Beyond RPV Design Life
A set of standard reactor pressure vessels (RPV) surveillance programs for monitoring the design life (up to 40 years of reactor operation) have been analyzed. In view of the improved test methods and embrittlement evaluation procedures, the necessity has been shown of introducing modifications...
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| Date: | 2004 |
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Інститут проблем міцності ім. Г.С. Писаренко НАН України
2004
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| Cite this: | Beyond RPV Design Life / A. Ballesteros, G. Garcia, L. Bogede, J. Bros // Проблемы прочности. — 2004. — № 1. — С. 15-22. — Бібліогр.: 9 назв. — англ. |
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Digital Library of Periodicals of National Academy of Sciences of Ukraine| _version_ | 1859650125491601408 |
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| author | Ballesteros, A. Garcia, G. Bogede, L. Bros, J. |
| author_facet | Ballesteros, A. Garcia, G. Bogede, L. Bros, J. |
| citation_txt | Beyond RPV Design Life / A. Ballesteros, G. Garcia, L. Bogede, J. Bros // Проблемы прочности. — 2004. — № 1. — С. 15-22. — Бібліогр.: 9 назв. — англ. |
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| description | A set of standard reactor pressure vessels (RPV)
surveillance programs for monitoring the design
life (up to 40 years of reactor operation) have been
analyzed. In view of the improved test methods
and embrittlement evaluation procedures, the necessity
has been shown of introducing modifications
in the present surveillance programs aiming
at a more precise RPV integrity evaluation facing
a possible service life of 60 years. Service life predictions
are performed for reactor pressure vessels,
based on the available surveillance data,
reconstitutioned Charpy specimens and Master
Curve testing.
Проанализированы стандартные программы мониторинга расчетного ресурса (до 40 лет
эксплуатации) корпусов атомных реакторов с использованием образцов-свидетелей. Ввиду
усовершенствования методов испытаний и оценки радиационного охрупчивания материалов
отмечается необходимость пересмотра действующих программ и проведения научного
исследования с целью уточнения ресурса прочности корпусных материалов (продления его
до 60 лет). Выполнено прогнозирование долговечности корпусных сталей на основании
результатов, полученных на образцах-свидетелях, образцах типа Шарпи, а также с использованием
метода "Master Curve”.
Проаналізовано стандартні програми моніторинга розрахункового ресурсу
(до 40 років експлуатації) корпусів атомних реакторів із використанням
зразків-свідків. Удосконалення методів випробувань та оцінки радіаційного
окрихчування матеріалів потребує перегляду діючих норм і проведення
наукового дослідження з метою уточнення ресурсу корпусних матеріалів
(подовження його до 60 років). Виконано прогнозування довговічності
корпусних сталей на основі результатів, що отримано на зразках-свідках та
на зразках типу Шарпі, а також із використанням методу “Master Curve”.
|
| first_indexed | 2025-12-07T13:32:50Z |
| format | Article |
| fulltext |
UDC 539.4
Beyond RPV Design Life
A. Ballesteros, G. G arcia, L. Bogede, and J. Bros
Tecnatom S.A., Madrid, Spain
УДК 539.4
О возможности превышения расчетного ресурса корпусов
атомных реакторов
А. Баллестерос, Г. Гарсиа, Л. Богеде, X. Брос
“Tecnatom S.A.”, Мадрид, Испания
Проанализированы стандартные программы мониторинга расчетного ресурса (до 40 лет
эксплуатации) корпусов атомных реакторов с использованием образцов-свидетелей. Ввиду
усовершенствования методов испытаний и оценки радиационного охрупчивания материалов
отмечается необходимость пересмотра действующих программ и проведения научного
исследования с целью уточнения ресурса прочности корпусных материалов (продления его
до 60 лет). Выполнено прогнозирование долговечности корпусных сталей на основании
результатов, полученных на образцах-свидетелях, образцах типа Шарпи, а также с исполь
зованием метода "Master Curve”.
Ключевые слова : корпусная реакторная сталь, программа наблюдения, рас
четный ресурс, прогноз ресурса (долговечности).
Introduction. The highest priority key Category 1 component identified in
all separate national categorization exercises has been the reactor pressure vessel
(RPV). This is because the RPV is considered irreplaceable or prohibitively
expensive to replace. This, in turn, means that if it degrades sufficiently, it could
be the operational life-limiting feature of the nuclear power plan (NPP). The RPV
houses the reactor core, and because of its function it has direct safety
significance.
The majority of the early Westinghouse designed plants had a “design basis
life,” as distinct from “physical life,” of 30-40 years. Specific “design basis life,”
such as 40 years, was not based on technical studies of material degradation in
general, but was based on fatigue usage factors, for the most part. Operation of
existing LWRs longer than originally intended has now become a relevant feature.
Radiation embrittlement is the most important degradation mechanism limiting
the RPV life. Neutron irradiation degrades the mechanical properties of RPV
steels, and the extent of the degradation is determined by a number of factors such
as neutron fluence, irradiation temperature, neutron flux, and the concentration of
deleterious elements in the steel.
As a result of technical and economic considerations, the operating life of an
NPP could be easily 50 or 60 years [1]. This might require upgrading of the RPV
surveillance programs and the use of modern techniques and approaches such as
Charpy reconstitution and Master curve testing to cover the extended operation
tim e.
© A. BALLESTEROS, G. GARCIA, L. BOGEDE, J. BROS, 2004
ISSN 0556-171X. Проблемы прочности, 2004, № 1 15
A. Ballesteros, G. Garcia, L. Bogede, J. Bros
RPV Surveillance Data. Changes in the material properties due to neutron
irradiation are monitored by means of surveillance programs. Every PWR and
BWR pressure vessel has an ongoing RPV material radiation surveillance
program. To date several hundreds surveillance capsules have been removed from
their host RPVs and tested. The results from these surveillance capsules have
been used to develop heatup and cooldown curves and to analyze all potential or
postulated accident or transient conditions.
The structural integrity evaluation of the RPVs of the Spanish reactors
follows the regulations, guidelines, codes, and standards developed in the US,
since the reactors were designed by Westinghouse and General Electric, and are
similar in design and operation to the American reactors. An exception is the
Trillo I reactor supplied by Siemens KWU (now Framatome ANP). Table 1 lists
the operating reactors in Spain. José Cabrera reactor will be decommissioned in
2006 after a useful life slightly lower than the design life. Its total gross electrical
production will be around 35.000 GWh by the year 2006. It can be observed in
Table 1 that, in general, the weld is the most limiting material of the beltline
region for the older reactors.
T a b l e 1
Spanish Reactors in Operation
Nuclear
power plant
Type
of reactor
Supplier Initiation
of operation
Beltline limiting
material for
brittle fracture
Surveillance
capsules
analyzed
in 04/2003
José Cabrera PWR Westinghouse 1968 Weld 4
Almaraz I PWR Westinghouse 1981 Base metal 4
Almaraz II PWR Westinghouse 1983 Base metal 3
Asco I PWR Westinghouse 1983 Base metal 3
Asco II PWR Westinghouse 1985 Base metal 3
Vandellos II PWR Westinghouse 1988 Weld 3
Trillo I PWR Siemens-KWU 1988 Weld 1
Santa Ma
de Garona
BWR General
Electric
1971 Weld 3
Cofrentes BWR General
Electric
1984 Weld 2
It is well known that certain residual elements, such as copper and
phosphorus, favor embrittlement. With a content of more than 1%, nickel also
causes embrittlement, although it has a beneficial effect since it produces a lower
initial value of the reference temperature RTn d t . The role played by other
residual elements such as tin, antimony, and arsenic is not clear. Copper produces
fine precipitates, which cause embrittlement by making the movement of
dislocations more difficult. Copper and nickel are believed to have a synergy
effect of hardening. Phosphorus embrittlement is a result of two mechanisms. On
the one hand, fine precipitates are formed in a manner analogous to the case of
copper, and on the other one, phosphorus precipitation is segregated at the grain
16 ISSN Ü556-171X. Проблемыг прочности, 2ÜÜ4, N 1
Beyond RPV Design Life
boundaries causing them to be weakened (non-hardening embrittlement) and
leading to the possibility of unstable crack growth. Table 2 shows the range of
concentration of Cu, Ni, and P measured in the RPV surveillance steels of the
Spanish LWR reactors.
T a b l e 2
Concentration of Deleterious Elements in Steel
Material Range of Cu (%) Range of Ni (%) Range of P (%)
Base metal 0.03-0.14 0.50-0.77 0.005-0.013
Weld 0.02-0.30 0.04-1.01 0.004-0.015
Because phosphorus can also contribute to hardening, it is not always clear if
its effect on the embrittlement is due to hardening or segregation, or both. A
predominant role for hardening would be consistent with the observed
characteristically low susceptibility of the Spanish PWR vessel steels to the
embrittlement phenomena caused by grain boundary segregation of impurities.
The ratio of the yield stress change to the increase in measured by
Charpy testing is frequently a good measure of whether non-hardening
embrittlement occurs. Figure 1 shows this ratio for the Spanish PWR vessel
steels. The scatter observed at low fluence values is not relevant since small
increases in the yield strength can produce high AT41j j Ay s ratios. The
conclusion from Figure 1 is that there are no non-hardening embrittlement
components in the Spanish PV steels under consideration.
AT41J / AYield vs. Fluence
10 20 30 40 50 60
Fluence (*1022n/m2)
Fig. 1. The absence of non-hardening embrittlement in the Spanish PWR PV steels.
Until now, a total of 26 surveillance capsules have been tested and analyzed
in Spain. The surveillance data allow one to verify the theoretical embrittlement
trend curve and to detect any anomaly in the irradiation conditions. Figure 2
shows a comparison of measured and predicted values using the Eason model [2].
The predictions are reasonably good at low fluence levels, but become more
dispersed at higher fluence levels.
ISSN 0556-171X. Проблемы прочности, 2004, № 1 17
'Чciaог<
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2.5
2
1.5
1
0.5
0
-0.5
-1
0
A. Ballesteros, G. Garcia, L. Bogede, J. Bros
0 50 100 150 200 250
M easured values of д T41j (°F)
Fig. 2. Comparison of measured and predicted shifts using Eason correlation.
Prediction to 40 Years. The steels used for reactor vessels become
progressively brittle throughout the service life of the component as a result of the
effect of neutron irradiation to which they are exposed. This progressive
degradation must be known in order to guarantee the structural integrity of the
vessel throughout its service life. There are two key parameters, which make it
possible to evaluate and quantify the vessel degradation. These parameters are the
USE (Upper Shelf Energy) and RTn d t (Reference Temperature). The initial
values of the RTn d t are obtained by means of Charpy and drop-weight tests,
while increases in the RTn d t and the value of USE are measured exclusively by
means of Charpy tests. The RTndt tends to increase throughout the service life of
the reactor, while the USE tends to decrease.
The American Code for a pressurized thermal shock (PTS) 10CFR50.61 [3]
establishes requirements on the ability of the reactor vessel in pressurized water
reactors to withstand events wherein the vessel is both rapidly overcooled
(thermally shocked) and pressurized (or repressurized). The Code requires
calculation of the projected values of the RTndt at the end of life (EOL) and
comparison with given limit values. The PTS screening criterion is 270°F for
plates, forging and axial weld materials, and 300°F for circumferential weld
materials.
On the other hand, Regulatory Guide 1.99 revision 2 [4] requires for RPV
beltline materials of new plants that the reference temperature RTndt at the 1/4T
position in the vessel wall at the end of life be less than 200°F. This could be also
considered a good recommendation for existing nuclear power plants.
Table 3 shows the projected values of the RTn d t at 32 EFPY (effective full
power years) for the most limiting beltline material of the Spanish RPVs at the
inside location of the reactor vessel wall. In the past, 32 EFPY were associated
with 40 calendar years of operation, representing an average availability factor of
80%. For PWRs operating in Spain, the projected RTn d t values are lower than
the PTS limit values established in 10CFR50.61. Moreover, all the RTndt values
for the PWRs and BWRs listed in Table 3 are below the recommended value of
200°F.
18 ISSN 0556-171X. Проблемы прочности, 2004, N 1
Beyond RPV Design Life
T a b l e 3
Values of ÆTNot and USE after 32 EFPY
RPV RTndt (°F) USE (J)
No. 1 146 71
No. 2 138 USE > 104
No. 3 156 USE > 68
No. 4 114 79
No. 5 107 95
No. 6 115 USE > 123
No. 7 106 163
No. 8 119 104
No. 9 90 114
As established in Appendix G of 10CFR50, initially reactor vessel beltline
materials must have Charpy upper shelf energy, USE, in the transverse direction
for the base material and along the weld for the weld material of no less than 102 J,
and must maintain the USE of no less than 68 J throughout the life of the vessel,
unless it is demonstrated that lower values of the USE can provide margins of
safety against fracture equivalent to those required by the ASME Code.
Regulatory Guide 1.161 [5] was developed by the US NRC to provide a
comprehensive guidance for evaluating RPVs when the Charpy upper shelf
energy falls below the 68 J limit of Appendix G to 10CFR50.
All the USE values at 32 EFPY listed in Table 3 are over the 68 J limit value.
In several cases it was difficult to get an accurate projected value of the upper
shelf energy at 32 EFPY, since the Charpy-V impact tests of unirradiated material
were performed at the temperature in the upper shelf region not high enough. For
these cases, the projected USE values were based on the surveillance data of
capsules, which accumulate higher neutron fluence than the projected one inside
the reactor vessel wall at 32 EFPY.
It is worth to mention two specific types of activities performed in Spain,
namely, analysis of the impact of power uprating on the RPV structural integrity
assessment and implementation of ASME Code Cases N-640 [6] and N-588 [7].
Several nuclear power reactors in Spain have undertaken a major initiative to
increase the economic value of their plants by increasing the license power at
which the plant is permitted to operate. The Spanish nuclear regulatory body,
CSN, has reviewed applications for power uprates in Almaraz, Asco, Vandellos,
and Cofrentes reactors in order to determine if adequate safety margins exist at
the increased power and to ensure that the regulatory limits are not exceeded. The
power uprating studies performed by the utilities include re-evaluation of the
projected neutron fluence at the end of life and, consequently, the projected
RTndt and USE values and the impact in the pressure-temperature limit curves.
At the Cofrentes nuclear power plant they made use of the ASME Code
Cases N-588 and N-640 to update the pressure-temperature limit curves. ASME
Code Case N-588 allows the use of an alternative procedure to calculate the
applied stress intensity factors of Appendix G of ASME XI for axial and
ISSN 0556-171X. npoôëeubi npounocmu, 2004, № 1 19
A. Ballesteros, G. Garcia, L. Bogede, J. Bros
circumferential welds. The ASME Code Case N-640 allows the use of K rather
than K ia to determine the pressure-temperature limit curves. The use of K ic
eliminates one of the conservatisms used to generate these curves.
Towards 60 Years of O peration. Section XI.M31 Reactor Vessel
Surveillance of NUREG 1801 (the GALL report) [8] contains recommended
actions and acceptable methods to evaluate the embrittlement status of the vessel
for a period of until 60 years. For instance, if a plant has a surveillance program
that consists of capsules with a projected fluence of less than the 60-year fluence
at the end of 40 years, at least one capsule is to remain in the reactor vessel, and it
should be tested during the period of extended operation. This is of application to
the Spanish BWRs since the lead factor is close to one.
One important remark in NUREG 1801 is the recommendation to remove the
standby capsules if the lead factors are relatively high. For example, in a reactor
with a lead factor of three, after 20 years, the capsule test specimens would have
received neutron exposure equivalent to what the reactor vessel would see in 60
years. Thus, the capsule is to be removed and placed in storage since further
exposure would not provide meaningful metallurgical data. The standby capsules
would be available for reinsertion into the reactor if additional license renewals
are sought (e.g., 80 years of operation). Although at present, the Spanish NPPs
are not contemplating any life extension beyond the 40-year term, Almaraz NPP
will be the first Spanish plant to follow this NUREG recommendation. Two
standby capsules from each unit will be removed from the vessel in the next
outage since by that time the standby capsules will have accumulated the neutron
fluence slightly higher than the projected one for the vessel at 60 years.
When all the surveillance capsules have been removed, some means must be
established to ensure that the on-going exposure of the reactor vessel is consistent
with the basis used to predict the effects of embrittlement to the end of life. It is
not possible to say now what the operating philosophies will be in the future. The
general recommendation is that, if possible, ex-vessel neutron dosimeter be
installed one or more fuel cycles prior to withdrawing the last (typically the
60-year) surveillance capsule. The dosimeter is then removed and analyzed
simultaneously with the surveillance capsule, and a replacement ex-vessel
dosimeter is installed. Simultaneous measurements inside and outside the reactor
vessel provide a larger amount of information to characterize the reactor vessel
exposure.
Facing a 60-year operation, the most promising techniques are the
reconstitution of surveillance specimens and Master curve testing. The former
allows us to solve the limitation in the amount of surveillance material available
for irradiation and testing, and the latter provides a technically sound approach
for defining a unique fracture toughness transition temperature T0 for ferritic
steels clearly superior to the old ^T NDT. The potential application of the Master
curve approach to the Spanish reactors and the benefits of its use were outlined in
[9].
In January 2002, the regulatory body CSN and the utilities represented by
UNESA started a 3-year project (CUPRIVA project) focused on the CT and
Charpy specimen reconstitution and Master curve testing. Two pilot plants are
participating in the project, Santa Maria de Garona (BWR) and Asco II (PWR),
20 ISSN 0556-171X. npo6n.eubi npounocmu, 2004, № 1
Beyond RPV Design Life
which provided surveillance material for investigation. The available irradiated
broken specimens from Asco II base material are used to machine reconstituted
precracked Charpy V-notch (PC-CVN) specimens. The Master-curve testing
results of the Asco II specimens will be compared with fracture toughness results
obtained by conventional testing of available irradiated compact test 1/2T CT
specimens. The base and weld surveillance materials of the Garona RPV are
investigated. Reconstituted CT and PC-CVN specimens are tested according to
the Master curve approach. For the Garona reactor it will be possible to compare
the results of the Master curve testing of unirradiated and irradiated specimens.
Preliminary results will be available by the end of the year 2003. The test results
from Asco and Garona reactors will be used to determine the remaining lives of
these reactors with a higher accuracy than previous evaluations. The possible
application of the ASME Code Cases N-629 and N-631 will be considered. The
study will include a comparison of the RTndt and RTT values. The latter
indexing temperature, as defined in the ASME Code Cases, is based on the
Master Curve concept, and assures that the K ic curve will bound the actual
material fracture toughness data by a functional equivalency of RTndt and RTTq .
As it is usual for the BWRs designed by the General Electric, the
surveillance practice of the two Spanish BWRs, Cofrentes and Garona reactors,
includes reinsertion in the vessel of capsules with reconstituted specimens. The
surveillance data provided by these non-mandatory capsules allows a better
definition of the embrittlement trend curves for these reactors, and a better
prediction of the projected RTn d t and USE values at the end of life. In addition
to the reinsertion of surveillance capsules, Garona NPP carried out a modification
of the surveillance holder to improve the lead factor of the surveillance capsules
that is now slightly higher than one. For Cofrentes NPP, the last capsule, which is
being manufactured and will be reinserted in the vessel in October 2003, will
include, in addition to the limiting beltline base and weld materials, a reference
steel or correlation monitor material, JRQ. This will allow us to detect and
analyze any anomaly or change in the irradiation conditions.
Conclusions. The reactor pressure vessel surveillance program is a key
factor in the life assessment of the Spanish nuclear reactors, since radiation
embrittlement of the RPV is the most life-limiting degradation mechanism, and
the RPV is the most important pressure boundary component of the nuclear power
plant. Similar design of the Spanish pressure vessels and their surveillance
programs are a favorable factor for the evaluation of the surveillance data, which
allows an easy identification of anomalies in any specific reactor.
The new embrittlement trend curves developed in the US have been applied
to Spanish reactors. These new correlations take into account new independent
variables like irradiation temperature, irradiation time in addition to the variables
taken into account in previous models (e.g., in Regulatory Guide 1.99, Rev. 2).
The results indicate that there is a good agreement between the theoretical values
(obtained using the Eason correlations) and the experimental values reported after
analyzing the surveillance capsules.
The projected values of RTndt and USE of the beltline limiting materials at
32 EFPY show that the pressure vessels could easily operate beyond the 40-year
ISSN 0556-171X. npo6n.eubi npounocmu, 2004, № 1 21
A. Ballesteros, G. Garcia, L. Bogede, J. Bros
design life. These key parameters were determined using the methodology
described in Regulatory Guide 1.99, Rev. 2, which is in force in Spain. Although
at present the Spanish NPPs are not contemplating any life extension beyond the
40-year term, the requirements established in NUREG-1801 are taken into
account. The main research activities in Spain are dealing with the surveillance
specimen reconstitution and Master curve testing.
Р е з ю м е
Проаналізовано стандартні програми моніторинга розрахункового ресурсу
(до 40 років експлуатації) корпусів атомних реакторів із використанням
зразків-свідків. Удосконалення методів випробувань та оцінки радіаційного
окрихчування матеріалів потребує перегляду діючих норм і проведення
наукового дослідження з метою уточнення ресурсу корпусних матеріалів
(подовження його до 60 років). Виконано прогнозування довговічності
корпусних сталей на основі результатів, що отримано на зразках-свідках та
на зразках типу Шарпі, а також із використанням методу “Master Curve”.
1. L. M. Davies and A. Ballesteros, “Aspects of operational life management of
nuclear power plants,” in: Proc. 3rd Int. Conf. on NDE in Relation to
Structural Integrity fo r Nuclear and Pressurized Components, Sevilla
(2001).
2. Charpy Embrittlement Correlation-Status o f Combined Mechanistic and
Statistical Bases fo r U.S. RPV Steels (MRP-45), PWR Materials Reliability
Program (PWRMRP), EPRI, Palo Alto, CA, 1000705 (2001).
3. Fracture Toughness Requirements fo r Protection against Pressurized
Thermal Shock Events, 10 Code of Federal Regulations 50.61, U.S. Nuclear
Regulatory Commission (1995).
4. Radiation Embrittlement o f Reactor Vessel M aterials, United States
Regulatory Commission, Regulatory Guide 1.99 Rev. 2 (1988).
5. Evaluation o f Reactor Pressure Vessels with Charpy Upper-Shelf Energy
Less than 50 Ft-Lb, Regulatory Guide 1.161 (1995).
6. Alternative Reference Fracture Toughness fo r Development o f P — T Limit
Curves, Section XI, Division 1, Code Case N-640 of the ASME Boiler &
Pressure Vessel Code (1999).
7. A lte rn a tiv e to R e feren ce F law O rien ta tion o f A p p en d ix G fo r
Circumferential Welds in Reactor Vessels, Section XI, Division 1, Code
Case N-588 of the ASME Boiler & Pressure Vessel Code, (1997).
8. NUREG 1801, Generic Aging Lessons Learned (GALL) Report (2001).
9. A. Ballesteros, “The Master curve approach and its significance,” in: Proc.
Int. Conf. PLIM-PLEX 99, Madrid (1999).
Received 26. 05. 2003
22 ISSN 0556-171X. Проблеми прочности, 2004, № 1
|
| id | nasplib_isofts_kiev_ua-123456789-47056 |
| institution | Digital Library of Periodicals of National Academy of Sciences of Ukraine |
| issn | 0556-171X |
| language | English |
| last_indexed | 2025-12-07T13:32:50Z |
| publishDate | 2004 |
| publisher | Інститут проблем міцності ім. Г.С. Писаренко НАН України |
| record_format | dspace |
| spelling | Ballesteros, A. Garcia, G. Bogede, L. Bros, J. 2013-07-09T13:18:27Z 2013-07-09T13:18:27Z 2004 Beyond RPV Design Life / A. Ballesteros, G. Garcia, L. Bogede, J. Bros // Проблемы прочности. — 2004. — № 1. — С. 15-22. — Бібліогр.: 9 назв. — англ. 0556-171X https://nasplib.isofts.kiev.ua/handle/123456789/47056 539.4 A set of standard reactor pressure vessels (RPV) surveillance programs for monitoring the design life (up to 40 years of reactor operation) have been analyzed. In view of the improved test methods and embrittlement evaluation procedures, the necessity has been shown of introducing modifications in the present surveillance programs aiming at a more precise RPV integrity evaluation facing a possible service life of 60 years. Service life predictions are performed for reactor pressure vessels, based on the available surveillance data, reconstitutioned Charpy specimens and Master Curve testing. Проанализированы стандартные программы мониторинга расчетного ресурса (до 40 лет эксплуатации) корпусов атомных реакторов с использованием образцов-свидетелей. Ввиду усовершенствования методов испытаний и оценки радиационного охрупчивания материалов отмечается необходимость пересмотра действующих программ и проведения научного исследования с целью уточнения ресурса прочности корпусных материалов (продления его до 60 лет). Выполнено прогнозирование долговечности корпусных сталей на основании результатов, полученных на образцах-свидетелях, образцах типа Шарпи, а также с использованием метода "Master Curve”. Проаналізовано стандартні програми моніторинга розрахункового ресурсу (до 40 років експлуатації) корпусів атомних реакторів із використанням зразків-свідків. Удосконалення методів випробувань та оцінки радіаційного окрихчування матеріалів потребує перегляду діючих норм і проведення наукового дослідження з метою уточнення ресурсу корпусних матеріалів (подовження його до 60 років). Виконано прогнозування довговічності корпусних сталей на основі результатів, що отримано на зразках-свідках та на зразках типу Шарпі, а також із використанням методу “Master Curve”. en Інститут проблем міцності ім. Г.С. Писаренко НАН України Проблемы прочности Научно-технический раздел Beyond RPV Design Life О возможности превышения расчетного ресурса корпусов атомных реакторов Article published earlier |
| spellingShingle | Beyond RPV Design Life Ballesteros, A. Garcia, G. Bogede, L. Bros, J. Научно-технический раздел |
| title | Beyond RPV Design Life |
| title_alt | О возможности превышения расчетного ресурса корпусов атомных реакторов |
| title_full | Beyond RPV Design Life |
| title_fullStr | Beyond RPV Design Life |
| title_full_unstemmed | Beyond RPV Design Life |
| title_short | Beyond RPV Design Life |
| title_sort | beyond rpv design life |
| topic | Научно-технический раздел |
| topic_facet | Научно-технический раздел |
| url | https://nasplib.isofts.kiev.ua/handle/123456789/47056 |
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