Ya.I.ShtrombakhYa.I.ShtrombakhControl of permanent-set structures state at

definition of nuclear power plants lifetime

Second International Symposium on Nuclear Power Plant life Management

SHANGHAI - CHINA - 2007

Russian Research Center "Kurchatov Institute“

NPP

RBMK VVER

EOL limiting features→ gas gap closure→ material properties degradation→ graphite bricks cracking→ graphite columns bowing

EOL limiting features→ RPV radiation embrittlement

Additional problems:→ Thermal ageing→ Internals→ Piping→ Concrete

END OF LIFE LIMITING FACTORS FOR RUSSIAN NPP’s

Nuclear power plants lifetime extension

Main task –state control and materials properties change prediction of permanent-set structures RBMK – graphite stack; VVER – reactor vessel.

For accomplishment of this task required is→ representative system of permanent-set structures state

monitoring; → reliable physically justified models of main

characteristics change, which define life of relevant structural elements, under the action of exploitation factors.

№

NPP № of power unit

Date of Commissi- oning

Тgraphite. оС

1 1 21.12.1973 500-600 2 02.07.1975 3 07.12.1979

Leningradskaya

4 09.02.1981 2 1 19.12.1976 650-750 2 28.01.1979 3 17.10.1983

Kurskaya

4 04.12.1985 3 1 09.12.1982 650-750 2 04.05.1985

Smolenskaya

3 04.02.1990

Justification of work capability of nuclear reactors permanent-set structures

1. RBMK Reactors11 units with RBMK-type reactors produce half of all the electric energy,

produced on Russian NPPs.

Graphite stack

Graphite stack, consisting of graphite bricks, is the permanent-set element of these reactors.In the process of reactor operation there take place thermal-radiation shrinkage of graphite bricks and enlargement of fuel channel diameter by means of irradiation creep.

Fragment of the “fuel channel-graphite stack” system

Large ring

Small ring

Graphite brick

Fuel channel

CRITERIA• Gas gap closure – radial (hoop) shrinkage of graphite

bricks• Critical value of neutron irradiation – critical neutron

fluence (Fcr). –the criterion of the ultimate irradiation loading on graphite as a material.• Ultimate value of vertical shrinkage of graphite

columns (not more then 225 mm)• Ultimate bowing of graphite columns (not more then

50 mm).

LIFETIME CRITERIA OF GRAPHITE STACK

The first criterion is gas gap exhaustion that defines time of FC substitution beginning (large-scale or stage-by-stage). This time will be different for different reactors because of the difference in graphite stack exploitation temperature (500-600 ºС – Leningrad NPP or 600-650ºС– Kursk and Smolensk NPP).

GAZ GAP CLOSURE

Rate of gas gap exhaustion for different NPPs

110.5

111.0

111.5

112.0

112.5

113.0

113.5

114.0

114.5

0 5000 10000 15000 20000Energy production per a cell, MW.days

Diam

eter,

mm

Kursk NPP, Unit 1 (Тirr=600-700оС)

Leningrad NPP, Unit 1 (Тirr=500-600оС)

Ignalina NPP, Unit 1 (Тirr=600-700оС)

The determination of critical neutron fluence

Neutron fluence, 1021 cm-2

0

5

10

15

20

25

30

300 400 500 600 700 800 900 1000 1100 1200 1300

Т обл , оС

F кри

т, 1

021 с

м-2

ГР-280

Leningrad NPP

Kursk NPP

Thermal dependence of graphite GR-280 neutron critical fluence

Arising of longitudinal crack in a graphite brick

0

400

800

1200

1600

0 4 8 12 16 20Years for period of prolongation of reactor operation

Num

ber o

f cel

ls w

ith c

rack

ed b

ricks

Cumulative curve of number of cells that have bricks with longitudinal cracks assessment

Graphite bricks longitudinal cracking resulting from arising irradiation-thermal stresses, opening of longitudinal cracks and following bending of graphite columns. Calculations have shown that size of columns deflection shouldn’t exceed 50 mm for the extension time.

Graphite brick cracking and bowing of graphite column

RBMK reactors graphite stacks lifetime prolongation.

After the assumed 30-year reactor operation commission inspection is executed that includes:

• measurement of cells geometrical parameters;• sampling of trepans (cores) from 5 cells with different energy

production along the whole height of graphite cells with diameter 10 mm and length 40 mm, and research of their physical-mechanical properties;

• execution of calculation of crack resistance of graphite bricks aimed at definition of time-point of their massive cracking.

• . Calculations have shown that size of columns deflection

shouldn’t exceed 50 mm for the extension time

Subsequent to the results of research conclusion on the condition of graphite stack and its maximum lifetime is made.

CONCLUSION (RBMK)

After the graphite stacks inspection, core graphite properties definition and stress-strain state of graphite stack definition conclusion on the condition of graphite stack and its maximum lifetime was made. At the present moment integrated survey of Units 1 and 2 of Leningrad NPP and Units 1 and 2 of Kursk NPP has been conducted. Subsequent to the results of conducted investigations a conclusion on working capacity of these reactors stacks within 10-15 years has been made and permission of the Russian Rostekhnadzor for their exploitation continuation has been received.

Three types of VVER units are in operation

VVER RPV EOL basically depends on weld seam Radiation Embrittlement (RE)

Low P and Cu contents:High P and Cu contents:

Generation 2Generation 1

~0.04 % CuUp to 0.22 % CuUp to 0.027 % P

Up to 1.9 % Ni< 0.3 % Ni

EOL depends on the primary REAll units were annealed –

EOL depends on reirradiation

RE depends on NiRE depends on P and Cu contents

~0.009 % PUp to 0.048 % P

VVER-1000 VVER-440/213VVER-440/230

VVER-440/230,179reactor pressure vessel materials

state evaluation

Radiation Embrittlement of the First Generation of VVER RPV Steels (VVER-440/179 and VVER-440/230)

0.180.140.17

19992001

1991, 1995, 20071991, 1995, 2003

Templet cutting

0.03919891974Kola-2 0.03419891973Kola-10.02919911972NVNPP-4

0.1350.0331987, 19911971NVNPP-3Cu, %P, %

Weld № 4AnnealingStartUnit

The absence of surveillance programswas compensated by taking templets

Templates are to small to use full-size 10×10×55 mm specimensSubsize impact bend specimens should be used for transition temperature evaluation for template metal

Maximum dimensions7,5×60×95 mm for WM→ 5×5×27.5 mm specimens5,5×60×95 mm for BM→ 3×4×27 mm specimens

For acquisition of additional data for justification of effectiveness of repeated materials embrittlement burn and kinetics the TACIS 91/1.1 Project has been realizedProject goals:− Define parameters Tk0 / Tk for reactor vessel material embrittlement monitoring of NVNPP-3 and 4 Units and KNPP-1 and 2 Units− Confirm safety of templates cutting from reactor vessel internal

surface− Justify annealing method and its control

Main conclusions− Execution of circle experiments provided concurrence of testing

methods− Study of reactor vessel material of NVNPP -3 and 4: defined

parameters of irradiation embrittlement prognosis;− Study of МКР of NVNPP -2 block confirmed annealing effectiveness;− Selection of templates is safe and provides valuable material

Sampling RPV of Unit 3 of Novovoronezh NPP

For optimization of templates cuts quantity a special program of templates material over-irradiation in channels for surveillance specimens of VVER 440/213 vessels has been developed

Irradiation embrittlement prognosis

Study of templates specimens, cut out from reactor vessels of Novovoronezhskaya NPP Unit 3 and 4 and Kola Units 1and 2 after over-irradiation in surveillance channels of Rovno NPP Unit 1 and Kola NPP Units 3 and 4.Result: irradiation embrittlement prognosis on the basis of advance irradiation that enabled to justify power units exploitation between templates cutting and evaluate the ultimate lifetime of RPVs.

Radiation embrittlement of RPV core weld No.4 of the unit 1 of Kola NPP

0.034%P; 0.14%Cu

Radiation embrittlement of RPV core weld No.4 of the unit 2 of Kola NPP

0.039%P; 0.18%Cu

Templets irradiation in VVER-440/213 surveillance channels along with periodical RPV sampling has been accepted in Russia as RPV RE monitoring programs for VVER-440 units of the 1st generation

0.180.140.17

19992001

1991, 1995, 20071991, 1995, 2003

Templet cutting

0.0392004→200919891974Kola-2 0.0342003→200819891973Kola-10.0292002→200719911972NVNPP-4

0.1350.0332001→2006 →20111987, 19911971NVNPP-3Cu, %P, %

Weld № 4Designed EOL // 5 years PLEX licenseAnnealingStartUnit

The studies made within the last 10 years enables NPPs to extend the lifetime of annealed units for 15 years with licensing for each 5 years and RE monitoring using templets

Factors, stipulating necessity of development of new irradiation embrittlement models for reactor vessel material

VVER-440/230

� Achieved results of microstructural research enable to definitely judge about leading mechanisms of irradiation embrittlement.� Recent occurrence of great number of representative data

on repeating irradiation embrittlement, what, in particular, is connected with planned periodical templates cut out from the internal surface of operating VVER-440/230 reactors and over-irradiation of templates materials in channels for surveillance specimens of VVER-440/213 reactors

CONCLUSION (VVER-440/1-st generation)

1. Using subsize impact bend specimens for DBTT evaluation provides possibility to study material of templets taken from operating RPVs.

2. Templets taken from RPVs are used in Russia for monitoring of the actual condition of RPV steels for the first generation of VVER-440 Units.

3. Irradiation of templet’s material in surveillance channels of VVER-440/213 Units is used for monitoring of re-irradiation embrittlement of RPV steels and validation of NPP lifetime.

4. Study of templets proved that using the lateral shift model for evaluation of re-irradiation embrittlement for the most of materials provided unreasonably conservative results.

VVER-440/213reactor pressure vessel materials

state evaluation

Radiation Embrittlement of the VVER-440/213 RPV Steels

0.0270.0280.0180.010P, %

Weld № 4

0.040.180.040.03

Cu, %

MAIN PROBLEM:

Representativity of surveillance specimens programms- by temperature;- by flux of neutrons.

International programsCOBRA – Surveillance specimens temperature measurementTAREG-2000

⇒ Development of surveillance specimens database, including revaluation of fluences and research results

⇒ Development of additional tests complex, enabling to raise surveillance specimens program representativity

⇒ Development of new normative dependences on the basis of a representative database on reactor vessel material irradiation embrittlement VVER-440/213

Direct measurements of irradiation temperature have shown that for surveillance specimens overheating, compared to internal surface of reactor vessel doesn’t exceed 5 °С

Thermal couple tracing schemeChain location in reactor scheme

The leading factor for VVER-440/213 surveillance CVN impact bend specimens is up to 20

Surveillance chains are located opposite the core in special channels on core barrel in water gap before RPV wall

30×1015 m-2s-14×1015 m-2s-1

The flux effect can result in significantunderestimation of radiation embrittlement

Precise WWER-440 SS neutron dosimetry is needed for adequate RPV lifetime prediction

Precise axial distribution of spectral index along of surveillance specimens

chain

Precise three-dimensional Discrete Ordinate Model

The leading factor for the two top containers in surveillance chain was found to be less than three

Surveillance data from the top containers can be used for VVER-440 RPV steels RE monitoring

Application of reconstitution technique was necessary to provide representativity of impact bend test results

Application of subsize specimens for study of Rovno-2 RPV steels

The surveillance set No.6 was exposed to irradiation from 24.12.1981 to 30.06.2003 during 21 fuel cycles. Total calendar irradiation time was 7858 days (188600 h). Total effective irradiation time was 6373 days.

Comparison of the subsize specimens test results with the standard surveillance specimens test results for Rovno-2 BM

Comparison of the subsize specimens test results with the standard surveillance specimens test results for Rovno-2 WM

Evolution of microstructure of VVER-440 weld (0.012%P, 0.04%Cu)

→ At low doses RE depends on Cu (and Р)→ Cu effect on RE has total saturation before 5⋅1023 m-2→ At higher doses RE depends on dislocation loops (and Р)

Distribution of Cu (�) and Р (�) atoms in irradiated VVER-440 RPV weld

Copper always combines with phosphorus

Model of VVER-440 RPV welds RE∆TТF=(1230P+207Cu)+{1.91+210(P–0.02)}F0.63P0=0.020%; Cu0=0.04% (σ=19.6°С; R=0.95)

∆TF=(APP+ACuCu)+(A0+APP)Fn

⊕ Threshold sensitivity of RE to P and Cu contents:if P< P0 , then P= P0; if Cu< Cu0 , then Cu=Cu0

Mechanism of RE associated with P

Mechanism of RE associated with Cu

Mechanism of RE associated with dislocation loops

CONCLUSION (VVER-440/2-nd generation)

1. The most of VVER-440/213 surveillance specimens, including all CVN specimens, are irradiated with leading factor up to 20 that can cause decreasing surveillance program representativity.

2. Significant neutron flux decrease at the top of surveillance chains provides an opportunity to use material from the top capsules of irradiation surveillance sets for correct evaluation of DBTT using subsize specimens along with reconstitution technique.

3. The study of materials irradiated in the top containers of Rovno-2 surveillance set No.6 and in-depth microstructural evaluations have shown that a new more adeqate model of VVER-440 RPV steels radiation embrittlement can be developed.

VVER-1000/187, 320reactor pressure vessel materials

state evaluation

Limitations of surveillance specimens VVER-1000 programs

� Complicated topology of iron-water environment makes it impossible to use two-dimension calculations of neutron fields� Inhomogenuity of surveillance specimens irradiation:

⇒ Number of specimens, irradiated in comparable conditions: 6 – 8

� Question about representativity of surveillance specimens programs⇒ on irradiation temperature ⇒ on irradiation homogeneity ⇒ on quality of specimens for fracture toughness.

The capsules with surveillance specimens are located above the core baffle in a place with complicated topology of iron-water surrounding and high flux gradient: the neutron flux variation through one level is of 80 – 100%

The reconstitution technique enables to provide the representativeness of the VVER-1000 surveillance programmes

International programs directions

TACIS-94 and TACIS-96- validation of neutron fluence three-dimension calculations (change of spectral indexes up to two times)- creation of Russian VVER-1000 surveillance specimens databaseTACIS -95 и TACIS -96- irradiation temperature and measurement and neutron dosimetry.

It has been ascertained that irradiation temperature doesn’t exceed 300°С.

TACIS -96- reconstruction method validation- validation of methods of fracture toughness determination.

The reconstitution technique enables to provide the representativeness of the VVER-1000 surveillance programmes

1 – the fragment under reconstitution (insert); 2 – corner tip; 3 – joint made using method of pulsed arc welding.

Upgrading of neutron dosimetry in WWER-1000 surveillance program

� Surveillance assemblies are located in high axial and azimuth fast neutron flux gradient� Neutron dosimeter sets are located not in each capsules with SS� Nb dosimeters are destroyed under irradiation because of small thickness

Direct measurements of 54Mn activity in each specimen in positions of notch and Retro-dosimetry procedure based on Nb extraction from SS and containersSpecial technique of measurements of destroyed Nb dosimeters (based on dissolving of detectors)Upgraded calculated-experimental procedure

Main results of surveillance programs improvement

1. Surveillance specimens programs are representative on irradiation temperature

2. Use of three-dimension calculations enables to conduct correct evaluation of fluences on surveillance specimens using the γ - scanning of each specimen

3. Implementation of reconstruction method enables to increase number of test specimens and provide representivity of VVER-1000 surveillance specimens programs

4. Implementation of reconstruction method enables to produce new specimens for tests on crack resistance

The standard reference dependence specified in the Russian Guide as for VVER-1000 RPV BM as for WM are not conservative

The standard reference dependence does not take into account Ni contents

Radiation embrittlement of VVER-1000 RPV welds increases with Ni contents

Radiation embrittlement of VVER-1000 RPV welds apparently depends on Mn contents

∆TTF=AFF1/3

Radiation embrittlement of welds with high Ni contents is rather high.

WWER-1000 lifetime management. Dosimetry aspects

� Precise neutron dosimetry for SS� Precise neutron dosimetry for RPV� Individual management for each RPV on

the base of fluence prognosis and kinetics of radiation embrittlement of material

SETS 1L-3L SETS 4L-6L

Leading factorLeading factor КК1,01,0

SS WWERSS WWER--10001000/320/320

0.501.001.502.002.503.003.504.00

-180 -120 -60 0 60 120 180angle,deg

Lead

facto

r

Leading factor obtained by RPV fluence calculation and investigation of different surveillance sets L2 & L4 positions (preliminary estimation)

Upper level

Low level

For most of VVER-1000 RPVS surveillance sets 1-3 are already unloaded!Surveillance specimens from sets 4-6 could not be used for forecast of RPV material radiation embrittlement

IT IS NECESSARY TO DESIGN NEW SURVEILLANCE IT IS NECESSARY TO DESIGN NEW SURVEILLANCE PROGRAM FOR VVERPROGRAM FOR VVER--1000 1000 RPVsRPVs ON THE BASE OF ALREADY ON THE BASE OF ALREADY IRRADIATED SPECIMENS FROM 4IRRADIATED SPECIMENS FROM 4--6 SETS6 SETS

WWER – 1000. Line of works

Additional problem of VVER-1000 surveillance specimens –sets 4-6 – single-store and are irradiating without advance compared to reactor vessel

+++++++++++++++++1

Tested (withdrawn) sets

+

+++++++(+)+++

+

+2

+++

+

+

++3

8985828886898786858480858493878785

Start Ni, %Cu, %P, %

0.0050.0070.0070.0060.0080.0090.0090.0080.0090.0050.0140.0080.0100.0070.0070.0090.009

Weld № 4

0.060.050.040.020.030.080.060.030.040.030.040.020.040.040.050.050.03

1.721.741.701.881.641.601.701.551.121.101.211.591.761.611.571.591.88

SUNPP-3+SUNPP-2

SUNPP-1++Khmelnitsk-1

+RovnoNPP-3+ZaporozhieNPP-5+ZaporozhieNPP-4+ZaporozhieNPP-3+ZaporozhieNPP-2

ZaporozhieNPP-1NV NPP-5

+KalininNPP-2KalininNPP-1BalakovoNPP-4BalakovoNPP-3

+BalakovoNPP-2BalakovoNPP-1

654Unit

Double-store Single-store

TAREG-2000

⇒ Development of database on surveillance specimens,including revaluation of fluences and research results

⇒ Development of additional test technics, enabling to raise surveillance specimens program representativity

⇒ Development of new normative dependences on the basis of a representative database on VVER-1000 reactor vessel material irradiation embrittlement

⇒ Evaluation of VVER-1000 reactors vessels integrity, including reference units calculation

CONCLUSION (VVER-1000)

1. Improvement of surveillance programmspermitted to obtain reliable data on RPVsmaterial embrittlement up to the end of design lifetime.

2. New models of RE kinetics for VVER-1000 steels should be specified, including dependences on chemistry content.

3. Only 3 of 6 surveillance sets can be used for RE prediction after lifetime extension and rearrangement of last three sets for each unit has to be implemented.

General conclusionsJustification of exploitation safety of NPP permanent-set structures beyond design lifetime leads to necessity of

1) development and implementation of principally novel means of their state control:• cut out of small metal pieces (templates) from the internal surface of 1st generation

VVER reactors non-clad vessels;• production and research of small-scale specimens from surveillance specimens, located

in higher containers of surveillance specimens chains, located on edge level of active zone;

• reampouling of single-store 4-6 sets of VVER-1000 surveillance specimens, and creation of novel, modernized surveillance specimens programs;

• drilling-out of core from graphite stack bricks.

2) implementation of a complex system of specification of irradiation loading on nuclear power plants elements, in particular, installation of dosimetric monitors system in off-vessel space.

3) conduction of additional research for justification of models of structural elements materials exploitation properties change beyond designed lifetime and introduction of relevant additions to accreditation reports for these materials, confirming possibility of the use of materials beyond previously defined maximal fluences and materials exploitation time.