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 о С; - Medium – water, water steam, liquid metal, gas, air; - Radiation intensity under stationary conditions:  fast and thermal neutron flux density – up to neutron/cm 2 s;  absorbed dose rate in structural materials – up to 10 5 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

RIAR RESEARCH REACTORS AND PROGRAMS

RIAR REASEARCH REACTORS AND PROGRAMS

INTEGRAL PARAMETERS OF NEUTRON FLUXES. COMPARISON OF REACTOR CAPABILITITES.

DAMAGE DOSE FOR PURE ELEMENTS BOR-60 AND SM REACTORS

Transmutations and Gas Generation in Zirconium

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. Shim rod Control rod Core rod in beryllium insert Core cell with FA FA with experimental cells  12 mm FA with experimental cells  25 mm Loop channel  68 mm

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

Repeated irradiation of refabricated and full-size fuel rods Tests of the WWER 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 5

Types and characteristics of tranducers for irradiation rigs and fuel rods ParameterDesign typeMeasurement rangeError Temperature of coolant and fuel rod cladding Chromel-alumel thermocouple, cable-type Up to 1100 о С0.75% Fuel temperatureChromel-alumel thermocouple, cable-typeUp to 1100 о С0.75% Fuel temperatureThermocouple WRe-5/20, casing Мо + ВеО Up to 2300 о С (up to 1750 о С*) ~ 1.5% Cladding elongation LDDT(0…5) mm± 30μm Diameter changeLDDT(0…200) μm± 2μm Change of gas pressure in fuel rod Bellows + LDDT(0…20) MPa~ % Neutron flux density (relative units) Neutron detector (ND) (Rh, V, Hf) … /m 2  s ~ 1% Volume steam content in coolant Cable-type20…100%10% 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.

Experi ment Composition, number and burnup of fuel rods in EFA Pressure in the primary circuit of a loop facility, MPa Temperature range, о С Drying duration, min Exposure at max. temperature, min Fuel rod state Unirra diated fuel rod Fuel rod with burnup, MWd/kgU TightFailed Experiments at increased pressure in the primary circuit of a loop facility (cladding compression) SL …950*72 + SL Up to SL-561/ … SL-5P61/496700… Experiment at decreased pressure in the primary circuit of a loop facility (cladding swelling) SL … 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 The main parameters of «SB LOCA» experiments *- short-term duration, non-instrumented corner fuel rod

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

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 DESIGNS MODERNIZATION OF OPERATING VVER REACTORS AND WWER-1500 DESIGN Basic 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 DESIGNS Russian 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 PROJECT Basic 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 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.