operating experiences and approaches for fatigue ... · pdf fileoperating experiences and...

Technical Meeting on Fatigue Assessment in Light Water Rector for Long Term Operation, 6.-8. July 2016, Erlangen, Germany

Operating experiences and approaches for fatigue assessment in operating

nuclear power plant

Ivan Lopoš, VUJE, Inc. Department of NPP Life-time Evaluation,

Division of Diagnostics of Nuclear Components

LTO of VVER 440 NPP units

Slovak NPP units

NPP unit

Reactor type

Start of operation

Lifetime end according to NPP design (expected LTO) Bohunice 1 WWER 440/V230

1978

shut down in 2006

Bohunice 2 WWER 440/V230

1980

shut down in 2008

Bohunice 3

WWER 440/V213

1984

2014, (2044 LTO)

Bohunice 4

WWER 440/V213

1985

2015, (2045 LTO)

Mochovce 1

WWER 440/V213

1998

2028,(2058 LTO)

Mochovce 2

WWER 440/V213

2000

2030, (2060 LTO)

Mochovce 3

WWER 440/V213

assumed 2017

2047, (2077 LTO)

Mochovce 4

WWER 440/V213

assumed 2018

2048, (2078 LTO)


Assessment of fatigue life

History: Fatigue life assessment is carried out by a decision of the Nuclear Regulatory Authority for all NPP units in Slovakia. It is implemented on an agreed range of equipment which are periodically evaluated on the basis of operation history monitoring since 1994.

Bohunice: EBO V2 – after each fuel campaign

Mochovce: EMO1,2 – after each fuel campaign

Range of monitored equipment : Selected NPP components cover most stressed devices and components: Primary circuit pipeline components - (nozzles, branches, elbows connected to the steam generator , pressurizer, etc.) Primary circuit main components - (reactor pressure vessel , main coolant pump, main isolating valve, steam generator, pressurizer)


Range of monitored equipment


Pipeline components pipeline double branch Ms 500/200 pipeline branch of low pressure reactor emergency cooling system for hot leg and cold leg pipeline branch of high pressure reactor emergency cooling system pipeline branch of pressurizer cold injection hot elbow under steam generator pressurizer surge line



Primary circuit components reactor pressure vessel reactor vessel hot nozzle sealing part of reactor vessel spherical closure head main coolant pump main isolating valve


Pressurizer components upper manhole body injection nozzle bottom zone with nozzle connection nozzle to safety valve



Steam generator components feedwater nozzle primary header SG shell steam header

header connection area


1. Knowledge of stress state in the assessed components in all operating modes and conditions (analytical or numerical calculation models).

2. Knowledge of all relevant material characteristics (minimum guaranteed material properties of the material standards).

3. Knowledge of operation history (a sequence of load cases for evaluation of CUF, realistic loads/real transients based on actual conditions measured during operation, monitoring of loacal loads at the fatigue relevant components by means of monitoring systems)

4. Assessment methodology or codes and standards (NTD SEV 4201-86 ÷ 4214-86 or Russian Code PNAE – G-7-002-86)

What is needed for the assessment of fatigue life?


operating conditions of NPP components are characterized by changes of operating parameters (internal overpressure, flow rate and coolant temperature)

operation loads arising from thermal stratification are monitored by means of special temperature measurement systems

mechanism of erosion-corrosion damage is negligible for austenitic structural materials but for carbon and ferritic steels (pressurizer, SG, feedwater and steam pipelines) is monitored by means of wall thickness losses and potential defects in critical areas

the results of all measurements are analyzed and taken into account in the analysis of lifetime

Calculations of fatigue damage

Operating modes and conditions :


continuous measurement of operating parameters in selected locations of assessed NPP components and processing of measured parameters boundary conditions definition for individual operating modes and states (load cases definition) numerical simulations of thermoelastic analyses using FEM analysis - resulting structural stresses stress tensor components determination or selection for the most exposed area (critical spots) of NPP component for individual load cases in terms of fatigue creation of stress-states database of selected critical spots for each one of representative LC (load case) calculation of fatigue damage cumulation on the basis of LC′s sequence (operation history)

Calculations of fatigue damage

Sequence of steps in the evaluation of fatigue damage:


Factors affecting fatigue damage assessment


Calculation models

Old less detailed New more detailed

example: SG steam header reduction of calculated max. total fatigue damage after 30. fuel campaign : old less detailed model: 3,77% new more detailed model: 1,33%

Damage for campaign Total damage

Fatig

ue d

amag

e [%

]

Fuel campaign


Fatigue damage cumulation in critical spots of feedwater nozzle NPP Unit

Number of fuel cycles

Node point

Maximal fatigue damage cumulation for steam generator [%]

1 2 3 4 5 6

Bohunice 3 18 1 18,3828 12,0754 23,3934 6,9426 12,5836 7,7793

4 17 1 26,2436 7,6465 8,5672 9,0552 26,2436 23,5576

Feedwater nozzle of steam generator In monitored cross-sections located near the feedwater nozzle were recorded occurrence of different thermal stratification (different number and type of stratification load cases) Different level of fatigue damage cumulation in analysed feedwater nozzles was caused by different operation history

Operation history



Operational load monitoring


Location of termocouples at cross-sections : 1,3,5,7,9,11

Location of termocouples at cross-sections : 2,4,6,8,10,12

Location of sensors for feed water piping (Steam Generators 1 – 6)

Turbine Hall

Containment Penetration

T-618

Non-Return Flap Valve

Location of sensors for feed water piping (Steam Generators 4)

monitoring of thermal load in feed water line was unique for each SG (different types and number of thermal cycles - cycling between stratified and non-stratified conditions)

new monitored cross-section located in the upstream side of non-return flap valve has confirmed the occurrence of stratification in the place where we did not expect (source: valve leakage)

these findings will lead to the installation of additional thermocouples in new pipe cross-sections because occurence of unforeseen (not considered in the design) operating loads (not monitored) on specific structural components could cause their permanent damage

Detection of load that has not been considered in the design


Operational load monitoring new monitoring system „MONEZ“ resulted from the recommendations of the evaluation process of transition to LTO

measuring range was extended to other piping systems where unspecified load occurrence is very probable

unspecified load occurrence was confirmed in the following pipe systems, for example:

• pipeline of pressurizer cold injection

• pipeline of passive low pressure reactor emergency cooling system

• pipeline of active low pressure reactor emergency cooling system


Pressurizer cold injection

Location of termocouples at cross-sections

Active low pressure reactor emergency cooling system

Location of termocouples at cross-sections

Passive low pressure reactor emergency cooling system

Location of termocouples at cross-sections : 1,5

Location of termocouples at cross-sections : 2,3,4



Installed measurement range

3VP2

3VP1

3VP6

3VP5

3VP3

3VP4

Old measurement range



Pressurizer surge line of Unit 4 of the V-2 Bohunice NPP

NFC Node Damage

[%]

17

1 1,1989

2 3,3844

4 1,4095

Fatigue damage cumulation in critical spots



Pressurizer surge line of Unit 3 of the V-2 Bohunice NPP

Fatigue damage cumulation in critical spots

NFC Node Damage

[%]

18 1 61,8218

2 94,2482

significant discrepancy between monitored stratification events on unit 3 of NPP Bohunice and other pressurizer surge lines

different location of the most damaged areas – critical spots




Location of sensors for pressurizer surge line

measurement was modified and extended in order to reduce the conservatism obtaining more detailed fluid temperature profiles over the length of the surge line during transient and steady state operation measurement was characterized by different number and type of stratification load cases for individual monitored cross sections the flow of coolant between pressurizer and cold leg resulted in asymmetric thermal load analysis confirmed the interdependence between the course of measured temperatures and the course of calculated stresses

New measurement range




Course of temperature and equivalent stress

Right branch of surge line Left branch of surge line


Activities associated with lifetime assessment

Assessment of defect acceptability

Determination of acceptable defect dimensions (for safe, quick and easy evaluation of indications/defects found during ISI) Evaluation of stability and acceptability of defects localised in piping systems and a pressure component body (indications revealed by means of ISI) applying methods of fracture mechanics and appropriate normative standards: VERLIFE document procedures British R6 method ASME Code Section XI

Every analysis of defect acceptability has to contain fatigue crack growth assessment caused by alternating operation states during NPP operation


Determination of critical and acceptable defect dimensions

In primary and secondary circuit NPP components (defects represent locations of stress concentration and are located in the weld metal and the surrounding base material) Each defect is conservatively assessed as a potential crack (elliptical cracks with different dimensions a and c) For evaluation of fatigue crack growth is necessary to take into account the load history

Assessment is carried out for important facilities of the primary and secondary circuit :

welds of piping systems: main coolant pipe, pressurizer surge line, main steam line, feedwater line ...

primary circuit components (reactor pressure vessel, reactor vessel hot nozzle, sealing part of reactor vessel, main coolant pump, main isolating valve, pressurizer, steam generator ...)

2c

t a

2c

t 2a

d


Surface crack Subsurface crack

Reactor vessel hot nozzle

Section 1

Section 2

Section 3

Section 4


Reassessment of indications revealed during ISI


If detected defect dimensions exceed the conservatively calculated acceptable dimensions of indications/defects it needs further detailed evaluation using more accurate calculation The following sequence of activities have to be realized for detailed evaluation (accurate calculation) : creation of calculation model boundary conditions proposal and calculation

model debugging definition of operating modes stability assessment of detected defect using

the results of stress state calculations consideration of fatigue crack growth through

simulation operation history (sequence of operating modes) representing the NPP design life

Nozzle weld of SG emergency feedwater supply

1. Significant deviations of actual service conditions from designed ones were found out by monitoring system.

2. Different outputs from individual units measurement revealed that unspecified loading (not considered in the design - stratification) is impossible to predict.

3. Archived data (history of plant operation) are an important source of information for next decision making about plant in future.

4. Assessment of some NPP components has to be reassessed because unforeseen complex loading and insufficient monitoring can lead to a risk of fatigue damage.

5. Reliable assessment of components critical to power generation mitigates risk of unplaned outages due to equipment failure and reduces operation and maintenance costs and optimize the availability of the plant.

Summary and conclusions



Thank you for your attention

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