irradiation embrittlement, thermal annealing and surveillance of reactor pressure vessels vienna ...

230 Pages·2009·6.83 MB·English

International Atomic Energy Agency lWGJtRPC-79/3
i
IAEA INTERNATIONAL WORKING GROUP
ON RELIABILITY OF REACTOR PRESSURE COMPONENTS
IRRADIATION EMBRITTLEMENT,
THERMAL ANNEALING AND
SURVEILLANCE OF REACTOR
PRESSURE VESSELS
VIENNA, AUSTRIA
26 FEBRUARY 1 MARCH 1979
Reproduced by the IAEA in Austria
December '"T-J
IRRADIATION EMBR1TTLEMENT,
THERMAL ANNEALING AND
SURVEILLANCE OF REACTOR
PRESSURE VESSITLS
VIENNA, AUSTRIA
26 FEBRUARY - 1 MARCH 1979
Chairman of the Meeting L.E. Steele
Head
Thermostructural Materials Branch
Materials Science and Technology
Division
Naval Research Laboratory
Washington DC 20875
USA
Scientific Secretary
of the IAEA IWG-RRPC IK. Terentiev,IAEA
CONTENTS
SUMMARY (Including Conclusions and Recommendations of the Meeting) 1
OPENING SESSION
Regulatory aspects of radiation embrittlement of reactor vessel steels 5
Pry or N. Randall
SESSION 1. RADIATION EMBRITTLEMENT
Radiation embnttlement predictions for reactor pressure vessels g
K.E. StaMkopf, T. U. Marston
Effect of residual elements and nickel on the sensitivity to irradiation embrittlement
of SA508 CL.3 pressure vessel steel and weld 18
P. Petrequin. P. Soulat, B. Houssin
Radiation damage and annealing of Cr-Mo-V steel 28
M. Vacek, S. Havel, M. Brumovsky
Irradiation embrittlement of a variety of RPV steel plates and weldments 35
LM. Davies, J.H. Venables, T.J. Williams
Size effect in radiation damage 48
At. Brumovsky
SESSION II. SURVEILLANCE I
Utility experience with reactor vessel surveillance 60
S. Rothstein
Current technical aspects concerning PWR reactor vessel surveillance programme 81
C. Buchalet
A Belgian operating experience in pressure vessel surveillance programme 87
Ph. Van Asbroeck, P. Gubel, H. Tourme, J. Van der Velde
Results of the steel surveillance programme for the pressure vessel of the
Dutch 50 MWE Dodewaard BWR 97
LB. Dufour, E. Lijbrink, W.L Zijp
Present status of reactor pressure vessel steels surveillance in Japan 112
S Miyazom. T. Kodaira
Recent experiences and problems in conducting pressure vessel surveillance
examinations 113
J.S. Perrin
SESSION 111 SURVEILLANCE II
Possibility of using small size specimens for toughness evaluation of pressure vessel
steels in surveillance programmes 119
P. Petrequin, P. Soulat
The assets and limitations of the Charpy V-notch approach to pressure vessel
safety in the surveillance programmes - A proposal for modifications 130
5. Rao
The analysis of reactor vessel surveillance program data 138
E.B. Norris
Analysis and interpretation of nuclear surveillance results 144
\LN. MCELROY et ai.
Characterization of the neutron environment for commercial LWR pressure vessel
surveillance programme 173
S.L. Anderson
SESSION IV. ANNEALING
Reactor vessel annealing to restore toughness — An engineering assessment 180
T.R. Mager, T. U. Marston
(Abstract only)
Survey of postirradiation heat treatment as a means to mitigate radiation embrittlement
of reactor vessel steels 181
J.R. Hawthorne
irradiation effects on reactor pressure vessel steels. Influence of flux and irradiation
annealing behaviour 195
P. Petrequin, P. Soulat
Hardness annealing behaviour of short- and long-term irradiated reactor pressure vessel
steel specimens 200
D. Pachur
Programme of post irradiation annealing studies 206
US. Sivaramakrishnan
A simplified theoretical approach to explain a possible, dose rate-dependent
saturation effect of irradiation embrittlement 214
C Leitz
LIST OF PARTICIPANTS 223
SUMMARY condition after neutron irradiation so as to assure vessel integrity for
the reactor's economic life. This applies to both current and futuie
light water reactor (LWR) systems. With this basis, the participants,
Introduction
organizers, authors, and discussors freely provided the latest available
The specialists' meeting on "Irradiation embrittlement, thermal information as well as ideas for added research and development to meet
annealing and surveillance of reactor pressure vassal steels" was organized the theme goal.
within the programme of activities of the IAEA International Working Group
on Reliability of Reactor Preasure Components (IWO-RRPC), and held at Papers from 3elgium, Chechoslovakia, Prance, the Federal Republic of
IASA Headquarters in Vienna, Austria, on 26-28 February 1979. Germany, India, Japan, The Netherlands, Sweden, the United Kingdom, and
the United States cjvered various topics within the framework of studies on
The subject meeting was followed by the Research Coordination meeting irradiation embrittlement, surveillance, and thermal annealing to correct
on "Analysis of the behaviour of Advanced Reactor Pressure Vessel Steels radiation damage. It -las agreed that the collective result in the long
mi would be to provide the jasic data on which to base criteria for
under Neutron Irradiation", held at the same place on 1-2 March 1979.
assurance of LWR vessel integrity. Such information is considered of
value both co those companies (or nations) now owning operating reactors
The specialists' aeeting was attended by 55 participants from
and to those companies (or nations) contemplating the acquisition of a
15 countries.
nuclear power plant or plants.
The research coordination meeting was attended by 12 participants
and 10 observers from 12 countries. The technical papers and technical discussion emphasized new
developments in our understanding of irradiation embrittlement, the
Some 27 reports and oral contributions were presented and discussed influence of individual component elements of vessel steel3 on sensitivity
at the specialists' meeting. General discussion at the end of the to neutron radiation, the fracture response of steels to conventional
(usually Charpy V-notch) fracture tests, the potential value of new, lesc
meeting identified the main findings and results and resulted in preparation
conventional fracture mechanics (quantitative) tests au applied both before
of the) recommendations.
and after irradiation, the current national or company approaches to
Burveillance of fracture response of vessel steels and the potential for
Dr. L.S.Steele, U.S.A, served as the Chairman of the meeting. In
radiation damage correction by postirradiation heat treatment or annealing.
cooperation with the participants he worked out the recommendations and
The latter included discussions of the factors responsible for the
susmarised conclusions which wars approved at the final session of the
engineering feasibility of this process which would be applied only as
neeting.
"last resort".
The most significant observations and recommendation? are as follows:
Summary of the reports and discussions and recommendations
of the meeting
1. Flux or Irradiation Rate Effect. Several authors described observations
(both from surveillance and experimental studies, mainly the former) sugtreBting
The basic theme of this meeting was the exposition of knowledge which the possibility for a significantly lower maximum level of embrittlement
will permit the evaluation or projection of the reactor pressure vessel for low fluxes. Evidence from test reactor studies suggested that the
phenomenon was more pronounced in steels of generally low radiation mainstay of current ( «*i future) surveillance programs the bent possible
sensitivity while surveillance studies showed no discrimination between use be made of 3uch data. One approach is to use computer based curve
steels. One theoretical analysis procedure was shown which had been fitting techniques and improved instrumented, load-deflection measurements.
tested on steels irradiated in a test reactor for as long as ten years. Results of surveillance programs containing new miniature "fracture
mechanics" (FW) speciments having potential for quantitative analysis
Recommendati on under linear-elastic conditions were reported. Similarly, results of
irradiations conducted using a variety of small Wi specimens were
The preliminary evidence must be evaluated by further data and reported though no standard testing and analysis methods beyond the
theoretical analyses including specially designed low flux linear elastic analysis of ASTM B-399 were provided. Extensive discu3!>ions
irradiation experiments and further correlative evaluations on fracture tests led to several points of general consensus. These aroj
of surveillance capsules containing identical steels or those
of the sane type and similar in composition. The fundamental Recommendations
physical processes involved must be explored and better explained
a. Surveillance capsules should contain FV" specimens which
2. Composition and Residual Element Effects. Data provided reinforced provide valid stress intensity values to temperatures (us
the fact that copper content (especially in the range above 0.20 percent) high as possible. Preferred is a 25-mm thickness specimen.
was most effective in causing embrittlement (both transition and toughness This is an effort to describe radiation induced increases
level). A less dominant yet unmistakable effect was shown for phosphorus in some critical reference temperature as carefully as
and nickel (especially above I.03S). Other evidence showed possible possible without overwhelming problems of temperature and
beneficial effects of silicon on "shelf" toughness and a detrimental fluence gradients. Meanwhile, the application of precracked
effect of arsenic on both transition and "shelf". Charpy specimens and standardized computer baaed curve
fitting procedures are endorsed to enhance the accuracy
Recotnaendati on (and adequacy) of small impact test specimens.
Continued efforts to produce pressure vessel steels low in the b. It is crucial that shelf ("ductile") toughness be defined
cited critical elements and in residual elements generally to operating temperature levels on a reliable qualitative
(clean steels) must be made. This alone should result in the basis but arbitrary Charpy energy criteria must be applied
minimization of radiation embrittlement and reduce the need until such tests are proven. When proven these new fracture
for extensive surveillance in future reactors. Nevertheless, tests must be used to cover the "full" upper shelf toughness
because a number of older reactors contain these elements including the whole regime of operating reactor temperatures.
(especially copper in welds), it is essential that quantitative Meanwhile, it is recommended that more Charpy specimens be
and fundamental bases (physical phenomena) for such effects be used in surveillance and their testing cover the range to
developed. reactor operating temperatures. (This need is made more
urgent by revelations of possible thermal and strain aging
3. Pract re Tests. Most irradiation results are reported in terms of propensity in some currently used vessel steels.)
the largely qualitative Charpy V-notch tests. As this specimen is the
Large section tests (test bars and simulated vessels) are results and aiding in the fundamen*al understanding of
encouraged for both unirradiated and irradiated conditions. radiation damage of ferritic steel::.
However, it is realized and accepted that such tests will d. Piirther use of the damage function analysis approach to
only be confirmatory - that is, provide a standard basis assess the relativo damaging potential (induced proportior
for small specimen modelling (and projection) of reactor change) of neutrons by energy groups was believed to be
operating conditions. important.
e. The amplication of standards for developing neutron dose
The use of systematic, statistically based methods of in terms of the more universal criterion, displacements
analysis must augment the foregoing, especially in per atom (DPA), was endorsed with the reservation that part
reconstructing older reactor conditions or making projections conventions for reporting neutrons greater than some
from the existing bank of Charpy impact data. energy level as u.l or 1,0 MeV be continued in parallel.
4. Neutron Dosimetry and Spectrum Analysis •j. Surveillance, The Techniques Therefor.
Several authors described extensive dosimetry programs for now The delegates and participants unanimously approved in principle
surveillance programs but much of the potential problem of embrittlement the inclusion of surveillance programs of even greater scope than in the
lies with older reactors wherein dosimetry was much less. r\jrther, past. The accepted thesis was that the cost of surveillance programs is
it was noted that the nature of a reactor and its operating history may negligible compared to possible reactor "downtime." h\irther, the suggested
change significantly over a period of years. These factors have led to analytical efforts in understanding radiation damage, steel fracture
a consensus that, as with fracture, we must be able to analytically behaviour and the neutron environmentr cited above, will provide vital,
augment limited data from older surveillance programs. This situation confirming complements to individual reactor surveillance studies.
is made less severe however by the fact that well documented reference
steels were placed in early surveillance capsules and that these nay
Recommendations.
serve as a reference standard for comparing results or for projecting to
a. Surveillance programs are essential.
the future if necessary. A clear call for advances in international
b. These programs should include the material projected to
standards for neutron dosimetry and related surveillance techniques was
bo limiting in terms of potential I'ai failu-o after
made by several speakers.
neutron degradation and the fullest possible complement
Recommendations. of neutron dosimeters,
a. Improved dosimetry packages should be applied in future c. Orwontional Charpy fracture specimens must be augmented
surveillance programs. by small fracture mechanic:; specimens oven though no
b. Systematic, statistical analyses of past data may help standard for interpretation of data therefrom is yet
define conditions for older reactors even if dosimetry available (:;oe item (a above). The number if .-.pocimon::
was minimal. sh-mld be increased and the temperature of tests extended
c. A US program with international support for a mock-up to that of reactor operation.
experimental facility to test and compare results on both d. Archive materials matching those in reactor should be
dosimetry and related changes in critical vessel steel retained for possible future clarifying tests or analyses.
properties was endorsed as one means for validating surveillance
e. Great care in capsule design, specimen and materials conditions and thus the possibility for taking actions to minimize the
documentation, dosimetry, and temperature monitoring is already small chance for catastrophic failure. Similarly, exhaustive
essential to any good surveillance effort. evaluations of composition effects and of the physical phenomena involved
in irradiation embrittlement are now being carried out in several nations.
f. .Vhere possible complement long-terra surveillance by
These efforts are highlighted by the IAEA sponsored coordinated research
analytical efforts, by parallel irradiations in sister
program which specifically seeks to analyze the potential (radiation
reactors or test reactor facilities.
resistance) of improved 3teels from steel makers in hYanco, Germany and
g. Related analytical studies of the neutron environment
Japan. This program, to be completed in 19 '2, 3hould prove the high
should focus on the projection of results in terms of
radiation resistance of advanced composition steels regardless of national
exposures anticipated at the one quarter thickness
source and also provide invaluable correlations of small fracture mechanic
location of the vessel wall, thereby reducing the
tests.
major conservatism inherent to the use of the inside
The general views of delegates and participants was that the meeting
vessel wall or capsule position for projecting
served as a positive step toward meeting the theme goal of assuring the
embrittlement.
structural integrity of all light water reactor pressure vessels.
o. Thermal Annealing to Correct Embrittlement
The effectiveness of thermal annealing for correcting amU it llement
was described by several authors and was universally recommended to be
studied thoroughly but used only as a "technique of last resort".
Conclusions
These points represent the most fully discussed matters, those
considered most important to reactor vessel integrity. Generally' the
discussions were addressed to older pressurized water reactors wherein
steels known to be relatively sensitive to embrittlement (especially welds)
were used. (Newer vessels are generally constructed of less sensitive
steels.)
A new and potentially more positive development was cited. This was
the observation of a possible flux effect resulting in an apparent
saturation of embrittlement afttr exposure at low flux rates. Full
documentation of the extent and piiysical nature of this phenomenon may
greatly reduce the concerns for radiation embrittlement, as well as the
need for extensive surveillance and annealing studies. Meanwhile,
analytical approaches for reworking old data, both fracture and neutron
environmental, should advance our confidence in the knowledge of vessel
II. SWAMPS AND ALLIGATORS 5
Obviously in the big picture, the swamp to be drained is to establish con­
fidence in the structural integrity of the reactor vessel. Some of my friends
Opening Session in the nuclear industry mirht say that NRC regulations are the alligators in
that swamp, which distract them from the objective. One might ask, why is there
regulation of nuclear power but not of fossil-fuel power, nor of oil refineries,
REGULATORY ASPECTS OF RADIATION chemical plants or highway bridges. I believe the main reason is that people
are not willing to let the nuclear industry develop its own rules for struc­
EMBR1TTLEMENT OF REACTOR VESSEL STEELS
tural integrity by learning from its mistakes, as the other industries have
done. The mistakes may cost too much. We hope the NRC provides another line
of defense against such mistakes. In trying to fulfill this role, it is NRC
PRYORN. RANDALL*
policy to make full use of the National Standards effort. An obvious example
USNRC here is the ASTM* Recommended Practice on Surveillance, £ 185, which is incor­
porated by reference in Appendix H to 10 CFR Part 50.
I. INTRODUCTION
The first of the technical distractions, which I categorize as alligators
or as the tips of icebergs, is the criterion for measurement of the adjustment
One purpose of this conference, as I see it, is to re-examine the conven­
of reference temperature. Traditionally, this temperature shift was measured
tional wisdom about neutron radiation erobrittlement and the'methods used to
at the 30 ft. lb. level of energy as determined by Charpy V-notch tests. Then
counteract embrittlement in reactor vessels. Perhaps, there have been suffi­
the ASME** Boiler and Pressure Vessel Code began to make use of lateral expan­
cient advances in fracture mechanics, core physics, dosimetry, and physical
sion measurements on Charpy specimens. In 1972, the reference temperature,
•etallurgy to permit a forward step in the quantitative treatment of the sub­
RT was defined in the ASME Code in terms of the nil-ductility transition
ject. Certainly this would be consistent with the position of the U.S. Nuclear temNDpTerature from dropweight tests, with 35 mils lateral expansion and 50 ft lb
Regulatory Commission (the NRC) in general. "There nas been a continued evolu­
energy from Charpy tests as supplomental criteria. Appendix G of our regula­
tion towaru increased specificity." This statement appeared in the response
tions, 10 CFR Part 50, adopted the two Charpy criteria as the measure of shift
prepared by the staff to a request from the Commission to explain how the staff
due to radiation to be consistent with the Code and to be conservative, However,
decides to apply a new requirement and to whom, i.e., to back-fit or forward-
as everyone knows, the 50 ft lb criterion becomes inoperative when the upper
fit-only or whatever.
shelf is below 50 ft lb and it is too conservative when the upper shelf energy
approaches 50 ft lb. This it has done for some radiation-sensitive materials
Pressure for increased specificity, i.e., for fleshing out general design
such as high-copper welds. Having no alternative, we are revising the regula­
criteria, co^es from "technical surprises" in the form of operating experiences
tion to drop the 35 mil lateral expansion criterion and reduce the 50 ft lb to
or froni research information, and from attempts to improve our confidence in 30 ft lb, but we do not regard this as the final word.
the safet> o* plants, especially new plants. Regulators don't like technical
surprises: not pany people do, except research types. Our goal is to have
anticipated an*, evaluated all possible modes of failure with sufficient quanti- Efforts to develop a more sophisticated method for characterizing the
tativeness that the probability of failure can be estimated with some accuracy. Charpy shift have begun. The first step is to standardize the curve-fittinij
Failing t^iis, regulators demand large margins of safety to cover our ignorance. technique for Charpy data. ASTM Subcommittee E 10.02*** has a task group
Although, »s ?--e often accused of having an insatiable appetite for large mar­ working to develop a recommended practice. A prime example of the proposed
gins, we rea;i> prefer to have less ignorance. methods is that of Dr. William Oldfield, who uses the hyperbolic tangent func­
tion to describe the transition curve mathematically. The four constants of
the "tanh-fit" are then used to describe the adjustment of reference temper"
This brings me to the subtitle of this talfc--"swamps and alligators". We
ature completely, rather than in terms of the temperature shift only.
have a saying, used by someone accused of overlooking the "big picture", "when
you are up to your. . .in alligators, it is hard to remember that your objective
was to drain the swamp". I want to classify the activities discussed at this
conference in terms of fighting alligators versus draining swamps. To use a
different anal-ogy, a lot of -our activity focuses on the tip of the iceberg. I American Society for Testing and Materials, 1916 Race Ft., Philadelphia, Pa.
want to remind you of the underlying objectives and the unresolved problems. 19103.
For each problem, I will discuss briefly the research efforts that the NRC is **
sponsoring and the additional work that I feel is needed. American Society of Mechanical Engineers, 345 East 47th St., New York, NY.
10017.
***
Office of Standards Development, U.S. Nuclear Regulatory Commission, Subcommittee on Behavior and Use of Nuclear Metallic Mate-ials in Nuclear
Washington, O.C. 20555. Systems of ASTM Committee E 10 on Nuclear Technology and Applications.