7th International Conference on Nuclear Engineering
Tokyo, Japan, April 19-23, 1999
ICONE-7473
IAEA CO-ORDINATED RESEARCH PROGRAMME ON
"ASSURING STRUCTURAL INTEGRITY OF REACTOR PRESSURE
VESSELS"
Milan BRUMOVSKY∗
Nuclear Research Institute Rez plc
250 68 Rez, Czech Republic
phone: +420-2-2094-1110
fax: + 420-2-2094-0519
e-mail: bru@nri.cz
Boris GUEORGUIEV
IAEA, Department of Nuclear Energy
Wagramer Strasse 5
P.O. Box 100
A-144 Vienna, Austria
phone: +43-1-2600-27791
fax:: +43-1-2600-29598
e-mail: B. Gueorguiev@iaea.org
This co-ordinated research programme (CRP) is a logical continuation of previous
three phases of the programme “Irradiation Embrittlement of Reactor Pressure Vessel Steels“.
This new programme started in 1996 and 25 organisations from 18 countries take part in it.
While the previous phases were concentrated mostly on radiation embrittlement of
reactor pressure vessel (RPV) steels and effects of different parameters, this new programme is
concentrated, in principle, on application of fracture mechanics approach to evaluation of RPV
steel damage.
The main objective of the CRP is to develop and validate a procedure for fracture
toughness testing small specimens applicable for surveillance programmes which could be used
internationally. These fracture toughness data will support a more precise evaluation of the
condition of nuclear power RPVs in the process of assessment of their structural integrity. For
this purpose, a „master curve“ approach has been chosen and is now being validated for
unirradiated as well as irradiated conditions.
Principal investigation is carried out on the IAEA „reference steel“ JRQ (of ASTM A
533-B type), additionally, one steel of domestic production should be also included by each
participant.
Paper describes this programme as well as preliminary results obtained.
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INTRODUCTION
During last 25 years, three phases of a Co-operative Research Programme on Irradiation
Embrittlement of Reactor Pressure Vessel Steels were organised by the International Atomic
Energy Agency. This programme started with eight countries in 1971 and finally 16 countries took
part in Phase III of the Programme in 1983. Several main efforts were put into implementation of
the programme, but the principal task was concentrated on an international comparison of radiation
damage characterisation by different laboratories as well as on development of small scale fracture
mechanics procedures applicable to reactor pressure vessel surveillance programmes.
In 1997 a new programme - Assuring Structural Integrity of Reactor Pressure Vessels - was
launched, concentrating mostly on small scale specimen fracture mechanics testing.
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PHASE 1
In 1971, the IAEA Working Group on Engineering Aspects of Irradiation
Embrittlement of Reactor Pressure Vessel Steels elaborated and approved, as an initial step,
a standard programme (Phase 1) of the Co-ordinated Research Programme on "Irradiation
Embrittlement of Reactor Pressure Vessel Steels" which should have been performed on a
reference steel ASTM A 533 Grade B Class 1 (HSST 03 Plate) deriving from the Heavy
Steel Section Technology Programme and provided to the IAEA by the Union Carbide
Corporation, USA.
Eight IAEA Member States (nine institutions) participated in the Phase 1 programme:
Czechoslovakia, Denmark, Federal Republic of Germany, France, Japan, Sweden, United
Kingdom, USA. The main goals were :
(1)
To establish the basis for describing embrittlement, and for performing the
measurement of neutron spectrum, fluence and mechanical properties with the aim to
be sufficiently standardised to permit direct intercomparison between international
programmes without major adjustment of the data.
(2)
To compare the embrittlement sensitivity of national steels with that of the standard
steel- ASTM A 533-B Plate 03.
No major discrepancies were observed in the results in spite of the use of unique
irradiation assemblies in nine different reactors with individual evaluations of neutron fluence
and neutron spectra. No major differences in mechanical test procedures and data
interpretation were observed. It was concluded that the differences in the results were probably
due to neutron measurements.
Main results from the programme including all reports from participated organisations
were published in the IAEA Report "Co-ordinated Research Programme on Irradiation
Embrittlement of Pressure Vessel Steels", IAEA-176, Vienna (1975).
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PHASE 2
After a detailed review and discussion of the CRP Phase 1, it was agreed that there was
a need to continue the research extending Phase 1. The general goal was "to demonstrate
that knowledge has advanced to the point that steel manufacture and welding for
nuclear technology can routinely produce steels for reactor pressure vessels of high
radiation resistance". In December 1976, the new programme (Phase 2), entitled "Analysis
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of the Behaviour of Advanced Reactor Pressure Vessel Steels under Neutron
Irradiation", was formally initiated.
Many organisations and countries provided steels for Phase 2 and during 1977-1979
the participants received test materials from advanced steels and welds typical for current
practice in France, the Federal Republic of Germany and Japan.
Nine IAEA Member States (ten organisations) participated in the programme:
Czechoslovakia, Denmark, Federal Republic of Germany, German Democratic Republic,
France, India, Japan, United Kingdom, USA.
The main goal was to undertake a comparative study of the irradiation embrittlement
behaviour of improved (advanced) steels produced in various countries. It was intended to
demonstrate that careful manufacturing specification of pressure vessel steel could eliminate or
significantly reduce the problem of neutron irradiation embrittlement, and to show that
knowledge had advanced to the point where steel manufacture and welding technology could
routinely produce steel reactor pressure vessels of high radiation resistance.
Within that programme, which lasted from 1977 to 1983, the following main
conclusions were obtained :
(1)
Results from this programme showed that modern pressure vessel materials (plates,
forgings and welds) possess high resistance to neutron irradiation damage.
(2)
In general, and for those mechanical tests which showed a response to neutron
irradiation, the results demonstrated that reducing the copper content (together with
low phosphorus content) of steels leads to an improvement in their irradiation
resistance. The copper content of modern steels is low, usually less than 0.1 mass %
while phosphorus content is usually lower than 0.012 mass %. Charpy transition shifts
of modern steels were generally lower than of those steels represented by the HSST 03
Plate.
(3)
There was no systematic variation of Charpy upper shelf energy change (decrease) with
neutron fluence.
(4)
The results of fracture toughness tests showed that modern steels are more resistant to
neutron irradiation than older pressure vessel steels. A good correlation was observed
between the Charpy 41J transition temperature increase and the shift in transition
temperature, defined at the 100 MPa.m1/2 level, from dynamic and static fracture
toughness tests.
(5)
The US NRC Regulatory Guide 1.99 (Rev.1), April 1977, was based primarily on the
results of accelerated irradiation in materials testing reactors. More recent data for
different compositions (including high-nickel) and from longer term surveillance
capsule irradiations have led to a reconsideration and probable revision of the
Regulatory Guide. Results from that programme underlined the shortcomings of the
initial Regulatory Guide approach, in particular with respect to high nickel content and
the description of upper shelf Charpy fracture energy decrease.
Further progress in the application of the fracture mechanics approach to radiation
damage assessment was achieved in that programme. Improvement and unification of neutron
dosimetry methods had provided better data with a smaller scatter.
All results together with their analysis and raw data were summarised in the IAEA
Technical Report Series No. 265 "Analysis of the Behaviour of Advanced Pressure Vessel
Steels under Neutron Irradiation", Vienna, 1986.
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PHASE 3
In June 1983 CRP participants agreed that a third phase of the Co-ordinated Research
Programme on neutron radiation effects on advanced pressure vessel steels should be initiated.
Success achieved in those two earlier phases provided the background for consolidation
and advancement of phase 3 to meet the principal goal of optimizing reactor pressure
vessel surveillance programmes and related methods of analysis for international
application.
The main objective of this phase of the programme was to consolidate the now
increasing body of knowledge on embrittlement and the technique used to determine its
significance. It was intended to establish guidelines for surveillance testing which could then be
used internationally.
To achieve this, four principal goals were defined for this phase :
(1)
Optimization of means for measuring fracture resistance.
(2)
Establishment of correlative methods for measuring irradiation response by using
different mechanical tests.
(3)
Seeking to understand the mechanisms responsible for embrittlement.
(4)
Establishment of means for ameliorating embrittlement.
In line with the four principal goals the primary emphasis of all participants was toward
the ultimate goal implicit in the agreed title - "Optimizing Reactor Pressure Vessel
Surveillance Programmes and Their Analyses". To this end, it was agreed to participate
with a key focus upon advancing quantitative fracture mechanics methodology and assuring the
extrapolation of qualitative fracture methods which have predominated in reactor vessel
surveillance during recent years.
From 32 steels offered by Japanese steelmakers, 27 were chosen for the study. Primary
interest was centred upon the group of Japanese laboratory melts to assess composition effects
and on a "radiation sensitive" correlation monitor from Japan. A crucial recommendation
was to provide the latter which may serve as a reference for this programme and for future
surveillance programmes throughout the world. This involved the procurement of a 20-25 ton
heat of this steel produced by Japanese manufacturer, which was designated as JRQ.
The Phase 3 of the programme was realised between 1983 and 1994 by 16 participants
- Member States: Argentina, Austria, Belgium, Czechoslovakia, Federal Republic of Germany,
France, Finland, German Democratic Republic, Hungary, India, Japan, Russia, Spain,
Switzerland, United Kingdom, USA. Within that programme, 23 different materials (17
specially prepared materials for this phase by Japanese companies, 3 materials from phase 2
and 3 from national programmes) were irradiated and tested. Irradiations were carried out in
20 different experimental and power reactors.
The main conclusions from that Phase were as follows:
(i)
The general aim of Phase 3 of the programme was to consolidate the now increasing
body of knowledge on embrittlement and the technique used to determine its
significance. Results from that programme showed that there was a comparable
knowledge, experience, irradiation and testing facilities in the Member States which
created a well established world-wide comparable centres for evaluation of the
behaviour of reactor pressure vessel materials under neutron irradiation damage.
(ii)
In general, comparison of results from mechanical testing gave quite reproducible
data suitable for a reliable assessment of reactor pressure vessel life.
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(iii)
Creation of a database of all experimental results obtained within the programme has
been found as a very usable instrument in analysis of all the data. This database was
also an initiating point for a wider IAEA database on surveillance programmes results.
Testing of "old" and "advanced" types of materials showed to an effective way of
decreasing the material susceptibility to radiation damage by decreasing phosphorus
and copper contents. Study of the effect of phosphorus, copper and nickel contents
on specially prepared experimental heats supported existing models on radiation
damage in steels of this type.
Comparison of experimental transition temperature shifts and upper shelf energy
decreases with those predicted by US, French, German and Japanese Guides showed
to the fact that none of them is fully applicable to the materials studied - they are
partially conservative and partially strongly non-conservative (up to 70 C).
Specially manufactured material JRQ (with higher content of copper and phosphorus)
has been studied by all participants. Analysis of the results obtained led to an approval
of a suggestion to use this steel as a "reference steel" for future surveillance as well as
research irradiation programmes.
This steel has been found as a well homogenous with reproducible results if
recommended conditions are fulfilled.
Mean trend lines have been determined for transition temperature shifts from impact
testing as well as for yield strength increase tested at room temperature, or after
irradiation at 290, resp. 270 C.
Progress in neutron dosimetry resulted in better instrumentation and characterisation
of irradiation experiments even though common uncertainty in neutron fluence
determination is still probably not better than 30 %.
(iv)
(v)
(vi)
(vii)
4
CRP ON "STRUCTURAL INTEGRITY OF REACTOR PRESSURE
VESSELS"
Further progress in the application of the fracture mechanics approach to radiation
damage assessment was achieved in Phase 3 of the programme. Nevertheless, improvement
and unification of the method has been found as an efficient way for improving precision and
reliability of reactor pressure vessel life evaluation based on surveillance specimens
programmes. Improvement and unification of neutron dosimetry methods have provided better
data with a smaller scatter but further steps seems still to be necessary.
Thus, in September 1996, a new IAEA CRP on "Assuring Structural Integrity of
Reactor Pressure Vessels" was founded, and 24 organisations from 20 different countries
joined it: Argentina, Austria, Belgium, Brazil, Switzerland, Czech Republic, Spain, Finland, France,
Germany, Hungary, India, Japan, Republic of Korea, Netherlands, Romania, Russia, Ukraine,
United States of America.
Purpose of the programme is to facilitate the international exchange of information and to
provide practical guidance in the field of monitoring reactor pressure vessels (RPV) materials
behaviour and to develop and assess a uniform procedure of testing small size (Charpy type)
specimens applicable for surveillance programmes with the aim of obtaining fracture toughness data
necessary for the assessment of RPV structural integrity.
The main objective of this Programme is to develop and validate a procedure for fracture
toughness testing small specimens applicable to surveillance programmes. This Programme could
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then be used internationally. These fracture toughness data will support a more precise assessment
of the condition of nuclear power RPVs in order to help assure their structural integrity.
To achieve this main objective, the Programme is divided into two parts :
(1) mandatory for each participant :
slow-bend testing of pre-cracked Charpy size specimens made from the IAEA reference
material JRQ and on at least one other RPV steel (base metal or weld) in the unirradiated
condition and additional instrumented impact testing of Charpy V-notch specimens from the
same materials;
(2) optional for each participant :
slow-bend testing of pre-cracked Charpy size specimens and instrumented impact testing of
Charpy V-notch specimens in both unirradiated and irradiated conditions with the
comparison of irradiation induced shifts preferably on reference material JRQ and on other
material deriving from the national programmes;
Such a partition of the programme has been chosen that organisations from countries
operating NPPs could take part even though they have no possibilities of irradiation and testing of
irradiated specimens.
First two meetings of the programme took place and results from unirradiated tests had
been presented and preliminary analysed.
Results obtained on the IAEA reference steel JRQ have showed:
relatively good homogeneity of this material,
consistency of standard notch impact Charpy testing as well as static fracture toughness
results,
possibility of „master curve“ application using small scale ( mostly pre-cracked Charpy size)
specimens.
Preliminary results from these two types of testing are summarised in the following Table 1
where distribution of two transition temperatures in different testing blocks (in 1/4 of the plate
thickness) of one plate (1,000 x 1,000 mm) is shown :
T28J - transition temperature from Charpy V-notch impact testing for criterion 28J,
T0 - transition temperature from static fracture toughness curve using „master curve“ approach
and adjusted for 25 mm thick specimens and criterion 100 Mpa.m0.5.
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Table 1. Distribution of transition temperatures T28J and To within the test block of JRQ
material (1/4 T).
5JRQ11
-33
-69
5JRQ21
-35
-85
5JRQ31
5JRQ41
-45
-73
5JRQ51
5JRQ12
-31
-59
5JRQ22
-42
-59
5JRQ32
-34
-75
5JRQ42
5JRQ13
5JRQ23
-42
-66
5JRQ33
-35
-79
5JRQ43
-37
-67
5JRQ14
-30
-72
5JRQ24
-42
-76
5JRQ34
5JRQ44
-34
-72
5JRQ15
-23
-73
5JRQ25
5JRQ35
-23
-69
5JRQ45
-43
-76
5JRQ16
-26
-68
5JRQ26
-29
-63
5JRQ36
5JRQ46
5JRQ52
-44
-75
This preliminary summarisation of data leads to the following mean values:
T28J
=
- 34.9 oC ± 6.8 oC
T0
=
- 70.9 oC ± 6.5 oC
which means that most of 18 laboratories obtained results within ± 1 σ. This is a very good result
when taking into account that materials from different parts of the plate have been tested by
different laboratories.
CONCLUSIONS
The International Atomic Energy Agency within the framework of its International Working
Group on Life Management of Nuclear Power Plants (IWG LMNNP) initiated and organised
several co-ordinated research programmes dealing with improvement and international comparison
of radiation damage testing results as well as fracture mechanics application to small scale
specimens used in reactor pressure vessel surveillance programmes.
These programmes has been found as very useful and provided a lot of important results
out of which the most significant are as follows :
radiation damage studies are internationally comparable,
IAEA reference steel JRQ is well homogenous and thus can be used as an international
standard/reference material,
results from these programmes provided a basis for the IAEA International Database on
reactor pressure vessel surveillance specimens programmes (IAEA International Database
on RPV Materials).
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