Limiting Characteristics of the New Accident Tolerant Fuel Cladding Concepts

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1 Current Nuclear Fuel System Limiting Characteristics of the New Accident Tolerant Fuel Cladding Concepts Study on the Dissertation CZECH TECHNICAL UNIVERSITY Faculty of Nuclear Sciences and Physical Engineering Department of Nuclear Reactors September 23, 2016 Supervisor: Ing. Mojmír Valach, CSc. Consultants: Dr. Ronald G. Ballinger, Ing. Ondřej Huml, Ph.D.

2 Current Nuclear Fuel System Traditional UO 2 +Zr Alloy Fuel Uranium (MOX) fuel in the form of pellets-in-cladding type fuel rods is used in all LWRs which constitute more than 90 % of the installed nuclear power There is a long-term (approx. 70 years) experience, research and databases available It proved to be eligible for reactor operation and performs very well in most of conditions except a few rare events Serves as a baseline and reference for new fuel development Represents huge financial investments so any change to the current design has to be well justified

3 Current Nuclear Fuel System Nuclear Fuel Cladding Alloy Sn (wt.%) Nb (wt.%) O (wt.%) Fe (wt.%) Cr (wt.%) Ni (wt.%) Zr Zry Balance Zry Balance ZIRLO Balance Low Sn ZIRLO Balance M Balance E Balance E Balance

4 Current Nuclear Fuel System

5 Current Nuclear Fuel System Accident Tolerant Fuels Accident Tolerant Fuels can tolerate a severe accident in the reactor core for a considerably longer time period than the current UO 2 Zr alloy fuel system, while maintaining or improving the fuel performance during normal operations and operational transients. Coping Time is the time to significant loss of geometry such that the fuel can no longer be cooled or cannot be removed from the reactor.

6 Cladding Coatings Current Nuclear Fuel System PVD Process Possible to deposit MAX phases; but limited to powders that are commercially available; Possible to deposit FeCrAl, if powders are available Slow deposition process; Needs vacuum chambers Coating thicknesses typically a few single microns; but up to 2 mm thickness is possible; Typically coarse grained microstructure Cold Spray Process Possible to deposit a wide range of MAX phases; sputter cathodes of elements need to be available; Possible to deposit FeCrAl; using alloy sputter cathode or individual element sputter cathodes High deposition rates; Under ambient conditions Coating thicknesses hundred microns or higher typical; with appropriate control coating thickness can be brought to 2 mm; Fine grained structure

7 Current Nuclear Fuel System Cladding Coatings to be Studied Based on evaluation metrics, preliminary tests and availability of materials three coatings will be studied. Two MAX phases: Ti-Al-C and Ti-Si-C and one metallic coating: FeCrAl

8 Related Work I Current Nuclear Fuel System Evaluation Metrics Applied to Accident Tolerant Fuel Cladding Concepts UO 2 +Th Heterogeneous Fuel Design High-Temperature Oxidation of Coated Zr-1Nb Alloy (electrolytic E110) - CrN, Cr coatings Fuel performance models - FRAPCON/FRAPTRAN; TRANURANUS; BISON;

9 Related Work II Current Nuclear Fuel System Irradiation damage of the Cladding Coatings Corrosion of the Zr-base Alloys Uranium Thermal Conductivity Measurements System models in Relap; TRACE as a source for boundary conditions (need for severe accident modelling - MELCOR, MAAP, SOKRAT...)

10 Current Nuclear Fuel System Manufacturing and Characterization of Coated Cladding Samples Manufacturing of ATF cladding coating samples using different techniques (PVD, cold-spray, evaporating) Thermal properties measurements Mechanical properties measurements Thermal conductivity, heat capacity and emissivity measurements Adhesion and adhesion strength Creep tests (thermal and mechanical), bend tests, elastic modulus measurements, ductility, hardness

11 Current Nuclear Fuel System Modeling of ATF Ion Irradiation Manufactured samples will be irradiated at the Texas A&M University by different accelerators Fast irradiation compared to reactor irradiation, very high dpa is achievable, different sources and energies can be combined Irradiation will be simulated by Monte Carlo based system SRIM/TRIM to determine levels of irradiation damage

12 Current Nuclear Fuel System Characterization of Coated Cladding Samples Autoclave Steady Corrosion Testing High Temperature Oxidation Thermal Shock Testing Unirradiated Propertiesion LWR Corro- High Temp. Thermal Thermal Mechanical Oxidation Shock Properties Properties Ti-Al-C Limited Limited Limited Yes Limited Ti-Si-C Limited Limited Limited Yes Limited FeCrAl Yes Yes Yes Yes Yes Zircaloy-4 Yes Yes Yes Yes Yes Irradiated Properties LWR Corrosion High Temp. Thermal Thermal Mechanical Oxidation Shock Properties Properties Ti-Al-C No No No No No Ti-Si-C No No No No No FeCrAl No No No Limited Limited Zircaloy-4 Yes Yes Yes Yes Yes

13 Current Nuclear Fuel System Fuel Performance Modeling Fuel performance modelling using traditional fuel performance codes (FRAPCON/FRAPTRAN, TRANSURANUS) Advanced multiphysics models in the MOOSE framework (BISON, MARMOT, RELAP7, RattleSnake) Implementation of correlations and characteristics measured for particular ATF concept Multi-code utilization for code-to-code V&V All results will be compared to traditional UO 2 Zr fuel system performance

14 Current Nuclear Fuel System Evaluation of ATF Cladding Concepts Based on developed evaluation metrics Consistent evaluation with the OECD/NEA evaluation process [3] Comparison of concepts and their characteristics Basic economic evaluation Recommendation for future improvements and determination of the best concept

15 Final Remarks Current Nuclear Fuel System Dissertation is tight to the MIT s project Development of Accident Tolerant Fuel Options For Near Term Applications which involves two industry collaborators (AREVA and Anatech) and three other universities (U. of Wisconsin, Texas A&M, and Penn State U.) Development of the ATF fuels included in the long-term strategy of the Department of Nuclear Reactors Student grant No. SGS16/252/OHK4/3T/14 focused on New Cladding Concepts for the Accident Tolerant Fuel and their Limiting Characteristics received in 01/2016 Nominated for Expert Group on Accident Tolerant Fuels for Light Water Reactors in the OECD/NEA where results and progress can be reported Applied for research contract at the IAEA within CRP project Analysis of Options and Experimental Examination of Fuels for Water-Cooled Reactors with Increase Accident Tolerance which will allow international cooperation with other world-class research teams Due to unavailability of experimental facilities (autoclaves, furnaces, microscopes etc.) other research organizations will be involved (UJP, ÚJV, CV)

16 1) Průběh ilustračních havárií Zasedání EGATFL - definice ilustračních scénářů pro hodnocení ATF paliv Report distribuován mezi členy k připomínkám; bude dostupný na jaře 2017 NEA Workshop on Nuclear fuel modelling in support of safety and performance enhancement of water-cooled reactors (March 2017) Dle článku Evaluation metrics... - definovány dva scénáře: nízko- a vysoko-tlaký Vysokotlaký - dlouhé SBO - do selhání reaktorové nádoby Nízkotlaký - LBLOCA bez ECCS Na základě těchto scénářů bude určen coping time, který je poskytnut konkrétním konceptem Kromě těchto havárií je samozřejmě nutné přepočítat klasické DBA - licenční proces

17 2) Zdroj kapitola 4.5 Development of High Thermal Conductivity UO 2 Th Heterogeneous Fuel Nuclear Engineering and Design; submitted 01Aug2016 Koncept vymyšlený na FS ČVUT, se kterým jsem autorovi pomáhal

18 3) Radiační poškození Ozařování těžkými ionty na Texas A&M Výhody - krátký čas, dostupné zařízení, možnost studia adheze tenké vrstvy, lze studovat změny chování materiálu při ozáření (pouze vrstva a přechod; vlastnosti substrátu jsou známé) Nevýhody - špatná přenositelnost do reálných podmínek; musí být stejně provedeny testy v reaktoru (HRP); takto lze spíše koncept vyloučit než potvrdit jeho komplexní chování při ozáření

19 4) Testy v autoklávech Vzorky v autoklávu - testy při fixních parametrech; lze měnit se chemii PO (PWR, VVER, BWR) nebo teplotu; parametry jsou však fixní Jde o testy zejména korozní odolnosti nikoliv testy TH parametrů paliva Cílem je zkoumání adheze; korozní kinetiky (hmotnostní přírůstek); koroze při narušení vrstvičky (oxidace probíhá pod nanesenou vrstvou) Ozářením těžkými ionty se změní materiálová struktura a korozní kinetika bude jiná; jde především o vliv na vrstvu nikoliv na substrát

20 Připomínky k práci Tepelná roztažnost paliva, tepelná vodivost paliva, vlivy vyhoření - okrajově kvůli zaměření na pokrytí (práce měla původně přes 120 stran) Vliv vyhoření na teplotu tavení - velmi sporné dle použitého zdroje Výběr a zdůvodnění LOCA a RIA - klasické licenční výpočty ověřující kritéria přijatelnosti pro palivový systém Širší rešerše v kapitolách 3. a 4. - kapitola 4 je souhrn vlastní práce nikoliv rešerše; kapitola 3 představuje tři vybrané materiály, možnosti nanášení a základní dostupné vlastnosti; Citování literatury - kapitoly 4, 5 a 6 jsou vlastní práce nikoliv rešerše; některé z článků ještě nebyly vydány, proto je jejich citace problematická; Výběr materiálů proběhl na základě vytvořených hodnotích metrik (jak je v práci uvedeno), které jsou popsány v kapitole 5; více viz. článek Evaluation Metrics Applied to Accident Tolerant Fuel Cladding Concepts for VVER Reactors

21 Fig.: Hodnotící metriky pro vyhodocení jednotlivých ATF konceptů

22 References I Lori Braase. Enhanced accident tolerant lwr fuels national metrics workshop report. Technical report, Idaho National Laboratory (INL), Shannon Bragg-Sitton. Development of advanced accident-tolerant fuels for commercial lwrs. Nuclear News, pages 83 91, Shannon M Bragg-Sitton, Jon Carmack, and Frank Goldner. Evaluation metrics applied to accident tolerant fuels. Technical report, Idaho National Laboratory (INL), D.G. Cacuci. Handbook of Nuclear Engineering: Vol. 1: Nuclear Engineering Fundamentals; Vol. 2: Reactor Design; Vol. 3: Reactor Analysis; Vol. 4: Reactors of Generations III and IV; Vol. 5: Fuel Cycles, Decommissioning, Waste Disposal and Safeguards. Handbook of Nuclear Engineering. Springer, LY Cheng, A Cuadra, and N Brown. PWR plant model to assess performance of accident tolerant fuel in anticipated transients and accidents. Technical report, Brookhaven National Laboratory (BNL), EGATFL. Light water reactor accident tolerant fuel: Evaluation metrics and technology readiness level definition. Technical report, OECD/NEA, 2016.

23 References II MT Farmer, Leonard Leibowitz, Kurt A Terrani, and Kevin R Robb. Scoping assessments of atf impact on late-stage accident progression including molten core concrete interaction. Journal of Nuclear Materials, 448(1): , KJ Geelhood and WG Luscher. FRAPCON-3.5: integral assessment Isabel Idarraga-Trujillo, Marion Le Flem, Jean-Christophe Brachet, Matthieu Le Saux, Didier Hamon, Sébastien Muller, Valérie Vandenberghe, Marc Tupin, Emilie Papin, Eric Monsifrot, et al. Assessment at CEA of coated nuclear fuel cladding for LWRs with increased margins in LOCA and beyond LOCA conditions. In Conference Paper) LWR Fuel Performance Meeting, Top Fuel 2013, volume 2, pages , Yang-Hyun Koo, Jae-Ho Yang, Jeong-Yong Park, Keon-Sik Kim, Hyun-Gil Kim, Dong-Joo Kim, Yang-Il Jung, and Kun-Woo Song. Kaeri s development of lwr accident-tolerant fuel. Nuclear Technology, 186(2): , Marion Le Flem, Xingmin Liu, Sylvie Doriot, Th Cozzika, Fabien Onimus, and Jean-Luc Bechade. Ti3(Si,Al)C2 for nuclear application: investigation of irradiation effects induced by charged particles Xingmin Liu, Marion Le Flem, Jean-Luc Béchade, and Isabelle Monnet. Nanoindentation investigation of heavy ion irradiated Ti3(Si,Al)C2.

24 References III Y Yamamoto, BA Pint, KA Terrani, KG Field, Y Yang, and LL Snead. Development and property evaluation of nuclear grade wrought fecral fuel cladding for light water reactors. Journal of Nuclear Materials, 467: , Steven J Zinkle, Kurt A Terrani, Jess C Gehin, Larry J Ott, and Lance Lewis Snead. Accident tolerant fuels for LWRs: A perspective. Journal of Nuclear Materials, 448(1): , [9], [2], [13], [14], [8], [11], [10], [6], [7], [5], [12],[1]