Research in Nuclear Engineering- opportunity for Ph.D students

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1 Warsaw University of Technology Faculty of Power and Aeronautical Engineering Institute of Heat Engineering Power Engineering Division Research in Nuclear Engineering- opportunity for Ph.D students September 25, 2014

2 The greater danger for most of us lies not in setting our aim too high and falling short; but in setting our aim too low, and achieving our mark. Michelangelo WARSAW UNIVERSITY OF TECHNOLOGY NUCLEAR ENGINEERING

3 Competences of the Institute of Heat Engineering WARSAW UNIVERSITY OF TECHNOLOGY FACULTY OF POWER AND AERONAUTICAL ENGINEERING INSTITUTE OF HEAT ENGINEERING General competences in heat, mechanical and electrical engineering Conventional Energy Technologies (fossil) Optimization of generation Modeling and simulation Sustainable energy Rational Use of Energy (RUE) Renewable Energy Sources (RES) Combine Heat and Power Production (CHP) Low Carbon Energy Technologies Nuclear Technology Fuel Cells Carbon Capture and Storage (CCS) Smart Grid Energy Storage Socioeconomic issues Energy law Liberalized Energy Market Impact of energy technologies on environment (climate change)

4 EDUCATION NUCLEAR POWER ENGINEERING AT WARSAW UNIVERSITY OF TECHNOLOGY

5 Short history of Nuclear Engineering at WUT Year Event No. Of Prof Initiation of Nuclear Power Engineering Education No. of PhD No. PhD Students Reactor Safety analysis of VVER Start of liquidation of the unfinished power plant Żarnowiec 1992 Education and Nuclear research slow decay 2006 Reactivation of Nuclear Power Engineering Education 2009 Full Master Program - Nuclear Power Engineering (4 semesters since 2013 in English) Creation of Center for Applied Modeling in Nuclear Engineering SARWUT Project INSPE Program Horizon 2020, SARWUT 2, OECD/PKL Phase 3, DOE?

Center for Applied Modeling in Nuclear Engineering at the Institute Heat Engineering Center for Applied Modeling in Nuclear Engineering Calculation codes:

1 Severe accident analysis: MELCOR 2.1, 1.8.6 RELAP/SCDAPSIM ASTEC V2 Monte Carlo simulations: MCNP5, MCNPX TRIPOLI-4 ver 4.7 Core physics: APOLLO2 ver. 2.8 CRONOS2, ver.

6 Center for Applied Modeling in Nuclear Engineering at the Institute Heat Engineering Center for Applied Modeling in Nuclear Engineering Calculation codes: Thermal-hydraulic analysis: RELAP5 MOD3.3 RELAP5 3D TRACE CATHARE 2 V25_2mod8.1 Severe accident analysis: MELCOR 2.1, RELAP/SCDAPSIM ASTEC V2 Monte Carlo simulations: MCNP5, MCNPX TRIPOLI-4 ver 4.7 Core physics: APOLLO2 ver. 2.8 CRONOS2, ver CFD: FLUENT FLICA4 ver Dose calculation, radioactive plume distribution: MACCS FLEXPART V9.02 Code Saturne Structural analysis ANSYS Code Aster ver. 12 Applications simplifying process of modeling: SNAP

7 INNOVATIVE Nuclear and Sustainable Power Engineering - PHD STUDIES PhD Studies funded by EU program, PO KL 4.1 Studies in English National partners: Warsaw University of Technology, Gdańsk University of Technology, National Centre for Nuclear Research, PGE EJ 1 Sp. z o.o. New laboratory, access to multidisciplinary calculation codes Scientific visits at national and foreign nuclear institution Program started in February 2014 Scientific visits at national and foreign nuclear institution, realization helped by international partners:

8 RESEARCH AT WARSAW UNIVERSITY OF TECHNOLOGY

Fuel pin design. Thermohydraulics. Enhancement of safety. Low or negative void effect.

9 Research: Nuclear Engineering at WUT Reactor Safety Analysis Modeling and simulation of power plant operation (from reactor up to Balance of Plant equipment) Studies of innovative designs Neutronic core design. Core configuration. Enrichment search. Fuel pin design. Thermohydraulics. Enhancement of safety. Low or negative void effect. Optimization of Shielding against radiation. ( ) * +, / IT systems including control and data acquisition systems and cyber security issues! ) ;;)? +, - 3.

Developing reactor models The development of reactor models It is a time consuming process and requires a a lot efforts and good understanding of the phenomena that occur in many reactor components /

10 Developing reactor models The development of reactor models It is a time consuming process and requires a a lot efforts and good understanding of the phenomena that occur in many reactor components / systems. The development of reactor models for the analysis the severe accidents, that take into considerations: reactor core meltdown, hydrogen production, transport of radioactive substances in the reactor containment and into environment. Nodalization of loop #1 of the primary side of the generic model of PWR 900 MWe in RELAP5. Nodalization scheme of reactor pressure vessel in MELCOR code.

Accident modeling WUT Team has modeled following accident using thermohydraulic codes: RELAP5, TRACE, CATHARE: Small break loss of coolant accident (SBLOCA) Large break loss of coolant accident

11 Accident modeling WUT Team has modeled following accident using thermohydraulic codes: RELAP5, TRACE, CATHARE: Small break loss of coolant accident (SBLOCA) Large break loss of coolant accident (LBLOCA) Main steam line break (MSLB) Steam Generator tube rupture (SGTR) The methodology for these accidents have been created. Michał Pawluczyk (WUT) GEH Wilmington (September December 2014) Thermal-hydraulic calculations for design and licensing BWRs, using TRACG system code. Possibly evaluate extended LOCA coping capability (ESBWR has a 7 day extended Station Blackout capability, objective is to determine corresponding response in LOCA). Develop corresponding LOCA evaluation in TRACE. This can involve applying a TRACG/TRACE conversion utility. SBLOCA, EPR: system pressure Piotr Mazgaj EDF SEPTEN (October November 2014) Assessment of the behavior of a two dimensional nodalization of the vessel downcomer, in particular during the refilling stage of the accident (accumulators refilling). The work will consist in : - the analysis of the experimental data, in particular with respect to the downcomer behavior - the handling of the LSTF ROSA 2 CATHARE input data deck realised by the CATHARE team - the realisation of sensitivity studies to identify and assess the most influential parameters and related uncertainties (nodalization, physical models, boundary conditions, geometry...)

12 Research topic two phase flow Two-phase CFD simulation using a commercial CFD codes with adding of some closure laws. Area: two-phase flow beyond bubbly flow. Kacper Samul and Michał Spirzewski expected to visit Ohio in 2015

Severe Accident modeling WUT Team is modeling the following severe accidents using MELCOR: Core meltdown accidents Loss of offsite power Station blackouts Large radioactive releases The methodology

13 Severe Accident modeling WUT Team is modeling the following severe accidents using MELCOR: Core meltdown accidents Loss of offsite power Station blackouts Large radioactive releases The methodology for these accidents have been created. NPP models, 14 models: RELAP5 :Generic PWR, 3-loop, Zion, PWR, 4-loop, CPY, ABWR model, EPR model CATHARE: Generic - PWR, 3-loop, CPY, Zion, PWR, 4- loop in development, EPR model TRACE: BWR, Peach Bottom, Generic 3-loop MELCOR: Peach Bottom, BWR, Zion, PWR, 4-loop, EPR model Piotr Darnowski (WUT) - GEH Wilmington (September December 2014) Severe accident, core melt, analysis of ESBWR. Subject to PIA GEH can provide a MELCOR 1.8 model for ESBWR, and MAAP severe accident scenario. 1) Update ESBWR MELCOR base deck used in the Fission Product removal analysis to the current GEH MELCOR version , select and re-execute one of the three scenarios in NEDE-33279P. 2) Execute a PRA Level 2 event and Benchmark MELCOR against an existing MAAP scenario. 3) Use 1 & 2 to evaluate the real time performance of MELCOR Core meltdown accident, total loss of power, EPR reactor s: large core degredation, lower bottom of RPV is begining to break

Severe Accident modeling Eleonora Skrzyperk (October 2014 March 2015) Thermohydraulic modelling of a steel metal layer on top of a corium pool in a PWR under severe accident conditions This project

14 Severe Accident modeling Eleonora Skrzyperk (October 2014 March 2015) Thermohydraulic modelling of a steel metal layer on top of a corium pool in a PWR under severe accident conditions This project will consist in performing detailed numerical calculations and analyses associated to these two phenomena (for example, using CFD for the first point) in order to help improving the associated 0D models developed in the PROCOR platform. Maciej Skrzyperk (October 2014 March 2015) Simplified thermo-mechanical modelling of the core support plate in a PWR under severe accident conditions The goal of the project is to develop a simplified model coupling thermal and mechanical models in the PROCOR framework. Then the thermal and mechanical stabilization of the corium pool above the core support plate will be studied for different scenarios using this simplified model. One of the perspectives of this modelling approach may be an extension to the vessel failure modelling taking into account the pressure in the vessel. Location: CEA-Cadarache, Severe Accidents Physics & Modelling Laboratory (LPMA)

Modeling impact of a severe accident at a Nuclear Power Station The MACCS

for modelling impact of nuclear incidents on the environment of a nuclear

Population Release specification Output data: Radioactive material

15 Modeling impact of a severe accident at a Nuclear Power Station The MACCS or Melcor Accident Consequence Code System, is a professional tool used for modelling impact of nuclear incidents on the environment of a nuclear facility Input date: Site localization, Spatial grid Weather condition Population Release specification Output data: Radioactive material concentrations Doses received by population Health consequences Contamination areas

SG, Heatexchangers Steamgenerator, heat exchangers and tanks Department Mathematical modeling of steam generators in steady states Beheviour of steam generators in transient processes

16 SG, Heatexchangers Steamgenerator, heat exchangers and tanks Department Mathematical modeling of steam generators in steady states Beheviour of steam generators in transient processes during normal operations Transient processes of steam generators during some accidents like Steam Line Break (SLB) Calculations of loads (forces) applied to internal parts of steam generators

Conceptual design of a 300 MWth lead fast reactor core. Design of the fuel rods, fuel assembles, control rods, core geometry.

17 Research in fast reactors Conceptual design of a 300 MWth lead fast reactor core. Analysis of the conceptual lead fast reactor ELSY (1500 MWth). Reactor was developed in the frame of the ELSY (European Lead SYstem) project initiated and cofunded by the FP6 EURATOM. Conceptual design of a 300 MWth lead fast reactor core. Design of the fuel rods, fuel assembles, control rods, core geometry. Calculation of the main parameters. Sodium Fast Reactors Gurgacz Sebastian (Ocotber 2014 March 2015) CEA Saclay Analyses of Natural Convection and Thermal Stratification Phenomena in Sodium Cooled Fast Breeder Reactors.

18 Gas Fast Reactors High Temperature Test Facility Test installation at Oregon State University Scaling ratios (Dimensions 1:4, Pressure 1:8, Temperatures 1:1) Power 2.2 MWt Ceramic core structure Graphite electric heaters Coolant helium Malwina Gradecka and Izabela Gutowska (joint PhD program WUT-OSU)

19 RESEARCH COMPLETED PROJECTS, WORKSHOP

SARWUT PROJECT SARWUT- Safety Analysis Report at Warsaw University of Technology Grant Issued by National Center for Research and Development (NCBiR) in the field of reactor safety National Center

20 SARWUT PROJECT SARWUT- Safety Analysis Report at Warsaw University of Technology Grant Issued by National Center for Research and Development (NCBiR) in the field of reactor safety National Center for Research and Development (NCBiR) is a state body that provides funding for research and development, especially ones that are focused on industrial applications. Topic of a grant: Elaboration of methods for the safety analysis of PWRs and BWRs in case of disturbances in coolant system (SBLOCA, LBLOCA) and serious accidents. Codes: RELAP5, TRACE, MELCOR. Grant Awarded Project started October 3rd 2012 Duration Oct 2012 August 2014 Possible prolongation of the Research --> Project SARWUT 2

Workshop #1: Familiarization with calculation codes application and safety analysis workshop with AREVA, 9-10 September 2013 Participants: AREVA, Warsaw University of Technology, National Atomic

21 Workshop #1: Familiarization with calculation codes application and safety analysis workshop with AREVA, 9-10 September 2013 Participants: AREVA, Warsaw University of Technology, National Atomic Energy Agency (PAA), National Center for Nuclear Research (NCBJ), Silesian University of Technology, Gdańsk University of Technology (PG), Polish Nuclear Society (PTN), Environmentalists For Nuclear Energy (SEREN Polska), Institute of Nuclear Chemistry and Technology (ICHTJ), Inspecta AB, Baltyn Consulting, PGE EJ 1 Sp. z o.o. The following scenarios was discussed: Scenario 1 Calculation codes: CATHARE, RELAP5 Small Break Loss of Coolant Accident (SBLOCA), 3-loop model, 900 Mwe, PWR Scenario 2 Calculation codes: CATHARE, RELAP5 EPR - 20 cm2 Cold Leg Leak Scenario 3: Calculation codes: MELCOR, MAAP LOOP (loss of offsite power) scenario

Workshop #2: SARWUT Reactor Safety Analysis Workshop #2 2-5 June 2014, Stockholm Participants: Warsaw University of Technology, Royal Institute of Technology (KTH) The workshop consisted of the

22 Workshop #2: SARWUT Reactor Safety Analysis Workshop #2 2-5 June 2014, Stockholm Participants: Warsaw University of Technology, Royal Institute of Technology (KTH) The workshop consisted of the following seminars: Simulations of Processes in Nuclear Reactors HPMC Training Seminar - High Performance Monte Carlo reactor core analysis project (HPMC) has been set up to develop and promote advanced high-performance Monte Carlo simulations of nuclear reactors. Exchange of experience in modeling reactor accidents

Workshop #3: SARWUT Reactor Safety Analysis Workshop #3 with AREVA 24 june 2014 Participants: AREVA, Warsaw University of

Technology, Gdańsk University of Technology (PG), Polish Nuclear Society (PTN), Environmentalists For Nuclear Energy

The following scenarios was discussed: Scenario 4 Calculation codes: CATHARE, RELAP5 CPY: Main feedwater regulation

23 Workshop #3: SARWUT Reactor Safety Analysis Workshop #3 with AREVA 24 june 2014 Participants: AREVA, Warsaw University of Technology, National Atomic Energy Agency (PAA), National Center for Nuclear Research (NCBJ), Silesian University of Technology, Gdańsk University of Technology (PG), Polish Nuclear Society (PTN), Environmentalists For Nuclear Energy (SEREN Polska), Institute of Nuclear Chemistry and Technology (ICHTJ), AGH University of Science and Technology. The following scenarios was discussed: Scenario 4 Calculation codes: CATHARE, RELAP5 CPY: Main feedwater regulation malfunction Scenario 5 Calculation codes: CATHARE, RELAP5 Severe accident, core melt, operation of core catcher, analysis of EPR (MELCOR, MAAP)

24 RESEARCH FUTURE PROJECTS/ACTIVITIES

25 SARWUT 2 - WUT-NCBJ SARWUT 2 Source of financing: National Center for Research and Development (NCBiR) Cooperation: WUT NCBJ Formal actions are ongoing to start the project. Timeline expectation: Advanced Nuclear Power Plant Safety Analysis for normal operation and accidents conditions with results validation and verification at the experimental facilities. 1. Conduction of the safety analysis a nuclear power plant for a set of accident conditions defined on the basis of a catalogue of postulated initiating events for the nuclear power plant and its location, design accidents set.

26 SARWUT 2 - WUT-NCBJ 2. Conduction of the safety analysis for design extension conditions of the nuclear plant for the following complex sequences shall be considered: 1) anticipated transients without scram, which may result in radioactive releases outside the reactor s primary containment; 2) entire loss of alternate current power supply; 3) accidents involving a containment bypass; 4) entire loss of the cooling system function; 5) rupture of the reactor cooling system piping with a simultaneous loss of one line of reactor core emergency cooling system; 6) uncontrolled water level decrease during low water level operation, especially during repair, renovation or refueling; 7) entire loss of functions of the indirect cooling systems of components important to nuclear safety and radiation protection in all installation states requiring the indirect cooling; 8) loss of capability to transfer heat to the ultimate heat sink; In design extension conditions, in case of PWR, the following shall be additionally considered: 1) uncontrolled dilution of boric acid in the PWR; 2) rupture of multiple heat exchange tubes in the steam generator of PWR; 3. Safety analysis of accident sequences involving a containment bypass, which may lead to direct radioactive release outside the primary containment, even without fuel melting.

27 SARWUT 2 - WUT-NCBJ 4. Conduction of the nuclear power plant safety analysis under severe accidents conditions which could lead to a damage of the primary containment at an early phase are prevented or it shall be indicated that the probability of their occurrence is so small that they do not need to be considered in the design: 1) hydrogen explosion; 2) damage of reactor vessel at a pressure possibly leading to: a) eruption of melted core material as well as direct primary containment heat, or b) formation of high energy fragments endangering integrity of reactor s primary containment; 3) steam explosion which could jeopardise the integrity of the primary containment; 4) reactivity accidents, including heterogenic boron dilution (in PWR). 5. Conduction of the safety analysis of the processes that are present during severe accidents, particularly: 1) retaining and cooling the melted reactor core; 2) limiting the effects of melted reactor core on concrete; 3) limiting the releases from reactor s containment, taking into consideration loads related to cladding oxidation, hydrogen combustion and other loads that are possible during severe accidents. 6. Probabilistic safety assessments of NPPs 1) level 1, where: 2) level 2, where the paths of possible release of radioactive material from the nuclear power plant to the environment are determined and the magnitude and frequency of such releases are estimated. 7. Estimation of exclusion zone around a nuclear power plant. Estimation of the effective dose during normal operation, design base accident and severe accidents

WUT and NCBJ to participate in Horizon 2020 EURATOM FISSION NFRP-2014 NARSIS -> New Approach to Reactor Safety Improvements WUT to do reactor safety analysis Proposal submitted 17 SEP 2014.

These advances would concern three main domains: The consideration of concomitant external events, either simultaneous-yet-independent hazards or cascading events The vulnerability of the elements to

28 WUT and NCBJ to participate in Horizon 2020 EURATOM FISSION NFRP-2014 NARSIS -> New Approach to Reactor Safety Improvements WUT to do reactor safety analysis Proposal submitted 17 SEP NARSIS project aims at making a significant scientific jump to propose some updates in the elements required for the safety assessment. These advances would concern three main domains: The consideration of concomitant external events, either simultaneous-yet-independent hazards or cascading events The vulnerability of the elements to complex aggressions (vector-based fragility surfaces The reduction of the uncertainties related to the integration of the expert judgment in the PSA The proposal encompasses 6 Work-packages: WP1: Hazard Characterisation and scenarios WP2: Vulnerability, robustness and resilience WP3: Integration and risk analysis WP4: Validation (Experiments + Simulations) WP5: Supporting Tool for Severe Accident Management WP6: Dissemination & training

Letter of Interest and Participation Proposal with Rensselaer Polytechnic Institute (RPI) Multiscale Modeling Approach

29 US Department of Energy s Office of Nuclear Energy US Department of Energy s Office of Nuclear Energy Funding Opportunity Announcement: DE-FOA Letter of Interest and Participation Proposal with Rensselaer Polytechnic Institute (RPI) Multiscale Modeling Approach to the Analysis of BWR RCIC Performance under Severe Accident Conditions Program RC-7, RCIC Performance under Severe Accident Conditions: Multi-Phase Analysis under the solicitation number noted above.

OECD PKL LOOP PKL III OECD Program Integral test facility, simulating a 1300 MW PWR Possible participation in the last experiments, but there is a need to provide a participation fee.

30 OECD PKL LOOP PKL III OECD Program Integral test facility, simulating a 1300 MW PWR Possible participation in the last experiments, but there is a need to provide a participation fee. Genuine knowledge, comparison of models with real data Conducted tests: H1/H2: Beyond-design-basis accidents with significant core heat up H1: SB-LOCA with failure of safety systems H2: Station Blackout (Performance of PWR- Instrumentation during beyond-design-basis accidents) H3: Accidents during cold shut-down conditions (failure of RHRS) H4: Cool down under asymmetric NC conditions (i.e. with isolated SGs) H5: Boron precipitation following LB-LOCA H6 H8: 3 Experiments connected to Heat transport in the steam generators with the core operating at high power (ATWS), Passive system, extensions to previous experiments)

TAEK/TETAS Project - Turkey Proposal for TETAS/TAEK WUT is subcontractort to: RWE Power International, RE GmbH Proposal concerning the Independent Modelling based on Russian used models for

Asymmetric insertion of reactivity due to steam line break 3. Asymmetric insertion of reactivity due to feed water line break 4. Main coolant pump shaft break 5.

31 TAEK/TETAS Project - Turkey Proposal for TETAS/TAEK WUT is subcontractort to: RWE Power International, RE GmbH Proposal concerning the Independent Modelling based on Russian used models for AES 2006: Final decision at the end of September 2014 List of Accident Scenarios to be modeled: 1. Asymmetric insertion of reactivity due to control rods, 2. Asymmetric insertion of reactivity due to steam line break 3. Asymmetric insertion of reactivity due to feed water line break 4. Main coolant pump shaft break 5. Large leak from primary to secondary in steam generator 6. Large break loss of coolant accident 7. Fuel storage and transportation 8. Release of radioactivity from systems and equipment 9. Beyond design basis accidents (with core melt outside pressure vessel

Trwające prace, analiza niepewności: Application of T/H code. A consistent application (development, qualification and application) of a thermalhydraulic system code. Ref. Petruzzi, A.

32 Trwające prace, analiza niepewności: Application of T/H code. A consistent application (development, qualification and application) of a thermalhydraulic system code. Ref. Petruzzi, A., & D'Auria, F. (2008). Thermal-Hydraulic System Codes in Nulcear Reactor Safety and Qualification Procedures. Science and Technology of Nuclear Installations, 2008(3), doi: /2008/460795

0) In-kind contribution to the SOFTWARE Assessment: Uncertainty studies on calculations of different SA scenarios for

33 ASTEC (Accident Source Term Evaluation Code) Calculation code: ASTEC (Accident Source Term Evaluation Code) Version 2 (currently V2.0) In-kind contribution to the SOFTWARE Assessment: Uncertainty studies on calculations of different SA scenarios for French PWR 900 with input decks that will be provided by IRSN. A few scenarios will be agreed with ASTEC IRSN Project Leader and will be calculated, and the associated uncertainty studies performed EPR severe accident scenario calculations

RELAP/SCDAPSIM Calculation code: RELAP/SCDAPSIM mainly used to conduct severe accident modeling We have been proposed to participate in the code development activities.

34 RELAP/SCDAPSIM Calculation code: RELAP/SCDAPSIM mainly used to conduct severe accident modeling We have been proposed to participate in the code development activities. Possible area of cooperation: (a) developing improved models for severe accidents for LWRs and CANDUs, (b) developing enhanced graphics options, (c) incorporating advanced fluid property and constitutive correlations for alternative fluid systems, and (d) analyzing experiments that have performed in the Germany Quench and CORA facilities.

Laboratories Radioprotection Laboratory at WUT Nuclear measurement Calculations of Dose Equivalent Rate (DER) Radioprotection

proposal for future experiments with students 2015 first experiments expected Laboratory of dynamics of transport and

35 Laboratories Radioprotection Laboratory at WUT Nuclear measurement Calculations of Dose Equivalent Rate (DER) Radioprotection Radiation absorption ratios Research Reactor at National Center for Nuclear Research MARIA is the sole research nuclear reactor currently operated in Poland. Its power amounts to 30 MW Ongoing discussions with NCBJ to start experiments for PhD Students advanced training Ongoing proposal for future experiments with students 2015 first experiments expected Laboratory of dynamics of transport and condensation of aerosols in the reactor containment: Conception works are ongoing Idea: to measure the influence of different paints at the concrete on the condensation rate at the severe accident conditions 2015 expected start of laboratory

Technology (Sweden) Technische Universität

Agency; PGE (Polish Energy Group) Nuclear

36 Nuclear Engineering Cooperation Close cooperation with: KTH-Royal Institute of Technology (Sweden) Technische Universität Dresden (Germany) Commissariat à l Énergie Atomique (France); National Centre for Nuclear Research National Atomic Energy Agency; PGE (Polish Energy Group) Nuclear Energy Areva; General Electric Hitachi. Westinghouse; AMEC, IRSN I2EN

37 THANK YOU FOR YOUR ATTENTION

ASTEC Model Development for the Severe Accident Progression in a Generic AP1000-Like

ASTEC Model Development for the Severe Accident Progression in a Generic AP1000-Like Lucas Albright a,b, Dr. Polina Wilhelm b, Dr. Tatjana Jevremovic a,c a Nuclear Engineering Program b Helmholtz-ZentrumDresden-Rossendorf