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State and Development of the RIAR
Techniques for In-Pile Investigation
of Mechanical Properties of Materials
and Products for Nuclear Engineering
A.Ya. Rogozyanov, R.R. Melder, А. А. Nuzhdov,
S.V. Seredkin and V.A. Starkov
FSUE “SSC RIAR”, Dimitrovgrad, Russia
The current state of the methods
1. Investigation
techniques of
creep,
deformability and
long-term
strength using the
“Neutron”
facilities
Features of the “Neutron” Facilities
SPECIMEN TO BE TESTED
FLAT, TUBULAR, CYLINDER
TEST MODES REGARDING LOADING,
STRAIN RATE AND TEMPERATURE
STATIC, STEPWISE
TENSILE FORCE, F (F,%)
NEUTRON FLUX DENSITY
(Е0,1 MEV)
Up to 5000 Н (1%)
RADIATION DAMAGE RATE
Up to 210-4 dpa/hr
(1–8)1013 cm-2s-1
ELONGFATION MEASUREMENT
RAHGE, L (L)
TESTING TEMPERATURE, Т (Т)
200 – 850 ОС (1%)
TENSION RATE RANGE, VP (Vp)
1 – 10-5 % / ч (5%)
DATA RECORDING
CONTINUOUS
IRRADIATION RIG
MULTI-USE MANUAL AND
REMOTE ASSEMBLING
TEST DURATION
TEST ENVIRONMENT
Up to 2 лет
HELIUM
6 mm (3 – 5 μm)
Current State of the Techniques
ОУ
2. Investigation
techniques based
on relaxation
tests
2.1. The UVIRIM
Facility
СИ
КН
СПН
ДТП
САРТ
СУ
Features of the UVIRIM Facility
SPECIMEN TO BE TESTED
ANNULAR, FLAT
MAXIMUM STRESS IN THE
SPECIMEN
Up to 0,2
NEUTRON FLUX DENSITY
(Е  0,1 MEV)
Up to1014 сm-2s-1
RADIATION DAMAGE RATE
Up to210-4 dpa/hr
TESTING TEMPERATURE (Т)
300 – 700 оС (1%)
TEST DURATION
Up to1,5 years
TEST ENVIRONMENT
HELIUM
Current State of the Techniques
2. Investigation
techniques based
on relaxation
tests
2.1. The URIP
facility
Aluminium
block
Springs
Central
core plane
(CCP)
Features of the URIP Facility
SPECIMENS TO BE TESTED
SPRINGS OF VVER FUEL
ASSEMBLIES
NUMBER OF SPRINGS TO BE TESTED
24
NEUTRON FLUX DENSITY
(Е  0,1 MeV)
1012-1013 cm-2s-1
DAMAGE RATE
up to310-5 dpa/hr
APPLIED LOADING, F (F)
up to4 kN (1 %)
TESTING TEMPERATURE,T (T)
О
С
300 – 350 оС (1 %)
LENGTH MEASUREMENT ERROR, μm
(10-20)
TEST DURATION
Up to 2 years
TEST ENVIRONMENT
HELIUM
Current State of Techniques
3. Investigation
methods of
creep,
deformability and
long-term
strength under
pressure
Specimen under
test
95
 9,15  0,7
101
Features of the UITO Facilities
SPECIMENS TO BE TESTED
TIBULAR, WELDED AND
CONNECTED TO THE FACILITY
NUMBER OF SPECIMENS
Up to 30
NEUTRON FLUX DENSITY (Е  0,1
MeV)
1012-21015 сm-2s-1
DAMAGE RATE
до 310-3 dpa/hr
STRESS ()
-200 – 400 MPa
(2 – 4 %)
SIZE MEASUREMENT ERROR
3 - 7 μm
TESTING TEMPERATURE, Т (Т)
250 – 600 оС (1 %)
TEST DURATION
Up to 5 years
TEST ENVIRONMENT
WATER, LIQUID SODIUM,
HELIUM
Сurrent State of Techniques
9
4. Investigation
techniques of
dispersed fuel
creep
The URIPT facility
8
7
6
5
4
3
2
1
Features of the URIPT Facility
SPECIMEN TO BE TESTED
CYLINDER
NEUTRON FLUX DENSITY
Up to 1014 сm-2s-1
LONGITUDAL COMPRESSION FORCE, F (F) Up to 3 kN (1,5 %)
TESTING TEMPERATURE, Т (Т)
TEST DURATION
300 – 600 оС
(1 %)
Up to 5000 hr
Further development trends
of in-pile techniques
• extension of temperature range for
structural materials tests up to 1200 оС
• development of methods for creep test
of high-burnup oxide fuel
• methodical support of the upgraded SM
reactor core cells
THE UPGRADED SM REACTOR
CORE
5
Д
-1
41
КО-1
АР-1
Channel and its No.
Shim rod
Automatic control
rod
Emergency protection control rod in
the beryllium insert
61
Core cell with a fuel
assembly (FA)
FA with
experimental cells
12 mm
FA with an
experimental cell
25 mm
Loop channel 68
mm
NEW POSSIBILITIES OF THE SM REACTOR
UPGRADING OF THE CORE TURNS THE SM REACTOR
INTO THE MOST ATTRACTIVE REACTOR FOR MATERIAL
SCIENCE INVESTIGATIONS, INCLUDING STUDIES ON
LONG-TERM MECHANICAL PROPERTIES:
- extension of the fast (up to 21015 сm-2s-1) and thermal (up to
31015 сm-2s-1) neutron density range and damage dose (up to 25
dpa/year), being superior to the same BOR-60 parameters
- wide temperature range (60-650оС) of high-flux irradiation in water,
boiling water, steam and inert gas environment
- instrumentation
of in-pile investigations on some material
properties under continuous monitoring and control over temperature,
load and strain
- target pressurization of specimens with gas, tensile and
compression loading, any combination of pressure and uniaxial
loading, relaxation bending tests
CAPSULE IRRADIATION RIGS (IR) LOCATED IN THE SM
CORE FOR TESTING AT 50 – 320ОС
Unsealed IR in FA,
cooled with primary
water
Boiling IR for testing
specimens under
longitudinal loading
Boiling IR for testing
tubes under pressure
To the stand of
high pressure
To the
stand
К стенду
высокого
of
high
давления
pressure
To the stand
Water
Вода
К стенду
высоof high
кого давления
pressure
Water
54
54
х33
338
Therm.
couple
Strain узел
Датчик
gauge
деформации
700
Specimens
Образец
Specimen
350
СПАЗ
CCP
 23,8
Separating
Разделительный
сильфон
sylphon
Датчик
Loading
нагрузки
gauge
CCP
100
Loading
Нагружающий
сильфон
sylphon
POSSIBILITIES OF THE SM AND RBT-6 REACTORS
regarding study on functional properties (strength,
ductility, irradiation-thermal creep, radiation growth) in
in-pile irradiation conditions and on their relationship
with material structure
Reactor
SM
Irradiation rig
Methods
Test conditions
Т  320оС, рwater  18,5 МPа,
φfn 21015сm-2s-1, K25 dpa/year
Investigation Water - distilled water.
on creep, long- Load type – pressure, bend;
2006
tension
or
term strength since
Loop or
and ductility compression, their combination
capsule type
mit pressure;
with boiling (in
load
mode
–
soft,
hard,
the reflector or
stationary and non-stationary
since 2006 in
Investigation Т 320оС, р
water  18,5 МPа,

the core)
on Delayed
15
-2 -1
φ
2
10
сm
s , K25 dpa/year
fn


Hydride
Water - distilled water with
Cracking
different stem content, VVER –
(DHC)
(р  100 МPа) type water-chemical conditions
after VP-3 upgrading
POSSIBILITIES OF THE SM AND RBT-6 REACTORS
Reactor Irradiation rig
SM
RBT-6
Loop or
capsule type
with boiling (in
the reflector or
since 2006 in
the core)
Methods
Out-of-pile
tests
Investigation
on creep, longterm strength
and ductility
Capsule type
including
(in the core)
those of preirradiated
specimens)
Test conditions
Т  320оС, рwater  18,5 МPа,
φfn 21015сm-2s-1, K25 dpa/year
Water - distilled water with
different stem content, VVER –
type water-chemical conditions
after VP-3 upgrading
Т = 250 - 450оС,
φfn 61013сm-2s-1, K210-4dpa/hr
Dose Ktpre-irrad. – unlimited;
medium – helium,
load type – tension, pressure or
their combination;
load mode – soft, hard,
stationary and non-stationary
Specimen gas pressurization rig
РД
MIDA
booste
r
He
Pр ≤ 15
MPa
to PC
- high pressure valve,
Рн=20 MPа;
- low pressure valve,
Рн=2.5 MPа;
Air
Pр=0.6 MPa
The rig provides
pressure up
to 100 MPa
(DHC, creep,
long-term
strength),
possibility to
model DНС of
claddings at high
burnup values.
The rig needs to
be equipped.
COMPREHENSIVE TECHNIQUES FOR IN-PILE INVESTIGATION OF MECHANICAL PROPERTIES
IMPACT FACTORS:
NEUTRON FLUX AND FLUENCE (DAMAGE DOSE AND RATE), TEMPERATURE,
TYPE,
TEST CONDITIONS,
ENVIRONMENT
UITO
PRESSURIZATION
URIPT
(SM, RBT-6,
BOR-60,
SM core)
RELAXATION
ABILITY
NEUTRON
COMPRESSION
TENSION
COMPRESSION
PRESSURE
UVIRIM
(SM)
(RBT-6, SM)
( RBT-6, SM)
BENDING
SCC, DHC
under
development
URIP
TWISTING
(RBT-6)
ARS
LOAD
CYCLIC TESTS
STATIC AND QUASI-STATIC TESTS
CREEP,
DUCTILITY,
LONG-TERM STRENGTH,
DEFORMABILITY
LOAD,
UITO(I)
PRESSURIZATION
(SM, RBT-6)
LOW / HIGH-CYCLE
FATIGUE
Under development
NEUTRON
TENSION
(SM, RBT-6)
UITO
PRESSURE
(SM,
RBT-6)
CHARACTERISITCS OF IN-PILE INVESTIGATION
METHODS OF MECHNICAL PROPERTIES
NUMBER OF SPECIMENS IN IR:
under longitudinal loading
1-2
under bending or twisting
1-25
under pressurization
or swelling composition
up to 20
NEUTRON FLUX DENSITY, cm-2s-1:
fast (Е0,1 МeВ)
1013-21015
thermal
1013-1015
DAMAGE DOSE ACCUMULATION RATE, dpa/year
0,2-25
MAXIMAL UNIAXIAL LOAD, kN
5 (1%)
MAXIMUM PRESSURE IN TUBULAR SPECIMENS, MPа
100 (1%)
PRESET DEFORMATION RATE, %/hr
1-10-5 (5%)
TESTING TEMPERATURE:
in water under pressure
55-320оС (1%)
in boiling water
200-320оС (1%)
in water with supercritical parameters 500-600оС (1%)
in helium
200-850оС (1%)
CONCLUSIONS ON THE METHODICAL
POSSIBILITIES OF THE RIAR REACTORS
THE REACTORS AND THEIR METHODS MAKE IT POSSIBLE:
• to determine radiation stability of basic materials under
modeled operation conditions of nuclear reactor materials
at high fuel burnup, specific loading types and thermalforce modes
• to provide reliable validation of
the main criteria (creep,
radiation growth, short-term properties, long-term
strength) for the choice of the most promising new reactor
materials
• to develop databases for creation adequate models of
operation parameters of selected alloys as applied to
calculation codes