation. The calculated FGR underestimate the measured data by about 2.5 cm 3 (measurements: cm 3 ).Thought this difference is within a reasonable extent considering the errors that would have been involved in the measurement of FGR and the deviation of the calculated power history. 5 Fission Gas Release (mm3) Experimental P.I Data ROFEM Results (group 1) ROFEM Results (group 2) ROFEM Results (group 3) Time (h) Figure 8. Fission gas release vs. time. 16

17 The prediction and measured cladding deformations are presented in Figures 9 to 14. The experimental data show the decrease of cladding diameter during irradiation and a slight increase at the power ramp. It can be noticed the satisfactory agreement between measured and calculated values in the midplane region at the end of the power history for elements group 1 and group 3.For the elements group 2 the calculated results slowly underestimate the measured data. In the pellet end region (ridge) calculated results are in satisfactory agreement with experimental data for elements group 1 and slowly higher then experimental data for elements group 3.For the elements group 2 the calculated results underestimate by about.17% measurements..7 Sheath Hoop Strain at Midplane (%) Experimental P.I. Data ROFEM Results Time (h) Figure 9 Sheath deformation at the pellet midplane region vs. time for group 1 NR bundle elements. 17

18 .7 Sheath Hoop Strain at Midplane (%) Experimental P.I. Data ROFEM Results Time (h) Figure 1. Sheath deformation at the pellet midplane region vs. for group 2 NR bundle elements..7 Sheath Hoop Strain at Midplane (%) Experimental P.I. Data ROFEM Results Time (h) Figure 11. Sheath deformation at the pellet midplane region vs. time for group 3 NR bundle elements. 18

19 1.2 Sheath Hoop Strain at Ridge (%) Experimental P.I. Data ROFEM Results Time (h) Figure 12 Sheath diameter at the pellet end region (ridge) vs. time for group 1 NR bundle elements. 1.6 Sheath Hoop Strain at Ridge (%) Experimental P.I. Data ROFEM Results Time (h) Figure 13. Sheath diameter at the pellet end region (ridge) vs. time for group 2 NR bundle elements. 19

20 Sheath Hoop Strain at Ridge (%) Experimental P.I. Data ROFEM Results Time (h) Figure 14. Sheath diameter at the pellet end region (ridge) vs. time for Table 5. Comparison between ROFEM code calculation results and P.I experimental data NR bundle group 1 NR bundle group 2 NR bundle group 3 Fission Gas Release Sheath Hoop Strain (%) (cm 3, STP) Pellet Midplane Pellet End (Ridge) calculated measured calculated measured * calculated measured * * Average values on axial direction Discussion Fission gas release behavior The FGR model of ROFEM code assumes that the grain boundary bubbles grow with an accumulation of influx of gas atoms and are connected to form tunnels to a free space. When the amount of fission gas retained in the grain boundary exceeds a 2

21 saturation level, an excess amount of gas is immediately released to the free space. This is the gas release criteria in the present model. In such model, a large amount of FGR in a short transient is allowed only in the limited situations where the following two situations are satisfied. (a) When a large amount of fission gas is accumulated at the grain boundary. (b) When a large and instantaneous decrease in the saturation value at the grain boundary takes place due to changes of the thermal and mechanical conditions. In the results of the case JC, however, the amount of accumulated fission gas at the grain boundary is small before the highest power ramp at 85MWh/kgUO 2, and the change of the fission gas saturation level caused by power ramps is small, too. In the other words, the rate/determining process of FGR is diffusion of gas atoms from the grain inside to the grain boundary in the present analyses during the whole irradiation period. Hence, the predicted FGR histories show only slow cumulative increases instead of the possible jumps caused by other mecanisms. This suggests that the actual rods subjected to JC tests had other mechanisms of fission gas release than diffusion process of gas atoms. Therefore the difference between predicted and measured pressure history is considered to be caused by such specific gas release mechanisms which are not taken into account in the present FGR model. This suggests that the actual FGR model used by ROFEM code must be improved in order to consider the specific gas release mechanisms which are not taken into account in the present FGR model. Sheath strain behavior CANDU fuel operated at burnups 4Mwh/kgU typically exhibits midpellet sheath strains up to.5% [ 2].Sheath strain is dependent on pellet geometry/density,fuel power and as-fabricated diametral clearance(between pellets and sheath).the maximum strain observed at a given burnup generally increase gradually from.5% at 4Mwh/kgU to 1.5% at 75 Mwh/kgU. This increase in strain appears to be related to two effects: a) Swelling of pellets due to the buildup of fission products in the ceramic matrix and associated PCI. b) Internal gas pressure above the gas over pressurization threshold. The JC and NR fuel elements appear to have been primarily strained as a result of PCI. The strains in NR group 2 and NR group 3 elements decreased proportionally with decreasing in as fabricated plenum volume and the code correctly predicted the effect of plenum volume on sheath strains Conclusions 1. JC and NR fuel bundles were succesfuly irradiated in X2 loop of NRU reactor. 2. The JC and NR fuel bundle was subjected to extensive post-irradiation examination (PIE) that included dimensional changes, fission gas release, fuel burnup analysis, and metallography that included grain size measurement. 21

22 3. The measured FGR and residual sheath strains are within the range that is expected for similar-operated commercial power reactors and reported in open literature. 4. The difference between predicted and measured internal pressure history at JC elements is considered to be caused by such specific gas release mechanisms which are not taken into account in the present FGR model. 5. This difference sugest that the actual FGR model used by ROFEM code must be improved in order to consider the specific gas release mechanisms which are not taken into account. 6. ROFEM calculated sheath strains compared satisfactory well with P.I. measurements. References [1] NEA Data Bank Data for FUMEX 3 Exercise.JC and NR fuel bundles irradiated in NRU reactor [2] M.R.Floyd, Extended-Burnup CANDU Fuel Performance,Proceedings of Seventh International Conference on CANU Fuel,21 September 23-27,Kingston,Ontario,Canada [3] Floyd M.R. at al: Performance of Two CANDU-6 Fuel Bundles Containing Elements with Pellet-Density and Clearance Variances, Proc.Sixth Int.Conf.CANDU Fuel, Niagara Falls, Canada, Case 2: EC 51 and EC 89 fuel elements irradiated in TRIGA reactor of INR Pitesti 1. Introduction Currently the INR Pitesti Nuclear Fuel R&D Program is focused on providing experimental data for development and validation of the fuel performance computer codes [1].The in-pile fission gas pressure measurements provide a wide data base for the evaluation of the fission gas release from the UO 2 pellet during power change operation. The results from the two instrumented fuel elements of different pellet microstructure which operated until Mwd/tU at significantly power levels in TRIGA Material Testing Reactor (TRIGA MTR) of INR Pitesti, are presented. Some analyses regarding these tests were performed using the ROFEM computer code and the results were compared with experimental data in the framework of the IAEA research project [2]. This 22

23 paper describes briefly measuring techniques developed and currently in use in INR Pitesti, and presents and discusses selected in-pile test results comparatively with computer code results. 2. Test Fuel Elements The test fuel elements have a nominal mm diameter pellets contained in a Zircaloy-4 cladding of mm inner diameter with.38 mm minimum wall thickness. The total length of the fuel stack is mm for EC51 fuel element and mm for EC89 fuel element. Fuel pellets were made by INR Pitesti from UO 2 powder enriched to 7.4 wt% U-235 for EC51 and 3.92 wt% U-235 for EC89 [3]. The level of fuel enrichment was selected to achieve a linear power output from each element higher then 55 KW/m in the TRIGA MTR. Apart from the differences in fuel enrichment and element length the only significant difference between test fuel elements appears to be in the fuel density and microstructure. The pellet average grain size was 14.5µm at the EC51 fuel element respectively 1.2µm at the EC89 fuel element. All of the fuel sheaths were coated on the inside surface with graphite layer. The fill gas was pure helium at.1 MPa pressure. Test fuel elements were instrumented with pressure transducers to measure the fission gas pressure changes during fuel irradiation. A small diameter 1.5 mm capillary tube was connected to the transducer to facilitate internal gas pressure measurement. The total volume of pressure transducer and capillary tube was approximately 1 mm3. Other relevant pellets and sheath information are presented in Table Irradiation Conditions Test fuel elements were irradiated in the C2 capsule of TRIGA MTR and an average exposure of 178.9MWh/KgU was achieved for EC51, respectively MWh/KgU for EC 89[4]. The power output of test elements was controlled by reactor power and was determined through calibration of the flux detectors as power sensors. The effective irradiation time was 2915 hours for EC 89 and 3864 hours for EC 51.The linear power varied during the course of irradiation and in the last irradiation 23

24 period the linear power increased until 55kW/m in EC 89 fuel element and until 59KW/m in EC 51 fuel element [4]. 4. Experimental Results The inner gas pressure evolution during irradiation is presented in Figures 1 and 2. During the first startup both fuel elements showed an abrupt rise in gas pressure, which gradually fell to a low equilibrium value after firs few days of irradiation. After this, the measured inner gas pressure, shown in these figures exhibited a progressive increase from approximately.7 Map to 3.9 Map starting from approximately 57 MWh/kgU but became more pronounced at the end of irradiation (5.4 MPa for EC89 and 9.9 MPa for EC 51 at the end of irradiation). 5. Post Irradiation Examination Results (PIE) Hot cell examinations on the test fuel elements were conducted at INR Pitesti. Visual examination of the elements did not reveal defects or abnormalities [7].Element EC89 had a 4-2µm thick oxide layer on the external cladding surface. Both elements have circumferential ridges. Elements EC51 and EC89 were gamma scanned after irradiation. The axial gamma scans are shown in Figures 3 and 4. Intensity dips are seen at the pellet interfaces, and there is clear distinction between pellet interfaces. The pellet interface dips are typical for CANDU fuel operated at powers higher than 55 Kw/m. The isotopic activity of the irradiated fuel elements showed a variation along the length of the element. These gradients reflect the axial power distribution during irradiation in the C2 capsule. Diameter profilometry is shown in Figures 3 and 4. Elements EC51 and EC89 were measured on three directions using dual-transducer profilometer. Each diameter profile was analyzed for ridge diameters at pellet interface locations and pellet mid-plane element diameters. Values for each of these were averaged for each profile as well as for all three directions. The mid-pellet residual sheath strain ranges from -.1% to.6% for EC89 element and from -.15% to.21% for EC51 element. The pellet interface residual sheath 24

25 strain ranges from.35% to.9% for EC89 element and from.1% to.65% for EC51 element. Post-irradiation measurements show that the bow of the irradiated element is slightly larger than the pre-irradiation bow for EC 51 and slightly lower than the preirradiation bow for EC 89.The observed dimensional performance is similar to that expected at the CANDU fuel operation in similar power conditions [8]. PIE puncturing analysis showed that fission gas released was1.8 cm3 STP and measured element void volume 1.22 cm3 at EC 89. The fission gas of EC51 was lost during puncturing. Three samples from every element were impregnated with resin and were subject to optical microscopy in order to observe fuel pellet microstructure. The typical crack pattern observed in the irradiated fuel pellets shows several radial cracks the number of which mainly depends on the linear heat generation rate.on the other hand, it could be confirmed from the longitudinal sectional photosthat the dish-shape of pellets has been changed from the as-manufactured conditions. Grain growth was present in the central part of the pellet indicating that the flux peaking was relatively high consistent with that observed in gamma scans. Metal fission products are still clearly visible in the columnar grains region. The grain size of the peripheral regions did not seem to have changed during irradiation. The cracking pattern and crystal grain grow regions are typical of the CANDU UO 2 fuel operating in similar power conditions [8]. The PIE results confirm a 6.5µm thick ZrO 2 layer on the EC 89 inner surface of the cladding. The PIE didn t reveal the ZrO 2 layer on the EC 51 inner surface of the cladding. 6. Discussion A possible explanation of the abrupt rise in gas pressure at first startup involves the water vapors released from fuel matrix and graphite layer and/or hydrogen trapped during pellets sintering. A similar mechanism has been proposed for a similar in-pile test performed at CRNL [5]. Fission gas starts to be released from the fuel after some burn-up and not at the beginning of irradiation. As seen, the measured rod pressure remains virtually unchanged up to 78 Mwh/kgU in the EC 51 fuel element (Figure 1) and up to 57 Mwh/kgU in the EC 89 fuel element (Figure 2), at which point a power increase causes a pressure increase. 25

26 One of the objectives of this experiment was to compare experimental data with ROFEM code results [2]. The comparison between the measured and calculated values is shown in Figures 1 and 2.As seen, in the first part of irradiation,the ROFEM code gives results very close to the measured value. The agreement could be considered satisfactorily considering the errors that would have involved in experimental data and ROFEM input data. In both fuel elements the measured pressures show an appreciable increase in the last power cycles, although the calculated increase in gas pressure is markedly lower than the measured increase. A possible explanation of this abrupt increase in gas pressure involves the presence of the pellet cracking pattern generated by rapid power increase. The fission gas has now possibility to be immediately released to the free space through the pellet cracks. The PI puncturing of the EC89 fuel element gave 1.3 cm 3 STP inner gas whereas ROFEM estimation at the end of irradiation was 7.2 cm 3 STP.This discrepancies could be mainly due to the underestimate of the fission gas release (FGR) by ROFEM code model. This is, in turn, equivalent to overestimation of gap conductance and has effect on fuel center temperature. Therefore the difference between predicted and measured pressure history is considered to be caused by such specific gas release mechanisms which are not taken into account in the present FGR model. The FGR model of ROFEM assumes that the grain boundary bubbles grow with an accumulation of influx of gas atoms and are connected to form tunnels to a free space [2]. When the amount of fission gas retained on the grain boundary exceeds a saturation level, an excess amount of gas is immediately released to the free space. This is the gas release criteria in the present code model. In such a model, a large amount of FGR in a short transient is allowed only in the limited situations.the results of the EC51, EC89 show that the amount of accumulated fission gas at the grain boundary is large before the final power ramps, and the predicted pressure histories show only cumulative increases instead of the measured jumps at end of irradiation. This suggests that the actual FGR model used by ROFEM code must be improved in order to consider the specific gas release mechanisms which are not taken into account in the present FGR model. An empirical correlation has been established at Halden for temperature burn-up threshold for fission gas release [6]. In the first part of irradiation period the fuel temperature in the EC 89 fuel element were below Halden temperature 26

27 threshold for most of the time and small amounts of released fission gases are expected in this period, which will have little effect on gap heat transfer and thus on the fuel center temperatures. Fission gas release can be strongly affected by the fuel microstructure. Performance improvements can be obtained by increasing the size of the sintered UO 2 grains, thereby increasing the length of the diffusion path and thus retarding precipitation at the grain boundaries.in these conditions the fuel element EC 51 with larger grain size presents a different FGR threshold. 7. Planned Future Experiments Two experimental fuel elements with larger grains (>3µm) are planned to be irradiated in C2 capsule of TRIGA MTR and a marked fission gas release reduction is expected in this case [1]. 8. Conclusions 1.The gas pressure developed inside operating UO 2 fuel elements were measured during irradiation up to a maximum burnup of 178.9Mwh/kgU.The performance of irradiated fuel elements are analised for linear power variation betwen 45 to 6 KW/m. 2. The anomalously abrupt rise in gas pressure during first startup is a direct consequence of water vapors released from UO 2 pellets. 3. The predictions of fuel performance code ROFEM in terms of internal gas pressure were compared with the experimental data. In both fuel elements the measured pressures show an appreciable step increase in the last power steps, although the calculated step increase is markedly lower than the measured ones. This suggests that the actual FGR model used by ROFEM code must be improved in order to consider the specific gas release mechanisms which are not taken into account in the present FGR model. 4. The detailed fuel temperature history has been determined by ROFEM code calculations when no temperature measurements were available. 5. Experimental results presented in this paper represent data relative to two fuel elements with different UO 2 grain size. The irradiation data shows that the release in an 27

28 larger grain size pellet fuel rod was smaller than in a small grain size fuel. Other two experimental fuel elements with larger UO 2 grains (>3µm) are planned to be irradiate in C2 capsule of TRIGA MTR and a marked fission gas release reduction is expected in this case. 6. Experimental results and calculated data showed that the fuel performance factors (e.g., fission gas release, sheath strain) were within the range expected for CANDU fuel operating in similar power conditions. References Horhoianu G.: Nuclear Fuel R&D Program at INR Pitesti for the Period 26-21, Internal Report No.7212/25, INR Pitesti. 2. Horhoianu G.,et al: Improvement of ROFEM and CAREB Fuel Behavior Codes and Utilization of These Codes in FUMEX 3 Exercise, Progress Report of the IAEA Research Project 14974,June 29,INR Pitesti 3. Covaci M, et al: Fabrication of the Test Fuel Elements, Technical Report No.4129/1985, SPEC-INR Pitesti. 4. Cicerone T,et al: Irradiation of the Experimental Fuel Elements in C2 Capsule of TRIGA MTR, Internal Report No 2375/1987, INR Pitesti. 5. M.J. F.Notley, Measurements of Fission Gas Pressures Developed in UO 2 Fuel Elements during Operation, Report AECL 2662, Chalk River, Vitanza C. et al: Fission Gas Release from In-Pile Pressure Measurements, Paper presented at the Enlarged Halden Programme Group Meeting, Loen, Norway, June Popov M. et al: Post-Irradiation Examination Results of EC 89 Fuel Element, Internal Report No 2537/1988, INR Pitesti. 28

29 7. Floyd M.R. at al: Performance of Two CANDU-6 Fuel Bundles Containing Elements with Pellet-Density and Clearance Variances, Proc.Sixth Int.Conf.CANDU Fuel, Niagara Falls, Canada, Table 1: Summary of EC51 and EC89 fuel elements design characteristics: FUEL ELEMENTS EC51 EC89 1. Pellet Sintered UO 2 Sintered UO 2 Enrichment U235 (%) Density (g/c.c.) Grain size average (µm) Roughness (µm, RMS) Stoichiometry (O/U) Pellet O.D. (mm) Pellet and geometry Chamfered and both end dished 2. Cladding Zircaloy-4 Zircaloy-4 Cladding I.D. (mm) ± ±.4 Wall thickness (mm) min..38 min..38 Diametral gap (mm) Fuel element Axial gap (mm) Active column length (mm) Number of pellets per column Filling gas He He Filling gas pressure (MPa).1.1 Graphite thickness (µm) Bearing pads and spacers 3 bearing pads and 2 spacers Instrumentation Pressure sensor Pressure sensor 29

30 Table2 Comparison between ROFEM code results and measurements. Internal Gas Pressure (MPa) Before the ramp* Maximum value during the iradiation time** Calculated Measured Calculated Measured EC EC * at MWh/kgU for EC89 and MWh/kgU for EC51 ** at MWh/kgU for EC89 and MWh/kgU for EC ROFEM Calculated Linear Power EC89 Measured Inner Gas Pressure ROFEM Calculated Inner Gas Pressure 6 5 Linear Power (KW/m) Inner Gas Pressure (MPa) Burnup (MWh/kgU) Figure 1. Measured and calculated internal gas pressure for EC 89 fuel element. 3

31 7 6 Linear Power EC51 Measured Inner Gas Pressure ROFEM Calculated Inner Gas Pressure 12 1 Linear Power (KW/m) Inner Gas Pressure (MPa) Burnup (MWh/kgU) Figure 2. Measured and calculated internal gas pressure for EC 51 fuel element. 5.3 Case 4:FIO 131 and FIO 13 LOCA tests in X2 loop of NRX reactor. Introduction The IAEA Vienna organized a Coordinated Research Project (CRP) on improvement the computer codes used for fuel behaviour simulation under the name: FUMEX III [1].The major research objective of this CRP would be to test and develop fuel modeling codes against experimental data and cases provided by IAEA and OECD/NEA [1,2]. Institute for Nuclear Research (INR) Pitesti participated at this CRP with ROFEM and CAREB computer codes [3, 4].The main aim of ROFEM code was to calculate fuel behaviour during steady state operating conditions [5].CAREB code was developed for fuel transients analyses such as LOCA and RIA [6].Recently both codes have been improved with new models in order to extend their capabilities [3 ]. The behaviour of the fuel elements during high-temperature transients is of importance to safety and licensing of the power reactor[7]. During the initial depressurization phase of a hypothetical loss of coolant accident in a pressurized water reactor, the fuel element will be subjected to rapid, high-temperature transient. The FIO- 131 and FIO-13 tests was performed to increase the knowledge base of CANDU fuel behavior under LOCA conditions and to provide additional data for validation of the transient fuel performance codes. 31

32 In this paper, the outline of the various models involved in the CAREB code and the results of the calculations on the FIO-131 and FIO-13 in-reactor tests conditions are summarized. The purpose of this study was to investigate the feasibility of using CAREB code to perform a thermal-mechanical analysis of the fuel element response under hypothetical loss of coolant accident conditions. The analysis of the FIO-131 and FIO-13 tests and the simulation of the test with ROFEM and CAREB codes was an important lesson learned about in-reactor LOCA tests. As a result, this lesson will be used in C2-LOCA tests planned to be performed in TRIGA research reactor of INR Pitesti. 1. CAREB Code The CAREB computer code was developed to simulate the thermal-mechanical response of a fuel element during rapid, high-temperature transients [6].The model assumed is a single UO 2 /Zircaloy sheath element with axi-symmetric properties. Physical effects considered in the code are: -Expansion, Contraction, cracking and melting of the fuel, -Variation of internal gas pressure, during the transient, -Changes in the fuel/sheath heat transfer, -Thermal, elastic and plastic sheath deformation (anisotropic) -Zr/H 2 O chemical reaction effects -Beryllium assisted crack penetration of the sheath (initiated from Be-brazed appendages) The new release of the code, CAREB. 1B, extends the capability of CAREB to model of sheath failure due to oxygen embrittlement upon rewetting and effect of oxide strengthening on sheath creep [3]. Other new features improved ROFEM-CAREB interface [3]. The equations used in the model to represent these physical effects use transient boundary conditions of coolant temperature, coolant pressure and sheath/coolant heat transfer coefficient evaluated by thermal/hydraulic codes. The conditions at the start of the transient are obtained from steady/state fuel performance code ROFEM [3].The CAREB code monitor conditions leading to sheath failure.several failure mechanisms are explicitly represented in the code: Sheath overstrain, Localized overstrain under oxide cracks Excessive sheath creep rate, Low ductility sheath failure, Beryllium-assisted crack penetration, High fuel enthalpy (>735 kj/kg UO 2 ), Oxygen embrittlement Beryllium-assisted crack penetration failure mechanism specific to CANDU fuel elements. Intergranular cracking of the Zircaloy fuel sheath can occur at bearing pad and spacer pad locations of the element brought on by the penetration of a beryllium-braze alloy in the presence of an applied hoop stress[8]. The thermal transient experienced by the sheath represent a complex heat treatment which cause changes in anisotropy, annealing, grain size, phase transformation(α-β) plus other changes in sheath microstructure which effect the plastic creep behaviour of the 32