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

2. WHAT IS “RADIATION STRENGTH” ● ● α кр α0α0  0 раз 00  кр Defect scope α0α0 brittle failure ● Simplified failure diagram The ability of a material or a structure to withstand failure during operation period under radiation exposure conditions Radiation strength The ability of a material or a structure to withstand failure during operation period Strength

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

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

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

7. A PLAN OF REACTOR VESSEL CALCULATION FOR BRITTLE FAILURE RESISTANCE (under load disturbance at emergency cooling conditions) Reactor vessel wall width Temperature Reactor vessel wall width Stresses Probability of a defect > a Upper zone 1.Stresses from internal pressure 2.Loads in the flange and support areas 3.Concentration of stresses in the zone of nozzles Static and cyclic loading Lower zone 1.Stresses from internal pressure 2.Irradiation metal embrittlement 3.Non-stationary temperature stresses during emergency cooling

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

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

Local approach is “a bridge", linking micromechanisms of a failure at atomic and dislocation levels with macrofracture of the material physical mechanism of a failure brittle failure is chipping or microchipping; viscous failure 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 application in the failure mechanics permits to calculate the limiting state and durability of structural elements local criterion local approach failure mechanics 10. LOCAL APPROACH IN FAILURE MECHANICS

Is used to find critical values of failure mechanics variables K IC, J C and the dependences describing the kinetics of cracks J R (  a), General principle for finding critical parameters 1.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 J or К) for which the failure criterion for an individual cell or a cell conglomeration is met. 11. LOCAL APPROACH IN CRACKS MECHANICS

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

Temperature dependency of failure viscosity for reactor vessel steels with different embrittlement rate at В=25 mm and P f =0.5 is described by the following equation MPa  m with =26 MPa  m; Т – temperature 0 С Parameter  is the only variable that depends on material embrittlement rate. Parameter  decreases with material embrittlement rate increase 13. «UNIFIED CURVE» Method

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

15. Schematic representation of the requirements for nuclear reactor vessel materials Т K  30°С at the end of operation term Ensured reactor vessel lifetime for 60 years minimum High resistance to brittle failure in initial state Т K0  -35°С High resistance to thermal and radiation embrittlement Ensuring needed weldability and manufacturability Ensuring needed material quality

16. RUSSIAN REACTOR VESSEL STEELS Steel brand, basic composition Used scince Types and number of reactors Strength category Maximum wall thickness, mm Initial brittleness critical temperature Т ко, °С Operation temperatur e, °С Designed fluence, n/cm 2 Brittleness critical temperature values at the end of operations °С 15Х2МФА 0,12 %С, 2,8 % Cr, 0,8 % Мо, 0,2 % V 1958 VVER pcs.; Navy installations  300pcs.; Nuclear icebreakers - 18 pcs. КП 00 270 (2  2,4)  Х2НМФА 0,12 % С; 2,3 % Cr Ni(1,0  1,5)% 0,8 %Mo; 0,15% V 1973 VVER pcs КП  (4  6)  Requirements for steels used in the manufacture of advanced reactor vessels 2007 All types of water-water reactors КП   ,4   30

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

18. EVOLTUION OF STEELS USED IN NUCLEAR POWER REACTOR VESSELS MANUFACTURE Decreased contents of detrimental impurities Rationing of Ni content ( %) replacement Mo for W Increased Ni content (0,6- 0.8%) Decreased contents of detrimental impurities Decreasing Ni down to 1.3% Сталь 15Х2МФА Сталь 15Х2МФА-А Steel 15Х2МФА-А мод. А Steel 15Х2МФА-А мод. Б Steel with fast drop of induced activity 15Х2В2ФА-А Сталь 15Х2НМФА Сталь 15Х2НМФА-А Сталь 15Х2НМФА кл. 1 КП-40 h max = 400 A F = 15.0 КП-40 h max = 400 A F = 14.0 КП-45 h max = 480 A F = 12.0 КП-45 h max = 520 A F = 12.0 КП-40  45 h max = 400 A F = 12.0 КП-45 h max = 400 A F = 21 КП-45 h max = 400 A F = 23 КП-45 h max = 400 A F = 29  30

19. MANUFACTURE OF A RING PIECE FOR 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”.

Steel of 15Х2МФА-А grade, mod. А ensures strength level corresponding to КП-45 strength category with a margin of 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 minus 75 - minus 95 0 С. 20. MECHANICAL PROPERTIES OF METAL ОЗП 1 test sample 2 test sample 1 test sample Basic t/t 2 test sample Basic t/t + min. PWHT cycle Basic t/t+ max. PWHT cycle Ultimate strength at С, Yield strength at С, Mpa Yield strength КП-45 Ultimate strength КП-45 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 MPa.

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 Overload capacity Т ка T К0 = -35°C years Steel 15Х2НМФА-А steels 15Х2МФА-А mod.А (Ni – 0,2÷0,4%) 15Х2МФА-А mod.Б (Ni – 0,6÷0,8%) Т К, °С A F = 12 A F = 23 Operation period, years EUR requirements: Т K = 30°C

22. DURABILITY OF VVER REACTOR VESSEL INTERNALS: NEW CHALLENGES Neutron irradiation +  -radiation Primary coolant Loading due to reactor heating and cooling + vibration resistance to cracking wear in corrosive environment corrosion cracking radiation swelling + radiation creep wear BASIC OPERATIONAL FACTORS INFLUENCING THE PERFORMANCE OF VESSEL INTERNALS Material of internals - steel Х18Н10Т

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 (J C ) by more than 10 times  decreased resistance to fatigue failures and corrosion cracking 23. WHY TO ASSESS STRUCTURAL STRENGTH AND PERFORMANCE OF VESSEL INTERNALS PERFORMANCE OF VESSEL INTERNALS CAN DEGRADE CAN DEGRADE

24. SOME PROBLEMS IN MATERIAL SCIENCE AND METHODOLOGY TO BE SOLVED IN ORDER TO DEVELOP A TECHNIQUE FOR THE CALCULATION OF VESSEL INTERNALS STRENGTH Material Science ProblemsMethodological tasks Radiation swelling (depending on temperature and neutron irradiation damaging dose) Calculation of strain-deformation state of the elements of vessel internals under thermal- mechanical loading with neutron irradiation considered. Radiation creep (depending on temperature and accumulation rate of neutron irradiation damaging dose) Viscosity of failure (depending on operation temperature, irradiation temperature and neutron irradiation damaging dose) Setting-up the marginal state of an internals’ element according to the criterion of non-steady crack development with phase change considered Strength according to the criterion of crack birth under corrosion cracking (depending on neutron irradiation damaging dose) A technique of damage calculation according to cracking mechanisms under long-time static loading and fatigue under cyclic loading Resistance to fatigue (depending on irradiation temperature and the damaging area taking into account corrosive environment) Rate of crack propagation according to corrosion cracking mechanism (depending on load and neutron irradiation damaging dose) Rate of fatigue crack propagation in air and in corrosive environment (depending on cycle load, loading frequency and neutron irradiation damaging dose) Conditions of phase change realization as swelling function

25. SWELLING OF Х18Н10Т STEEL n=1,88; Т max = 470 о С; с=1,035  r h =1,804  o C -2 r l =1,5  o C -2 Damaging dose, dpa

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, T irr = °C a) T test = 20°Cb) T test = 495°C

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

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

29. DESIGN REQUIREMENTS TO VESSEL INTERNALS IN TERMS OF γ → α TRANSMUTATION Reflection shield scheme x coolant (water) coolant (water ) Distribution of temperature and neutron fluence along the reflection shield wall thickness F, T irr x T irr F α +γα +γ γ 12 T irr, °C D, dpa design ofinternals γ → α transmutationbrittle- viscous transition A good and a bad example of the design of internals in terms of γ → α transmutation, resulting in brittle- viscous transition: good design 1 – good design bad design 2 – bad design

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

1.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. 2.Intended set of complex operational features of the steels is provided for by metal rough parts with thickness up to 525 mm. 3.All new reactor steels, weld materials and welding technologies have been put into industrial production and can be used to manufacture reactor vessels soonest. 31. CONCLUSIONS

32. MAIN ACTIVITY AREAS 1.Industrial development and comprehensive certification of steels and welds for big- and medium-power reactors. 2. Development and wide-range examination of highly radiation resistant steel meant for use in vessel internals of big-power reactors. 3.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. 4.Using material science findings to support lifetime extension of operating different-purpose nuclear power installations.

THANK YOU!