MAINTAINING RADIATION STRENGTH OF REACTOR VESSELS AND VESSEL INTERNALS IN VVERs

1. FSUE CRI KM «Prometey» MAINTAININGRADIATION STRENGTH OF REACTOR VESSELS AND VESSEL INTERNALS IN VVERs G. P. Karzov Deputy General Director, Doctor of Engineering, Professor

2. 2. WHAT IS “RADIATION STRENGTH” brittle failure ● α0 αкр ● αкр ● Defect scope α0 0раз 0 кр кр Strength The ability of a material or a structure to withstand failure during operation period Simplified failure diagram Radiation strength The ability of a material or a structure to withstand failure during operation period under radiation exposure conditions

3. 3. MAINTAINING RADIATION STRENGTH • Radiation strength can be ensured by a system of back action measures aimed against failure development that are applied at all stages of a structure life-time: • its design • manufacture, and • operation • The main characteristics of radiation strength is no-failure life-time of a structure

4. 4. BASIC COMPONENTS OF A NUCLEAR POWER REACTOR Control devices (upper unit) Reactor dome Vessel Vessel internals Core

5. 5. A SYSTEM OF RADIATION STRENGTH ENSURING - ENSURED SAFE LIFETIME OPERATION OF REACTOR VESSEL AND VESSEL INTERNALS Design Manufacture Operation Scientific-engineering support Advanced metallurgical technologies, increased metal purity Studies of metal radiation-caused failure physics Structural optimization Field inspection of metal defects Studies of metal failure mechanisms and plotting models Advanced welding technologies and thermal treatment methods A program of testifying samples. Monitoring metal radiation failure development rate. Choice of material Metal quality control. Non-destructive control of production defects Loading mode optimization Development and implementation of compensating measures: annealing, pack-shields Development of methods for safe operation period calculation Setting-up actual time period of safe operation Calculating safe operation period during design stage and in operation process Is the only, though indirect assessment way. Should definitely be less than actual operation period. Not known!!! Cannot be defined by direct experimenting

6. 6. A scheme of reactor vessel strength and durability validation based on experiment-calculated methods Technological operations promoting the material resistance to various operational damages and destructions Operational factors promoting the material degradation and destruction Conditions of metal load External loads Plastic treatment Thermal treatment Welding Molding Thermal ageing Temperature loads Neutron irradiation Residual stresses Assessment of loads in the areas of stress concentrations Corrosion environment impact Computed assessment of the safe operational life The design life is not ensured The design life is ensured Decrease of the stress concentration levels Development of technical documentation for manufacturing and production quality control Development of technical decisions to ensure the design life Introduction of a favorable residual stresses system Improvement of production technology Development of the ways to improve the material features in the most loaded facility areas

7. 7. A PLAN OF REACTOR VESSEL CALCULATION FOR BRITTLE FAILURE RESISTANCE (under load disturbance at emergency cooling conditions) Reactor vessel wall width Temperature • Upper zone • Stresses from internal pressure • Loads in the flange and support areas • Concentration of stresses in the zone of nozzles • Static and cyclic loading Reactor vessel wall width • Lower zone • Stresses from internal pressure • Irradiation metal embrittlement • Non-stationary temperature stresses during emergency cooling Stresses Probability of a defect > a Reactor vessel wall width

8. 8. A DIAGRAM OF REACTOR VESSEL CALCULATION FOR BRITTLE FAILURE RESISTANCE (finding subcritical neutron fluence) Maximum curve KJC (T) in irradiated state in initial state KJ (T) (load under the thermal shock Maximum displacement – embrittlement temperature Life-time: Fluence limit: Ф – neutron flux

9. 9. SOME CHALLENGES OF KJC(T) DEPENDENCY FORECASTING FOR HEAVILY EMBRITTLED MATERIALS Steel 15Х2НМФА, heavily embrittled state (thermal treatment) Weld joint material KS01, heavily embrittled state (neutron irradiation) Curves – a prognosis made by the methods applying horizontal shift condition Dots – experimental data For heavily embrittled materials the shape of KJC(T) is changed, hence there is a need to use methods accounting for this change

10. 10. LOCAL APPROACH IN FAILURE MECHANICS Local approach is “a bridge", linking micromechanismsof a failure at atomic and dislocation levels with macrofracture of the material physical mechanism of a failure brittle failure is chipping or microchipping; viscousfailure is the development of microvoids; fatiguefailure is fatigue wear damage; creep failure is intergranular cavitation failure failure local criterion is a failure criterion expressed in the terms of deformable body mechanics with internal variables related to failure physical mechanisms and material structure. local approach local criterion failure mechanics Local approach applicationin the failure mechanics permits to calculate the limiting state and durability of structural elements

11. 11. LOCAL APPROACH IN CRACKS MECHANICS Is used to find critical values of failure mechanics variables KIC, JCand the dependences describing the kinetics of cracksJR(a), General principle for finding critical parameters • A material is viewed as a conglomeration of elementary cells with a failure local criterion set up for all of them. • 2. Load-strain state is calculated for the crack tip and the load parameters are found (for example Jor К) for which the failure criterion for an individual cell or a cell conglomeration is met.

12. 12. «UNIFIED CURVE» Method Calculation results obtained by a local criterion of brittle failure for different embrittlement rates Method МКc-КР-2000 (РД ЭО 0350-02) Similarity of curves (left-side figure): when normalized to a certain level KJC= the curves are “reduced” into a “Unified curve”

13. 13. «UNIFIED CURVE» Method MPam with =26 MPam; Т – temperature0С Temperature dependency of failure viscosity for reactor vessel steels with different embrittlement rate at В=25 mmandPf=0.5is described by the following equation Parameteris the only variable that depends on material embrittlement rate. Parameter decreases with material embrittlement rate increase

14. 14. COMPARISON OF KJC(T) DEPENDENCIES OBTAINED BY «MASTER CURVE» AND «UNIFIED CURVE» METHODS FOR REACTOR VESSEL MATERIALS WITH HIGH EMBRITTLEMENT RATE облученный материал (ORNL, США) охрупченный материал (ЦНИИ КМ «Прометей») облученный материал (VTT, Финляндия) MASTER CURVE UNIFIED CURVE Embrittlement material (CRI KM Prometey) Irradiated material (VTT, Finland) Irradiated material (ORNL, USA) Embrittlement material(CRI KM Prometey

15. REQUIREMENTS FOR REACTOR VE SSEL MATERIALS Strength level КП - 45 ³ R 44 0 М П а ; p 0,2 ³ R 5 40 МПа m at Т = 350 ° С 15. Schematic representation of the requirements for nuclear reactor vessel materials Ensuring needed weldability and manufacturability Ensuring needed material quality High resistance to High resistance brittle failure to thermal and radiation in initial state embrittlement £ Т - 3 5 °С K 0 £ Т 30 °С K at the end of operation term Ensured reactor vessel lifetime for 60 years minimum

16. 16. RUSSIAN REACTOR VESSEL STEELS

17. 17.INCREASED RADIATION STRENGTH OF Cr-Mo-V STEELS AS A RESULT OF PandCuIMPURITY DECREASE Impact of impurities on embrittlement of steels Decreased impurities content resulting from advanced steel melting technologies Radiation embrittlement coefficient Af Concentration of Cu, SнP (x10) Irradiation temperature, 0C a) b)

18. Сталь  15Х2МФА Сталь  15Х2НМФА КП-40 hmax = 400 AF = 15.0 КП-45 hmax = 400 AF = 29ё30 Decreased contents of detrimental impurities Decreased contents of detrimental impurities Сталь  15Х2МФА-А Сталь  15Х2НМФА-А КП-40 hmax = 400 AF = 14.0 КП-45 hmax = 400 AF = 23 Decreased contents of detrimental impurities Rationing of Ni content (0.2-0.4%) DecreasingNi down to 1.3% Сталь  15Х2НМФА кл. 1 КП-45 hmax = 400 AF = 21 Steel  15Х2МФА-А мод. А КП-45 hmax = 480 AF = 12.0 replacementMo for W IncreasedNi content (0,6-0.8%) Steel with fast drop of induced activity 15Х2В2ФА-А Steel  15Х2МФА-А мод. Б КП-40ё45 hmax = 400 AF = 12.0 КП-45 hmax = 520 AF = 12.0 18. EVOLTUION OF STEELS USED IN NUCLEAR POWER REACTOR VESSELS MANUFACTURE

19. 19. MANUFACTURE OF A RING PIECEFOR CONNECTING PIPES SECTION An experimental-industrial ring piece for VVER-1000 reactor connecting pipes was manufactured from a 235,0 ton 15Х2МФА-АВ steel ingot. The activity was done within the framework of the project called “Integrated research and manufacturing activities to study the feasibility of 15Х2МФА-АВ steel , modification A with strength category КП-45, for VVER reactor vessels manufacture”, funded by “Rosenergoatom”.

20. 20. MECHANICAL PROPERTIES OF METAL ОЗП Steel of 15Х2МФА-А grade, mod.А ensures strength level corresponding to КП-45 strength category with a margin of 50-30 MPa after basic thermal treatment and additional tempering in minimum and maximum cycles. Hence there is a possibility for additional manufacturing tempering (for example, when a structure is made more complex or during maintenance). Initial critical brittleness temperature is minus75 - minus950С. The steel has good tempering stability - the degradation of its strength features after additional tempering is 50 MPa maximum; The difference in strength characteristics after minimal and maximal manufacturing tempering operations is 10-20 MPa. Ultimate strength at 350 0С, Yield strength at 350 0С, Mpa Ultimate strength КП-45 Yield strength КП-45 Basic t/t+ min. PWHT cycle Basic t/t+ max. PWHT cycle Basic t/t 1 test sample 2 test sample 1 test sample 2 test sample

21. ТК, °С 80 Overload capacity Тка Steel 15Х2НМФА-А 40 AF = 23 EUR requirements: ТK = 30°C AF = 12 steels15Х2МФА-А mod.А (Ni – 0,2÷0,4%)15Х2МФА-А mod.Б (Ni – 0,6÷0,8%) 0 TК0 = -35°C -40 50 100 Operation period, years 60 years 21. COMPARISON DATA OF RADIATION EMBRITTLEMENT OF STEEL 15Х2НМФА-А,15Х2МФА-А . mod. А(Ni – 0,2÷0,4%) and mod. Б(Ni – 0,6÷0,8%) IN AES-2006 REACTOR OPERATION CONDITIONS

22. 22. DURABILITY OF VVER REACTOR VESSEL INTERNALS:NEW CHALLENGES BASIC OPERATIONAL FACTORS INFLUENCING THE PERFORMANCE OF VESSEL INTERNALS Material of internals - steelХ18Н10Т Primary coolant Loading due to reactor heating and cooling + vibration Neutron irradiation + -radiation resistance to cracking wear in corrosive environment corrosion cracking radiation swelling + radiation creep wear

23. 23. WHY TO ASSESS STRUCTURAL STRENGTH AND PERFORMANCE OF VESSEL INTERNALS • Heavy neutron irradiation results in: • development of serious strain due to swelling gradient • objectionable distortion of elements’ shapes and dimensions • decrease in resistance to cracking (JC) by more than 10 times • decreased resistance to fatigue failures and corrosion cracking PERFORMANCE OF VESSEL INTERNALS CAN DEGRADE

24. 24. SOME PROBLEMS IN MATERIAL SCIENCE AND METHODOLOGY TO BE SOLVED IN ORDER TO DEVELOP A TECHNIQUE FOR THE CALCULATION OF VESSEL INTERNALS STRENGTH

25. 25. SWELLING OF Х18Н10Т STEEL Damaging dose, dpa Damaging dose, dpa rh =1,80410-4oC-2 rl=1,510-4oC-2 n=1,88;Тmax= 470оС;с=1,03510-4

26. 26. SERIOUS DEGRADING OF INTERNALS’ MATERIAL BEHAVIOR. RELATIONSHIP BETWEEN RADIATION EMBRITTLEMENT AND SWELLING SURFACE OF DAMAGED SAMPLES MADE FROM PARENT METAL OF Х18Н10Т STEEL IRRRADIATED BY A DOSE OF 49 dpa, Tirr=400-450°C a) Ttest = 20°C b) Ttest = 495°C

27. 27. A MODEL OF MATERIAL FAILURE AFTER γ → α TRANSMUTATION

28. 28. IRRADIATION PARAMETERS RESULTING IN γ → α TRANSMUTATION AND IN DEVELOPMENT OF BRITTLE-VISCOUS TRANSITION IN AUSTENITIC STEELS (Sw)с = 7% D, сна In parent γ -phase(no brittle-viscous transition) (α+ γ) PHASES(there is brittle-viscous transition) Tirr, °C for flux dpa/s: Equation variables for Х18Н01Т steel: СD=1.035·10-4, n=1.88 Tmax=470°C r=1,5·10-4 °С-2

29. 29. DESIGN REQUIREMENTS TO VESSEL INTERNALS IN TERMS OF γ → α TRANSMUTATION coolant (water) x α+γ coolant (water) D, dpa 1 2 F, Tirr Tirr γ Tirr, °C F x Reflection shield scheme A good and a bad example of the design ofinternals in terms of γ→ αtransmutation, resulting in brittle-viscous transition: 1 –good design 2 –bad design Distribution of temperature and neutron fluence along the reflection shield wall thickness

30. 30. MATERIALS FOR VESSEL INTERNALS IN VVER-TYPE REACTORS Maximum permissible swelling swelling Prospective material swelling Applied material Maximum permissible plasticity Period of irradiation, years current material— steel Х18Н10Т potential material—steel with increased nickel content and nanostructure in the form of short range ordering domains

31. 31. CONCLUSIONS • Newly developed reactor vessel steels together with corresponding welds can remove any restrictions limiting the lifetime of nuclear reactor vessels because of metal radiation embrittlement. • Intended set of complex operational features of the steels is provided for by metal rough parts with thickness up to 525 mm. • All new reactor steels, weld materials and welding technologies have been put into industrial production and can be used to manufacture reactor vessels soonest.

32. 32. MAIN ACTIVITY AREAS • Industrial development and comprehensive certification of steels and welds for big- and medium-power reactors. • Development and wide-range examination of highly radiation resistant steel meant for use in vessel internals of big-power reactors. • Improving calculation analysis methods used to assess the deterioration of structural materials working in various nuclear reactors and creating calculation methods to support the verification of their safe operational lifetime. • Using material science findings to support lifetime extension of operating different-purpose nuclear power installations.

33. THANK YOU!