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Korea Atomic Energy Research Institute 2004. 11.09-11 THE PRELIMINARY PERFORMANCE ANALYSIS OF THE TRANSMUTATION FUEL FOR HYPER THE PRELIMINARY PERFORMANCE ANALYSIS OF THE TRANSMUTATION FUEL FOR HYPER B.O.Lee, W.S.Park, Y.Kim, T.Y.Song OECD/NEA 8 th Information Exchange Meetings on Actinide and Fission Product Partitioning and Transmutation Las Vegas, Nevada, USA 9 – 11 November 2004

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Korea Atomic Energy Research Institute Contents - Contents- - Introduction - Code description - Design Parameter - Fuel Temperature Prediction - Fission Gas Release and He Release Rate Insertion - Strain Limits Analysis - Cumulative Damage Fraction Analysis - Conclusions 2

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Korea Atomic Energy Research Institute Introduction 3 U-TRU-(40-60)Zr metallic fuel for HYPER in Korea HYPER ( HYbrid Power Extraction Reactor ) : ADS & sub-critical system U-TRU-15Zr metallic fuel for critical system Steady-state performance computer code development MACSIS-H (A Metallic fuel performance Analysis Code for Simulating In-reactor behavior under Steady-state conditions) Parametric study for the selection of nominal design characteristics and operating limits Material data of U-Pu-Zr : used for those of U-TRU-Zr fuel Fuel temperature distribution : check the operational limits Constituent migration analysis : by the quasi-binary U-Zr model He production rates : inserted into the swelling/FGR routine Burnup limits : derive the design concept CDF(cumulative damage fraction) : estimate the failure probabilities

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Korea Atomic Energy Research Institute MACSIS-H Code Description 4 ● Main structure - Fuel temp. calculation routine - Swelling and FGR calculation routine - Cladding deformation calculation routine ● Main Function - Axial and radial temperature distribution - Fuel slug swelling - Fission gas release including He release - Fuel constituent migration - Cladding deformation by plenum pressure - Cumulative damage fraction by the input of other program - Cladding wastage effect by eutectic melting (now developing) - FCMI by solid fission product (now developing) MACSIS-H Flow Chart

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Korea Atomic Energy Research Institute Design Parameter 5 Sub-critical (HYPER)Critical Fuel Slug Contents (wt%)8U-34.3Pu-4.2Am-1.6Cm- 1.7Np-0.2RE-50Zr 65.5U-18Pu-0.5Am-0.4Cm- 0.5Np-0.5RE-14.6Zr 241 Am Content (wt%)2.660.1 Fuel Slug Diameter (mm)5.636.63 Smeared Density (%)75 Pin Outer Diameter (mm)7.78.8 Cladding Thickness (mm)0.60.57 Fuel Slug Length (mm)1,5001,240 Peak Linear Power (kW/m)28.528.9 Coolant Outlet Temperature (°C)490540 Cladding MaterialHT9 ● Key Design Parameter

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Korea Atomic Energy Research Institute 6 ● Transmutation Fuel Composition (w/o) in an Equilibrium Cycle HYPER IsotopeFeedChargeDischarge U-234 U-235 U-236 U-238 Np-237 Pu-238 Pu-239 Pu-240 Pu-241 Pu-242 Am-241 Am-242m Am-243 Cm-242 Cm-243 Cm-244 Cm-245 Cm-246 Re FP* 1.7E-3 0.060 0.040 6.46 6.25 2.09 46.96 20.80 4.00 4.24 6.73 0.016 1.12 3.8E-5 0.003 0.22 0.023 0.0031 0.98 0.00 0.68 0.18 0.32 15.35 3.35 4.89 23.13 28.10 4.22 8.16 5.31 0.30 2.71 0.018 0.019 1.91 0.61 0.38 0.37 0.00 0.62 0.17 0.31 14.15 2.17 4.36 14.30 24.08 3.75 7.36 3.76 0.30 2.51 0.21 0.018 1.99 0.61 0.38 3.72 15.25 Critical IsotopeFeedChargeDischarge U-234 U-235 U-236 U-238 Np-237 Pu-238 Pu-239 Pu-240 Pu-241 Pu-242 Am-241 Am-242m Am-243 Cm-242 Cm-243 Cm-244 Cm-245 Cm-246 Re FP* 0.017 0.60 0.40 64.80 2.09 0.70 15.70 6.95 1.34 1.42 2.25 0.0053 0.37 1.28E-5 0.0010 0.075 0.0078 0.0011 3.28 0.00 0.17 0.15 0.43 75.24 0.62 0.87 10.48 7.36 0.83 1.44 0.10 0.056 0.43 0.0020 0.26 0.079 0.045 0.55 0.00 0.15 0.087 0.39 68.86 0.41 0.79 8.92 6.65 0.76 1.30 0.71 0.056 0.39 0.027 0.0021 0.27 0.079 0.045 2.26 7.84 * Without RE Design Parameter

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Korea Atomic Energy Research Institute 7 Fuel Temperature Prediction ● Fuel Temperature Prediction Thermal conductivity for the unirradiated U-TRU-Zr alloy by Billone et al. Thermal conductivity for more than 20wt% of Zr or more than 30wt% TRU by AAA Report (LA-UR-02-2630) The porosities and sodium infiltration effect by Bauer et al The predicted temperatures of U-Zr metal fuel by MACSIS-H versus LIFE-M by Billone, Pahl, Hofman et al. reasonably good capability in predicting fuel pin rod temperatures

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Korea Atomic Energy Research Institute ● Fuel temperature limits on fuel melting calculated solidus temperature of U-42TRU-50Zr and U-20TRU-14.6Zr : 1090 and 1295 o C, respectively calculated power-to-melt for HYPER fuel: 420 W/cm (in case of 50 % degraded thermal conductivity and at hot channel) calculated power-to-melt for critical system fuel : 500 W/cm 8 The operating limits on linear power rate for metallic fuel pin in HYPER and critical system Fuel Temperature Prediction

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Korea Atomic Energy Research Institute 9 ● Fuel Constituent Migration Based on the Ishida’s model and Hofman’s theory Reconstruct the quasi-binary U-Zr phase diagram by Ishida’s Concept Assumption of the diffusion coefficient by Hofman’s theory Flow Chart of Calculation Scheme Diffusion equation for constituent migration Single phase Multi-phase Boundary Fuel Temperature Prediction

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Korea Atomic Energy Research Institute ● Calculated and measured radial profile of Zr for the U-19Pu-10Zr driving forces acting on the Zr migration : molar enthalpy of solution ( △ Hs), heat of transport (Q*), and concentration gradient The main reason for redistribution : radial solubility change of Zr The heat of transport plays a role in the redistribution discrepancy was small : in case of Q* : -97,000kJ/mole significant amount of Zr is depleted in the middle zone 10 Calculated and measured radial profile of Zr for U-19Pu-10Zr Fuel Temperature Prediction

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Korea Atomic Energy Research Institute 11 ● Calculated radial profile of Zr for the U-20TRU-14.6Zr Zr fraction in fuel center at around 670 o C : 0.5 fuel centerline melting : retarded by addition of Zr Sharp Zr depletion at the upper limit of ( + ) phase boundary melting temperature and eutectic-melting point decreased This phenomenon has not been confirmed yet experimentally Calculated Radial profile of Zr for U-20TRU-15Zr Fuel Temperature Prediction

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Korea Atomic Energy Research Institute ● Margin to slug centerline melting Centerline temperature : very lower than the solidus temperature Sufficient Margin to the slug centerline melting → same for the case of the fuel constituent migration 12 radial temperature for U-20TRU-14.6Zr radial temperature for U-42TRU-50Zr Fuel Temperature Prediction

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Korea Atomic Energy Research Institute ● Fission Gas Release (FGR) Prediction Behavior of intragranular FGR : diffusion theory by Booth Behavior of grain boundary : multiple bubble distribution model by Hwang Percentage gas releases according to burnup variation by MACSIS-H Largely increases at around 1 to 2 at% burnup Fission gas release at a burnup of 10at% : about 70~80 %. The predictions by MACSIS-H with the semi-theoretical models agree comparatively well with the experimental results from ANL 13 Fission Gas Release and He Release Rate Insertion Fission gas release data by Pahl Calculated fission gas release

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Korea Atomic Energy Research Institute ● He generation rates insertion Assumption - Helium production rates from 6-40 wt% 241 Am : 50 ml He per gram of transmuted americium by Meyer - 241 Am weights and the He generation rates : calculated by fuel design spec 14 241 Am weights and the He generation rates Inserting the He generation rates into the code In the MACSIS-H - the volume of fission gas generated are recalculated including the He generation rate Fuel typeContent (wt%) 241 Am weight (g)He generation rate U-42TRU-50Zr 2.669.140.0 ml/165day U-20TRU-14.6Zr 0.10.59.36 ml/165day * time required to achieve 50% transmutation of 241 Am : 2.5 years by Walker Fission Gas Release and He Release Rate Insertion

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Korea Atomic Energy Research Institute ● Strain limits analysis Cladding strain comparison with He effects as a function of the plenum-to-fuel ratio for HYPER fuel - The effects on the strain with different plenum sizes :analyzed by the MACSIS-H code cladding strain by the plenum pressure stress alone He effects will be a very important factor - the burnup limit for the plenum-to-fuel ratio of 1.75 : 33at% by the thermal creep strain limit of 1% - major deformation mechanism : thermal creep strain in the metallic fuel - the thermal creep strain of 1% : used for the burnup limit criteria for metallic fuel : swelling and irradiation creep of HT9 : very small 1.5 and 1.75 times of the plenum-to-fuel ratios : conservative for satisfying the discharge burnup goal 15 Cladding strain of HYPER fuel according to plenum- to-fuel ratio plenum-to-fuel ratio thermal creep strain Without He Thermal creep strain With He 1.500.14% at 25at%0.4% at 25at% 1.750.09% at 25at%0.19% at 25at% Strain limits analysis

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Korea Atomic Energy Research Institute Cladding strain comparison with He effects as a function of the plenum-to-fuel ratio for the critical system fuel - the HT9 cladding is not conservative for satisfying the discharge burnup goal : because of the high coolant outlet temperature - In the metallic fuel of the critical system, : replacement of cladding material with higher thermal creep resistance may be needed 16 plenum-to-fuel ratio Thermal creep strain Without He Thermal creep strain With He 1.502.27% at 9.8at%4.8% at 9.8at% 1.750.64% at 9.8at%1.2% at 9.8at% Burnup limits analysis Cladding strain of critical system fuel according to plenum-to-fuel ratio

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Korea Atomic Energy Research Institute Two kinds of specific design limits strain limit approach strongly dependent on temperature and strain rate cumulative damage fraction (CDF) method based on linear summation of creep damage appropriate for non-stationary stress and temperature loading conditions Weibull analysis probabilistic method of analyzing life-test experiments Useful for the evaluation of component reliability 17 Probabilistic estimation of the CDF (Cumulative Damage Fraction) Derivation of failure distribution functions Evaluation of CDF of the X447 fuel pins Estimation of fuel pin performances Weibull analysis MACSIS-H code HYPER and critical system condition Calculation Scheme for the steady-state conditions Calculation Scheme for the transient conditions Evaluation of CDF of WPF test fuel pin FCTT failure correlation Evaluation of cladding performance WPF transient test data Derivation of failure distribution functions WPF test condition Weibull analysis Estimate of fuel pin performances

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Korea Atomic Energy Research Institute limit on the fuel pin failure rate : less than 0.01% the CDF limit of 0.001 was reasonable Failure probability of the HYPER fuel pin during steady-state condition fuel pin failure rates for the 1.5 and 1.75 plenum-to-fuel ratios : 0.017 and 0.003%: 1.75 times of the plenum-to-fuel ratio : conservative by CDF limit Failure probability of the HYPER fuel pin during transient condition: lower than that of the WPF pin Because of a higher plenum-fuel volume ratio and lower cladding inner radius vs. thickness ratio 18 Fuel pin performance by Weibull distribution Probabilistic estimation of the CDF (Cumulative Damage Fraction)

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Korea Atomic Energy Research Institute 19 Performance analysis for the transmutation fuel by MACSIS-H code Margin to the slug melting temperature Sufficient margin for the fuel of HYPER and critical system Eutectic melting limit One of the main issue, but analysis model is now developing Detailed analysis needed for the fuel of critical system, because of significant amount of Zr depletion in the middle zone by the constituent migration Cladding Strain limit The He effects will be a important factor Replacement of cladding material may be needed for the fuel of critical system CDF limit 1.75 times of the plenum-to-fuel ratio was conservative for the fuel of HYPER at steady-state condition Failure probability of the HYPER fuel pin was low at transient condition Some experimental test are needed for clarifying the uncertainties of the fuel modeling Conclusion

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