Research the behaviour and properties of WWER type fuel claddings from Zr1%Nb alloy in loss of the coolant accident
The paper presents safety criteria placed on fuel rod condition in loss of the coolant accident (LOCA) conditions as applied to reactor plants with WWER. The paper reveals briefly experimental studies carried out to validate safety criteria (acceptance criteria). The scope of the data experimentally...
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| Опубліковано в: : | Вопросы атомной науки и техники |
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| Дата: | 2021 |
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| Мова: | Англійська |
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Національний науковий центр «Харківський фізико-технічний інститут» НАН України
2021
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| Назва журналу: | Digital Library of Periodicals of National Academy of Sciences of Ukraine |
| Цитувати: | Research the behaviour and properties of WWER type fuel claddings from Zr1%Nb alloy in loss of the coolant accident / M.M. Semerak, S.S. Lys // Problems of Atomic Science and Technology. — 2021. — № 2. — С. 80-86. — Бібліогр.: 9 назв. — англ. |
Репозитарії
Digital Library of Periodicals of National Academy of Sciences of Ukraine| _version_ | 1859874807233904640 |
|---|---|
| author | Semerak, M.M. Lys, S.S. |
| author_facet | Semerak, M.M. Lys, S.S. |
| citation_txt | Research the behaviour and properties of WWER type fuel claddings from Zr1%Nb alloy in loss of the coolant accident / M.M. Semerak, S.S. Lys // Problems of Atomic Science and Technology. — 2021. — № 2. — С. 80-86. — Бібліогр.: 9 назв. — англ. |
| collection | DSpace DC |
| container_title | Вопросы атомной науки и техники |
| description | The paper presents safety criteria placed on fuel rod condition in loss of the coolant accident (LOCA) conditions as applied to reactor plants with WWER. The paper reveals briefly experimental studies carried out to validate safety criteria (acceptance criteria). The scope of the data experimentally obtained by research the behaviour and properties of WWER type fuel claddings from Zr1%Nb alloy under loading conditions simulating the stage of core flooding with water in LOCA suffices to judge the character and numerical value of criterional parameters of the embrittlement criterion in terms of the cladding stability upon flooding and subsequent implementation of fuel assembly (FA) unloading and transportation. Accidents are considered involving loss of coolant by primary circuit which are characterized by conditions of degraded heat transfer from fuels. During accidents loss of tightness by fuel rod cladding is tolerable, however, in this case, the cooling of a distorted fuel rod and the dismantling (unloading) of the core after an accident have to be feasible.
Наведено критерії безпеки, що пред’являються до стану твелів в умовах проектних аварій з втратою теплоносія для реакторних установок з ВВЕР. Представлений короткий огляд експериментальних досліджень, проведених з метою обґрунтування критеріїв безпеки. Обсяг експериментальних даних, отриманих при дослідженні поведінки і властивостей оболонок твелів реактора типу ВВЕР зі сплаву Zr1%Nb в умовах навантаження, імітуючих стадію затоплення активної зони водою при аварії з втратою теплоносія, достатній для судження про характер і чисельне значення критеріальних параметрів окрихчування з точки зору стійкості оболонок під час заливу і видаленні ТВЗ та транспортуванні. Розглядаються аварії з втратою теплоносія першого контуру, для яких характерне погіршення умов тепловідведення від твелів. У процесі аварій допустима розгерметизація оболонки твела, однак при цьому повинна зберігатися можливість охолодження твела зі зміненою геометрією і можливість розбирання (вивантаження) активної зони після аварії.
Приведены критерии безопасности, предъявляемые к состоянию твэлов в условиях проектных аварий с потерей теплоносителя для реакторных установок с ВВЭР. Представлен краткий обзор экспериментальных исследований, проведенных с целью обоснования критериев безопасности. Объем экспериментальных данных, полученных при исследовании поведения и свойств оболочек твэлов реактора типа ВВЭР из сплава Zr1%Nb в условиях нагружения, имитирующих стадию залива активной зоны водой при аварии с потерей теплоносителя, достаточен для суждения о характере и численном значении критериальных параметров охрупчивания с точки зрения стойкости оболочек при заливе и извлечении ТВС и транспортировке. Рассматриваются аварии с потерей теплоносителя первого контура, для которых характерно ухудшение условий теплоотвода от твэлов. В процессе аварий допустима разгерметизация оболочки твэла, однако при этом должны сохраняться возможности охлаждения твэла с измененной геометрией и разборки (выгрузки) активной зоны после аварии.
|
| first_indexed | 2025-12-07T15:51:07Z |
| format | Article |
| fulltext |
80 ISSN 1562-6016. ВАНТ. 2021. №2(132)
https://doi.org/10.46813/2021-132-080
UDC 621.039.548
621.039.586
RESEARCH THE BEHAVIOUR AND PROPERTIES OF WWER TYPE
FUEL CLADDINGS FROM Zr1%Nb ALLOY
IN LOSS OF THE COOLANT ACCIDENT
M.M. Semerak, S.S. Lys
Lviv Polytechnic National University, Lviv, Ukraine
E-mail: lysss@ukr.net
The paper presents safety criteria placed on fuel rod condition in loss of the coolant accident (LOCA) conditions
as applied to reactor plants with WWER. The paper reveals briefly experimental studies carried out to validate
safety criteria (acceptance criteria). The scope of the data experimentally obtained by research the behaviour and
properties of WWER type fuel claddings from Zr1%Nb alloy under loading conditions simulating the stage of core
flooding with water in LOCA suffices to judge the character and numerical value of criterional parameters of the
embrittlement criterion in terms of the cladding stability upon flooding and subsequent implementation of fuel
assembly (FA) unloading and transportation. Accidents are considered involving loss of coolant by primary circuit
which are characterized by conditions of degraded heat transfer from fuels. During accidents loss of tightness by
fuel rod cladding is tolerable, however, in this case, the cooling of a distorted fuel rod and the dismantling
(unloading) of the core after an accident have to be feasible.
INTRODUCTION
Validation of reactor plant (RP) safety under
accident conditions is an indispensable component of
work to license NPP operation. The most important
element of RP safety validation is analysis of behaviour
of nuclear fuel (FS, fuel rods) in design basis
(postulated) accidents.
The objective of the work is research the behaviour
and properties of WWER type fuel claddings from
Zr1%Nb alloy under loading conditions simulating the
stage of core flooding with water in LOCA. A design
basis accident is the one for which the design defines
initial events and final states as well as contemplates
safety systems that with the account for a single failure
of safety systems or a single staff error, independent of
an initial event, provide for the limitation of its
consequences via limits established for accidents of this
type [1, 2].
Accidents are considered involving loss of coolant
by primary circuit which are characterized by conditions
of degraded heat transfer from fuels. Under these
conditions, the temperature of fuel rods (fuel and/or
cladding) rises compared to that under normal operating
conditions. During accidents loss of tightness by fuel
rod cladding is tolerable, however, in this case, the
cooling of a distorted fuel rod and the dismantling
(unloading) of the core after an accident have to be
feasible.
These requirements are provided via introducing
some limitations on the parameters of the major
processes proceeding in a fuel rod in design basis
accidents, namely:
embrittlement of fuel rod cladding (decreased
ductility) effected by its intensive oxidation at elevated
temperatures that can lead to a fuel rod destruction into
several fragments (fragmentation) at the stage of core
cooling down;
hydrogen release that may lead to a hydrogen
explosion in the core and is also related to intensive
oxidation;
fuel melting.
FORMULATION OF THE PROBLEM
The major requirements for the RP safety are called
upon to provide the permanent cooling and the
feasibility of a core discharge [1, 2].
In LOCA these requirements are met when the
maximal design limit of fuel rod damage is not
exceeded [2]. The fuel rods safety state in LOCA is
stipulated by the following acceptance criteria:
Maximal design limit
of fuel
Maximum temperature of cladding does not
exceed 1200 °C
Maximum local depth of cladding oxidation is
not higher than 18% of its original thickness
Limitation of
embrittlement
rod damage Fraction of steam reacted Zr in core is not higher
than 1% of its mass in fuel cladding
Limitation of hydrogen
release
Fuel temperature does not exceed fuel melting
temperature
Absence of melted
fuel-cladding
interaction
ISSN 1562-6016. ВАНТ. 2021. №2(132) 81
The validation and check-up of safety criteria are
performed via:
experimentally implemented validation of safety
criteria;
design analysis of fuel rod state in design basis
accidents at the detailed design stage;
check-up of fulfilling (not exceeding) safety
criteria.
ANALYSIS OF CRITERION LIMITING
EMBRITTLEMENT
OF Zr1%Nb FUEL CLADDING MATERIAL
In process of LOCA, the inadequate cooling of fuel
claddings gives rise to their heating to reach high
temperatures. Under these conditions, their intensive
steam oxidation is feasible.
The degree of the cladding oxidation is basically
governed by the levels of temperatures and pressure,
time of oxidation in steam and deformation.
As a results of the cladding-steam interaction, the
cladding material becomes brittle.
Because of the cladding material embrittlement, the
Zr1%Nb alloy changes its original thermophysical and
mechanical properties, and the ductility characteristics
become lower. The thickness of the unoxidized cladding
material is reduced.
Temperature stresses of cladding arising at the stage
of core flooding with water from the emergency cooling
system as well as mechanical loads effected by fuel
assembly (FA) unloading and transportation may lead
to the fragmentation of an embrittled cladding.
The degree of Zr1%Nb cladding-coolant interaction
is stipulated by the maximum limit of a fuel rod damage
in terms of the tolerable depth of cladding oxidation,
namely, the embrittlement criterion [1].
The degree of the Zr1%Nb cladding embrittlement is
limited by two criterional parameters
lim
cladТ and local
oxidation depth of cladding (equivalent cladding
reacted – ECR) lim
cladECR :
the maximal cladding temperature Tclad does not
exceed 1200 °C
Тclad <
lim
cladТ = 1200 °С.
The ultimate value of the fuel cladding temperature
lim
cladТ characterizes the temperature with the exceeding
of which a self-sustaining Zr-steam reaction may begin.
the maximum ECR is not more than 18% of its
original thickness lim
cladECR
ECR < lim
cladECR = 18%.
The ultimate magnitude of ECR reflects the degree
of the embrittlement of a fuel cladding material and, if it
is exceeded, a cladding may fracture in brittle mode as
a result of loads produced by the emergency core
cooling, discharge, handling and transportation of FA.
ECR is an overall thickness of an equivalent Zr
layer (that would react with steam assuming the whole
locally absorbed oxygen would be spent to form
stoichiometric Zr, ZrO2) related to the original cladding
thickness.
If a fuel rod is ruptured (depressurized), the
oxidation of both outer and inner surfaces of cladding
are taken into account.
The degree of oxidation as applied to Zr1%Nb
claddings is assessed from either the oxygen specific
weight gain, ∆m, or ECR.
ECR and weight gain, ∆m, are related by the ratio
[4]
ECR = N ∙ ( e/o ) ∙ 100 %,
wehre N – is the coefficient that takes into account the
oxidation of cladding at both the surfaces (N = 2 for a
ruptured); δo – is original thickness of specimen, cm;
δe – is thickness of equivalent layer (calculated
thickness of Zr layer that would be spent to form ZrO2),
cm;
e = Zr /O2 ∙ 1/Zr ∙ m,
Zr, O2 – are molecular masses of zirconium and oxygen,
respectively; Zr – is zirconium density, mg/cm
3
; m – is
specific weight gain, mg/cm
2
; m = М/Sо; М – is
oxygen weight gain, mg; Sо – is area of original cladding
surface, cm
2
.
It is to be pointed out that the specific weight gain of
a cladding is found by dividing the weight gain by the
area of the original cladding surface. This gives a
conservative assessment of the specific weight gain. The
real oxidized area may be larger at the expense of
possible cladding deformation [4].
The ration to determine ECR has the form of
ECR = N 4.355 ∙ 10
-2
∙ m/o %. (1)
It is recommended to determine the oxygen weight
gain of Zr1%Nb claddings via the parabolic ratio [5]
m = 920 ∙ exp (-10410/T) ∙ , (2)
where m – is specific weight gain, mg/cm
2
; Т – is
temperature, K; – is time, s.
(2)
The relation (2) recommended for the assessment the
degree of Zr1%Nb cladding oxidation is conservative at
the temperatures of 900…1200 °C and the time of the
alloy-steam interaction up to 900 s for the case of the
availability of a hydrogen additive in steam,
deformation of cladding and irradiation.
RESULTS ON EXPERIMENTALLY
VALIDATED EMBRITTLEMENT
CRITERION
The embrittlement criterion pertaining to Zr1%Nb
fuel claddings is validated by the results of the
following experiments [3–7]:
research of fuel cladding ability to withstand
quenching (thermal shock tests);
estimation of the physico-mechanical state of
oxidized specimens after flooding with cool water or
after quick removing into cool water (estimation of
impact elasticity and residual ductility), metallographic
examinations of oxidized specimens subjected to a
thermal shock.
Thermal shock tests
Requirements for the procedure used to test for
thermal shock (Table 1) comprise the following:
– indirect heating of specimens in well
thermostatically-controlled facility;
82 ISSN 1562-6016. ВАНТ. 2021. №2(132)
– isothermal exposure; time and temperature being
recorded;
– quick removing of a specimen into cool water (or
the flooding of a specimen from bottom);
– analysis of a tested simulator state;
– formation of simulator failure map.
Table 1
The parameters of the thermal shock tests
Parameters
Fac il i ty
UNOPRO TEFSAI UVS
Simulator cladding temperature,
о
С 900…1200 900…1300 1000…1200
Simulator heating rate, deg/s
var
10…20
1…3
(from 800 °С)
Steam pressure, MPa 0.1 0.1 0.1
Steam specific transfer, mg/cm
2
/s 7 50 2 g/min
Flow rate of carrier gas (argon), cm
3
/min – – 140 6
Temperature of flooded water (immersion),
о
С 20 20 25…35
Rate of flooding from bottom, m/s
(immersion), s
0.2
0.2
0.5…0.8
Cooling rate, deg/s ~ 100 ~ 100 ~ 100
In the experiments carried out with short-length
simulators (Table 2), account was taken of many
loading factors involved in LOCA. The use of a short-
length simulator ensured a uniform in height and time
controlled temperature field. The availability of UO2
pellets (or sintered Al2O3) gave essentially real values of
temperature effected stresses arising in a cladding upon
flooding (quick immersion) with cool water. The tests
were implemented using both integer fuel rod
simulators internally pressurized with an inert gas
(deformation of cladding was simulated) and those
having unsealed ends (not pressurized).
Table 2
Сomponents of types 1–4 simulators
Component
Facility
UNOPRO TEFSAI UVS
Cladding Material Zr1%Nb Zr1%Nb Zr1%Nb
Length 60 mm 120 mm 60 mm
Pellet Material Al2O3 UO2 UO2
Shape Standard – WWER type
Plug Material Zr1%Nb – Zr1%Nb
The experiments with unirradiated claddings were
carried out in two facilities:
with specimen heating in a shaft-type furnace;
the weight gain of a specimen being continuously
recorded during of the experiment – UNOPRO;
with heating simulator cladding placed in the
central hole of UO2 pellets – TEFSAI.
The experiments with irradiated claddings were
carried out in all-purpose rigs UVS [3].
The experiments to study the thermal resistance
were carried out in a wide temperature range
900…1200 °C; the cladding oxidation degree was
different (from 18 to 60% ECR).
The data were acquired on the heat resistance of
WWER type fuel claddings operated to reach the burn-
up of ~ 50 (MW day)/kg U.
Simulators having unirradiated (types 1–3) and
irradiated (type 4) claddings used in tests for heat
resistance are schematically depicted in Fig. 1.
The results of the thermal shock tests are presented
as maps illustrating the ability of Zr1%Nb claddings to
resist thermal shock – failure maps. The results of the
thermal shock tests are plotted on the failure map as
temperature of oxidation vs oxidation time.
Table 3
The boundaries of the allowable ranges of cladding state
Maximally allowable temperature
lim
cladТ = 1200 °С Vertical straight line
Maximally allowable degree of oxidation lim
cladECR = 18%
(∆m is calculated from ratio (2))
Inclined straight line
The results of the tests for thermal shock evidence
that the claddings of the types 1–4 simulators fractured
outside the range of the allowable cladding state (Figs.
2–5).
The embrittlement criterion “1200 °C – 18% ECR”
(Тable 3) for Zr1%Nb claddings of WWER type fuel
rods was corroborated experimentally.
ISSN 1562-6016. ВАНТ. 2021. №2(132) 83
Type 1 Type 4
Type 3 (simulation of cladding deformation)
Fig. 1. Schemes of simulators: 1 – cladding; 2 – pellet (AL2O3 – types 1, 3; UO2 – type 2); 3 – tungsten heater
Fig. 2. Cladding failure map.
Simulator of type 1
Fig. 3. Cladding failure map.
Simulator of type 2
1
2
3
4
2
3
5
BL
TL
Lower
sample
BU
TU
Upper
sample
1 - simulator head
2 - UO pellets
3 - cladding
4 - adaptor
5 - pellets holder
2
Type 2
84 ISSN 1562-6016. ВАНТ. 2021. №2(132)
Fig. 4. Cladding failure map.
Simulator of type 3
Fig. 5. Cladding failure map.
Simulator of type 4
ESTIMATION OF MECHANICAL
PROPERTIES OF OXIDIZED CLADDINGS
SUBJECTED TO THERMAL SHOCK
The mechanical properties of oxidized claddings of
cold water quenched fuel rod simulators were estimated
from the results of impact testing (estimation of impact
elasticity) and compression testing in the direction
normal to the specimen symmetry axis (estimation of
residual ductility).
Compression tests
To assess the mechanical properties of oxidized
Zr1%Nb claddings subjected to a thermal shock, use
was made of the results of compression testing 30 or
50 mm long pieces cut off from pre-oxidized WWER
tubes ( 9.15 x7.72 mm). The compression tests were
carried out at room and elevated temperatures in the
Instron-TT-DM machine and in a high temperature
vacuum facility 1246P – 2/2300, respectively. The grip
displacement rate is 2 mm/min. The strain temperature
is 20…900 °C.
The results of the compression tests revealed a clear-
cut distinction between the specimens having a ductility
margin (complete ductility, low ductility) and those that
fractured in a brittle mode (Fig. 6).
Fig. 6. Compression tests (Ttest = 20 °C)
A strong dependence of the “ductility boundary” on
a testing temperature and a specimen weight gain can be
seen [7].
The results of the compression testing showed that
the ductility characteristics of Zr1%Nb and Zry-4 alloys
differed at a room temperature [8]. This difference
reduces with the rise of the compression testing
temperature.
Impact tests
In our view, the estimation of the oxidized Zr1%Nb
cladding elasticity from the results of impact tests is
most unbiassed.
In the experiments assessing the impact elasticity of
cold water quenched oxidized claddings, original
specimens were used; they were prepared from
claddings of fuel rod type 2 simulators that remained
integer after testing for thermal shock.
A notch 0.5 mm wide and 1.0…1.5 mm deep was
cut in a simulator cladding 100 mm long using a
diamond disc. The impact elasticity tests were carried
out using a pendulum hammer of the PSV-1.5 type:
Testing temperature…………………..20 °С;
Maximum impact energy…………….~ 15 J;
Impact speed………………………….~ 3…4 m/s.
The impact strength was calculated via the formula
ak = W / F,
where W – is fracture strength, J; F – is a specimen
cross area at a point of shock load application, cm
2
.
The impact elasticity of Zr1%Nb specimens in the
original state equals 64.2…89.3 J/cm
2
. This value is a
factor of ~ 2 higher than that of Zry-4F specimens.
The tough fracture mode shown by Zr1%Nb
claddings oxidized at 900…1300 °C takes place upon
oxidation to 5% ECR in impact tests (Fig. 7).
At a higher oxidation degree, claddings fracture in a
mixed mode while at the oxidation degree more than
7% ECR, the fracture of claddings is brittle.
As far as Zry-4F claddings, the critical oxidation
degree at the tough to brittle fracture transition is not
less than 7% ECR. The same procedure was used to test
the specimens of both the alloys.
According to [9], the critical impact elasticity of
Zry-4 claddings was 1.25 J/cm
2
. If this value is
normalized to the cross section area of the transformed
-Zry it will be ~ 1 J/cm
2
. Fig. 8 indicates the impact
elasticity range within which Zr1%Nb cladding does not
fracture when cold quenched (subjected to thermal
shock). At the degree of Zr1%Nb cladding oxidation to
18% ECR, the impact strength remains not lower than
1 J/cm
2
.
0 5 10 15 20 25
Weight gain, mg/cm
600
700
800
900
1000
1100
1200
1300
T
e
m
p
e
ra
tu
re
o
f
o
x
id
a
ti
o
n
,
Cо
2
Zr1%Nb alloy
presence of ductility ( > 4%)
low ductility ( < 4%)
brittle
e
D
e
D
.
ISSN 1562-6016. ВАНТ. 2021. №2(132) 85
Fig. 7. Impact elasticity of oxidized claddings vs
oxidation degree
Metallographic examinations
Examinations of the cladding microstructure allowed
determination of the thickness of Zr1%Nb – steam
interaction layers (ZrO2, -Zr(O), -Zr) formed as a
result of the oxidation. The estimation of the degree of
the cladding oxidation based on the results of the
metallographic measurements of interaction layer
thicknesses is an indirect one.
The knowledge of the oxygen distribution within the
interaction layer thickness gives an idea of the influence
produced by absorbed oxygen on the ductile (plastic)
properties of claddings. The indirect estimation of the
oxygen content and distribution within interaction layers
was carried out based on the results of measuring the
microhardness of layers.
The microhardness of claddings was measured with
PMT-3 microhardness tester at the load of 50 g. The
error of the measurements was ± 8 kg/mm
2
.
Metallographic examinations revealed that the
(metal + -Zr(O)) layer thickness averaged hardness
H50 of the claddings of types 1–4 simulators oxidized at
temperatures 1050 °C to 43% ECR amounts up to
600 kg/mm
2
at the cladding thickness up to two times
reduced compared to the original thickness (Fig. 8).
The microhardness of the metal part of the Zr1%Nb
claddings oxidized at 1050 °C increases with the
oxidation degree.
Fig. 8. Variation in (metal + -Zr(O)) layer thickness
averaged microhardness of Zr1%Nb claddings vs
oxidation temperature
The microhardness of claddings oxidized at
1100…1200 °C is weakly dependent upon the oxidation
degree (see Fig. 8).
It is to be noted that as far as Zr1%Nb alloy, the rate
of the -phase microhardness increase is higher
compared to Zry-4 (particularly, at the early stages of
oxidation). This difference may result from different
ultimate solubilities of oxygen in the -phase of the two
alloys at identical temperatures of their oxidation.
CONCLUSIONS
The analysis of the data experimentally obtained by
research the behaviour and properties of WWER type
fuel claddings from Zr1%Nb alloy under loading
conditions simulating the stage of core flooding with
water in LOCA suffices to shows the character and
numerical value of criterional parameters of the
embrittlement criterion in terms of the cladding stability
upon flooding and subsequent implementation of FA
unloading and transportation.
The representative maximal design limit of fuel rod
damage in terms of oxidation (embrittlement criterion)
comprises all together the maximum temperature of the
fuel cladding heating and the local depth of cladding
oxidation.
The numerical values of the criterional parameters
lim
cladТ =1200 °С and lim
cladECR =18% are validated by the
experimental data obtained from studies into the
kinetics of the Zr-steam reaction and from the specially
implemented thermal shock experiments.
It is demonstrated that the mechanical properties of
oxidized claddings upon a thermal shock (impact
elasticity, residual ductility) are adequate for claddings
to be stable upon flooding and subsequent handling FA
(unloading and transportation).
REFERENCES
1. Obshchiye polozheniya obespecheniya
bezopasnosti atomnykh stantsiy OPB-88/97. NP-001-97
(PNAE G-01-011-97): Utverzhdeny postanovleniyem
Gosatomnadzora Rossii №9 ot 14.11.97 (in Russian).
2. Pravila yadernoy bezopasnosti reaktornykh
ustanovok atomnykh stantsiy. PBYA PU AS-89.
Vvedeny v deystviye s 01.09.90 (in Russian).
3. Yu.K. Bibilashvili, N.B. Sokolov, L.N. Andreeva-
Andrievskaya, V.Yu. Tonkov, A.V. Salatov, A.M. Mo-
rosov, V.P. Smirnov. The oxydized Zr1%Nb WWER-
type fuel rod claddings heat resistance during quenching
in loss of the coolant accident conditions // 6
th
International QUENCH Workshop, Forschungszentrum
Karlsruhe, October 10-12, 2000.
4. Yu.K. Bibilashvili, N.B. Sokolov, V.V. Dranenko,
et al. Vliyaniye vysokotemperaturnogo okisleniya i
teplovykh udarov na deformatsiyu do razrusheniya
obolochek tvelov iz splavov na osnove tsirkoniya //
Voprosy atomnoy nauki i tekhniki. Seriya
“Materialovedeniye i novyye materialy”. 1991, N 2(42),
p. 34-39 (in Russian).
5. Yu.K. Bibilashvily, N.B. Sokolov, L.N. Andreye-
va-Andrievskaya, A.V. Salatov, A.M. Morozov. High-
temperature Interaction of Fuel Rod Cladding Material
(Zr1%Nb alloy) with Oxygen-containing Mediums //
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CLADDING
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18% ECR
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30% ECR
30% ECR
48% ECR
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86 ISSN 1562-6016. ВАНТ. 2021. №2(132)
Proceedings of IAEA Technical Committee on
Behaviour of LWR Core Materials under Accident
Conditions, held in Dimitrovgrad, Russia, on 9–13
October, 1995. IAEA-TECDOC-921, Vienna, 1996,
p. 117-128.
6. V.Yu. Tonkov, A.M. Kaptel'tsev, I.V. Golikov, et
al. Vliyaniye okisleniya v vodyanom pare na
mekhanicheskiye svoystva splava Zr1%Nb v
temperaturnom intervale 20-1000 °C // Voprosy
atomnoy nauki i tekhniki. Seriya “Materialovedeniye i
novyye materialy”. 1990, N 4 (38), p. 7-11 (in Russian).
7. Yu. Bibilashvily, N. Sokolov, L. Andreeva-
Andrievskaya, Yu. Vlasov, O. Nechaeva, A. Salatov.
Assesment of WWER Fuel Condition in Design Basis
Accident // Proceedings of an International Seminar,
held in St. Constantine, Varna, Bulgaria, on 7–11
November, 1994.
8. S. Kawasaki. A review of studies on behaviour of
fuel cladding under under LOCA
s
// Proceedings of
Japan-USSR seminar on LWR Fuels, held in Tokyo,
Japan, 29–31 October, 1990.
9. H.M. Chang, A.M. Garde, T.F. Kassner.
Development of oxygen embrittlement criterion for
zircaloy cladding applicable to loss-of-coolant accident
conditions in light-water reactors // Zirconium in the
Nuclear Industry (Fourth Conference), ASTM STP 681,
ASTM, 1979, p. 600-627.
Статья поступила в редакцию 08.02.2021 г.
ИССЛЕДОВАНИЕ ПОВЕДЕНИЯ И СВОЙСТВ ОБОЛОЧЕК ТВЭЛОВ РЕАКТОРА ТИПА
ВВЭР ИЗ СПЛАВА Zr1%Nb В УСЛОВИЯХ АВАРИИ С ПОТЕРЕЙ ТЕПЛОНОСИТЕЛЯ
М.М. Семерак, С.С. Лыс
Приведены критерии безопасности, предъявляемые к состоянию твэлов в условиях проектных аварий с
потерей теплоносителя для реакторных установок с ВВЭР. Представлен краткий обзор экспериментальных
исследований, проведенных с целью обоснования критериев безопасности. Объем экспериментальных
данных, полученных при исследовании поведения и свойств оболочек твэлов реактора типа ВВЭР из сплава
Zr1%Nb в условиях нагружения, имитирующих стадию залива активной зоны водой при аварии с потерей
теплоносителя, достаточен для суждения о характере и численном значении критериальных параметров
охрупчивания с точки зрения стойкости оболочек при заливе и извлечении ТВС и транспортировке.
Рассматриваются аварии с потерей теплоносителя первого контура, для которых характерно ухудшение
условий теплоотвода от твэлов. В процессе аварий допустима разгерметизация оболочки твэла, однако при
этом должны сохраняться возможности охлаждения твэла с измененной геометрией и разборки (выгрузки)
активной зоны после аварии.
ДОСЛІДЖЕННЯ ПОВЕДІНКИ І ВЛАСТИВОСТЕЙ ОБОЛОНОК ТВЕЛІВ РЕАКТОРА
ТИПУ ВВЕР ЗІ СПЛАВУ Zr1%Nb В УМОВАХ АВАРІЇ З ВТРАТОЮ ТЕПЛОНОСІЯ
М.М. Семерак, С.С. Лис
Наведено критерії безпеки, що пред'являються до стану твелів в умовах проектних аварій з втратою
теплоносія для реакторних установок з ВВЕР. Представлений короткий огляд експериментальних
досліджень, проведених з метою обґрунтування критеріїв безпеки. Обсяг експериментальних даних,
отриманих при дослідженні поведінки і властивостей оболонок твелів реактора типу ВВЕР зі сплаву
Zr1%Nb в умовах навантаження, імітуючих стадію затоплення активної зони водою при аварії з втратою
теплоносія, достатній для судження про характер і чисельне значення критеріальних параметрів
окрихчування з точки зору стійкості оболонок під час заливу і видаленні ТВЗ та транспортуванні.
Розглядаються аварії з втратою теплоносія першого контуру, для яких характерне погіршення умов
тепловідведення від твелів. У процесі аварій допустима розгерметизація оболонки твела, однак при цьому
повинна зберігатися можливість охолодження твела зі зміненою геометрією і можливість розбирання
(вивантаження) активної зони після аварії.
|
| id | nasplib_isofts_kiev_ua-123456789-194896 |
| institution | Digital Library of Periodicals of National Academy of Sciences of Ukraine |
| issn | 1562-6016 |
| language | English |
| last_indexed | 2025-12-07T15:51:07Z |
| publishDate | 2021 |
| publisher | Національний науковий центр «Харківський фізико-технічний інститут» НАН України |
| record_format | dspace |
| spelling | Semerak, M.M. Lys, S.S. 2023-12-01T13:18:05Z 2023-12-01T13:18:05Z 2021 Research the behaviour and properties of WWER type fuel claddings from Zr1%Nb alloy in loss of the coolant accident / M.M. Semerak, S.S. Lys // Problems of Atomic Science and Technology. — 2021. — № 2. — С. 80-86. — Бібліогр.: 9 назв. — англ. 1562-6016 DOI: https://doi.org/10.46813/2021-132-080 https://nasplib.isofts.kiev.ua/handle/123456789/194896 621.039.548 621.039.586 The paper presents safety criteria placed on fuel rod condition in loss of the coolant accident (LOCA) conditions as applied to reactor plants with WWER. The paper reveals briefly experimental studies carried out to validate safety criteria (acceptance criteria). The scope of the data experimentally obtained by research the behaviour and properties of WWER type fuel claddings from Zr1%Nb alloy under loading conditions simulating the stage of core flooding with water in LOCA suffices to judge the character and numerical value of criterional parameters of the embrittlement criterion in terms of the cladding stability upon flooding and subsequent implementation of fuel assembly (FA) unloading and transportation. Accidents are considered involving loss of coolant by primary circuit which are characterized by conditions of degraded heat transfer from fuels. During accidents loss of tightness by fuel rod cladding is tolerable, however, in this case, the cooling of a distorted fuel rod and the dismantling (unloading) of the core after an accident have to be feasible. Наведено критерії безпеки, що пред’являються до стану твелів в умовах проектних аварій з втратою теплоносія для реакторних установок з ВВЕР. Представлений короткий огляд експериментальних досліджень, проведених з метою обґрунтування критеріїв безпеки. Обсяг експериментальних даних, отриманих при дослідженні поведінки і властивостей оболонок твелів реактора типу ВВЕР зі сплаву Zr1%Nb в умовах навантаження, імітуючих стадію затоплення активної зони водою при аварії з втратою теплоносія, достатній для судження про характер і чисельне значення критеріальних параметрів окрихчування з точки зору стійкості оболонок під час заливу і видаленні ТВЗ та транспортуванні. Розглядаються аварії з втратою теплоносія першого контуру, для яких характерне погіршення умов тепловідведення від твелів. У процесі аварій допустима розгерметизація оболонки твела, однак при цьому повинна зберігатися можливість охолодження твела зі зміненою геометрією і можливість розбирання (вивантаження) активної зони після аварії. Приведены критерии безопасности, предъявляемые к состоянию твэлов в условиях проектных аварий с потерей теплоносителя для реакторных установок с ВВЭР. Представлен краткий обзор экспериментальных исследований, проведенных с целью обоснования критериев безопасности. Объем экспериментальных данных, полученных при исследовании поведения и свойств оболочек твэлов реактора типа ВВЭР из сплава Zr1%Nb в условиях нагружения, имитирующих стадию залива активной зоны водой при аварии с потерей теплоносителя, достаточен для суждения о характере и численном значении критериальных параметров охрупчивания с точки зрения стойкости оболочек при заливе и извлечении ТВС и транспортировке. Рассматриваются аварии с потерей теплоносителя первого контура, для которых характерно ухудшение условий теплоотвода от твэлов. В процессе аварий допустима разгерметизация оболочки твэла, однако при этом должны сохраняться возможности охлаждения твэла с измененной геометрией и разборки (выгрузки) активной зоны после аварии. en Національний науковий центр «Харківський фізико-технічний інститут» НАН України Вопросы атомной науки и техники Thermal and fast reactor materials Research the behaviour and properties of WWER type fuel claddings from Zr1%Nb alloy in loss of the coolant accident Дослідження поведінки і властивостей оболонок твелів реактора типу ВВЕР зі сплаву Zr1%Nb в умовах аварії з втратою теплоносія Исследование поведения и свойств оболочек твэлов реактора типа ВВЭР из сплава Zr1%Nb в условиях аварии с потерей теплоносителя Article published earlier |
| spellingShingle | Research the behaviour and properties of WWER type fuel claddings from Zr1%Nb alloy in loss of the coolant accident Semerak, M.M. Lys, S.S. Thermal and fast reactor materials |
| title | Research the behaviour and properties of WWER type fuel claddings from Zr1%Nb alloy in loss of the coolant accident |
| title_alt | Дослідження поведінки і властивостей оболонок твелів реактора типу ВВЕР зі сплаву Zr1%Nb в умовах аварії з втратою теплоносія Исследование поведения и свойств оболочек твэлов реактора типа ВВЭР из сплава Zr1%Nb в условиях аварии с потерей теплоносителя |
| title_full | Research the behaviour and properties of WWER type fuel claddings from Zr1%Nb alloy in loss of the coolant accident |
| title_fullStr | Research the behaviour and properties of WWER type fuel claddings from Zr1%Nb alloy in loss of the coolant accident |
| title_full_unstemmed | Research the behaviour and properties of WWER type fuel claddings from Zr1%Nb alloy in loss of the coolant accident |
| title_short | Research the behaviour and properties of WWER type fuel claddings from Zr1%Nb alloy in loss of the coolant accident |
| title_sort | research the behaviour and properties of wwer type fuel claddings from zr1%nb alloy in loss of the coolant accident |
| topic | Thermal and fast reactor materials |
| topic_facet | Thermal and fast reactor materials |
| url | https://nasplib.isofts.kiev.ua/handle/123456789/194896 |
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