METHODICAL CAPABILITITES OF SM, BOR-

60, RBT, AND MIR REACTORS FOR TESTING OF FUEL RODS AND NUCLEAR ENGINEERING MATERIALS

V.Golovanov, V.Efimov, N.Kalinina, A.Klinov, E.Lebedeva, V.Makhin, R.Melder, A.Rogozyanov,

S.Seryodkin, V.Starkov, G.Shimansky, V.Tsykanov

FSUE “SSC RIAR”, Dimitrovgrad, Russian Federation

FSUE“SSC RIAR”

OPERATING CONDITIONS:- Temperature – up to 3000 оС;

- Medium – water, water steam, liquid metal, gas, air;

- Radiation intensity under stationary conditions: fast and thermal neutron flux density – up to 5 1015 neutron/cm2 s;

absorbed dose rate in structural materials – up to 105 Gy/s.

EFFECT OF DIFFERENT RADIATIONS AND

CHANGES IN MATERIALS PROPERTIES:

- “Immediate” effects disappearing after radiation termination and occurring only under intensive radiation;

- “Integral” radiation effects lasting after radiation.

Material Behavior in Intensive Reactor Radiation Fields

RIAR IS THE LARGEST MATERIAL TESTING CENTER having:

- High-flux reactor SM;

- Fuel rod and assemblies testing reactor MIR;

- Three research reactors RBT;

- Pilot fast neutron reactor BOR-60;

- Pilot vessel-type boiling reactor VK-50 with natural coolant circulation;

- Largest in Europe “hot” material testing laboratory.

RIAR EXPERIMENTAL CAPABILITITES

Reactor Experimental capabilities

Research program Prospectives,

application pro-posals

Notes

SM high-flux, water-cooled (Р = 5 MPa), vessel-type reactor, power -100 MW (Starts in 1961, last modernization 1991-1992)

- 20 channels in reflector;

- Core cells; - Central neu-

tron “trap”

- Testing of materi-als, fuel rods and absorbing ele-ments;

- GT-MHR, ITER programs, …. - Radionuclides accumulation

Core is being modernized for high-dose testing

Critical facil-ity is available

MIR Loop-type experimen-tal channel reactor, core in water pool, power – up to 100 MW (commissioned in 1966, modernized in 1976)

- Up to 11 loop channels with water and gas coolants;

- 5 loop facili-ties with wa-ter;

- One loop with gas coolant

Testing of fuel, fuel rods, FA fragments in support of new VVER designs

Analysis and jus-tification of ad-vanced designs, such as GT-MHR, RMWR, etc.

Critical facil-ity, shielded cells with equipment for primary ex-aminations

RIAR RESEARCH REACTORS AND PROGRAMS

Reactor Experimental capabilities

Research program Prospectives,

application pro-posals

Notes

BOR-60 Pilot fast neutron reac-tor, sodium coolant, power – 60 MW, com-missioned in 1969

- 20 core cells for testing,

- 3 dry horizontal and 9 vertical channels

Testing of materi-als and fuel, opti-mization of fuel cycle and coolant technologies

Analysis and jus-tification of ad-vanced designs

RBТ 3 pool-type water reac-tors consuming spent SM reactor FA, power 6 and 10 MW (commis-sioned in 1975, 1983 and 1984)

- 8-10 core chan-nels;

- 6-17 reflector ampoule chan-nels

Testing of struc-tural materials and fuel compositions

Testing to extend lifetime of the operating units, justification of evolutionary and advanced de-signs (GT-MHR)

VK-50 Boiling water vessel-type reactor, power – 200 MW, P = 5Mpa; natural coolant circula-tion

In-core testing Testing to justify boiling water reac-tor designs

Justification of boiling water re-actor designs

Database on previous in-vestigations is being cre-ated

RIAR REASEARCH REACTORS AND PROGRAMS

INTEGRAL PARAMETERS OF NEUTRON FLUXES.

COMPARISON OF REACTOR CAPABILITITES.

Neutron flux, 1015 sm-2s-1, energy E, MeV Irradiation position

E>0 0<E<5.10-7 5.10-7<E<0.1 0.1<E<1 E>1

BOR-60, cell D-23 2.31 0 0.39 1.37 0.55

SM, core, cell 52 4.24 0.26 1.65 1.20 1.14

SM, core, cell 44 2.96 0.22 1.18 0.81 0.74

SM, channel 4, gas 2.07 0.52 0.84 0.45 0.26

SM, channel 4, water

2.04 1.20 0.47 0.22 0.15

DAMAGE DOSE FOR PURE ELEMENTS BOR-60 AND SM REACTORS

Damage dose, dpa, 1 effective power year

Ele-ment BOR-60,

Cell D-23 SM,

core, cell 44 SM,

core, cell 52 SM, channel

4, gas SM, channel 4,

water

Ed,eV

Be 32.9 26.0 38.9 12.9 6.5 31 C 42.2 34.7 52.0 16.3 8.4 31 Al 65.3 62.0 93.4 26.3 14.2 27 Ti 31.4 36.5 55.2 14.8 8.6 40 V 38.5 41.6 62.7 17.3 9.8 40 Cr 33.6 37.0 55.9 14.7 8.5 40 Mn 35.7 38.9 58.6 16.4 10.1 40

Fe 30.3 33.0 49.8 13.1 7.4 40 Ni 33.4 34.9 52.7 14.5 8.4 40 Cu 61.2 64.2 96.9 26.2 14.9 20 Zr 36.1 35.9 54.1 14.8 8.0 40 Mo 24.0 24.2 36.5 10.3 5.5 60 W 8.4 9.0 13.6 3.5 2.0 90

Transmutations and Gas Generation in Zirconium

Irradiation po-sition

Target, bur-nup, appm

Transmutants accumulation, appm

Gas accumula-tion, appm

Period Zr Y Nb Mo Tc Ru H He

BOR-60, 1 year 240 1 31 210 <1 <1 4 1

Cell D-23 5 years 1400 6 33 1400 <1 <1 20 4

SM, core, 1 year 980 4 55 920 <1 <1 13 2

cell 52 5 years 5100 16 57 5000 20 5 63 10

SM, core, 1 year 710 3 40 660 <1 <1 8 1

cell 44 5 years 3700 10 41 3600 8 2 41 6

SM, channel 4, 1 year 520 1 35 480 <1 <1 3 <1

gas 5 years 2800 4 36 2800 4 1 12 2

SM, channel 4, 1 year 450 1 43 400 <1 <1 2 <1

water 5 years 2600 3 45 2500 2 1 8 1

IN-REACTOR TESTING PROCEDURES DEVELOPED AT: - testing of material mechanical properties under irradiation,

- determination of thermal and electric conductivity of materials, electrophysical properties of insulation and piezoelectric materials;

- investigations of oxidation in water steam (zirconium alloys), etc.

SPECIAL IMPORTANCE IS PAID TO FUEL ROD TESTING, in particular:

- lifetime ( including re-irradiation of standard spent fuel rods);

- simulating transient and special conditions of power maneuvering NPP;

- LOCA and RIA.

COMPLEX OF DATABASES DEVELOPED FOR REACTOR MATERIAL TESTING EXPERIMENTS:

We have 3 databases: 1.“Catalog of methods for reactor testing of materials

and nuclear engineering items” (database MERI);

2.“Russian research reactors. Factual information and experimental capabilities” (database IRR),

3.“Atlas of shielded cells" (database AZK).

SYSTEMATIZATION OF DEVELOPED METHODS AND THEIR APPLICATION

RIAR NEW EXPERIMENTAL HIGH-DOSE TESTING CAPABILITITES

High-dose testing materials are carried out in the BOR-60 reactor. At present the high-flux SM reactor core is being modernized for irradiation of structural materials by the damage dose of 25 dpa per year. Tests are performed in the core using the loop channels (up to two channels in the core) or ampoules.

New experimental capabilities of the SM reactor including new equipment and methods for instrumented testing are important for justification of advanced designs. Comparative high-dose irradiation tests of materials and evolutionary designs (i.e. new zirconium alloys) are of practical interest.

Рис. 3. Модернизированная активная зона реактора

Modernized Reactor Core

Channel No.Channel No.

Shim rodShim rod

Control rodControl rod

Core rod in Core rod in beryllium insertberyllium insert

Core cell with FACore cell with FA

FA with FA with experimental cellsexperimental cells 12 12 mmmm

FA with FA with experimental cellsexperimental cells 25 25 mmmm

Д -15

4 1К О -1

А Р -1

6 1

Loop channelLoop channel 68 68 mmmm

Tests in the water of different

pressures (supercritical)

• A complex of methods was developed and successfully tested in RIAR for capsule testing of materials in the pressurized and boiling water at temperature of 350 oС. Upgrading of the methods has started lately to expand their possibilities for tests in water of supercritical parameters. The developed technique allows carrying out experiments in the reflector channels closest to the reactor core.

Ø43

Ø8*1

Ø44*2

Ø3

1

2

3

4

Core center

5

6

7

Capsule for samples irradiation: 1-vessel; 2-block; 3-capsule; 4-sample; 5, 6- pipes; 7-thermocouple.

FSUE SSC RF RIAR

Repeated irradiation of refabricated and full-size fuel rods

Tests of theWWER high-burnup fuel rods in the

MIR reactor

Power ramping (RAMP) and stepwise

increase of power (FGR)

Testing under design-basis RIA conditions

Testing under fuel rod drying, overheating

and flooding conditions (LOCA)

Testing under power cycling conditions

Testing of defective fuel rods

Lay-out of the WWER experimental fuel rods in irradiation rigs FSUE SSC RF RIAR 5

62.2

12.75

12.75

60

~300

mm

50

0 m

m

500

mm

640

mm

64

0 m

m

460

mm

49

0 m

m

200

mm

200

mm

Cor

e h

eigh

t

500

mm

50

0 m

m

Core average

plane

WWER fuel rod dummy

(fuel column)

FSFR (fuel column)

RFR (fuel column)

Square 42

Types and characteristics of tranducers for irradiation rigs and fuel rods Parameter Design type Measurement range Error

Temperature of coolant and fuel rod cladding

Chromel-alumel thermocouple,cable-type

Up to 1100 оС 0.75%

Fuel temperature Chromel-alumel thermocouple, cable-type Up to 1100 оС 0.75%

Fuel temperature Thermocouple WRe-5/20,casing Мо + ВеО

Up to 2300 оС(up to 1750 оС*)

~ 1.5%

Cladding elongation

LDDT (0…5) mm ± 30μm

Diameter change LDDT (0…200) μm ± 2μm

Change of gas pressure in fuel rod

Bellows + LDDT (0…20) MPa ~ 1.5-5 %

Neutron flux density (relative units)

Neutron detector (ND)(Rh, V, Hf)

1015…1019 1/m2s ~ 1%

Volume steam content in coolant

Cable-type 20…100% 10%

FSUE SSC RF RIAR 6

* - experimental data for high-burnup fuel rods

The MIR reactor is a channel-type, pool-type and beryllium-

moderated reactor. It has several high-temperature loop facilities,

which provide necessary coolant parameters for WWER fuel testing.

Experiment

Composition, number and

burnup of fuel rods in EFA

Pressure in the

primary circuit of

a loop facility,

MPa

Temperature range, оС

Drying duration,

min

Exposure at max.

temperature, min

Fuel rod state

Unirradiated

fuel rod

Fuel rod with

burnup, MWd/kgU

Tight Failed

Experiments at increased pressure in the primary circuit of a loop facility (cladding compression)

SL-1 18 - 12 530…950* 72 72 +

SL-2 19 - 12 Up to 1200 100 3 +

SL-5 6 1/52 4.9 750…1250 40 2 +

SL-5P 6 1/49 6 700…930 40 40 +

Experiment at decreased pressure in the primary circuit of a loop facility (cladding swelling)

SL-3 19 - 4 650…730 25 25 +

The WWER-1000 fuel assembly fragments were tested in the SL-1, SL-2 and SL-3 experiments; the WWER-440 fuel assembly fragments were tested in the SL-5 and SL-5P experiments.

FSUE SSC RF RIAR 17

The main parameters of «SB LOCA» experiments

*- short-term duration, non-instrumented corner fuel rod

FSUE SSC RF RIAR 23

Impulse shape in the MIR reactor( - exposure time at maximal LP)

Schematic diagram of the irradiation rig designed for RIA test in the MIR reactor

1 – fuel rods, 2 - conductor pipes, 3 – shroud,4 –upper shield, 5 –lower shield, 6 – loop channel

vessel

Core average

plane

200

12.75,

triangular step

6

1

2

3

5

4

0

1

2

3

4

5

6

7

8

0 2 4 6 8 10Time, s

En

ergy

rel

ease

, rel

ativ

e u

nit

s

RESEARCHES IN SUPPORT OF ADVANCED (EVOLUTIONATY) AND INNOVATIVE

DESIGNS (NEXT GENERATION NUCLEAR

REACTORS)

Proposals on application of RIAR capabilities for justification of advanced

(3rd generation) and 4th generation designs of power reactors were put

forward. The proposals were reviewed and approved by INPRO Board of

Directors (may, 2005)

EVOLUTIONARY DESIGNSMODERNIZATION OF OPERATING VVER REACTORS

AND WWER-1500 DESIGNBasic Tasks Accounting High Burn-Up:

- determination of standard fuel rod capacity limits under stationary, transient (including power maneuvering) and designed accident conditions at high burn-up;

- determination of capacity limits of vibropacked fuel rods (including MOX fuel rods) under stationary, transient (including power maneuvering) and designed accident conditions at high burn-up; - lifetime testing of fuel rods cladded with new Zr alloys;

- investigation of reasons and mechanisms of fuel rods failures, as well as consequences of cladding leakage;

- determination of composition and activity of the radionuclides releasing from fuel rods into the primary circuit coolant under regular operating conditions (leaky fuel rods) and under accident conditions (designed accidents);

- development of recommendations on optimization of the technology for fuel production, fuel rod operation conditions, and spent fuel storage.

The object of investigation is standard fuel of high burn-up.

TESTING IN RESEARCH REACTORS

- changes in form, strength and corrosion resistance of zirconium alloy cladding tubes up to the damage dose of 40 dpa with simulation of typical load types and conditions;

- high fuel burn-up with maximum possible load of fuel rods for regular operating conditions (experiments “BURNUP” with post-irradiation examinations of the fuel rods of high burn-up for special experiments);

- simulation of transient operating conditions;

- LOCA RIA simulation (leak-tight and leaky fuel rods);

- experiments with leaky fuel rods (artificial cladding defects);

- experiments with fuel rod specimens of different burn-up and different RIM layer thickness to determine integral thermophysical characteristics.

Migration of radionuclides in the circuit and influence of coolant parameters on radionuclide yield and migration from leaky fuel rods are studied along with investigations of fuel rod state.

INNOVATIVE DESIGNSRussian experts design fast neutron reactors consuming

nitride fuel and having a fuel cycle in equilibrium state under INPRO Program.

BREST reactor with lead coolant is being designed. Foreign BN-type reactors with lead and lead-bismuth coolant are analyzed.

Joint U.S./Russian Gas-Cooled Reactor Project is being implemented. Gas-cooled reactors are designed in some other countries, too.

Designing boiling water reactors with “decreased neutron moderation” (RMWR, Japan) and reactors with supercritical coolant parameters supports development of water-cooled power reactors.

Analysis of these tasks shows that they can be resolved using RIAR

experimental capabilities.

GT-MHR PROJECTBasic tasks for fuel testing:

- Investigation of fuel radiation resistance and matrix graphite;

- radionuclide release from fuel compositions depending on testing conditions;

- radionuclide migration in circuit.

At present testing methods for solution of these tasks are being

developed.

• INVESTIGATIONS IN SUPPORT OF REACTOR WITH COOLANT SUPERCRITICAL PARAMETERS

BN-800 REACTOR PROJECT Commissioning of fast sodium-cooled reactor BN-800 is scheduled for 2012.

Tasks that can be resolved at the BOR-60 reactor facility:

• analysis and justification of nitride fuel performance under stationary and transient conditions;

• investigations of fuel rod behavior under accident conditions; • creation and reactor testing of ultrasonoscopy elements and reactor diagnostics;

• optimization of sodium coolant technology with respect of 40-year experience in sodium operation;

• optimization of closed nitride fuel cycle elements.

RIAR has experience in fabrication of various experimental devices for solution of these tasks (dismountable FA, loops-ampoules, sodium purification devices, in-reactor monitoring and diagnostics systems and elements).

BREST PROJECT

Designing of reactors of this type arises a lot more problems than that of BN-800.

Special loop tests in the BOR-60 reactor were carried out to investigate performance of BREST-OD-300 fuel rod mockups in lead environment.

Dismountable FAs for testing of similar fuel rods in sodium environment to study under-cladding processes were designed to increase representativeness of the tests.

RIAR has started fuel cycle optimization work using BREST fuel

rods.

Available RIAR capabilities allow putting forward an international project proposal “Testing of nuclear

facilities with 4th generation reactors”

- analysis of reactor designs and nuclear engineering development concept;

- selection of prospective designs for testing; .

- development of testing proposals, their consideration by wide discussions and implementation; .

- analysis of the obtained results and correction of researches, if necessary. .

The developed methodology for reactor testing, technology and 40-year experience in reactor experiments, engineering capabilities for fuel cycle optimization make good basis for preparation and implementation of the basic Project. .

The Project may be prepared and implemented by a group of different specialists. The project incorporates:

Conclusions

1. Water-cooled reactors SM, MIR, RBT (3 reactors), VK-50 and fast neutron sodium-cooled reactor BOR-60 having well-developed testing and post-irradiation examination capabilities provide potentialities for researches in support of designs of various power facilities. RIAR process and engineering capabilities and experience in fuel cycle investigations are of great significance. .

.

Conclusions

• 2. Over 40 years RIAR has developed special-purpose in-reactor procedures for investigation of material properties and NPP characteristics. The methods have been systematized, and experiment planning databases have been created. RIAR analyses prospectives for application and modernization of research reactors. At present the SM reactor core is being modernized for high-dose irradiation of materials by the damage dose of 25 dpa per year.

Conclusions

• 3. Proposals were elaborated on use of the reactors for experiments in support of evolutionary and innovative NPP designs. These experiments are important both for development of Russian power engineering in the XXI century, and international cooperation.