Transcript Methodical Possibilities of SM, BOR

FSUE“SSC RIAR”
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
Material Behavior in Intensive Reactor Radiation Fields
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.
RIAR EXPERIMENTAL CAPABILITITES
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 RESEARCH REACTORS
AND PROGRAMS
Reactor
Experimental
capabilities
Research program
- Testing of materiSM
als, fuel rods and
- 20 channels in
absorbing elehigh-flux, water-cooled
reflector;
(Р = 5 MPa), vesselments;
- Core cells;
- GT-MHR, ITER
type reactor, power - Central neu100 MW (Starts in 1961,
programs, ….
tron “trap”
- Radionuclides
last modernization
1991-1992)
accumulation
- Up to 11 loop
channels with
Loop-type experimen- water and gas
tal channel reactor, coolants;
core in water pool, - 5 loop facilipower – up to 100 MW ties with wa(commissioned
in ter;
1966, modernized in - One loop with
1976)
gas coolant
Prospectives,
application proposals
Notes
Core is being Critical facilmodernized for ity is available
high-dose testing
MIR
Analysis and jusTesting of fuel, fuel
tification of adrods, FA fragments
vanced designs,
in support of new
such as GT-MHR,
VVER designs
RMWR, etc.
Critical facility, shielded
cells with
equipment for
primary examinations
RIAR REASEARCH REACTORS AND PROGRAMS
Reactor
Experimental
capabilities
Research program
Prospectives,
application proposals
Notes
BOR-60
Pilot fast neutron reac- - 20 core cells for
tor, sodium coolant, testing,
power – 60 MW, com- - 3 dry horizontal
missioned in 1969
and 9 vertical
channels
Testing of materials and fuel, opti- Analysis and jusmization of fuel tification of adcycle and coolant vanced designs
technologies
RBТ
3 pool-type water reac- - 8-10 core chantors consuming spent nels;
Testing of strucSM reactor FA, power - 6-17 reflector tural materials and
6 and 10 MW (commis- ampoule chan- fuel compositions
sioned in 1975, 1983 nels
and 1984)
VK-50
Boiling water vesseltype reactor, power – In-core testing
200 MW, P = 5Mpa;
natural coolant circulation
Testing to extend
lifetime of the
operating units,
justification
of
evolutionary and
advanced
designs (GT-MHR)
Database on
Testing to justify Justification
of previous inboiling water reac- boiling water re- vestigations
tor designs
actor designs
is being created
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
Element
Be
C
Al
Ti
V
Cr
Mn
SM,
core, cell 44
SM,
core, cell 52
32.9
42.2
65.3
31.4
38.5
33.6
35.7
26.0
34.7
62.0
36.5
41.6
37.0
38.9
38.9
52.0
93.4
55.2
62.7
55.9
58.6
12.9
16.3
26.3
14.8
17.3
14.7
16.4
6.5
8.4
14.2
8.6
9.8
8.5
10.1
31
31
27
40
40
40
40
33.0
49.8
13.1
7.4
40
34.9
64.2
35.9
24.2
52.7
96.9
54.1
36.5
14.5
26.2
14.8
10.3
8.4
14.9
8.0
5.5
40
20
40
60
Fe 30.3
Ni
Cu
Zr
Mo
33.4
61.2
36.1
24.0
SM, channel SM, channel 4,
4, gas
water
Ed,eV
BOR-60,
Cell D-23
Transmutations and Gas Generation in
Zirconium
Irradiation position
Period
Target, burnup, appm
Transmutants accumulation,
appm
Gas accumulation, appm
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
5 years
2800
4
36
2800
4
1
12
2
SM, channel 4, 1 year
450
1
43
400
<1
<1
2
<1
2600
3
45
2500
2
1
8
1
gas
water
5 years
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.
SYSTEMATIZATION OF DEVELOPED METHODS AND
THEIR APPLICATION
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).
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.
Modernized Reactor Core
5
Д
-1
41
Channel No.
Shim rod
Рис. 3. Модернизированная
КО-1
активная зонаControl
реактора
rod
АР-1
Core rod in
beryllium insert
Core cell with FA
61
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.
6
1
Ø44*2
2
Ø43
Core
center
Ø8*1
3
4
Ø3
7
5
Capsule for samples irradiation: 1-vessel;
2-block; 3-capsule; 4-sample; 5, 6pipes; 7-thermocouple.
FSUE SSC RF RIAR
Repeated irradiation
of refabricated and
full-size fuel rods
Testing under designbasis RIA conditions
Tests of the
WWER highburnup fuel
rods in the
MIR reactor
Testing under fuel rod
drying, overheating
and flooding
conditions (LOCA)
Testing of defective
fuel rods
Power ramping
(RAMP) and stepwise
increase of power
(FGR)
Testing under power
cycling conditions
5
FSUE SSC RF RIAR
Lay-out of the WWER experimental fuel rods in irradiation rigs
62.2
WWER fuel rod
dummy
(fuel column)
RFR
(fuel column)
FSFR
(fuel column)
Core height
500 mm
500 mm
~300 mm
640 mm
640 mm
460 mm
490 mm
Square  42
200 mm 200 mm
12.75
500 mm
Core
average
plane
500 mm
60
12.75
6
FSUE SSC RF RIAR
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
Bellows + LDDT
(0…20) MPa
~ 1.5-5 %
Neutron detector (ND)
(Rh, V, Hf)
1015…1019 1/m2s
~ 1%
Cable-type
20…100%
10%
Change of gas
pressure in fuel rod
Neutron flux
density (relative
units)
Volume steam
content in coolant
* - experimental data for high-burnup fuel rods
The MIR reactor is a channel-type, pool-type and berylliummoderated reactor. It has several high-temperature loop
facilities,
which provide necessary coolant parameters for WWER fuel
testing.
17
FSUE SSC RF RIAR
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
Experi
ment
Composition,
number and
burnup of fuel rods
in EFA
Unirra
diated
fuel
rod
Fuel rod
with
burnup,
MWd/kgU
Pressure
in the
primary
circuit of
a loop
facility,
MPa
Temperature
range, оС
Drying
duration,
min
Exposure at
max.
temperature,
min
Fuel rod state
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
*- short-term duration, non-instrumented corner fuel rod
25
25
+
23
FSUE SSC RF RIAR
Impulse shape in the MIR reactor
200
1
Core
average
plane
4
2
3
5
12.75,
triangular step
Energy release, relative units
( - exposure time at maximal LP)
8
7
6

5
4
3
2
1
0
0
2
4
Time, s
6
Schematic diagram of the irradiation rig
designed for RIA test in the MIR reactor
6
1 – fuel rods, 2 - conductor pipes, 3 – shroud,
4 –upper shield, 5 –lower shield, 6 – loop channel vessel
8
10
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 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”
The Project may be prepared and implemented by a group of different
specialists. The project incorporates:
- 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.
.
Conclusions
1. Water-cooled reactors SM, MIR, RBT (3 reactors), VK50 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 inreactor 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.