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Session
5:High Temperature In-core Instrumentation and Irradiation Facilities
5-1
IRRADIATION
AND POST-IRRADIATION CAPABILITIES AT SCK•CEN
A.
Verwimp, J. Dekeyser, M. Verwerft, Ch De Raedt, M. Decréton
SCK•CEN,
Mol, Belgium
The Belgian Nuclear Research
Centre (SCK•CEN) is an institute of public utility under the tutorial
of the Belgian Federal Minister in charge of Energy. More than 600 highly
qualified researchers and technicians are working in various nuclear oriented
projects. One of the main research areas is the investigation of reactor
materials, fuel as well as structural materials.
Since 1963, SCK•CEN
in Mol operates the materials testing reactor BR2. The reactor is cooled
by light water under a pressure of 12 bar. The core is composed of beryllium
blocks with central channels. A total number of 79 channels is available
for loading driver fuel and/or experimental devices or control rods. The
core configuration can be modified in function of the experimental loading.
Fast neutron fluxes > 2x1014 n/cm2.s (E >
1 MeV) and thermal neutron fluxes up to 1015 n/cm2.s
can be achieved. Gamma heating can reach intensities up to 15 W/g Al.
BR2 is being used for
testing fuel and materials for different types of reactors, existing and
under design, including fusion reactors. Dedicated rigs and loops have been
designed, built and irradiated. For instance: in the past coated fuel particles
in graphite fuel compacts for the THTR have been tested in swept capsules
(eg. the MOPS test series) up to a temperature of 1500°C. Presently
tests are under preparation to study the electrical behaviour of selected
ceramic materials, like alumina, under fast neutron flux, vacuum and high
temperatures. The objective is to detect the so-called RIED (Radiation Induced
Electrical Degradation) effect involving a sudden increase in conductivity
after a given neutron fluence threshold.
The Reactor Physics
Department has the necessary tools available for predicting in core behaviour
of experiments and for optimising the core configuration as a function
of the experimental loading.
After performing in-pile
testing, the experimental facilities are transferred to the adjacent hot
cells of BR2 for recuperation of the irradiated samples. Mechanical property
and microstructure analyses are done at SCK•CEN's Laboratory for
High and Medium Level Activity (LHMA). It has the necessary shielded infrastructure,
and appropriate expertise to perform investigations related to irradiated
nuclear fuel, reactor pressure vessel and primary core materials and various
other services to the nuclear industry. The laboratory is internationally
renowned for post irradiation examinations. Mechanical testing (instrumented
impact, tensile, fracture toughness, fatigue) and microstructural examinations
(optical microscopy, scanning electron microscopy and electron probe microanalysis)
on irradiated material are performed routinely. More specialised techniques
are used in the frame of the own research activities of this laboratory:
internal friction, positron annihilation lifetime spectroscopy, and transmission
electron microscopy. In the near future, also slow strain rate testing
in autoclave will become available. Sample preparation can be performed
either in the hot-cell facilities or in specialised workshops.
SCK•CEN is interested
in performing dedicated and representative irradiation experiments in
the BR2 reactor in the framework of high temperature nuclear engineering
and performing state-of-the-art post-irradiation analysis and research
on the experimental samples.
5-2
EXPERIMENTAL
CAPABILITIES OF IVV-2M REACTOR AND ITS RESEARCH COMPLEX TO DETERMINE WORKABILITY
OF COATED PARTICLES AND FUEL ELEMENTS OF HIGH TEMPERATURE GAS-COOLED REACTORS
K.
N. Koshcheev, L. I. Menkin, V. I. Tokarev
SF
NIKIET, Russia
A.
S.Chernikov
NPO
“Lutch”, Russia
O.
G. Karlov, V. I. Vasilyev
RRC
KI, Russia
The IVV-2M water-moderated
water-cooled research reactor operates at a power of 15 MW. The reactor
power varies depending on the requirements of experiments being carried
out. The reactor is used in the following areas of research:
- a testing of structural
materials and fuel elements of power reactors of various application;
- a study of materials
exposed to irradiation;
- the solid state
physics.
The reactor core is 500
mm high.
The maximum neutron
flux density:
- thermal neutrons
(undistrurbed) 5·1014 cm-2·s-1
- fast neutrons (E
> 0.1 MeV) 2·1014
cm-2·s-1
60 cavities in the reactor
core and reflector with the diameter from 24 to 120 mm accommodate vertical
experimental channels (ampoules) for testing fuel element samples, structural
elements, etc.
The IVV-2M reactor operates continuously during 500 hours at rated power,
with shut down for 48 hours.
The effective time of the reactor operation at the rated power of 15 MW
is 7500 hours a year.
The complex of the IVV-2M reactor irradiation devices was created to solve
two main problems for fuel HTGR:
- a performance of
comparison tests of coated particles (CP) of various design and
fabrication technology for selection of the most optimum solution (“CP”
and “ASU“ channels);
- a performance of
qualification lifetime tests of full-scale fuel elements at various
temperatures characteristic of HTGRs (“Vostok” channels).
The channels of “CP”
type allow to irradiate coated particles included into graphite disks, matrix
graphite cylinders, graphite capsules with coated particles loosely arranged
in cylindrical openings. Depending on objectives of experiments, such channels
can be used for:
a) the passive
(without measurements of fission gas release) irradiation of coated particles
in graphite capsules in the temperature range from 1000 to 2000°C,
with individual capsules removed at different burn-up levels (usually
5, 10 and 15 % Fima);
b) the irradiation of up to 20 samples (coated particles, compacts) in
one ampoule, but in different capsules at temperatures up to 1600°C,
with simultaneous fission gas sampling.
The fuel irradiation temperature
is varied by moving the capsules along the core height, or changing the
gas mixture (He-Ne). The maximum thermal neutron flux density is 7·1013
cm-2·s-1.
The channels of “Vostok”
type allow to irradiate fuel elements in capsules with individual sampling
of fission gas.
To monitor the test
parameters, all of the above mentioned devices are provided with thermocouples
attached to samples and capsule components, thermal-neutron detectors
and thermal-emission sensors based on Rh, their indications are continuously
registered and analyzed by the neutron- and thermal-physical mathematical
models with the measuring-computational facility.
Post-irradiation investigation
of CP and fuel elements are performed in hot-cells laboratory. Fuel HTGR
of each channel was investigated according to its individual program.
The defect of protective
claddings of CP is estimated by three principal methods:
- IMGA (Irradiation
Microsphere Gamma Analyzer);
- post-irradiation
annealing;
- leaching (dissolution)
in nitric acid.
A distribution of fission
products and uranium in CP cladding is determined by gamma-spectrometric
and track analysis of spherical layers of PyC and SiC claddings. A sequential
stripping of the layers is made with the following techniques:
1) physical:
a laser uniform – surface evaporation of spherical layers of cladding;
2) chemical: an etching of the material by selective solutions.
Investigations of other
kinds are made by conventional methodology.
For the HTGR study
10 channels of “CP” type, 4 channels of “ASU-8”
type and 4 channels of “Vostok” type were irradiated and investigated
during 10-11 years.
5-3
DEVELOPMENT
OF A NEW METHOD FOR HIGH TEMPERATURE IN-CORE CHARACTERIZATION OF SOLID
SURFACES
M.
Yamawaki1, K. Yamaguchi2, A. Suzuki1,
T. Yokota1, G. N. Luo1 and K. Hayashi3
1:
Department of Quantum Engineering and Systems Science, University of
Tokyo, Tokyo, Japan
2: Nuclear Engineering Research Laboratory, University of Tokyo, Ibaraki,
Japan
3: High Temperature Irradiation Laboratory, Japan Atomic Energy Research
Institute, Ibaraki, Japan
In the in-pile irradiation
tests of ceramic breeder materials for fusion reactor blanket such as Li2O,
Li4SiO4, Li2TiO3, Li2ZrO3
and LiAlO2, the effect of irradiation and sweep gas atmosphere
upon the tritium extraction kinetics is among the most important subjects
of such experimental studies. In situ measurement of the gas-surface
reaction with regard to the lattice defects formation and adsorption/desorption
equilibria can be performed by using a high temperature Kelvin probe, with
which the work function at high temperature and under controlled atmosphere
can be evaluated. In case of Li4SiO4 and Li2ZrO3,
dependence of the measured work function on oxygen pressure suggested the
formation of oxygen vacancies. While, in case of Li2O,
Li2TiO3 and LiAlO2, such vacancy formation
was not observed but the adsorption / desorption reactions were. On
the other hand, the work function measured on gold under proton beam irradiation
showed a steep drop in the work function during the initial irradiation,
while it gradually recovered after the end of irradiation. The second-time
irradiation gave rise to a smaller value of the work function of gold.
These results support a possibility of adopting the high temperature Kelvin
probe for the purpose of monitoring the solid surfaces under irradiation
in nuclear reactors and other fascilities as to the defects formation in
near-surface region of solid specimens. The recent developments of
this technique are to be presented.
5-4
IRRADIATION
CAPABILITIES AND IN-CORE INSTRUMENTATION AT HALDEN
RELEVANCE TOGETHER WITH POSSIBLE APPLICATIONS, AND LIMITATIONS FOR
HIGH-TEMPERATURE ENVIRONMENTS
C.
Vitanza, H. Thoresen
OECDHalden
Reactor Project
This paper overviews key
fuel and materials irradiation capabilities at the Halden reactor. Such
Irradiations are normally characterised by the use of in-core sensors which
provide unique insights on the test specimen performance during service.
The function and use of these sensors are to some extent described in the
paper, together with the possible applicability in the high temperature
range. The experience at the Halden in high temperature irradiations and
in gas environment is also outlined.
5-5
DEVELOPMENT
OF HEAT- AND RADIATION-RESISTANT OPTICAL FIBERS
T.
Shikama
Tohoku
University, Japan
M.
Ishihara, T. Kakuta, K. Hayashi
Japan
Atomic Energy Research Institute, Japan
S.
Ishino
Tokai
University, Japan
The Innovative basic
research on new materials development and so on is to be carried out using
the High Temperature Engineering Test Reactor (HTTR), a helium gas cooled
graphite moderated high-temperature reactor. Several researchers
have pointed out the availability of optical fiber as optical diagnostics,
dosimetry etc. Our interest applying the optical fiber is a measurement
of temperature etc. of the core and/or test samples in the HTTR, high
temperature neutron irradiation condition; the temperature is up to around
1000°C and the maximum fast-neutron flux around 2x1017
m-2s-1. However, there is no available optical
fiber under the high temperature irradiation condition. Therefore,
an experimental study had been carried out to develop the heat- and radiation-resistant
optical fibers. To study thermal effects several kinds of optical fibers,
which were selected from a radiation resistivity point of view at a low
temperature region (below 200°C), were heated in an electric furnace
maximum at about 800°C for 20 days, and optical characteristics such
as optical transmissivity, optical loss etc. were measured. Furthermore,
to study radiation effects several fibers were irradiated in Japan Materials
Testing Reactor, where the fast neutron flux was up to 1.5x1018m-2
and the temperature up to about 800°C.
The present paper describes
the experimental results on the development of heat- and radiation-resist
optical fibers. Moreover, the thermal and radiation effects on the
optical properties are discussed.
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