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

WARSAW UNIVERSITY OF TECHNOLOGY – NUCLEAR ENGINEERING

“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 ourmark.” ― Michelangelo

FACULTY OF POWER AND AERONAUTICAL ENGINEERING

INSTITUTE OF HEAT ENGINEERING

WARSAW UNIVERSITY OF TECHNOLOGY

• 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)

Competences of the Institute of Heat Engineering

EDUCATION – NUCLEAR POWER ENGINEERING

AT WARSAW UNIVERSITY OF TECHNOLOGY

Short history of Nuclear Engineering at WUT

Year Event No. OfProf.

No. ofPhD

No. PhDStudents

1959 Initiation of Nuclear Power Engineering Education

Reactor Safety analysis of VVER 440

1990 Start of liquidation of the unfinished power plant –Żarnowiec

1992 Education and Nuclear research – slow decay

2006 Reactivation of Nuclear Power Engineering Education

1 1

2009 Full Master Program - Nuclear Power Engineering (4 semesters since 2013 in English)

1 3

2011 Creation of Center for Applied Modeling in NuclearEngineering

1 3 1

2012 SARWUT Project 1 3 3

2 3 9

2014 INSPE Program 2 3 12

2015 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 EngineeringCalculation codes:

• Thermal-hydraulic analysis:• RELAP5 MOD3.3• RELAP5 3D• TRACE• CATHARE 2 V25_2mod8.1

• 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. 2.10

• CFD:• FLUENT• FLICA4 ver. 1.10

• Dose calculation, radioactive plume distribution:• MACCS• FLEXPART V9.02• Code Saturne 2.0.7

• Structural analysis• ANSYS• Code Aster ver. 12

• Applications simplifying process of modeling:• SNAP

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:

RESEARCHAT WARSAW UNIVERSITY OF TECHNOLOGY

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

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Developing reactor models

The development of reactor modelsIt is a time consuming process and requires a a lot efforts and good understanding of the phenomena that occur in many reactor components / systems.

Nodalization of loop #1 of the primary side of the genericmodel of PWR 900 MWe in RELAP5.

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 scheme of reactor pressure vessel in MELCOR code.

Accident modeling

SBLOCA, EPR: system pressure

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.

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 ...) ”

Research topic – two phase flow

Two-phase CFD simulation using acommercial CFD codes with adding of someclosure laws.

Area: two-phase flow beyond bubbly flow.

Kacper Samul and Michał Spirzewskiexpected to visit Ohio in 2015

Severe Accident modeling

Core meltdown accident, total loss of power,EPR reactor.

19511 s: large core degredation, lowerbottom of RPV is begining to break

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 1.8.2 base deck used in the

Fission Product removal analysis to the current GEH MELCOR version 2.1.4803, 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 2.1.4803.

Eleonora Skrzyperk (October 2014 – March 2015)Thermohydraulic modelling of a steel metal layer ontop of a corium pool in a PWR under severe accidentconditions

This project will consist in performing detailed numericalcalculations and analyses associated to these two phenomena (forexample, using CFD for the first point) in order to help improving theassociated 0D models developed in the PROCOR platform.

Maciej Skrzyperk (October 2014 – March 2015)Simplified thermo-mechanical modelling of the coresupport plate in a PWR under severe accidentconditions

The goal of the project is to develop a simplified model couplingthermal and mechanical models in the PROCOR framework. Then thethermal and mechanical stabilization of the corium pool above thecore support plate will be studied for different scenarios using thissimplified model. One of the perspectives of this modelling approachmay be an extension to the vessel failure modelling taking intoaccount the pressure in the vessel.

Location: CEA-Cadarache, Severe Accidents Physics &Modelling Laboratory (LPMA)

Severe Accident modeling

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

Output data:• Radioactive material concentrations• Doses received by population• Health consequences• Contamination areas

Input date:• Site localization, • Spatial grid• Weather condition• Population• Release specification

Modeling impact of a severe accident at a Nuclear Power Station

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

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 ThermalStratification Phenomena in Sodium CooledFast Breeder Reactors.

Gas Fast Reactors

High Temperature Test Facility

• Test installation at Oregon StateUniversity

• Scaling ratios (Dimensions 1:4, Pressure1:8, 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)

RESEARCHCOMPLETED 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 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 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 #22-5 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 AREVA24 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)

RESEARCHFUTURE PROJECTS/ACTIVITIES

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: 2015 – 2017

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.

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.

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, orb) 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 NPPs1) 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

SARWUT 2 - WUT-NCBJ

• EURATOM FISSION NFRP-2014

• NARSIS -> New Approach to Reactor Safety Improvements

• WUT to do reactor safety analysis

• Proposal submitted – 17 SEP 2014.

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

WUT and NCBJ to participate in Horizon 2020

• US Department of Energy’s Office of Nuclear Energy

• Funding Opportunity Announcement: DE-FOA-0000998. 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.

US Department of Energy’s Office of Nuclear Energy

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 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 break3. Asymmetric insertion of reactivity due to feed water line break4. Main coolant pump shaft break5. Large leak from primary to secondary in steam generator6. Large break loss of coolant accident7. Fuel storage and transportation8. Release of radioactivity from systems and equipment9. 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., & D'Auria, F. (2008). Thermal-Hydraulic System Codes in Nulcear Reactor Safety and Qualification Procedures. Science and Technology of Nuclear Installations, 2008(3), 1–16. doi:10.1155/2008/460795

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.

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• Radiation absorption ratios

Research Reactor at National Center for Nuclear ResearchMARIA is the sole research nuclear reactor currentlyoperated 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 ofaerosols 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 accidentconditions

• 2015 – expected start of laboratory

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

Nuclear Engineering– Cooperation

THANK YOU FOR YOUR ATTENTION