Lifetime Analysis of WWER Reactor Pressure Vessel Internals Concerning Material Degradation
Внутренние элементы реакторов подвергаются воздействию трех основных эксплуатационных факторов: нейтронного и гамма-излучений; статических и динамических механических напряжений и химических веществ, используемых для охлаждения реактора. Исследовано влияние этих факторов на расчетную долговечность...
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Інститут проблем міцності ім. Г.С. Писаренко НАН України
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Dudra, Ju. Szávai, Sz. 2017-01-13T14:10:18Z 2017-01-13T14:10:18Z 2010 Lifetime Analysis of WWER Reactor Pressure Vessel Internals Concerning Material Degradation / Ju. Dudra, Sz. Szávai // Проблемы прочности. — 2010. — № 1. — С.70-79. — Бібліогр.: 6 назв. — англ. 0556-171X https://nasplib.isofts.kiev.ua/handle/123456789/111647 539.4 Внутренние элементы реакторов подвергаются воздействию трех основных эксплуатационных факторов: нейтронного и гамма-излучений; статических и динамических механических напряжений и химических веществ, используемых для охлаждения реактора. Исследовано влияние этих факторов на расчетную долговечность внутренних элементов реакторов ВВЭР 440 с целью продления срока их эксплуатации. Внутрішні елементи реакторів зазнають впливу трьох основних експлуатаційних чинників: нейтронного і гамма-випромінювання; статичних і динамічних напружень та хімічних речовин, що використовуються для охолодження реактора. Досліджено вплив цих чинників на розрахункову довговічність внутрішніх елементів реакторів ВВЕР 440 із метою подовження терміну їх експлуатації. en Інститут проблем міцності ім. Г.С. Писаренко НАН України Проблемы прочности Научно-технический раздел Lifetime Analysis of WWER Reactor Pressure Vessel Internals Concerning Material Degradation Расчет долговечности внутренних элементов сосудов давления реакторов типа ВВЭР с учетом деградации материала Article published earlier |
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Lifetime Analysis of WWER Reactor Pressure Vessel Internals Concerning Material Degradation |
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Lifetime Analysis of WWER Reactor Pressure Vessel Internals Concerning Material Degradation Dudra, Ju. Szávai, Sz. Научно-технический раздел |
| title_short |
Lifetime Analysis of WWER Reactor Pressure Vessel Internals Concerning Material Degradation |
| title_full |
Lifetime Analysis of WWER Reactor Pressure Vessel Internals Concerning Material Degradation |
| title_fullStr |
Lifetime Analysis of WWER Reactor Pressure Vessel Internals Concerning Material Degradation |
| title_full_unstemmed |
Lifetime Analysis of WWER Reactor Pressure Vessel Internals Concerning Material Degradation |
| title_sort |
lifetime analysis of wwer reactor pressure vessel internals concerning material degradation |
| author |
Dudra, Ju. Szávai, Sz. |
| author_facet |
Dudra, Ju. Szávai, Sz. |
| topic |
Научно-технический раздел |
| topic_facet |
Научно-технический раздел |
| publishDate |
2010 |
| language |
English |
| container_title |
Проблемы прочности |
| publisher |
Інститут проблем міцності ім. Г.С. Писаренко НАН України |
| format |
Article |
| title_alt |
Расчет долговечности внутренних элементов сосудов давления реакторов типа ВВЭР с учетом деградации материала |
| description |
Внутренние элементы реакторов подвергаются воздействию трех основных эксплуатационных факторов: нейтронного и гамма-излучений; статических и динамических механических
напряжений и химических веществ, используемых для охлаждения реактора. Исследовано
влияние этих факторов на расчетную долговечность внутренних элементов реакторов ВВЭР
440 с целью продления срока их эксплуатации.
Внутрішні елементи реакторів зазнають впливу трьох основних експлуатаційних чинників: нейтронного і гамма-випромінювання; статичних і динамічних напружень та хімічних речовин, що використовуються для охолодження
реактора. Досліджено вплив цих чинників на розрахункову довговічність
внутрішніх елементів реакторів ВВЕР 440 із метою подовження терміну їх
експлуатації.
|
| issn |
0556-171X |
| url |
https://nasplib.isofts.kiev.ua/handle/123456789/111647 |
| citation_txt |
Lifetime Analysis of WWER Reactor Pressure Vessel Internals Concerning Material Degradation / Ju. Dudra, Sz. Szávai // Проблемы прочности. — 2010. — № 1. — С.70-79. — Бібліогр.: 6 назв. — англ. |
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2025-11-26T16:26:56Z |
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| fulltext |
UDC 539.4
Lifetime Analysis of WWER Reactor Pressure Vessel Internals
Concerning Material Degradation
Ju. Dudra and Sz. Szávai
Bay Zoltan Foundation for Applied Research, Institute for Logistics and Production
System, Department of Structural Integrity, Miskolc-Tapolca, Hungary
ÓÄÊ 539.4
Ðàñ÷åò äîëãîâå÷íîñòè âíóòðåííèõ ýëåìåíòîâ ñîñóäîâ äàâëåíèÿ
ðåàêòîðîâ òèïà ÂÂÝÐ ñ ó÷åòîì äåãðàäàöèè ìàòåðèàëà
Þ. Äóäðà, Ø. Øàâàè
Ôîíä ïðèêëàäíûõ èññëåäîâàíèé èì. Çîëòàíà Áàÿ, Èíñòèòóò ëîãèñòèêè è ïðîìûøëåí-
íûõ ñèñòåì, Îòäåë ïðî÷íîñòè êîíñòðóêöèé, Ìèøêîëüö-Òàïîëüöà, Âåíãðèÿ
Âíóòðåííèå ýëåìåíòû ðåàêòîðîâ ïîäâåðãàþòñÿ âîçäåéñòâèþ òðåõ îñíîâíûõ ýêñïëóàòàöèîí-
íûõ ôàêòîðîâ: íåéòðîííîãî è ãàììà-èçëó÷åíèé; ñòàòè÷åñêèõ è äèíàìè÷åñêèõ ìåõàíè÷åñêèõ
íàïðÿæåíèé è õèìè÷åñêèõ âåùåñòâ, èñïîëüçóåìûõ äëÿ îõëàæäåíèÿ ðåàêòîðà. Èññëåäîâàíî
âëèÿíèå ýòèõ ôàêòîðîâ íà ðàñ÷åòíóþ äîëãîâå÷íîñòü âíóòðåííèõ ýëåìåíòîâ ðåàêòîðîâ ÂÂÝÐ
440 ñ öåëüþ ïðîäëåíèÿ ñðîêà èõ ýêñïëóàòàöèè.
Êëþ÷åâûå ñëîâà: âíóòðåííèå ýëåìåíòû ðåàêòîðîâ, ñòàðåíèå, ðàäèàöèîííîå
îõðóï÷èâàíèå, êîððîçèîííîå ðàñòðåñêèâàíèå, ðàñ÷åò äîëãîâå÷íîñòè.
Background. The internal structures of water reactors play an important role as
they contain the core of the reactor, channel the water flow inside the vessel, support
and guide the instrumentation necessary for controlling and monitoring the reactor.
Long-term ageing at WWER operating temperatures results in changes in
microstructure and phase composition, and, hence, mechanical properties of
structural steels for WWER. Design basis for reactor internals in WWER has been
calculated with no account of ageing of these structures due to that time unknown
mechanisms of degradation.
This study aims to extend the operational lifetime of WWER 440 type units.
Technical margins, the inherent safety and the realized safety improvement
measures provide an opportunity for the extension of operational lifetime of
nuclear power plant.
Namely, the purpose of this task is to perform evaluative analyses of
presumable changes in material properties of WWER 440 reactor vessel internals
(RVI). In accordance with these, the study address the issue of availability of
degradation mechanisms which limit the lifetime of RVI structural elements.
Nowadays, it is a very important research area, but the available literature
contain scarce information on failures of WWER RVI. Within this framework, we
consider the following:
© Ju. DUDRA, Sz. SZÁVAI, 2010
70 ISSN 0556-171X. Ïðîáëåìû ïðî÷íîñòè, 2010, ¹ 1
(i) systematization of structural plans and basic data required for the analysis,
determination of initial data for analyses;
(ii) investigation of the presumable changes in material properties of the RVI;
(iii) perform strength and fatigue analyses of critical structural elements of
RVI;
(iv) estimate the technically possible operation time of the structural elements
considering the operation load of the 50 and 60 years operation time.
Basically, the strength and fatigue analyses should be performed in accordance
with the provisions of the ASME (ASME III (2001)). Calculation methodologies
for these structures include strength analysis, fatigue analysis and thermal
stratification analysis for systems and components.
Structural Elements. Boundaries of Analyses. The subjects of this study are
the WWER 440 reactor vessel internals, namely the guide tubes, core barrel, core
basket and lower core barrel structures (Fig. 1.). Analyzed boundaries are the edge
boundaries of the WWER 440 RVI and its nonwelded joints.
Potential Degradation, Material Models and Evaluations. In the following,
we summarize the possible ageing mechanisms of WWER 440 RVI (Fig. 2.).
Evaluation of these mechanisms is based on service experience, pertinent laboratory
data and relevant experience from other industries.
During the evaluation we examine the following: effect of changes of the
material characteristics not only on the strength and fatigue analysis, but also on
the allowable crack size and crack propagation rates, in case of operation or
breakdown.
Fig. 1. Scheme of WWER 440 reactor pressure vessels and internals.
ISSN 0556-171X. Ïðîáëåìû ïðî÷íîñòè, 2010, ¹ 1 71
Lifetime Analysis of WWER Reactor Pressure Vessel Internals ...
The austenitic steels and the base material of the reactor internals are
characterized by the ductile fracture mode. At the same time, the effect of
irradiation induces the embrittlement mechanisms in the material, which can lead
to deterioration of the material toughness. These embrittlement mechanisms can
decrease the Charpy test values and the J–R curve (and J 0 2. ). The J–R curve can
be applied (in addition to the strength material characteristics) to investigation of
crack stability or stable crack propagation. For evaluation of the crack propagation
stability two parameters of failure assessment diagram (FAD) are applicable.
Loads and Material Characteristics. Within the framework of the study, the
analysis was made by means of the load catalogue (above 100 and 108% power
values), which identify the loads, determine load curves and load cycle numbers for
cyclic loads expected during the planned extended lifetime.
The main structural material used for WWER 440 for RVI is titanium
stabilized austenitic stainless steel 08Ch18N10T (equivalent to A-321 steel); minor
parts were made from nickel alloy ChN35VT-VD (this material is used for studs
and is tungsten-alloyed) and steel 14Ch17N2. The weld metals Sv08Ch19N10G2B
and Sv04Ch19N11M3 were used.
Static Strength Material Characteristics and Crack Growth under Static Load.
The RVI are susceptible to several degradation mechanisms in course of reactor
operation, the most serious problem being the change of material properties and
J–R curve due to neutron irradiation. By the requested data to characterize the
operational change of material properties are meant the material changes due to
irradiation. To perform the calculations the material characteristics and J–R curve
should be determined concerning 50 and 60 years relevant to the expected
maximum fluence.
Fig. 2. Evaluation methodology.
72 ISSN 0556-171X. Ïðîáëåìû ïðî÷íîñòè, 2010, ¹ 1
Ju. Dudra and Sz. Szávai
Material Characteristics Concerning Resistance of the Fatigue Crack
Propagation. According to the ASME Code XI:2001 C 3210-1 and other
experimental results of the investigation, which were performed in inner atmosphere
(in cooling water) of the WWER 440 and on laboratory air, the ratio of the velocity of
the crack propagation was around 2.5. Making use of this constant ratio, we
determined the curve of the crack propagation concerning the heat-transfer media.
Material Characteristics Concerning Resistance of the Stress Corrosion Crack
Propagation. After longer operation, the appearance of stress corrosion cracking
cannot be excluded in the reactor vessel internals. This crack propagation is
function of the temperature, the media and other outer effects, like the neutron
irradiation. To investigate the possibility of the crack propagation, the threshold of
the K SCCI stress intensity factor (SIF) is required.
Functional Temperature and Fluence Information. The nominal inlet and
outlet temperatures in the reactor vessel are, respectively: 267/296�C for 100%
power and 267/299�C for 108% power.
Former research results of neutron physics calculations concern the position of
the reactor vessel wall. The fluence of the test sample estimated for 60 years
practically equals to the load of the outer surface of the core barrel, and it is ~ 12 dpa.
On the basis of study [1, 2] it can be proved that the fluence value of the core
boundary can be assessed as that of the core barrel multiplied by ~ 5–10 dpa,
namely it can be estimated as 35–60 dpa for 60 years operation time.
According to the NRI study [3], which concerns fracture analysis of the basket
of the Greifswald Unit 1, the fluence value of the basket is the following. The
calculated maximum dose for the baffle component of Greifswald Unit 1 internals
after 15th operation cycle, according to the first calculation method is 12 1026. � m�2
(E �0 5. MeV), i.e., 11.4 dpa [4], and according to the second method is13 1026. � m�2
(E �1 0. ), i.e., 19.4 dpa [5].
In the analysis, the first step is to define the maximum neutron fluence values
of the critical place of the RVI after 50 and 60 years’ operation time. The resulting
material characteristics relevant to the maximum fluence will be used in the
calculations.
The KFKI Atomic Energy Research Institute has performed calculations
concerning the critical environments of the structural elements of RVI in order to
determine the expected maximum fluence value. The analyses were made for 30,
50, and 60 years’ operation time periods. The irradiation dose was defined in dpa
units (displacement per atom), and Table 1 shows the results of the measurements.
Based on our estimations, the results show 20% decrease in the bolt head and
40% decline in the shank of the bolt in radial direction. The estimated irradiation
value at the bolt neck is about 85 dpa.
Available Material Data and Material Characteristics Concerning 50 and
60 Years’ Operation Time. The base metals (A304, A304L, A316, A316L,
A316Ti) of RVI of Western type reactors are nonstabilized and/or nontitanium
stabilized austenitic steels, and cannot be considered equivalent to the materials
used in WWER type reactors. On the basis of the previous overview of the
available analytical and research results [2, 3], the danger cannot be excluded that
higher neutron fluence can cause the embrittlement which changes the mechanism
ISSN 0556-171X. Ïðîáëåìû ïðî÷íîñòè, 2010, ¹ 1 73
Lifetime Analysis of WWER Reactor Pressure Vessel Internals ...
of the austenitic steel fracture and invokes of the intergranular type fracture
mechanism in the material. To avoid appearance of these particular examination
results, material characteristics are required for high fluence values.
Available examination results regarding RVI of NRI Rez and Prometey
Institute for 50–60 dpa are the following:
Tensile test � � �� , YS , UTS , A, RA, and fracture mode.
Instrumented impact � KV and KCV .
Fracture resistance � J 0 2. , J R� , and fracture mode.
During calculations we have defined the material characteristics of the core
barrel, core basket, shank of the bolt, bolt head and core plate concerning 50 and 60
years operation time relevant to the expected maximum fluence (based on the NRI,
Prometey measurement and other data available in the literature).
In accordance with the above-mentioned notions, we have determined:
a) Strength characteristics (YS, UTS, and RA) as a function of the size of
irradiation. These material characteristics show an increase (except RA, which
shows a decrease) until these trends turn into saturation (after 60 dpa).
b) Fatigue curves based on strength characteristics. Our investigations show
that the calculated fatigue curves run above the one constructed according to the
PNAÉ [5]. Thus, the earlier fatigue analyses will be valid for the 50–60 years
operation time. In the following, we refer only to the curve corresponding to the 60
years’ operation time according to r �0 (see Fig. 3).
c) We have also determined the J c static crack growth resistance and the
K Jc stress intensity factor as a function of the irradiation scale. For example, for
30 dpa J c �15 kJ/m2 and K Jc �51 MPa m (for the base metal), and kinetics of
both material characteristics manifests saturation over 30–35 dpa (both in case of
the base metal and the weld material).
d) We have defined the swelling characteristics of the materials under study
after 50–60 years’ operation time. The swelling level for 30 dpa in the course of the
lower core barrel is 0.3%, in the core basket for 57 dpa is 1.1% , while the swelling
level in the bolt head for 75 dpa is 1.8%.
In case of the material characteristics (K SCCI , da/dt–K) concerning crack
propagation and crack initiation, these are affected by the environmental effect:
determination of these material characteristics are based on particular measurement
results. However, there are no measurement data in the literature. Therefore we set
out from the earlier conservative approach, where 3 10 11
�
� m/s value for the crack
propagation rate is used, irrelevant of the load value.
T a b l e 1
The Irradiation Dose in the Site with Maximum Dose
Operation
time
[year]
Irradiation [dpa]
Bolt head,
baffle
Shank
of the bolt
Core basket Lower core
barrel
Core plate 2–3
30 42 36 34 14 36
50 75 61 58 24 61
60 91 73 70 29 73
74 ISSN 0556-171X. Ïðîáëåìû ïðî÷íîñòè, 2010, ¹ 1
Ju. Dudra and Sz. Szávai
Previous Assessment of the Presumably Alternative Mechanisms of the
RVI Material Properties. The degradation of the material characteristics and the
degradation of the components are the result of complex environmental effects.
The hardening and embrittlement, fatigue and environmental effects (corrosion
cracking and propagation) appear as the effect of the irradiation, those are
degradation mechanisms, and analysis of these mechanisms is urgently required.
Boundary Analysis. During the analysis we have selected at least one most
critical environment, in case of any structural elements, for which we have
performed the fracture mechanics calculations. If the cross section did not coincide
with height of the zone where the largest irradiation was expected, we would select
the second critical cross section of the structural element with the height of the
zone.
Necessary Calculations – Analysis of Stable Crack Propagation. As we have
already mentioned, the effect of irradiation induces the embrittlement mechanisms
in the material, which can lead to deterioration of the material toughness. These
mechanisms can decrease the Charpy specimens’ test values and affect the J–R
curve (and J 0 2. ). The J–R curve can be applied (beside the strength material
characteristics) to the investigation of the stableness of the crack or the stable crack
propagation. To evaluate the stability of the crack propagation two parameters FAD
are applicable.
While the RVI are not pressure elements, it is practical to assume that the
material contains a through crack. The ASME standard contains no information on
the calculation of through cracks, therefore we have applied the FITNET European
Union Standard Collection in our calculations.
The basis of the determination of the allowable crack length was the stress
intensity factor K I . We have determined the stress intensity factor supposing a
different crack length for every single dangerous environment. Using this
Fig. 3. Fatigue curves ñorresponding to 60 years’ operation time ( ).r � 0
ISSN 0556-171X. Ïðîáëåìû ïðî÷íîñòè, 2010, ¹ 1 75
Lifetime Analysis of WWER Reactor Pressure Vessel Internals ...
information we were able to define the crack size which is tolerable in the structure
under 60 years operation. In possession of the FAD, calculation of the allowable
crack size can be performed for the critical environment of the structural element.
Table 2 contains the allowable crack lengths for some elements.
The allowable crack sizes are still large enough to be visually detected.
However, our examinations did not reveal such large cracks neither in the marked
environments, nor anywhere else in the RVI components.
Necessary Calculations – Fatigue Crack Propagation Analyses. For obtaining
data relative to fatigue crack propagation, particular measurement results are
needed. Since the operating load cycles include neither significant load, nor large
number of cycles, these do not cause fatigue crack propagation. The crack
propagation can be resulted only from the vibration of the reactor internals, if the
load vibration induces larger stress intensity factor than the threshold SIF of the
fatigue crack propagation ( K th ). The vibration-caused number of cycles is so
large that even small crack propagation rate can gradually provide a larger crack
size. Therefore the allowable crack size has to be such that the vibration load
would not cause crack propagation.
According to the calculations, the fatigue crack propagation preclude so large
a crack that could imply stable or instable crack propagation, in case of zone basket
and lower core plate. The currently known examinations did not reveal cracks in
the dangerous cross section of the control rod guide tubes and lower core barrel.
Stress Corrosion Crack Propagation Analyses. After longer operation the
reactor vessel internals cannot be excluded from the appearance of the stress
corrosion cracking. We have performed stable crack propagation analysis of a crack
appearing in fastener bolts. Since the ASME standard contains no information
regarding calculation of the crack appearing in bolt, calculations were performed
according to the FITNET standard, based on the BS 7910 procedure. The stress of
the shank of a bolt equals to the allowable maximum tensile stress value (according
to the PNAÉ standard [5]), which is half of the planned yield stress.
During calculations we have determined the SIF value, from which we have
deduced the J-integral and fracture toughness of the material as a function of
irradiation dose. We have also determined the J 0 2. value corresponding to
fracture.
T a b l e 2
The Allowable Crack Sizes in Some Critical Environments
Environment T , �C Irradiation
[dpa]
Operation
condition
Allowable crack
size [mm]
Lower core barrel K205 295 9.7 tv10* 544
Control rod guide tubes K602 262 9.7 tv10* 1280
Lower core barrel K620-3 275 29.0 tv10* 542
Core basket K301-1 332 70.0 NA500 heat
circuit fracture*
82
* Final safety report / Load Catalogue Data.
76 ISSN 0556-171X. Ïðîáëåìû ïðî÷íîñòè, 2010, ¹ 1
Ju. Dudra and Sz. Szávai
We have determined the time interval between the initial crack appearance and
the moment when the fatigue-corrosion crack propagation reaches the section of
the stable crack propagation. The operation time of the bolt significantly depends
on the crack shape. When we consider the worst case, namely a circular crack, it
takes approximately 3.5 years to reach the stable crack propagation phase in a RVI
bolt (Fig. 4.). When we consider a straight crack front, the respective period is
approximately 6.1 years (Fig. 4).
Swelling Effect. Swelling appears primarily in components exposed to high
gamma irradiation. We have determined the level of swelling appearing in the
basket wall, in order to demonstrate the distribution of the wall swelling. In the
basket wall (at the half of the zone height), swelling along the wall can be
described using the swelling relation given by the Prometey Institute [6]), assuming
that the value of the irradiation can be decreased by 40% in the wall.
Fig. 4. Time required to reach an allowable stress corrosion crack size in a RVI component.
Fig. 5. Swelling effect in the basket wall after 50–60 years’ operation.
ISSN 0556-171X. Ïðîáëåìû ïðî÷íîñòè, 2010, ¹ 1 77
Lifetime Analysis of WWER Reactor Pressure Vessel Internals ...
Figure 5 shows the swelling distribution in the basket wall after 50–60 years’
operation time. The swelling effect can induce the second principal stress (and
thus, stress biaxiality) in the material.
Conclusions. We have examined the presumable changes in the RVI material
properties and estimated the technically possible operation time of the structural
elements considering the operation load of the 50 and 60 years’ operation time.
The results show that the swelling and the irradiation-induced stress corrosion
cracking are the most significant degradation mechanisms in the course of
prolongation of the RVI operation time.
From the point of view of the swelling, the critical environments are: basket
component at the middle part of the zone, baffle, fastener bolts, and cylinder of the
core basket.
Based on measurement results and the presumable irradiation and temperature
data, it cannot be excluded that 0.5–1.0% swelling will cause deformation and
tightening in the fasteners bolts after 30 years’ operation time. The deformation of
the baffle can influence the installation, assemblage and joints as well. The
appearance of abrasion places on the matt surfaces will be the first sign of this
deformation.
From the point of view of the irradiation-induced stress corrosion cracking,
the critical sites are the basket fastener bolts which can be characterized by fast
irradiated neutron fluence and the stagnant environments. The swelling causes
hindered expansion of the bolt, while possible stress corrosion cracking of the bolts
can lead to fracture. During the examination, for lack of other information, we
supposed the preset load according to the PNAÉ maximum, which is rather a
conservative approach with a significant safety factor. For determination of the
crack propagation rate for the selected environments further analysis is needed.
Ð å ç þ ì å
Âíóòð³øí³ åëåìåíòè ðåàêòîð³â çàçíàþòü âïëèâó òðüîõ îñíîâíèõ åêñïëóàòà-
ö³éíèõ ÷èííèê³â: íåéòðîííîãî ³ ãàììà-âèïðîì³íþâàííÿ; ñòàòè÷íèõ ³ äèíàì³÷-
íèõ íàïðóæåíü òà õ³ì³÷íèõ ðå÷îâèí, ùî âèêîðèñòîâóþòüñÿ äëÿ îõîëîäæåííÿ
ðåàêòîðà. Äîñë³äæåíî âïëèâ öèõ ÷èííèê³â íà ðîçðàõóíêîâó äîâãîâ³÷í³ñòü
âíóòð³øí³õ åëåìåíò³â ðåàêòîð³â ÂÂÅÐ 440 ³ç ìåòîþ ïîäîâæåííÿ òåðì³íó ¿õ
åêñïëóàòàö³¿.
1. Strength Analysis, Fatigue Analysis and Thermal Stratification Analysis for
Systems, Structures, and Components (SSCs) – Reactor Pressure Vessel,
Steam Generator, Pressurizer, Main Circulating Pump, Main Gate Valve,
Pipelines, Vessels, Pumps, Heat Exchangers and Valves – Classified into the
Safety Classes 1 and 2, Methodology and Criteria Documentation.
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Evaluation and Studies (AMES), Report No. 11, Paris (1997).
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Materials from Greifswald Active Samples/Active Material Database/Core
Barrel, Report NRI Rez, DITI 302/419 Rev. 3 (2007).
78 ISSN 0556-171X. Ïðîáëåìû ïðî÷íîñòè, 2010, ¹ 1
Ju. Dudra and Sz. Szávai
4. CEA, TECNATOM, and VTT, Effect of Irradiation on Water Reactors
Internals, AMES Report No. 11, EUR 17694 EN, European Commission,
Brussels–Luxemburg (1997).
5. PNAÉ G-7-002-87. Regulations for Strength Analysis of Equipment and
Piping of Nuclear Power Plants [in Russian], Énergoatomizdat, Moscow
(1989).
6. B. Z. Margolin, I. P. Kursevich, A. G. Gulenko, et al., Generalization of
Experimental Data and Development of Methods for Prediction of Physical-
Mechanical Properties of Irradiated Material for Pressure Vessel Internals
Concerning the Blocks 1–4 of the Paks Nuclear Power Plant [in Russian],
Final Connon Report Rev01, Saint-Petersburg (2008).
Received 21. 06. 2009
ISSN 0556-171X. Ïðîáëåìû ïðî÷íîñòè, 2010, ¹ 1 79
Lifetime Analysis of WWER Reactor Pressure Vessel Internals ...
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