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Operation and Maintenance

&

Research and Development

Activities in Czech Republic at the field

of LTO and AM

Technical Working Group on Life

Management of Nuclear Power Plants 22-24 February 2017, IAEA ,Vienna

Miroslav Žamboch

22. 2. 2017

Structure

1. Czech NPPs LTO implementation status

2. PTS Re-evaluation of Czech RPVs

3. Implementation of AMP RVI

4. Evaluation of the Swelling development

NPP Dukovany

4 PWR units - VVER 440/213 type

Operation since 1985

Power up-rate to 500 MWe finished in 2012

6 reactor coolant loops with horizontal steam generators, 2 turbine generators

Original projected life time 30 years (excluding RPV which has 40 years)

2

NPP Temelin

2 PWR units - VVER 1000 V320 type

Operation since 2002

Power up-rate to 1055 MW

4 reactor coolant loops with horizontal steam generators, 1 turbine generator

Original projected life time 30 years (excluding RPV which has 40 years)

3

Structure

1. Czech NPPs LTO implementation status




In 2016 NPP Dukovany Unit 1 acquired permission to operate beyond design life time – The permission is conditioned by requirements of SONS

that must by fulfilled according to agreed time schedule

– In the end of the 2016 the owner asked for permission for Unit 2 – According to the Czech law this must be decided during

following six months

For LTO implementation the Operator has been using approach and guidance presented in corresponding document of IAEA

LTO status in Czech Republic - Dukovany

5

NPP Temelin is in the half of design lifetime

– But the first work for LTO preparation has started,

specifically elaboration of strategic (most general) control documentation

– Preparation of tools for AMR (database of materials used, medium, TLAA etc.)

– Goal is to make LTO implementation swifter, easier and cheaper than at EDU

LTO status in Czech Republic – Temelin

6

Structure

1. NPP Dukovany LTO implementation




5

TNR VVER 440 TNR VVER 1000

oblast svaru 0.1.4

oblast svaru 3

oblast svaru 4

Evaluated locations at VVER 440 and VVER 1000

oblast nátrubku

oblast

nátrubku

History of PTS Evaluation in Czech Republic

9

The beginning in 1996

On the basis of today outdated methodologies:

Normative Technical Documentation of the Association of Mechanical

Engineers, section IV. – Lifetime Assessment of VVER components and

piping

Guidelines and recommendation for lifetime assessment of RPV and

internals during operation, SONS (State Office for Nuclear Energy)

1998

Guidelines on Pressurized Thermal Shock Analysis for WWER Nuclear

Power Plants, IAEA-EBP-WWER-08, MAAE, Vídeň, 1997

First evaluation performed even for surface crack! (later recalculated)

History of PTS, ETE example

10

72 system thermal – hydraulic calculation was performed in 2001-2004

On their basis the 22 PTS scenario was chosen and consequently analysed according to at that time valid methodologies

(VERLIFE 2003 - later NTD ASI 2004)

Scenarios were devided into 4 groups:

LOCA … Loss of Coolant Accident

MSLB … Main Steam Line Break

PRISE … Primary to Secondary Leak

Others

Review of analyzed PTS scenarios:

History of PTS ETE, Example

11

Group of

Accident

System Thermal –

Hydraulic Analysis

Mixing Therma

Hydraulic Analysis and

Structural Analysis

LOCA 18 5

MSLB 13 4

PRISE 15 2

OTHERS 26 11 (+2)

SUMMARY 72 22 (+2)

Structural calculation were performed for

Postulated underclad crack with depth 1/10 s (wall thickness) –

qualified value at that time for NDT inspection ETE ≈ depth 20 mm

With help of elastic–plastic fracture mechanics

Result of the PTS Evaluation ETE - Worst Scenarios,

Input into New Assesment

Group of

Accidents

PTS

scenario

Tka

[ºC]

Crack

depth

[mm]

Approach Welds

no.

Orientation a/c

C32min 64,6 20 WPS 4 obvodová 0,7

LOCA PP210min 76,1 20 TEČNÝ 4 obvodová 0,7

H300min 90,1 20 WPS 4 obvodová 0,7

H850 102,7 20 WPS 4 obvodová 0,3

SLB1A 126,9 20 WPS 3 osová 0,3

MSLB SLB1B 108,6 20 WPS 3 osová 0,3

SLB1C 111,2 20 WPS 3 osová 0,3

PRISE 3SGT 66,3 20 WPS 3 osová 0,3

SGH1 89,4 20 WPS 4 obvodová 0,7

PSV31 64,1 20 WPS 4 obvodová 0,7

PSV31zra 48,6 13 TEČNÝ 3 osová 0,3

ostatní PSV33 67,7 20 TEČNÝ 4 obvodová 0,3

PSV33zr 60,8 13 TEČNÝ 3 osová 0,3

FB1 75,2 20 WPS 4 obvodová 0,3

Note: Majority of the results is from original project Evaluation of the PTS ETE from 2001-2004 for cracks with depth 20 mm, only PSV31zr and PSV33zr are from evaluation performed for power up rate of the ETE = 13+1 mm

With experience from the existing analysis

According to planned document VERLIFE – That document is going to be transformed in new edition of

NTD ASI (The Association of Czech Mechanical Engineers)

Evaluation is going to be performed even for reference temperature T0 according to Master Curve approach(ASTM E1921)

– With help CFD application FLUENT for thermo hydraulic

mixing analysis (where it is possible)

– For postulated crack length corresponding to todays quality of qulified NDT inspections (13 + 1) mm

New program of the Evaluation of the

RPVs resistance against fast fracture

13

Important changes in new VERLIFE methodology

with respect to the VERLIFE 2008 (PTS chapters)

14

1. New definition of residual stresses (according to MRKR-

SCHR-2004)

2. Residual stresses are included for normal operational

condition and heat affected zone

3. Alternativ [KIC] curve (according to MRKR-SCHR-2004)

4. Warm Pre-stresing approach is accepted even for non –

monotonic unloading

1. Residual Stresses According to IAEA VERLIFE

15

Z, X, N is the local coordinate system;

N is the normal to surface of RPV element;

Z is the tangent to surface of RPV

element.

FIG. 5.1. Schematization of distribution

of residual stresses clres and W

res on

cross-section of the weld of RPV caused

by cladding and post-weld tempering.

FIG. 5.2. The dependence of residual stress Wres

and cladres on duration and temperature of

tempering.

res clad

W

Sclad

0o

N0

N

z

0

NN

res

res

0

100

200

300

400

0 20 40 60 80 100

T=650 oC, T=670 oC

duration of tempering, h

T=670 oC

cladres

Wres

T=650oC

σWres = 56 MPA (EDU)

σWres = 140 MPA (ETE)

Influence of changes in residual stresses postulation for

PTS outputs, example: VVER 440

16

0

10

20

30

40

50

60

70

80

20 40 60 80 100 120 140 160 180 200 220 240 260 280 300

KI,

KIc

[M

Pa

m1/2]

teplota [°C]

Comparison of PTS H200b vs H200_2016 pro circumferential crack a/c=0.3, tangent approach

H200b, 15mm podnávarová H200_2016, Verlife 2008, 13+1mm H200_2016, Verlife 2016, 13+1mm

Tka=114.9°C

Tka=116.7°C Tka=105.4°C

Warm pre-stressing (WPS) principle

0

10

20

30

40

50

60

70

80

90

100

-180 -130 -80 -30 20 70

K [MPa.m1/2]

T [°C]

Schematické zobrazení vlivu WPS na lomovou houževnatost

teplotní závislost lomové houževnatosti bez WPS (virgin material)

LUCF

LPUCF

LCF

Přetíženíza tepla

LPUCF

LUCF

LCF

Example of non monotonic WPS benefit for PSV31zr

scenario

0

10

20

30

40

50

60

70

80

90

20 30 40 50 60 70 80 90 100110 120130140150160170180190200210220230240250260270280290300

[MPa.m1/2]

teplota [°C]

PSV31 (se znovuuzavřením), svar 3, a/c = 0,3, osová trhlina

KI, [KIC]3 v závislosti na teplotě

KI K_IC pro tečný přístup K_IC pro WPS 90% KI_max

T_KA=49,2 C T_KA=71,6 C

According to new methodology it is possible to use WPS approach for this

non monotonic unload scenario, Tka was increased about 20°C (conditions

for using this method– paragraph 5.10 of the VERLIFE procedure were

fulfiled )

Progress in the state of art of PTS Evaluation

The new qualification of the NDT control allows to postulate underclad crack with depth 13 mm (+1 mm into cladding)

In the past it was 20 + 1 mm (ETE) or 15 + 1 (EDU) Advance in hardware development and advance in thermo hydraulic modeling allows use of the new software for TH mixing calculation : FLUENT application

Software application used in the past: RELAP 5d REMIX/NEWMIX, occasionally CATHARE, SYSTUS

Software application for present evaluation: RELAP 5D, updated model of NPP FLUENT (where possible or RELAP 5D) SYSTUS

Software approved by SONS

Fluent advantages: PTS ETE – PSV71zra

20

The FLUENT temperature curves comparison with values from older software NEWMIX

that was used in the past.

It can be seen that outputs from NEWMIX are much more conservative, calculated

temperatures in downcomer have faster and deeper decrease . Reason is more simple

modeling of mixing in NEWMIX, that could not include flow in loops (from SG direction)

where NEWMIX assume total stagnancy.

Fluent benefits and disadvantages

21

Benefits:

-Better simulation of mixing

-In majority cases higher temperatures of RPV walls during PTS (as

consequence higher values of Tka)

Disadvantages -Extensively demanding on computational time (weeks for 1 analysis)

-Difficult to use for LOCA accident analysis and other scenarios with

large leak from primary circuit with turbulent flow, two phase medium and

changes in direction of the flow

-It is not possible to guess benefits prior to analysis itself, in some case

the expectation could be to high

On the basis of the former results the following scenarios were selected 1) Basic series of TH Relap analysis for determination of worst scenarios with

uncontrolled reclosure of pressurizer safety valve 2) PSV71zra

3) PSV73 4) PSV73 zr

5) PSV83 6) LOCA C32min

7) PRISE 3SGT 8) FB1

9) LOCA PP210min 10) PRISE SGH1

11) H300min 12) H850

13) MSLB SLB1B 14) MSLB SLB1C

15) MSLB SLB1A 16) The worst 1 to 2 cases from scenarios above – variation in effectivness of

emergency core cooling branches and output power of the reactor

Present program of the PTS ETE Evaluation

Structure





RVI VVER 1000 – components and materials

Ti stabilized

austenitic SS

08Ch18N10T

Fe-Ni-Cr alloy

ChN35VT-VD

24

Protective

tube unit

Reactor

core barrel

Core basket

NPP ETE parameters

25

Material: 08Ch18N10T

Temperature: max. T (RVI ETE) = 434°C (inner surface of

the reactor core barrel at the axial max. absorbed dose)

Radiation damage (considering 30% uncertainty):

dpa barrel basket

ETE 1 actual 13. fuel cycles 21,9 2,4

prediction 80. fuel cycles 124,7 14,2

ETE 2 actual 12. fuel cycles 20,1 2,3

prediction 80. fuel cycles 133,6 16,6

year

dpa1.7

dt

dF

RVI VVER 1000 – role

Provide support, guidance and

protection to the core

Provide a path for reactor coolant

flow to the core

Provide a path for control elements

and instrumentation

Provide gamma and neutron

shielding for the vessel

26

RVI VVER 1000 – degradation due to irradiation

RVI – focusing to 08Ch18N10T

During operation – influence of thermal, mechanical,

corrosion and radiation attack – degradation of mechanical

properties

The main degradation mechanisms due to radiation

Radiation hardening and loss of fracture toughness

Irradiation swelling

H2 and He generation

Radiation induced segregation

Stress relaxation (creep)

Feγ → Feα transformation

SCC and IASCC

27

Microstructure induced by irradiation

28

Consequences of irradiation on mechanical

properties

29

New phases generation:

G-phase,

carbides

Generation of dislocation loops

Additional internal stresses in the components of

internals

Decrease of cohesive strength of grains

boundaries

Partial Feγ→Feα transformation,

resulting in brittle-to-ductile

transition in austenitic steel

Grain boundary sliding under

radiation creep

Redistribution of alloy and impurity elements

Helium and hydrogen generation

Increase of voids nucleation rate under

deformation. Increasing of strain

localization

Sharp decrease of fracture strain, fracture toughness and ultimate strength for ductile fracture (S>15%)

Decrease of plasticity and sharp decrease of fracture toughness

Decrease of stress corrosion cracking

resistance

Radiation swelling (vacancy

voids)

1. Increase of yield strength

2. Decrease of strength of interphases

3. Decrease of strain hardening

RVI assessment provided by UJV

According to the Czech normative document NTD A.S.I. 2013 (VERLIFE 2008, „Guidelines for Integrity and Lifetime Assessment of Components and Piping in WWER Nuclear Power Plants“) focused on RPVs and NPP pipe lines, no RVI

According to the IAEA VERLIFE

Includes APPENDIX C „INTEGRITY AND LIFETIME ASSESSMENT PROCEDURE OF RPV INTERNALS IN WWER NPP’S DURING OPERATION“

This procedure covers RVI components of WWER water-moderated water-cooled power reactors. It is used to assess RVI lifetime according to the following criteria:

— Crack nucleation and growth by different degradation mechanisms;

— Inadmissible change of RVI component geometrical sizes

30

IAEA VERLIFE procedure for RVI assessment

31

This procedure is applicable to neutron irradiation of material that is characterized by the following ranges of neutron damage dose and irradiation temperature: F ≤ 70 dpa; Tirr=270 to 450 °С

Strength of a component is considered as resistance to loss of carrying capacity and resistance to unstable crack propagation. Analysis of unstable crack growth and loss of carrying capacity is carried out for components with an initial crack. Nucleation and crack growth may happen via IASCC, fatigue mechanisms, and/or due to formation of a limit embrittlement area (LEA).

The following component conditions are considered critical events with respect to component failure

Fatigue crack initiation

IASCC crack initiation

Formation of limit embrittlement area

Loss of component carrying capacity

Inadmissible change of component geometrical sizes

IAEA VERLIFE – List of annexes to Appendix C

32

Annex A MECHANICAL PROPERTIES AND STRESS-STRAIN CURVES FOR RVI

MATERIALS

Annex B CONSTRUCTION OF LOW-CYCLE FATIGUE CURVES FOR RVI MATERIALS

Annex C DOSE-TIME DEPENDENCE OF IASCC INITIATION

Annex D FRACTURE TOUGHNESS OF RVI MATERIALS

Annex E FATIGUE CRACK GROWTH RATE FOR RVI MATERIALS

Annex F TIME-DEPENDENT CORROSIVE CRACK GROWTH RATE FOR RVI MATERIALS IN

WWER ENVIRONMENT

Annex G SWELLING OF RVI MATERIALS

Annex H RADIATION CREEP OF RVI MATERIALS

Annex I A PROCEDURE OF CYCLE FORMATION UNDER NON-RADIAL LOADING

Annex J PROCEDURE OF REFERENCE STRESS CALCULATION

Annex K CONSTITUTIVE EQUATIONS TO CALCULATE VISCOELASTIC-PLASTIC PROBLEMS

BY FEM

Annex L DETERMINATION OF A LIMIT EMBRITTLEMENT AREA AND DIAGRAM OF A

POSTULATED CRACK

Application 1 VIBRATION LOADING AND WEAR OF RVI COMPONENTS

Structure of AMP for RVI

33

Structure





The componet most influenced by swelling

Core baffle

35

Degradation of Core Baffle

Degradation mechanism "radiation swelling„ was detected in core baffle of type of VVER 1000.

Today’s assessment of the state - only computational

36

Temperature distribution for the

12th campaign of the 1st block of

NPP Temelin.

Degradation mechanism caused by a

combination of fluence (dpa) and

temperature (heat from gamma radiation)

distributed in the bulk of the core baffle.

Degradation of the crystal lattice ,

misplaced atoms, micro-void, voids,

change of fracture mechanical properties

Core Baffle – simulation of the degradation by swelling

37

Development degradation mechanism "Radiation swelling"

predicted based on calculations for 60 years of operation

(Gidropress, NRI Rez, a. s.).

Critical locations of core baffle The center of the segment (1 point) - a move towards

center of core (will 4 mm).

The outer surface on the threaded rod (178 point)

- shift towards the reactor shaft (will 2.5 - 10 mm).

Assessment of deformation for ETE Report DITI 2301/332 „Hodnocení VČR ETE

s hodnocením creepu a swellingu“, 2013

Distribution of the radial displacement of the

campaign. No. 60 (Block No. 1)

178 1

Node no. 1 - near of two small canals.

Node no. 100 - in the middle of the area near the grand canal.

38

The positions of selected nodes on the "inner surface" core

baffle of the 1st block Temelin NPP.

Measurement of CB – selection of critical points

1 100

178

Gg

Measurement of Core Baffle

39

Benefits of the core baffle assessment using NDT - dimensional inspection

• Provides the information about the actual state of component and demonstrate the functionality of the device for the next operation.

• Repeated measurement gives information about dimension changes, and help predict the effect of DM „radiation swelling“.

• Verify calculations of DM „radiation swelling“ development.

• Make possible to plan corrective measures well in advance of their actual implementation,

• Verification of corrective actions.

Displacement of node No. 100 is 7,15 mm after

60 years, 0,82 mm after 20 years.

The measured change of diameter should by

about 1,6 mm. Displacement is away from the

axis of the reactor.

The predicted shifts of selected nodes

40

0.000

0.002

0.004

0.006

0.008

0.010

0.012

0.014

0 5 10 15 20 25 30 35 40 45 50 55 60 65

rad

iáln

í po

suv

[m]

č. kampaně

Časový vývoj radiálních posuvů uzlů 178 a 100 za studena (1. blok). Posuv je směrem od osy reaktoru.

uzel 178 uzel 100

0

0.0005

0.001

0.0015

0.002

0.0025

0.003

0.0035

0.004

0.0045

0 5 10 15 20 25 30 35 40 45 50 55 60 65 70

abso

lutn

í hod

nota

rad

iáln

ího

posu

vu[m

]

č. kampaně

Časový vývoj radiálního posuvu (bez teplotních dilatací) uzlu s max. posunutím na vnitřním povrchu PAZ (2. blok)

Uzel 1, posuv směrem k ose reaktoru

Displacement of Node. 1 is about 3,6 mm after 60

years, about 0,6 mm after 20 years.

The measured change of diameter 1,2 mm in the

direction to the axis of the reactor.


The measurement principle

Measurement of dimensions at specific

locations (areas near threaded rod and

the middle segment).

Rotation 2x 60 ° provides a measurement

of the entire periphery of the ring.

Measuring dimension core baffle at

selected height levels - each ring will be

measured at two levels (proposal about

200 mm from the top and 200 mm from

the bottom edge of each ring).

This planned measurement corresponds

to the control measurements carried out

during assembly of the CB.

41

Measurement of Core Baffle – Design Model

42

Parts of the measuring device:

43

Design Model - Requirements

Requirements for the proposed measuring sensors

Frequency measurement every 3 years

Life of the equipment at least 20 years.

Measuring about 10 m below the water surface with the admixture of boric acid (H3BO3) in

a concentration of 15 g / l.

Equipment must withstand ionizing radiation.

Incremental optical sensors must allow pneumatic ejection test probes (to ensure an

adequate pressure during measurement) and subsequent retraction of the probe spikes.

In case of accidental failure of the system must ensure the automatic retraction of probe

spikes using a spring or other means to avoid jamming of the measuring equipment in core

baffle.

Dimensions exploded measuring devices for transport and

storage (L x W x H): 3300 x 600 x 1100 mm

44

Measurement of Core Baffle – Polar Crane

Swelling degradation – present status in the

Czech Republic

The swelling was accepted as DM influencing lifetime of

internals by operator

Design of the measurement device was agreed with NPP

Design phase is ongoing

The necessary hardware material is being purchased

The development of state of art in the field is continuously monitored

Methodology for swelling evaluation is developed with aim to introduce it into Czech Technical Normative Documentation


The project schedule

46

2016 - Engineering design of measuring equipment structural design measuring instrument

development of measurement technology and data processing

2017 - Design and manufacture of various parts of the equipment,

qualification of measuring equipment

2018 - Manufacturing of the prototype, production test specimen for

qualification, initiation tests to verify the measurement technology

2019 - Qualification and calibration of measuring equipment, completion of

documentation

2020 - Submission of final documentation of equipment and documentation

for measurement in NPP

2021 (2022 - 2023) - the realization of measurement (outside of this project)

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