Nuclear Engineering and Design 238 (2008) 1026–1061
The multiple pressure tube rupture (MPTR) issuein RBMK safety technology
F. D’Auria a,∗, B. Gabaraev b, O. Novoselsky b, V. Radkevich b, V.N. Filinov b, D. Mazzini a,F. Moretti a, F. Pierro a, A. Vigni a, L. Parafilo b, D. Kryuchkov c
a DIMNP, University of Pisa, Pisa, Italyb NIKIET, Moscow, Russia
c PhEI, Obninsk, Russia
Received 22 September 2006; received in revised form 24 February 2007; accepted 1 March 2007
bstract
The RBMK core is constituted by more than one-thousand pressurized channels housed into stacked graphite blocks and connected at the bottomnd at the top by small diameter (D) and long length (L) pipes (less than 0.01 and more than 10 m, respectively) that end up into headers andrum separators. Control valves are installed in the bottom lines. Due to the large L/D value and to the presence of valves and other geometriciscontinuities along the lines connecting with the pressure channels, the Fuel Channel Blockage (FCB) event is possible and already occurred inwo documented NPP events. Pressure tube rupture occurred in a third NPP event not originated by FCB. Previous investigations, have shown theelevance of these events for the safety technology, and the availability of proper computational technique for the analysis, see the first and thehird companion paper in this journal issue, respectively.
The occurrence of the FCB event remains undetected for a few tens of seconds because of the lack of full monitoring for the individual channels,ourth companion paper in this journal issue. Therefore, fission power continues to be produced in the absence of cooling. This brings in subsequentimes to fuel rod overheating, pressure tube failure, damage of the neighbouring graphite brick and ejection of damaged fuel.
Following the pressure tube rupture, reactor cavity pressurization, radioactivity release into the same area and change of fluid properties occurhat allow the detection of the event and cause the reactor scram at a time of a few tens of seconds depending upon the channel working conditionsnd the severity of the blockage.
Notwithstanding the [delayed] scram and the full capability of the reactor designed safety features to keep cooled the core, the multiple pressureube rupture (MPTR) issue is raised. The question to be answered is whether the ‘explosion’ of the blocked pressure tube damages not only theeighbour graphite bricks but propagates to other channels causing the potential for several channel failure.
In order to address the MPTR issue fuel channel thermal-hydraulics and three-dimensional (3D) neutron kinetics analyses have been performed,
s well structural mechanics calculations for the graphite bricks and rings (graphite rings surround the pressure tube to accommodate for thermalnd radiation induced expansions).
The bases for the analysis and the results of the study are presented. The conclusion, not reported within a licensing based format, is that thePTR consequences are not expected to be relevant for the safety of the RBMK installations. This is supported by the analysis of experiments
ear M
an
erformed at the TKR facility available at the EREC research Centre n2007 Elsevier B.V. All rights reserved.
. Introduction
The RBMK (Reactor Bolshoy Moshchnosty Kipyashiy) is aoiling light water cooled, graphite moderated thermal reactor.lightly enriched uranium fuel is adopted for fuel rods that are
∗ Corresponding author.E-mail address: [email protected] (F. D’Auria).
ssembled in groups of 18 to constitute pressurized Fuel Chan-els (FC). A zirconium–niobium tube envelopes the channelo sustain the coolant pressure and is embedded into square-ross-section graphite blocks. More than 1600 graphite stacksith embedded fuel channel constitute the core that is bounded
y a steel tank enclosed into a pressure resistant reactor cav-ty.
Established fundamental principles, already valid in the 50s,re at the basis of the design of the reactor system that nowadays,
(a) overheated fuel is ejected from the channel, and
F. D’Auria et al. / Nuclear Enginee
ollowing an experience of around 360 reactor-years, shows suit-ble operational and safety records with the noticeable exceptionf the Chernobyl unit 4 event in 1986.
The detailed knowledge of the RBMK system configurationas not spread in the Western world till the 1986 event. After-ards, “information batches” of RBMK technology became
vailable and were unavoidably evaluated in the light of thehernobyl event. This caused a search for and a characteriza-
ion of inadequacies not counterbalanced by the identificationf the acceptable safety features, ending-up in an overall neg-tive judgement from the reactor safety viewpoint. The lack orhe inadequacy of a comprehensive safety related documenta-ion from the Soviet Union, also connected with the uses ofhe reactor, contributed to this judgement. Furthermore, geo-
etric and material features of the reactor and primarily ofhe core, were not consistent with capabilities or with the val-dation domain of computational tools adopted in the Westernorld to assess the fulfilment of standard safety requirements,
ctually preventing a sound and (Russian) independent evalua-ion.
The results of recently completed project sponsored byuropean Commission (EC), with the participation of RBMKesigners in Russia and the supervision of the national utilitynd the regulatory authority (D’Auria et al., 2005), allow to given idea of RBMK current safety characteristics.
The project has been made possible owing to the avail-bility of sophisticate computational tools developed andualified in the last decade. These include powerful com-uters, advanced numerical solution methods, techniques foreveloping input decks and for proving the qualificationevel.
The general subject of the project is the deterministic accidentnalysis where emphasis is given to the phenomena occurringuring the expected transient scenarios rather than to the rigoreeded within a nuclear reactor licensing process. Followinghe identification and the characterization of bounding scenariosssuming to envelope all accident conditions relevant to RBMKafety technology, two main chains of codes have been set-upnd utilized to perform safety analyses.
The main achievements from the project are criticallyeviewed in the set of six companion papers including theresent one and are supported by recent literature documents,.g., Sorokin et al. (2006) and Uspuras and Kaliatka (2006).he objectives of the series of six papers can be summarized as
ollows:
1) To present numerical techniques and computational tools,including qualification levels and results from the applica-tions, suitable for deterministic safety analysis of RBMK.
2) To demonstrate the results of computational analyses, whichallow making conclusions about the current safety charac-teristics of the plants with RBMK reactors.
The former objective is primarily pursued in the present papernd in the papers by D’Auria et al. (2008b,c,d, e) (see also’Auria et al., 2005) that constitute the support for the con-
lusions that are derived in the first paper of the series, e.g.,
(
nd Design 238 (2008) 1026–1061 1027
’Auria et al. (2008a). The content of those papers can be sum-arized as follows with ‘qualification of computational tools’
onstituting a common issue:
overall perspective and status for deterministic accident anal-ysis in RBMK (D’Auria et al., 2008a),thermal-hydraulic performance of the primary system ofRBMK following selected accidents (D’Auria et al., 2008b),thermal-hydraulic performance of confinement system ofRBMK, following selected accidents (D’Auria et al., 2008c),the use and the relevance of 3D neutron kinetics coupled withthermal-hydraulics in RBMK accident analysis (D’Auria etal., 2008d),addressing the multiple pressure tube rupture (MPTR) issue,present paper,the proposal for the individual channel monitoring (ICM) sys-tem to prevent pressure tube rupture following FC blockage(D’Auria et al., 2008e).
The latter objective is pursued primarily in the paper by’Auria et al. (2008a) that makes use of results documented
n the remaining five companion papers. The background andhe rationale for achieving the selected objectives are also partf that paper. This also includes an arbitrarily defined list ofopics derived from a spot-based investigation within the safetyomain of water cooled reactors including RBMK. It must alsoe premised that well established Probabilistic Safety Assess-ent (PSA) results have been used, but no investigation has
een carried out to demonstrate the validity or the quality ofhose results.
Data, analyses and conclusions in the six companion papersre related to the current configuration of the Smolensk-3PP (some reported analyses also relate to Ignalina-2 NPP)
nd no effort is made to provide any evaluation of safety forBMK where the innovation or modernisation feedbacks for
he Smolensk-3 plant are not applicable.The configuration of the RBMK core with more than 1600
ressurized Fuel Channels (FC) make the designers aware of theroblem connected with the rupture of one single tube. This mayappen owing to a technological defect (even though the proba-ility of such event could be negligible as discussed by D’Auriat al., 2008a), or due to the coolant blockage at the channelnlet, or due to power excursion (this event has been documentedwo times in existing reactors originated by coolant blockage,’Auria et al., 2005, see also below). The consequences of thereak of a single FC are handled by the overall protection sys-em of the reactor with radiological consequences within thecceptance limits prescribed by the Regulatory Authority (see’Auria et al., 2008a,b,c,d). However, in such conditions, i.e.
fter the FC break,
b) the FC neighbouring the broken channel are loaded byhydraulic forces to an extent that can cause the failure thustriggering, through a domino type effect, the potential catas-trophic failure of the core.
1 ring and Design 238 (2008) 1026–1061
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The last item, commonly known with the terms multipleressure tube rupture (MPTR), constitutes the main topic ofhis paper. The RBMK designers were aware of the issue andnalized the core design to prevent the MPTR occurrence. Theecently available safety technology software and recent exper-ments, e.g., Medvedeva et al. (2004a), made it possible toubstantiate the safety margins of the RBMK in relation to the
PTR according to the modern exigencies in the domain ofuclear reactor safety. The issue has been thoroughly studied inhe past, e.g., NIKIET (1983, 1998) and constituted the subjectf international conferences, e.g., Simonov et al. (1994), and ofrojects supported by the European Commission, e.g., Sureaut al. (1996) and TACIS (1996).
Therefore, the objective of the present paper is two-fold:n the one hand to present the capabilities of computationalools and to outline the results of experiments dealing with the
PTR, on the other hand to show results from the applicationf those tools that bring (preliminary results have been obtainedithin the present framework that need further qualification) to
he exclusion of the possibility of MPTR in currently operatingBMK NPP.
. The boundary conditions for the MPTR issue
The overall framework for the study can be found in D’Auriat al. (2008a), with more details given in D’Auria et al. (2005). Inhe present paper the attention is focused toward the accident sce-ario originated by the fuel channel blockage (FC-BLOCKAGEy D’Auria et al., 2008b,d) making reference to boundary con-itions in the Smolensk-3 NPP unit.
The key elements for achieving the objectives assigned forhe paper can be found in chapters 4–6. Hereafter, chapters 2 and, the background for addressing the multidisciplinary problemrising from the FC-BLOCKAGE and the MPTR issues is intro-uced. This includes the presentation of following aspects: (a)ketches of components and zones of the RBMK core region toake clear the concerned accident scenario, (b) the characteriza-
ion of the steady state operation of the reference RBMK boilinghannel, (c) the experience from the pressure tube (PT) rupturevents in RBMK NPP, (d) the phenomenological evolution ofhe transient, (e) the differences between the FC-BLOCKAGEnd the FC-LOCA scenarios and (f) the licensing environment.
.1. Elements of the RBMK core layout relevant to thePTR
The overall RBMK system, the primary loop, the confinementncluding the reactor cavity and the core region are described inhe companion papers by D’Auria et al. (2008a,b,c,d), respec-ively. More details can be found in taken from the referenceslmenas et al. (1998), Uspuras and Kaliatka (2006) and D’Auria
t al. (2005). A few elements are reported below that are relevantor the present study.
.1.1. The overall core configurationReferring to the Smolensk-3 plant data the reactor core is
omposed by 2488 graphite columns or stacks, Fig. 1, of which:
•
ig. 1. Radial cross-section of RBMK core (Smolensk-3 NPP). ‘White’ (ormpty) squares are the (1570) fuel channels.
a) 1570 fuel channel columns, Fig. 2, (b) 314 non-fuel channelsolumns (211 are control rods channels, part of the Control androtection System, CPS, regularly distributed over the core lat-
ice) and (c) 604 radial reflector channels. The following shoulde noted (relevant aspects for the present context):
‘Hot’ and ‘cold’ graphite stacks are part of the core. A typicalcell consisting of one CPS channel and of five fuel channels isillustrated in Fig. 3 (top). In the same figure (bottom) typicaltemperature values in one fuel channel ‘graphite cell’ duringnominal core operation are reported (Parafilo et al., 2000).In all channels thermal power is generated in the graphitedue to the neutron moderation process and is transferred tothe central cooled channel (either FC or CPS channel, thickarrows in Fig. 3). Furthermore, graphite blocks envelopingFC are warmer than graphite blocks enveloping CPS channels,thus heat transfer occurs across gaps (narrow arrows in Fig. 3)as discussed by Uspuras and Kaliatka (2004). Coolant in theCPS is kept below the boiling point by suitable circulationflow.Temperature of graphite is not only a function of the radialcoordinates (above bullet) but also of the axial coordinate withlower average temperature at the bottom and at the top of thecore (see results of steady-state analyses, for instance Fig. 18cin D’Auria et al., 2008b). Therefore, axial heat transfer occursbetween ‘piled-up’ graphite blocks to a lower extent than
radial heat transfer owing to smaller temperature differences.The thickness of the gaps between adjacent graphite blocks,Fig. 3, depends upon temperature differences, not uniform allover the core, and upon neutron fluence: irradiation causes
F. D’Auria et al. / Nuclear Engineering and Design 238 (2008) 1026–1061 1029
Fig. 2. Fuel channel and f
Fig. 3. Graphite stacks in RBMK: (a) ‘module’ including five fuel channels andone CPS channel (red circle); arrows indicate the versus of the heat transfer;(b) typical graphite temperatures during nominal operation. (For interpretationof the references to color in this figure legend, the reader is referred to the webversion of the article.)
•
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uel rod of a RBMK.
changes in the geometric dimensions of graphite blocks otherthan in material properties, e.g., IAEA (2000).Other gaps exist between pressure tubes and graphite rings andbetween graphite rings and graphite stacks (Fig. 4). The gapdimensions are largely a function of neutron fluence, pressureand temperature, e.g., operating conditions and core life.The gas, helium or nitrogen or a mixture helium–nitrogen,circulates inside the gaps among graphite stacks (see also thedevoted paragraph below). Its main roles for the safety andoperation of RBMK are: (a) detection of high radioactivity orhumidity in the reactor cavity (irrelevant for the present con-text of analysis); (b) transferring thermal power from graphiteto pressure tube; (c) transferring thermal power between adja-cent graphite columns, i.e. heat transferred by the gas removalsystem is negligible compared with the heat passing from‘hot’ stacks to ‘cold’ stacks.Mechanical, thermal and neutron related properties of mate-rials constituting the core are largely a function of core life(neutron fluence) and of operating temperature, e.g., IAEA(2000). This must be considered in the analyses.The deformation characteristic of the fuel channels and asso-ciated graphite column depend upon the type of load (uniform,punctual, etc.) and upon the location. Furthermore, other thanthe pressure tube that constitutes the most resistant part ofthe ensemble, the graphite blocks, the fuel bundles and thecentral bar contribute to the overall stiffness.
.1.2. The reactor cavity and the hydraulic connectionsPhenomena occurring in the confinement including the reac-
or cavity (RC), are discussed by D’Auria et al. (2008c). The
ketch of the RC of the Smolensk-3 NPP with main hydrauliconnections is given in Fig. 5. The connection of both the RCnd the accident localisation system (ALS) with the environmenthould be noted.
1030 F. D’Auria et al. / Nuclear Engineering and Design 238 (2008) 1026–1061
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.1.3. The ensemble of graphite columns, the envelopingank and the structural constraints
About 2500 graphite stacks constitute the core, with about500 fuel channels as described above. Each stack, composedf several blocks (one block shown in Fig. 4), is centred over ahick-walled, pressure resistant, zirconium (plus 2.5% niobium)ube. Each tube is constrained at the top and at the bottom in
complex way as given in Fig. 6a where possible constraintodelling is also shown.The ensemble of stacks is surrounded by a cylindrical steel
ank (“KZh” structure in Fig. 6b; internal diameter 14.50 m,hickness 0.016 m and height 9.75 m), that together with the topnd bottom metal structures (“E” and “OR” in Fig. 6b) formshe sealed region for the reactor cavity. In order to compensateor axial thermal expansion, the tank is provided with a bellows
ompensator. The tank is designed to resist to relatively smallressures (in the order of 0.6 MPa), and not to local loads dueo the hard contact with bending columns. Nevertheless, in thease that a peripheral column is pushed against the tank, this
ao
f the RBMK core and main geometric dimensions.
s expected to hinder further displacements of the columns andhus to play a role in mitigating the deformation of the involvedtack ensemble.
.1.4. The gas removal systemThe gas removal system has already been introduced in the
aper by D’Auria et al. (2008c) because of its role in confinementnalyses. The related sketch is given in Fig. 7. The main goal forhe system design is to prevent graphite oxidation and to con-rol pressure tubes sealed condition. Performing a best estimatenalysis for this specific scenario, and for similar ones withinhe RBMK safety technology, may imply the consideration ofhe gas removal system.
.2. The reference quantity values
A variety of data is needed to calculate the complex scenariorising as a consequence of the fuel channel blockage. An ideaf the parameters that characterize the transient performance
F. D’Auria et al. / Nuclear Engineering and Design 238 (2008) 1026–1061 1031
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Fig. 6. Structural configuration of the RBMK core: (a) pressure tube constraints;(b) mechanically resistant elements.
ig. 5. Sketch of the reactor cavity of the Smolensk-3 RBMK NPP with relatedain hydraulic connections.
f a fuel channel can be derived from Table 1. These includearameters relevant to:
thermal-hydraulics (related to primary system and confine-ment),structural mechanics,neutron kinetics.
Material properties and dependencies of some of theelected parameters upon temperature and fluence consti-ute additional sets of parameters needed for performing thenalyses.
.3. Experience from the occurred PT rupture events inBMK NPP
The individual channel flow blockage accidents happenednd have been documented in the following RBMK NPP (yearlso reported):
Chernobyl-1 (1982),Leningrad-3 (1992).
A third pressure tube rupture event occurred in Leningrad-(1975) and was initiated by local power excursion because
he additional scram system (denominated ‘LAR-LAZ’), now
nstalled in all RBMK NPP, was absent. During this event,lthough the mismatch between the flow rate and the increasedower was experienced by more than one fuel channel, only oneressure tube failed. Fig. 7. Sketch of the gas removal system from the RBMK reactor cavity.
1032 F. D’Auria et al. / Nuclear Engineering and Design 238 (2008) 1026–1061
Table 1Key quantities relevant for the analysis of the FC-BLOCKAGE scenario in RBMK
No. Item Unit Value Notes
1 Fuel generated powerMW
2.01 Typical value. Maximum allowed value is 1.6 timeslarger
2 Power coming from graphite blocks 0.123 Power removed by the gas cooling system per fuel
bundle<0.001
4 Fuel bundle active length m 7.0 See Fig. 1. Two fuel bundles are inserted into thechannel
5 No. of fuel rods per bundle – 186 Diameter of the pressure tube
m0.08
7 Thickness of pressure tube 0.0048 Edge of the square graphite block 0.25 Equivalent cylinder radius (for modelling purposes)
is 0.141 m9 FC inlet flow-rate kg/s 6.1 Average value
10 FC inlet sub-cooling K 2511 FC inlet pressure MPa 7.812 FC outlet quality/void fraction −/− 0.15/0.813 FC total pressure drop MPa 0.6514 Mass of coolant in one FC kg ∼815 Ratio flow-rate/power kg/MW ∼316 Thickness of the gap between graphite blocks m 0.0015 Affected by irradiation and by thermal expansion17 Young module of pressure tube
MPa × 103
at300 ◦C
75.4 90.4 at 80 ◦C18 Young module of central bar19 Young module of fuel bundle20 Young module of graphite stack 145.0 160.0 at 80 ◦C21 Mass of UO2 in one FC
kg
141.822 Mass of zirconium in fuel assembly 29.223 Mass of zirconium in FC wall 53.024 Producible H2 in one FC 0.95 and 1.7 For fuel bundle and tube25 Mass of graphite associated to FC 743.26 Reactivity associated with the FC voiding
$+0.3 Assuming β = 0.007. Calculation by MCNP
27 Reactivity associated with the increase of 200 K foraverage fuel temperature in the bundle
−0.44 Assuming β = 0.007. Calculation by MCNP. In caseof 300 K the value is −0.65
28 Pressure in reactor cavityMPa
0.1 Nominal operation29 Overpressure for scram in reactor cavity 0.0075 Case of Leningrad-3 NPP30 Average gas temperature in reactor cavity K 82331 Fuel enrichment % 2.0 and 2.432 Burn-up (with above enrichment) MWd/t 10000 Average value33 Axial and radial temperature distribution for the
graphiteK * *See Fig. 3 above and Fig. 18c in D’Auria et al.
(2008b)34 Beginning and end-of-life gas pressure in the fuel
pin-hot conditionMPa 1.6 and 1.9
35 Radioactivity in one FC, fuelCi
∼57000 See also Table 636 Radioactivity in one FC, gap ∼25037 Yield stress for zirconium tube
MPa380–430 Typical range given, affected by temperature and
fluence3
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8 Yield stress for graphite
The first of the above listed events, Chernobyl-1 (1982), was aonsequence of erroneous actions of the operator who, in adjust-ng the flow rates in the fuel channels, fully closed the controlnd isolation valve (CIV) of one FC. The designers responded tohis event by equipping all CIV at all power units with restrainersf the CIV stem travel in order to prevent FC flow rate reductionuring adjustment below the permissible minimum.
The scenario for the second of the above listed events,
eningrad-3 NPP, is discussed below, see NIKIET (1992) andedosov et al. (1994).
Status of the plant before the event: nominal operating con-itions, 3150 MW. The blocked channel is characterized by the
prss
6–9
oordinates 52-16, located in the left upper part of the reactorap.On 24 March 1992 at 2 h 34 min 45 s (event start), a pressure
ncrease signal from the reactor cavity activated the fast act-ng emergency protection system and the reactor was scrammedy rod insertion within 2.5 s. This time (with an error of abouts) can be considered the time of the pressure tube rup-
ure. Because of the time lag of 40 s between two recorded
oint from the flow-rate transducers the time of the flow-ate reduction occurrence is known with an error of 40 s. Auitable number of withdrawn control rods was available forcram.
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The signal for shut-down system activation is triggered whenhe cavity pressure increases by 0.075 bar. Two seconds after thecram the relative pressure in the cavity reached 0.13 bar andubsequently decreased to the value of 0.1 bar after 5 s. After-ards the pressure again started to rise with a slower rate and
eached a peak of 0.19 bar 23–25 s after the event start. Then theavity pressure steadily decreased.
The cause for the accident was the flow decrease at the inletriginated by the failure of the control valve. In particular, ahrottling device of the valve was partially destroyed and closedhe flow area. The resulting abrupt flow rate decrease led toritical heat flux and then, subsequently, to fuel, clad and pres-ure tube overheating. The loss of strength of the pressure tubeaterial caused the rupture. After the rupture flow reversal flow
rom the steam drum to the broken channel occurred causinghe cooling of the upper part of the fuel channel and the furtherressurization of the reactor cavity.
About 14 h after the event, it was attempted to remove theroken channel. Only the hanger and a 0.54 m long part of theentral supporting tube could be removed. The remaining part ofhe fuel assembly remained in the channel. Further investigationhowed that the channel had ruptured in the upper part of the corepproximately 6 m above the bottom. The graphite rings aroundhe rupture location were partially destroyed and the graphitelocks were damaged. The fuel rods were mechanically bowedn the direction of the breach.
On 28 March the extraction of the channel following the nor-al procedure began. Only the top part of the channel, with
ength of 4945 mm, was extracted. Based on the visual analysisf the broken channel the picture in Fig. 8 has been created. Theraphite block and the fuel assembly were destroyed.
However, the neighbouring fuel channels were not subject tony damages and were left in operation without any restrictions.ost accident examination (performed at Paul Scherrer Institute
n Switzerland) showed that the pressure tube rupture occurred
(
Fig. 8. Leningrad-3 RBMK NPP: sketch of the dama
nd Design 238 (2008) 1026–1061 1033
hen the tube temperature was in the range 797–847 ◦C (i.e.uite larger than the current licensing limit of 650 ◦C).
.4. The licensing environment
The licensing environment for accident analysis in RBMKas been discussed in the companion papers (D’Auria et al.,008a,b,c,d), pointing out aspects related to the overall system,o the confinement and to the fuel. Therefore the attention isocused here on the following key topics:
The consideration of the FC-BLOCKAGE within the licens-ing framework.The pressure tube rupture curve.The phenomena based list of scenarios relevant for RBMKsafety analysis.
.4.1. The FC-BLOCKAGE event and the licensingramework
From the licensing point view, the following statements applyo the accident scenario in RBMK originated by blockage in oneuel channel:
a) The probability of occurrence, if one considers only theevents documented in existing NPP, is of the order of10−2 per reactor-year. Because of this, the event should beclassified as design basis accident (DBA).
b) The event implies the severe damage of substantial amount offuel, even though this is limited to one fuel channel thereforeto an order of magnitude that is a fraction less than 1/1000of the fuel mass in the core. Because of this the event should
be classified as Beyond DBA (BDBA).
c) The event is not explicitly mentioned in the ‘list of events’proposed by the IAEA (IAEA, 2005), that reflects the levelof knowledge and agreement among specialists at the date,
ged fuel channel following the 1992 accident.
1 ring and Design 238 (2008) 1026–1061
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even though the event ‘break of a channel tube within thereactor cavity’ is a classified DBA (more details given inSection 3.2 below).
d) Following Uspuras (1999), the event ‘reduction or loss offlow in one fuel channel’ is ‘. . . one accidents initiated byequipment failure [that] should be also analysed . . .’ accord-ing to GAN (1987).
e) During the event, the DBA thresholds defined for RBMKaccident analysis, e.g., fuel temperature below 1200 ◦C,IAEA (2005) (see also D’Auria et al., 2008a,b), are over-passed.
f) The acronym Ultimate DBA (UDBA) is used for RBMKaccident analysis, e.g., Lillington et al. (1997), but not inrelation to the fuel channel blockage event.
Therefore, if the fuel channel blockage event is not consideredDBA, a contradiction with item (a) might be envisaged. If
he event is considered a DBA a contradiction raises in relationo item (e). In this case, the leakage of damaged nuclear fuelutside the pressure boundary brings an important differenceetween accident evolutions in RBMK related to other waterooled reactors as already pointed out by D’Auria et al. (2008a).
The recommendation from the present study is to considerhe flow blockage event in a RBMK fuel channel as a DBA ando offer measures capable to reduce probability of the event, fornstance see D’Auria et al. (2008e).
.4.2. The acceptability limits for pressure tubeThe pressure tube failure constitutes the most important
ccurrence during the event: on the one side it allows the detec-ion of the event, on the other side it causes the release ofadioactive material outside the pressure barrier, but also it bringso the possibility of cooling for the damaged channel and causeshe potential for the MPTR.
The failure of the pressure tube and of the associated graphiteings and graphite blocks is the result of a complex mechanical,hermo-hydraulic interaction process that is discussed into detailn chapter 3 below. Several quantities contribute to the process
aking the process itself a multidimensional problem. Com-ined criteria are strictly necessary to demonstrate the pressureube rupture in a best estimate way, as discussed by Novoselskynd Filinov (1997a,b,c) (see also D’Auria et al., 2005).
A synthesis (simplified and conservative) approach for theressure tube rupture ‘continuous’ acceptability threshold isased on the diagram in Fig. 9. The pressure tube temperaturet which the rupture is expected is reported as a function of theifferential pressure across the tube walls. Any working condi-ion below the dashed region is ‘safe’, while rupture is expectedbove the dashed region. The following should be noted:
CANDU and RBMK pressure tubes behave in a similar way.The rupture is affected by the gradient of temperature rise
and occurs ‘earlier’, i.e. at lower temperatures at low gradientcompared with the higher gradient.Graphite bounded tube have a slightly higher resistance thanbare tubes.
t3aB
ntegrity of pressure tubes of RBMK.
The licensing rupture criterion, T < 650 ◦C, is conservative,primarily at low differential pressures.
.4.3. The FC-BLOCKAGE and the phenomena based listf relevant scenarios
A phenomena based list of events has been proposed for theeterministic safety analysis of RBMK by D’Auria et al. (2005),lso reported in D’Auria et al. (2008a). This is given in Table 2,here the FC-BLOCKAGE and its key role within the contextf RBMK safety technology is underlined.
Each of the letters ‘A’ to ‘F’ in the first column in Table 2,dentifies classes of accidents characterized by bounding phys-cal phenomena suitable to assess capabilities of computationalools.
The class ‘D’, of concern within the present paper, deals withhe class ‘FC rupture and MPTR’ and related phenomena areescribed in chapter 3 below. Key differences between the twocenarios in class ‘D’ (FC-BLOCKAGE and FC-LOCA) are alsoddressed in chapter 3.
. The multidisciplinary problem associated with theC-BLOCKAGE scenario
An overview is given below of the phenomenological aspectsssociated with the scenario originated by the blockage ofne fuel channel in the RBMK NPP (i.e. FC-BLOCKAGEvent). To this aim, phenomena are identified that characterizehe progression of the event together with differences betweenhe concerned scenario and the FC-LOCA (Sections 3.1 and
.2, respectively). The failure map for RBMK pressure tubesnd the probable position for break elevation following FC-LOCKAGE are described in Section 3.3.
F. D’Auria et al. / Nuclear Engineering and Design 238 (2008) 1026–1061 1035
Table 2Phenomena based list of accident scenarios suitable for deterministic accident analysis in RBMK (D’Auria et al., 2008a)
Identification Codes adopted for achieving resultsdocumented in the present paper
Reasons for the selection
No.a Acronym explanation
A1 LOCA-PH-FIGDH: LOCA in pressureheader with failure to isolate GDH
Relap5
Largest primary system break with singlefailure. Challenging core cooling and theECCS design
A2 LOCA-SL: LOCA originated by a break insteam line
Highest depressurization rate. Challengingcore cooling and the ECCS design
A3 LOOP-ATWS: loss of on site power with theATWS condition
Challenging core cooling and the neutronkinetics model of the thermal-hydraulicsystem codes
A4 GDH-BLOCKAGE: full blockage of theGDH
Check of the capability of the ‘ECCSbypass’ to cool the core
B1 GDH-BLOCKAGE-SA: Full blockage of theGDH with the ‘Severe Accident’ assumptionof no bypass line available
Cocosys & Relap5 Challenging the venting capability of thereactor cavity (part of the confinement)
B2 LOCA-PH-FIGDH: see A1 Contain & Relap5 Challenging the ALS (part of theconfinement) structural resistance (same asA1)
B3 LOCA-SL: see A2 Contain Challenging the reactor building (part of theconfinement) venting capability (same asA2)
various phenomenological areas and codesD2b FC-LOCA: rupture of one FC Contain & Relap5 Fluent-Ansys Korsar-Rapta To assess the ballooning model in the fuel
pin mechanics areaE1 FC-BLOCKAGE: see C1 Cocosys
Mel-cor
To assess the hydrogen and the fissionproducts source term and transport (same asD1)
E2 GDH-BLOCKAGE-SA: see B1 To assess the hydrogen and the fissionproducts source term and transport in oneextreme conditions (same as B1)
F1 FC-BLOCKAGE: see C1 Relap5 To formulate the ICM proposal (same as D1)
s of aruptu
pf
3
Bdhja
(
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otsnr(
ic
a See Table 3 in D’Auria et al. (2008a). In particular, ‘A’ to ‘F’ identify classeb The class of accidents ‘D’, of concern for the present paper, deals with ‘FC
The multidisciplinary nature and the demonstration of com-lexity for the concerned scenario constitute important outcomesrom the description.
.1. Expected phenomena
The main phenomena expected following the FC-LOCKAGE event and the qualitative time succession can beerived from the diagram in Fig. 10, main arrow at the top. Otherorizontal arrows in the diagram indicate technological sub-ects relevant during the concerned time frame. The two verticalrrows indicate:
1) The potential role of the individual channel monitoring(ICM) system in preventing the progression of the accident(see D’Auria et al., 2008e).
rp
b
ccidents characterized by bounding phenomena.re and MPTR’.
2) The assessed role of the tank (see chapter 5) in contributingto limit the possibility of MPTR.
Thermal-hydraulic phenomena are relevant at the beginningf the transient, like dry-out, and counter-current flow limita-ion at the outlet of the channel that prevents water from thetem drum to enter the blocked channel. Neutron kinetics phe-omena are also relevant since the beginning because of theeactivity coefficient associated with coolant void formation andconsequent) temperature increase of the fuel.
Phenomena associated with fuel performance becomemportant because of the unavoidable high temperatures andonsequent metal water reaction, clad collapse and substantial
od deformation. Radiation heat transfer is also relevant at thisoint in time.
Break of the pressure tube and of the neighbouring graphitelocks occur by ductile and fragile mechanisms, respectively. At
1036 F. D’Auria et al. / Nuclear Engineering and Design 238 (2008) 1026–1061
xpec
(paigta
shuaccm
cfibt<
3F
esTsFu
sT
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Fig. 10. Main phenomena and qualitative time evolution e
nearly) the same time of the pressure tube rupture event, two-hase critical flow occurs at the break together with flow reversalt channel top and ejection of damaged fuel and possible chem-cal reactions between the damaged fuel and the surroundingraphite. The ejection of fuel is caused by hydraulic forces athe break caused by pressure wave propagation at first (tenths ofsecond time-scale) and to the high fluid speed later on.
Pressurization of reactor cavity occurs as well as mixing ofteam and non-condensable gas, therefore confinement thermal-ydraulics is relevant. Mechanical stresses are generated occurpon neighbouring graphite stacks causing elastic and plasticnd deformation for the associated fuel channels (potential pre-ursor phenomena for the MPTR). The rupture of pressure tubesauses heavily damaged fuel to enter the reactor cavity, thusaking relevant the area of fission products transport.A detailed list of phenomena is provided in Table 3 , where
odes suitable for the evaluation of those phenomena are identi-ed, too. The codes have been used within the project describedy D’Auria et al. (2005). In the case a quality proof is availableo the authors in relation to the concerned phenomenon, the letterQ> is added to in the fourth column of Table 3.
.2. The difference between FC-BLOCKAGE andC-LOCA
In order to better focus on selected important phenom-na, differences between FC-LOCA and FC-BLOCKAGE, e.g.,cenarios D1 and D2 in Table 2, are discussed hereafter.
he discussion should also aim at presenting the FC-LOCAcenario even though with a level of detail lower than theC-BLOCKAGE (same level of detail would also cause ann-necessary longer paper).
F
•
ted following the FC-BLOCKAGE event in RBMK NPP.
By FC-LOCA the scenario is meant here originated by theudden break of one pressure tube inside the graphite stacks.herefore, the FC-LOCA acronym is not an indication of:
A break in the pressure tube inside the reactor cavity in theregion outside the graphite blocks (i.e. inside the free spaceof the reactor cavity).A break in the pipelines connecting the pressure tube withthe Group Distribution Header (channel inlet region) of thesteam drum (channel outlet region).
Both of these set of events are relevant in the safety assess-ent of RBMK and are part of the list of DBA according to
AEA (2005) (see also D’Auria et al., 2008a). In the aboveases scram is generated early during the transient with a sit-ation of undamaged fuel and good cooling conditions cane kept, by the existing emergency safety features during thentire course of the transient (i.e. till the full recovery of thePP).The origin of a FC-LOCA is typically a defect of the pres-
ure tube in the active region. The related probability has beenstimated as negligible in the recent paper by Lee et al. (2006)see also D’Auria et al., 2008a), but it should be investigated asDBA according to the document at reference IAEA (2005).
The similarity between FC-LOCA and FC-BLOCKAGE liesn the fact that in both cases a pressure tube is broken, thusmplying the damage of the graphite block(s) close to the ruptureegion. The following differences exist between FC-LOCA and
C-BLOCKAGE:
In the case of FC-LOCA the scram occurs at a time whenthe fuel is not overheated (originated by the same signal as in
F.D’A
uriaetal./N
uclearE
ngineeringand
Design
238(2008)
1026–10611037
Table 3List of expected phenomena following the FC-BLOCKAGE event in RBMK NPP and suitable codes for the evaluation
No. Event Key phenomena Codesa Notes
1
Blockage at channel inlet
Flow-rate decrease Korsar & Relap5 <Q> The blockage may be partial or total2 Void increase in the FC3 Change in pressure drop distribution along the
FC axis4 Flooding (and possible CCFL) at FC outlet5 Neutron kinetics feedback due to loss of coolant Bars & Nestle <Q> Coolant is an absorber, see Fig. 17 in D’Auria et al.
(2008d)6
Dry-out
Rod surface temperature excursion Korsar & Relap5 <Q>7 Steam superheating generation and transport Relevant to ICM design8 Neutron kinetics feedback due to Doppler effect Bars & Nestle <Q> See Fig. 17 in D’Auria et al. (this issue-d)9 Flooding (and possible CCFL) at FC outlet Korsar & Relap5 <Q> Phenomena at rows 2 and 3 also apply
10 Change in fission power generation Bars & Nestle <Q> Owing to phenomena at rows 5 and 811
Over-passing the threshold for clad integrity
Clad collapse Rapta & Frap <Q> Loss of integrity expected for the fuel rod12 Releases of gaseous FP Refp & Melcor <Q>13 Change of bundle geometry and of cooling
characteristicsRelap5 Minor effect
14 Radiation heat transfer Relap5 Relevant during later phases, too15
Over-passing the threshold for significant H2 production
in the fuel region (e.g., candling corium oxidation,etc.)
17 Fuel bundle damage (loss of geometric integrity)18 Temperature increase in PT Korsar & Relap5 <Q>19 Creep of the PT U Stackb & Ansys <Q> Owing to the phenomenon at the row above20
Rupture of the PT
Mechanical interaction of PT with graphite ringsand with graphite blocks in the region of creep
Axial location of the creeping region dependingupon BIC like % of FC blockage
21 Flow-rate (reversal) from steam drumestablished
Korsar & Relap5 <Q>
22 Cooling of the bundle region at an elevationhigher than the break
Owing to the phenomenon at the row above
23 Simultaneous break of PT, graphite rings andblocks in the creeping region
U Stack & Ansys <Q> Ductile and fragile mechanisms for PT and graphite,respectively
24 TPCF at the break and at a top location of theaffected FC
Korsar & Relap5 <Q>
25 Hydraulic loads on thermally damaged fuelbundle
Fluent
26 Loss of damaged fuel from the break Fluent & Melcor Large radioactivity release to RC27 Chemical interaction between molten-damaged
fuel and graphite– High temperature chemical interactions not
investigated28 Continuous H2 generation Korsar & Relap5 <Q> From claddings and possibly from PT walls29
Pressurization of reactor cavity and scram occurrencePressurization of the RC Cocosys & Relap5 <Q> Because of the high pressure in the reactor cavity
30 Cooling (quench) of both halves of fuel bundlesseparated by the break
Korsar & Relap5 <Q> Liquid penetration possible owing to the stop ofpower generation. Stop of H2 production
31 Heat transfer through graphite gaps Cocosys & Relap5 Relevant during later phases, too
1038F.D
’Auria
etal./Nuclear
Engineering
andD
esign238
(2008)1026–1061
Table 3 (Continued)
No. Event Key phenomena Codesa Notes
32
Deformation of neighboring graphite stacks
Mechanical load due to displacement of thebroken blocks
Ansys & U Stack Limited by the displacement associated with thecreep deformation before rupture for the broken PT
33 Dynamic forces associated to the pressure wavepropagation
Relap5 <Q> Situation of no propagation of the rupture (noMPTR)
34 Dynamic forces associated with flashing Negligible in case of superheated steam in the FC34 Differential pressurization of the graphite gaps
on opposite faces of neighboring graphite blocksU Stack & Relap5 <Q>
36 Elasto-plastic deformation of several FC Ansys & U Stack <Q> This constitutes a constraint to the propagation ofthe PT rupture
37 Touching of fuel stacks and tank Possible. The tank may contribute to increase theconstraint to the propagation of the PT rupture
38 Transport of H2 from broken FC to graphitegaps and to RC free space
Relap5 & Melcor <Q> Risk of deflagration negligible in case of functioningof the gas cooling system
39 Transport of FP from broken FC to graphite gapsand to RC free space
Melcor Various chemical and physical processes relevant.H2 (see above) and FP may enter the pool and finishinto the ALS
40 Cooling (including quench) of graphite blocksby water coming from the break and flooding ofRC
U Stack & Relap5 Potential interaction with dissociated O2 notinvestigated
ALS = accident localisation system (part of the RBMK confinement), BIC = boundary and initial conditions, CCFL = counter current flow limitation, FC = fuel channel, FP = fission products, ICM = individual channelmonitoring, PT = pressure tube, <Q> = qualification evaluated within the present context, RC = reactor cavity and TPCF = two phase critical flow-rate.
a Adopted codes within the present framework, D’Auria et al. (2008a), see also D’Auria et al. (2005).b Katran code is embedded into U Stack.
Fig.11.
Differences
between
FC-L
OC
cladradius
calculatedby
Korsar-R
apta;
thecase
ofFC
-BL
OC
KA
GE
reactorcavity).
•In
thecase
ofFC
-LO
CA
afterthe
breakoccurrence,the
fuelis
imm
ediatelycooled
byhigh
velocityfluid
coming
fromthe
bottomand
thetop
ofthe
channel(only
flowreversal
fromchanneltop
isavailable
inthe
caseof
FC-B
LO
CK
AG
E).
•B
ecauseofthe
above,nodry-outcondition
isexpected
inFC
-L
OC
Aafter
thebreak
occurrence:despite
theirincrease
theclad
surfacetem
peraturesof
thefuelrods
remain
wellbelow
thestated
safetylim
it.•
Inthe
caseof
significantdelay
inscram
signal,overheatingof
fuelcladis
expectedafter
thepressure
tuberupture
event.T
hem
echanismfor
claddam
ageis
the‘ballooning’
inthe
caseof
FC-L
OC
A,
andis
the‘collapse’
inthe
caseof
theFC
-BL
OC
KA
GE
andoccurs
beforethe
pressuretube
ruptureevent.
This
isillustrated
inFig.
11,w
hereresults
fromtw
ocouples
of[coupled]codesare
shown:(a)the
increasein
cladradius
versustim
e(500
sis
thetransientstarttim
e)calculatedby
Korsar-R
aptacode
isgiven
inFig.
11aand
iscaused
bythe
increasingpressure
differenceacross
theclad
(rodinternal
pressurenearly
constantanddecreasing
externalpressure);(b)the
decreasein
gapsize
versustim
e(100
sis
thetransientstart
time)calculated
byR
elap5/Frapcode
isgiven
inFig.11b
and
Aand
FC-B
LO
CK
AG
Escenarios:
(a)(b)
gapsize
calculatedby
Relap5-Frap.
,i.e.
over-pressurizationof
the
ring a
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F. D’Auria et al. / Nuclear Enginee
is caused by the increasing clad temperature (with pressuredifferential across the clad remaining nearly constant duringthe calculated transient period). The fuel rods can be dam-aged by hydraulic loads caused by the high speed flow at thebreak. Accordingly, radioactivity source term inside the reac-tor cavity can be larger in case of FC-LOCA compared withFC-BLOCKAGE.The break flow is greater in the case of FC-LOCA comparedwith FLOW-BLOCKAGE. Energy outflow from the breakcan also be greater in the former case: this difference hasbeen found to have a negligible effect in case of evaluationsfor possible MPTR.
.3. The failure-map and the break location for RBMKhannels
Once the fuel channel blockage event triggers, the followingystem parameters or occurrences are relevant for determininghe scenario:
a) nominal operating power of the affected channel,b) axial power shape and/or linear power peaking factor of the
affected channel,c) percentage of the blockage,d) axial and azimuthal position of the break,e) area of the break.
The nominal channel power and the percentage of blockagealues are linked by the ‘FC failure map’ and the axial powerhape and the power peak factor contribute to the calculation ofhe break axial position, both of these discussed below.
The azimuthal position of the break has a role in the estima-
ion of the possibility of MPTR, but it is a statistic quantity withniform distribution (i.e. as a function of the angle identified byhe break axis in the horizontal plane and any reference radiusn the same plane centred on the axis of the fuel bundle).
•
Fig. 12. Failure map for RBMK fuel channels following
nd Design 238 (2008) 1026–1061 1039
The area of the break has been found to have a minor rolepon the long term overall scenario progression, provided its (much) larger than the flow area associated with the mini-
um cross-section at the top of the fuel channel. In this case,ritical flow establishes at the top of the affected fuel channelnd rupture area is irrelevant for the mass and energy releaserom the break. However, break area is relevant for the estima-ion of ‘prompt’ hydraulic loads upon the fuel bundle and theeighbouring channels and related graphite stacks. The ‘prompt’ydraulic loads including pressure wave propagation vanishesn tenths of a second.
All the parameters (a) to (e) are considered in the analyses dis-ussed in chapter 5 In addition, both the RBMK channel failureap and the characterization of the axial break position follow-
ng FC-BLOCKAGE constitute significant results (see chapter) from the application of the computational tools presented inhapter 4. These are discussed hereafter in advance, in ordero complete the phenomenological picture of the fuel channellockage scenario.
.3.1. The failure map for RBMK fuel channelsA fuel channel failure map has been derived as a signif-
cant by-product of the analysis conducted in relation to theC-BLOCKAGE (D’Auria et al., 2005). The FC failure map isiven in Fig. 12 and shows the boundaries for pressure tube andlad damage. The following should be noted:
The map is derived based on the results from ‘discrete’ cal-culations of pressure tube failure in the domain fuel channelinitial power versus the percentage of blockage of the channelinlet pipe. The percentage of blockage is defined in terms of
% blocked free flow area.Overheating of the clad is a prerequisite for the occurrenceof the pressure tube failure, but not in all cases fuel rod over-heating implies pressure tube rupture.
the FC-BLOCKAGE event (D’Auria et al., 2005).
1 ring and Design 238 (2008) 1026–1061
•
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040 F. D’Auria et al. / Nuclear Enginee
The blue region represents the space of un-failed non-overheated rods. Three different colours are used tocharacterize the damage area for rods only (all three regions)or for pressure tube and rods (the two regions on the extremeright of the diagram).Oxidation is the reason for failure in the extreme right regionof the map, oxidation and collapse occur in the central damageregion and collapse occurs in the left region of the damagearea. In addition, hydraulic loads are at the origin of furtherpin failure (or of the ‘increase’ in the fuel pin damage level)in the two right regions of the damage area.
The map has been derived by a dozen code runs. Thereforehe boundaries among the various zones are approximate. Morealuable is the procedure adopted to derive the map and thembedded qualitative information.
.3.2. The characterization of the elevation of the breaklong the fuel channel axis
The full channel blockage causes a fast voiding in the chan-el, i.e. the order of magnitude for the channel emptying timess one second. Therefore, all zones of the fuel bundle remainarly in the transient without cooling and fuel and rod surfaceemperature growths are faster where linear power is higher.his typically happens in the central region of the bundlelong the axis. High rod temperatures imply high tempera-ures for pressure tube owing to radiation and contact heatransfer and consequent trigger of creep mechanism for tubeupture.
Partial channel blockage is expected to cause dry-out owingo liquid film depletion earlier in the upper region of the bundleith bottom region partially cooled. In this situation, high tem-eratures with rods and, subsequently, tube wall damage occur
n the upper part of the bundle.
The overall phenomenological picture related to the possi-le elevation of the break along the fuel channel axis and theiscussion above are summarized in the sketch of Fig. 13.
(l
able 4lassification of codes adopted for the analysis of the FC-BLOCKAGE scenario in R
o. Technological safety area
System Thermal-HydraulicsPrimary System
FuelConfinement
Computational Fluid DynamicsStructural Mechanics
Neutron KineticsGeneration of average p(e.g., macroscopic cros‘λ-functions’)3D transient neutron flu
Fission ProductsGenerationTransport
a U Stack code has capabilities to handle technological safety areas 1, 2, 3 and 5.
ig. 13. Sketch to illustrate the parameters affecting the break location alonghe fuel channel axis.
. The computational tools and the qualification level toalculate the FC-BLOCKAGE
The variety of computational tools necessary to calculatehe scenario consequent to the blockage of one RBMK fuelhannel can be derived from the previous chapter. In particular,he numerical codes that have been adopted within the presentontext are listed in Table 3. An outline of the main featuresf adopted codes and nodalizations is given below. Additionaletails related to codes can be found in D’Auria et al. (2005).
.1. The codes
The numerical codes adopted within the present frameworkTable 3) are distinguished in Table 4 according to the techno-ogical areas discussed by D’Auria et al. (2008a) and a short
BMK according to technological areas relevant in nuclear reactor safety
Codes Notes
Relap5, Korsar Also includes H2generation and transport
Frap, RaptaCocosys, Relap5FluentKatran, Ansys (U Stack)a The MPTR issue is
addressed by the U Stackcode and by a procedure(given in chapter 6)
arameterss-sections or
Njoy, Unk, Helios
x Bars, NestleRefp, MelcorCocosys, Melcor
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escription is provided hereafter in relation to codes not dis-ussed in companion papers (D’Auria et al., 2008a,b,c,d). It shalle noted that all technological areas identified as characterizinghe deterministic analysis sector within the nuclear reactor safetyre relevant for evaluating the consequences of the fuel channellockage scenario in RBMK.
.1.1. System Thermal-Hydraulics: primary systemRelap5 and Korsar codes are adopted to calculate the thermal-
ydraulic parameters following the FC-BLOCKAGE event inhe Main Coolant Circuit of the RBMK. Relap5 is the widelynown and diffused code developed by Idaho National Labo-atory in the US already available since the end of 70s. Korsars based on the same principles and equations at the basis ofelap5 and was developed and qualified in the last few years at
he St Petersburg Institute NITI in Russia. Information can beound in D’Auria et al. (2008b).
.1.2. System Thermal-Hydraulics: fuelRapta and Frap codes are adopted to calculate the fuel per-
ormance parameters including rod deformation following theC-BLOCKAGE event in the RBMK fuel bundle.
The code Rapta, includes basic models of processes and phe-omena inherent to behaviour of fuel rods with oxide fuel andladding made of zirconium alloy in various transient and emer-ency regimes of RBMK reactor. Thermo-mechanical modulesere developed making reference to one single-equivalent (i.e.
he most loaded in terms of thermal power) fuel rod. The codes qualified against experimental data, e.g., Goncharov et al.2005).
The code Frap is a well established international code devel-ped at Pacific Northwest National Laboratory in the US aimingt the evaluation of fuel rod performance. Two main modules areart of the code: Frapcon, e.g., Lanning et al. (1997), and Frap-ran e.g., Cunningham et al. (2003), that calculate the steadytate and the transient performance, respectively. The qualifica-ion level can be recognized from the references of the aboveocuments. The code is embedded into the Relap5 code.
.1.3. System Thermal-Hydraulics: confinementCocosys and Relap5 codes are adopted to calculate the
hermal-hydraulic parameters in the RBMK confinement follow-ng the FC-BLOCKAGE event. Relap5 code has already been
entioned above. Cocosys code has been developed and qual-fied in the last few years at the GRS in Germany. Informationbout the codes and their qualification level in the concernedrea can be found in D’Auria et al. (2008c) (information aboutocosys as a fission product transport code is also given below).
.1.4. Computational fluid dynamicsFluent code has been adopted to calculate hydraulic loads
cting upon the fuel rods following the rupture of the pressure
ube occurring during the FC-BLOCKAGE event. Fluent is aommercial ‘finite volume’ based code, e.g., Fluent Inc. (2003),or modelling fluid flow and heat transfer in complex systems.t can be used for (not an exhaustive list) the analysis of: (a)
cdot
nd Design 238 (2008) 1026–1061 1041
ncompressible or compressible, steady-state or transient, in-iscid, laminar, and turbulent flows; (b) flows with Newtonians well as non-Newtonian rheology; (c) convective, natural andorced, coupled conduction-convective-radiation heat transfer;d) moving reference frames, including sliding mesh interfacesnd mixing planes; (e) flows of chemical species, mixing andeaction, including combustion and surface deposition reac-ions; (f) flows with arbitrary volumetric sources of heat, mass,
omentum and turbulence; (g) two-phase flows, including cav-tation.
.1.5. Neutron kinetics: generation of average parametersNjoy code is used to extract information from material related
ibraries for microscopic cross-sections functions of energye.g., ENDF). Helios and Unk codes are used to derive macro-copic cross-sections or ‘�-matrices’, respectively that are useds input by 3D neutron kinetics codes (see below). Informationbout Njoy, Helios and Unk codes can be found in D’Auria etl. (2008c).
.1.6. Neutron kinetics: 3D transient neutron fluxNestle and Bars codes respectively coupled to Relap5 and
orsar codes are adopted to calculate the neutron kinetics param-ters in the individual fuel channel and associated graphite stackollowing the FC-BLOCKAGE event. Both Nestle and Bars areidely used and qualified codes within US and Russia, respec-
ively, in the area of their application, i.e. the transient neutroninetics in water cooled nuclear reactors. Information can beound in D’Auria et al. (2008c).
.1.7. Structural mechanics and the MPTR issueKatran and Ansys codes have been adopted to calculate
tresses and strains in the pressure tube and in the graphite blocksollowing the rupture of the pressure tube occurring during theC-BLOCKAGE event.
Katran is a special code developed within the RBMK designechnology, e.g., Parafilo et al. (2000) and Soloviev et al. (2003).he code models a spatial axis symmetric problem of viscous-lastic deformation of the pressure tube loaded with internalressure that is uniform along the azimuthal angle and vari-ble together with temperature along the height. Anisotropy ofhe tube material properties is taken into account. Calculationf deformations is made for any axial section according to therofile of temperature. At a certain stage of the deformation pro-ess, the interaction of a pressure tube with the graphite columns taken into account. After the occurrence of extended contactetween a tube and the graphite, the blocks of graphite columnre loaded with internal pressure minus the “resistant reaction”f the tube. The occurrence of a critical pressure for the graphitelock causes the formation of cracks under the simultaneousccurrence of pressure tube ballooning. The full loss of integrityor the graphite blocks occurs when the cracks cover the entire
ross-section. After the destruction of graphite blocks the furthereformation of a tube before break is calculated. The estimationf the integrity of pressure tubes is carried out by consideringemperatures, deformations, force and power failure criteria.
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Ansys is a commercial ‘finite element’ based code, e.g.,nsys Inc. (2002), for performing static and dynamic analysesf linear and non-linear problems (due to materials proper-ies, geometry, contact between surfaces, etc.) in many fields ofpplication (structural, thermal, electromagnetic, fluid-dynamic,tc.). It is possible to solve coupled problems in the areas ofuid-structure and thermal–mechanical interactions.
The MPTR related analysis is performed (as mentioned inable 4) by the U-Stack code and, independently, by a proce-ure making use of different codes as described in chapter 6.he 3D U Stack code, Baldin et al. (2004) and Parafilo et al.
2004), developed by PhEI and NIKIET Institutions in Russia,s an integrated computational tool used for the simulation ofrocesses taking place in the graphite stack, the reactor cav-ty, the gas circuit and the main circulation circuit of a RBMKn case of a pressure tube leakage. The capabilities of Katrans well as of a thermal-hydraulic system code like Relap 5 arembedded into the U Stack. This code is capable of assessinghe venting system efficiency of the reactor cavity in dealingith small leakages. In case of large breaks of pressure tubesith coolant discharge leading to graphite stack deformation,
he code allows the evaluation of the stack deformation and thessessment of the possibility of propagation of the rupture topntact fuel channels.
.1.8. Fission products: generationRefp code is used to calculate the source term associated with
he operation of a fuel channel of the RBMK, i.e. the amountf radioactivity that is released during the progression of theC-BLOCKAGE event. The program simulate the behaviourf five isotopes of radioiodine, two isotopes of caesium andve noble gases. It provides a step by step calculations of: (a)
he releases from fuel and coolant including the sub-divisionetween gaseous and liquid volumes inside the confinement, (b)he changes of chemistry and of physical form of iodine, (c) theemoval from the confinement atmosphere and the absorptionn the walls and (d) the revolatilization and adsorption of iodinend leakage outside the confinement (Moskalev and Jankowski,004). However, the last capabilities of the code are not exploitedithin the present framework and related analyses are carriedut by Cocosys and Melcor, as described in the next paragraph.
.1.9. Fission products: transportCocosys and Melcor codes are used to calculate the transport
f the fission products generated as a consequence of the meltingnd the damage of a RBMK fuel bundle during the progressionf the FC-BLOCKAGE event. Fission products are transportednside the primary circuit and, to the largest extent, from thereak region to the reactor cavity, to the pool, to the Accidentocalization System and to the environment (flow paths dis-ussed in D’Auria et al., 2008c). Cocosys and Melcor, developednd qualified at GRS in Germany (already mentioned) and atandia National Laboratory in US, respectively, are well estab-
ished codes widely used by the international community. Theualification level in the area of interest is demonstrated in theapers by Ahrens et al. (2003) and by Nagasaka et al. (1998),or Cocosys and Melcor codes, respectively.
T
to
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.2. The nodalisations
Nodalizations or input decks have been developed and qual-fied (as far as possible) for each of the code listed in Table 4.n some cases different input decks for the same code were pre-ared to address specific objectives of the analysis (e.g., theechanical resistance of the bare pressure tube, of the pressure
ube plus the graphite rings and graphite blocks, with differentssumptions for the constraints, etc.). At the end, more than 20nput decks were prepared and used to obtain a comprehensiveiew of the FC-BLOCKAGE scenario.
A detailed description of all of these can be found in D’Auriat al. (2005) and is far beyond the scope of the journal paper.owever, selected sketches representative of input decks devel-ped in relation to the areas mentioned in Section 4.1 are outlinedereafter. Approximate dimensions of nodalizations, e.g., num-er of nodes or of elements for the various cases, are given in’Auria et al. (2008a) and are not repeated here. Notes about
he qualification level of the nodalizations are given below notn a systematic way.
.2.1. The TH input decks covering the areas ‘Systemhermal-Hydraulics: Primary System And Confinement’
The thermal-hydraulic nodalization for the overall RBMKystem, main coolant circuit and confinement are discussed by’Auria et al. (2008b,c), respectively. The fuel channel nodal-
zation with the connected portion of the reactor cavity is givenn Fig. 14. The same level of qualification applicable for theverall core model, i.e. above papers, is valid here.
Several valves are part of the input deck and are ‘installed’erpendicular to the channel axis. Each valve opening controlakes input from the pressure difference across the tube wallsnd from the wall temperature in the region where the valve isnstalled. The curve in Fig. 9 is modelled and the (simplified)pproach described in Section 2.4 is used to determine the con-itions for valve opening. Once opened the valve does neverlose. The implemented logic is such to allow the calculation ofhe break location (location of the valve along the axis) and thereak area (number of valves open). Several studies have beenone to optimize the input deck also aimed at identifying theest value for the individual valve area.
The two ‘columns’ on the left of the figure represent theeactor cavity gaps. Namely, the gap around the affected chan-el and the overall gap area (and volume) associated with eighteighbouring channels, as shown in the sketch on the top rightf the figure, are simulated. This allows the calculation of theifferential pressure across the neighbouring channel stack.
The first column on the left represents the remaining space inhe reactor cavity and allows, together with the simulated upper,ower and side cavity free volume, the calculation of the absoluteressure in the cavity.
.2.2. The FU input decks covering the area ‘System
hermal-Hydraulics: Fuel’
The fuel in RBMK does not differ from the fuel in otherypes of water cooled reactors, including the material propertiesf interest. Therefore, ‘well established’ (limited independent
F. D’Auria et al. / Nuclear Engineering and Design 238 (2008) 1026–1061 1043
Fig. 14. The TH input deck covering the areas ‘System Thermal-Hydraulics, Primary System and Confinement’ to model the FC-BLOCKAGE in RBMK: Relap5code used to model the affected channel and two rows of neighbouring graphite stacks in the confinement.
Fig. 15. The FU input deck covering the area ‘System Thermal-Hydraulics, Fuel’ to model the FC-BLOCKAGE in RBMK: Rapta and Frap codes nodalizations (leftand right, respectively) of the fuel rod.
1044 F. D’Auria et al. / Nuclear Engineering a
Ftt
qa
pttb
4‘
eaiFibo
spatad
4KT
af3N
mbcD
4M
eifor the Ansys code. The input deck includes the pressure tubeand the associated graphite ring and bricks, as shown in Fig. 17.
Extensive validation was conducted and documented in thementioned report mainly in relation to the creep model and to the
ig. 16. The CFD input deck covering the area ‘Computational Fluid Dynamics’o model the FC-BLOCKAGE in RBMK: Fluent model of affected region ofhe pressure tube to calculate hydraulic loads acting upon the fuel bundle.
ualification needed) nodalizations with suitable level of detailre used for Rapta and Frap as shown in the sketches of Fig. 15.
The nodalizations are used to determine temperature andressure (for the internal gas) transients inside the fuel rods andhe deformation of the clad till the possible rupture event. Theiming of such occurrences are also calculated. Results obtainedy Rapta have already been shown in Fig. 11.
.2.3. The CFD input deck covering the areaComputational Fluid Dynamics’
A region of the fuel bundle of the RBMK fuel channelnclosed by two horizontal planes (orthogonal to the channelxis) has been modelled for the application of the Fluent code,ncluding a simplified form for the break (consequence of theC-BLOCKAGE event). The sketch of the nodalization is given
n Fig. 16. Convergence analyses were carried out for the num-er of nodes as well as sensitivity studies related to the choicef the turbulence model and the assumed free area for the break.
Single phase flow (gas) is used to simulated the superheatedteam exiting the break. The gas speed causes forces (nearly) per-endicular to the rod axis and resulting hydraulic forces. These
re applied to the individual rods in the bundle and, based uponhe related constraints (i.e. grids), the material properties and thectual temperature values, the conditions are calculated for rodisruption (caused by hydraulic loads).
Ftt
nd Design 238 (2008) 1026–1061
.2.4. The NK input decks covering the areas ‘Neutroninetics: Generation of Average Parameters and 3Dransient Neutron Flux’
The neutron kinetics nodalizations suitable for the gener-tion of macroscopic cross-section and ‘λ-functions’ (neededor Nestle and Bars code) and for performing transient coupledD neutron kinetics thermal-hydraulic analyses (by Relap5/3D-estle and Korsar-Bars) are described by D’Auria et al. (2008d).The NK nodalization is used, as described in the above-
entioned paper, to calculate the actual fission power in theundle affected by the FC-BLOCKAGE event considering theounterfeiting effect of voiding (positive reactivity) and ofoppler (negative reactivity).
.2.5. The SM input deck covering the area ‘Structuralechanics’As in the case of the FU input deck discussed above, a ‘well
stablished’ (limited independent qualification needed) nodal-zation with suitable level of detail has been developed suitable
ig. 17. The SM input deck covering the area ‘Structural Mechanics’ to modelhe FC-BLOCKAGE in RBMK: Ansys model of the break region of the pressureube plus graphite ring plus graphite block.
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F. D’Auria et al. / Nuclear Enginee
ptimum number of meshes. Sensitivity studies were performedn relation to the effects of fluence, gap thicknesses (across theraphite ring), graphite average temperature, heating rate andardening of the pressure tube.
The nodalization was used to determine the loads (primarilynternal pressure for the zirconium tube) and the conditions (pri-
arily wall temperature for the pressure tube) under which theompound formed by the pressure tube, the ring and the graphiterick fail. Information about the azimuthal location of the breakould also be attained.
.2.6. The MPTR input deck covering the area ‘MPTR’In order to address the MPTR issue, two independent ways
ave been pursued as already mentioned: (1) use of a procedureased on the application of nodalizations described above; (2)se of the U Stack code. The procedure at item (1) is discussedn chapter 6 and the U Stack nodalization, item (2), is outlinedelow.
The geometric arrangement of the core, the thermal-hydraulic
onditions of the primary loop and of the confinement, theonfiguration of the constraints and the material properties con-titute the main input dataset necessary to develop a suitable-Stack nodalization. A simplified sketch is given in Fig. 18.
ig. 18. The MPTR input deck covering the area ‘MPTR’ to model the FC-LOCKAGE in RBMK: U-Stack model of a portion of the core. ‘Crackropagation valves’ can be seen in the upper diagram and arrow indicating flowaths across graphite stacks in the lower diagram.
eaL
4P
pmiRRn
teamat
5
tDflca
5
athD
nd Design 238 (2008) 1026–1061 1045
ypical relevant aspects of the MPTR scenario that are consid-red in the nodalization, giving an idea of the complexity of theroblem are:
Translational movements of graphite bricks affected by: (a)inertia of the brick, pressure on lateral edges, (b) contact forcesfrom neighbouring stacks, (c) deformation of the pressuretube and (d) friction in the contact between adjacent bricks.Rotational movements of graphite bricks affected by: (a)moment of inertia, (b) torque caused by non-uniform pressuredistribution and (c) torque from contact and friction forces.Various cases of contact between pressure tube and graphitestack include: (a) no contact, (b) one contact region, (c) two orthree contact regions and (d) contact along the entire surfaceof the tube.
The U-Stack nodalization could benefit of the qualificationased on the experiments performed in various versions of theKR facility installed at Electrogorsk (Ru), e.g., Medvedevat al. (2004b). The qualification domain for the U Stack codend nodalization also included the simulation of the event ineningrad-3 NPP described above, NIKIET and PhEI (2006).
.2.7. The FP input decks covering the areas ‘Fissionroducts: Generation and Transport’
The main path for the fission product release, including thehenomena of generation and transport, is inside the confine-ent. Therefore, the adopted input decks are those outlined
n D’Auria et al. (2008c) specifically related to Cocosys andelap5. In addition, the developed Melcor nodalization for theBMK NPP includes features that are similar to the Cocosysodalization.
The FP nodalizations are used to calculate the fission productransport throughout the confinement and the source term to thenvironment. The main output is constituted by the radioactivitymount in each region of the confinement and in the environ-ent. The source term to environment is characterized by timing
nd elevation of the release and is suitable for providing inputo the environmental impact codes.
. The FC-BLOCKAGE predicted scenario
Results related to the FC-BLOCKAGE scenario, accordingo the information available from Table 2 has been given by’Auria et al. (2008a,b,c,d) with all transients outlined in the
ormer case and with focus put on 3D neutron kinetics in theatter case. Hereafter a complete picture of the scenario is given,onsidering the technological safety areas already defined anddopted in chapter 4.
.1.1. Boundary and initial conditions
The nominal operating conditions for Smolensk-3 RBMK
re assumed (see D’Auria et al., 2008a) as boundary and ini-ial conditions for the analysis. The considered blocked channelas vertical and horizontal coordinates 27–32 e.g., Fig. 10 in’Auria et al. (2008d), with axial power distribution assumed
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046 F. D’Auria et al. / Nuclear Enginee
s the core average power distribution, Fig. 11 of the same paper.he channel is characterized by an initial power of 2.1 MWthnd by a ‘flow-rate/power’ ratio equal to 3 kg/MW s. Other inputata, namely, inlet and outlet pressure, inlet sub-cooling, pres-ure of the reactor cavity, axial and radial distribution for theraphite temperature are given in Table 1.
As an exception, the results for the area ‘System Thermal-ydraulics, Confinement’, obtained by the U Stack and Cocosys
odes relate, respectively:
to the blocked channel with coordinates 52–16 of theLeningrad-3 core where the initial power is 2.0 MWth andthe ‘flow-rate/power’ ratio is to 3.8 kg/MW-s; in this case theflow-blockage event brings the core inlet flow-rate from 7.5to 0.25 kg/s.
nmpea
able 5esulting sequence of main events and related key-quantities values for the FC-BLO
afety analysis
o. Event Time (s)a
1 Full blockage occurrence 02 Full blockage completed 0.13 Dry-out occurrence 1–404 10 K superheated steam appears at the channel
outlet3
5 Fission power attains 90% of its initial value ∼106 Only steam present in the channel 127 Clad surface temperature achieves 700 ◦C 20–70
8 Pressure tube heating rate before break (K/s) 10–129 The pressure tube temperature achieves 700 ◦C 710 ‘Prompt’ radioactivity release to MCC (Ci) 2501 Clad achieves 1204 ◦C 652 Occurrence of the ‘first’ break 74 (tB)3 Elevation of the first break (m) 54 Overall break area (m2) **5 Break opening time 0.056 Flow reversal occurs at channel outlet tB + 0.57 Time of maximum load acting upon
neighboring graphite stacktB + 2
8 Value of the maximum load on neighboringgraphite stack (t)
22
9 Scram occurrence tB + 2.20 Quench of wall temperatures above break axis tB + 51 Max pressurization rate in gas-gap (MPa/s) ∼0.62 Pressurization rate in reactor cavity (MPa/s) 0.043 Maximum hydraulic load acting upon (still)
intact rods (kN)2.8
4 Maximum stress in graphite before crackpropagation (MPa)
6.6
5 Value of fission power before the scram (%) ∼856 Quantity of damaged fuel (%) due to thermal
loads and hydraulic loads∼40
7 Maximum pressure in the reactor cavity (MPa) <0.138 Quantity of H2 produced (kg) 0.069 Radioactivity in the reactor cavity 200 s after
the break occurrence (Ci/kg)4.1e3/5.7e
0 Radioactivity in the ALS 200 s after the breakoccurrence (Ci/kg)
2.9e3/5e−
1 Radioactivity to the environment 200 s after thebreak occurrence (Ci)
to the maximum power channel of Smolensk-3 characterizedby an initial power of 2.9 MWth.
Boundary and initial condition values for quantities noteported in the above two dashed items coincide with the valuesescribed in the previous paragraph.
. Results
The resulting sequence of main events (Smolensk-3 chan-el with coordinates 27–32) is given in Table 5. Reference is
ade to the safety technology area identified below and to the
henomena listed in Table 3. It shall be noted that for somevents or related quantities, like evaluation of hydraulic loadscting upon the fuel rods following the break of the pressure
CKAGE scenario considering the various technological areas relevant for the
Notes
Range given for the various axial locationsRelevant to ICM design (D’Auria et al., 2008e)
3D coupled NK TH analysis
Clad collapse occurs, Fig. 11b. Range given for the variousaxial locationsRelevant for calculating tube rupture timeStart of creep mechanismFission gases releaseStart of significant H2 production
Starting from the bottom of the active fuel**Value > > than flow area of outlet pipeBased on creep analysis
Including dynamic load. See also Fig. 34b
Pressure increase in RC = 0.075 MPa
Average value for 5 s after the break startThis causes rupture of eventually intact rods at the location ofbreak occurrenceThe value largely depends on parameters like fluence
% of initial value.% of initial value at 5 s after scram. This corresponds to about50 kgFree space regionDuring the predicted transient duration
−3 Radioactivity values are derived from radioisotopesmasses. H2 to the environment has not been calculated.
3
F. D’Auria et al. / Nuclear Engineering and Design 238 (2008) 1026–1061 1047
Fig. 19. FC-BLOCKAGE scenario, area ‘System Thermal-Hydraulics, PrimaryStfl
tfot
6
HR
FSs
afcflpb
transient, Fig. 19b: (1) soon after the blockage, pressure drop
ystem’, Relap5 results for the affected channel, short term: (a) flow-rates athe inlet and at the outlet; (b) pressure in different positions along the axis; (c)ow-rates across the ‘break’ simulation valves.
ube, the ‘nearly’ steady-state assumption has been made. There-ore, timing is related to the connected event that is the resultf the ‘dynamic’ calculation (in this case the rupture openingime).
.1.1. Area ‘System Thermal-Hydraulics, Primary System’
Results for quantities related to the area ‘System Thermal-ydraulics, Primary System’ obtained by the application of theelap5 code are given in Figs. 19–21.
rsa
ig. 20. FC-BLOCKAGE scenario, area ‘System Thermal-Hydraulics, Primaryystem’, Relap5 temperature results for the affected channel, long term: (a) rodurface; (b) fluid; (c) pressure tube.
In the channel inlet region, Fig. 19a, the flow-rate vanishest the transient start owing to the blockage. Immediately after,ollowing the vaporization of the existing coolant mass in thehannel, flow-rate at the channel outlet also vanishes. However,ow reversal occurs after the rupture occurrence in the upperart of the channel as identified by the horizontal part of thereak.
The pressure in the channel changes two times during the
edistribution occurs; (2) after the tube rupture occurrence pres-ure in the lower part of the channel and in the region of the breakchieve the reactor cavity pressure (or very close to it) and pres-
1048 F. D’Auria et al. / Nuclear Engineering and Design 238 (2008) 1026–1061
F Systef e tube
stt
TsHstasLtlb
oflotftF
ab
ig. 21. FC-BLOCKAGE scenario, area ‘System Thermal-Hydraulics, Primaryraction along the axis; (b) trajectory (bold dotted line) to determine the pressur
ure in the upper region is determined by pressure drops alonghe flow path from the steam drum (where pressure remains athe initial value) and the break.
The opening of three break simulators can be seen in Fig. 19c.he result show that conditions for channel rupture are reachedimultaneously in different parts of the channel along the axis.owever, the rupture at one level immediately releases the pres-
ure at other levels and the overall break area calculated inhis way is consistent with the experimental values (Baldin etl., 2004; Medvedeva et al., 2004b) and with the values mea-ured in post-accident examination of Chernobyl-1 (1982) and
ively. In particular, the predicted value for the ‘equivalent breakength’ along the channel axis is 0.7 m approximately and thereak area is roughly 10 times the value of the cross-section area
•
•
m’, Relap5 results for the affected channel: (a) overall channel power and voidrupture time.
f the fuel channel outlet pipeline. Therefore, the overall massow-rate is only affected by the cross-section of the channelutlet pipeline (i.e. not by the break area itself). The horizon-al break axis is centred around the elevation of 5 m startingrom the bottom of the active fuel. The sum of the flows thoughhe breaks in Fig. 19c equals the amount of flow reversal inig. 19a.
Rod surface, fluid and pressure tube temperatures at differentxial elevations are reported in Fig. 20a–c. The following shoulde noted (see also Table 5):
Dry-out phenomenon and superheated steam at the channeloutlet are calculated immediately after the blockage event.In a few tens of seconds, after the blockage event, temper-atures for rod damage (clad collapse mechanism), pressure
ring and Design 238 (2008) 1026–1061 1049
•
•
agvshas
taatp
6
HcFnis
mbp‘Ra
6
HUtt
gatT
cc
Fig. 22. FC-BLOCKAGE scenario, area ‘System Thermal-Hydraulics, Confine-mta
aThe difference between pressures in Fig. 22b give are needed
to calculate the mechanical load acting upon the neighbouringchannels (i.e. to estimate the MPTR possibility).
F. D’Auria et al. / Nuclear Enginee
tube damage (break opening) and significant H2 productionoccur.The occurrence of the break event causes, in about one second,‘turnaround’ of all temperatures in the channel region abovethe break axis.After the break occurrence, quenching of metal temperaturesin the region above the break axis is completed in a few sec-onds, while quenching of the bottom part of the channel iscompleted in a few hundreds seconds.
The time for channel emptying can be derived from Fig. 21a,s well as the “assigned” power curve. The curve can be distin-uished into three periods: (a) the steady state nominal imposedalue for the channel, i.e. till t = 0 s, (b) the transient till thecram, determined by the coupled 3D neutron kinetics thermal-ydraulic analysis derived from results in D’Auria et al. (2008d)nd (c) the decay power period, fixed after the estimation of thecram time.
The graphical support for estimating the break occurrenceime can be found in Fig. 21b. For the considered scenario,
nearly vertical trajectory is calculated (pressure differencecross the tube wall remains constant with increasing tempera-ure) that ‘hits’ the available threshold curve when break time isredicted.
.1.2. Area ‘System Thermal-Hydraulics, Fuel’
Results for quantities related to the area ‘System Thermal-ydraulics, Fuel’ obtained by the application of the Relap5/Frap
ode are similar to results already discussed in relation toig. 11b: the clad rupture is caused by the collapse mecha-ism with constant overpressure acting across the clad wall andncreasing rod surface temperature (thus, decreasing materialtrength conditions).
In performing the fuel performance analysis including theechanical deformation of the clad by the Frap code, it has
een ensured that the thermal transient (e.g., rod surface datarovided in Fig. 20a and fuel pellet temperatures) during theperiod of un-deformed rod’, produced the same results byelap5 and Frap codes (documentation available in D’Auria etl., 2005).
.1.3. Area ‘System Thermal-Hydraulics, Confinement’
Results for quantities related to the area ‘System Thermal-ydraulics, Confinement’, obtained by the application of theStack and Cocosys codes are given in Figs. 22 and 23, respec-
ively. In relation to both the analyses, time ‘t = 0 s’ correspondso the rupture opening time.
Pressure evolutions in different positions along the axis areiven in Fig. 22a including the ‘open’ region of the reactor cavitynd pressure evolutions at same elevation in different radial posi-ions (i.e. in different gas-graphite gaps) are given in Fig. 22b.
he following can be observed:
A ‘quasi-static’ pressure peak as high as 0.6 MPa is cal-ulated in the gas graphite gap adjacent top the brokenhannel.
Fm
ent’, U Stack results for pressure evolutions in the reactor cavity: (a) alonghe axis of the affected channel including ‘free-space’; (b) in different gas-gapst 6.4 m elevation.
The pressure peak of the open space of the cavity (needed tossess the integrity of the cavity walls) remains below 0.16 MPa.
ig. 23. FC-BLOCKAGE scenario, area ‘System Thermal-Hydraulics, Confine-ent’, Cocosys results for pressure evolutions in affected confinement regions.
1 ring a
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6
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Fa
050 F. D’Auria et al. / Nuclear Enginee
The pressure time trends of the various involved zones ofhe confinement can be derived from Fig. 23 for the long-termransient evolution. The reference nodalization is presented inig. 4 of D’Auria et al. (2008c). An average pressure is calculatedor the affected channel (gas) gap region in this case, red curve in
ig. 23. In the long term period of the transient, i.e. after abouth since the transient start, the predicted pressure increase is
ess than 0.03 MPa.
tia
ig. 24. FC-BLOCKAGE scenario, area ‘Computational Fluid Dynamics’, Fluent resxis.
nd Design 238 (2008) 1026–1061
.1.4. Area ‘Computational Fluid Dynamics’
Results for quantities related to the area ‘Computational Fluidynamics’ obtained by the application of the Fluent code areiven in Fig. 24. By adopting the input deck outlined in Fig. 16,
he velocity and the pressure fields are derived at first, follow-ng the break occurrence. Then, three constraint assumptionsre considered to model the spacer grids (schemes A, B and C
ults for mechanical loads acting on fuel rods in the neighborhoods of the break
ring a
ia‘i
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ctiad
6
NNti
Ftsd
tf
ototp
nfi(eeitpower increases after the blockages event. However, the overalleffect for the overall bundle is a decrease in generated fissionpower.
F. D’Auria et al. / Nuclear Enginee
n Fig. 24). This allowed the evaluation of the hydraulic loadcting upon the selected region of the rod. Allowable load (orcritical load’, last row of the table in Fig. 24), was determinedndependently considering geometry and material properties.
The result is given in the bottom part of Fig. 24, related tohe overall bundle section: hydraulic loads are such to cause theupture of at least 5 (over 18) rods, whereas nine rods appearsafe’ and the status of four remaining rods needs more accuratepecification of boundary and initial conditions in order to bevaluated.
The documented one is the result of a ‘static’ analysis withonstant fluid speed, set at the maximum value corresponding tohe critical flow that establishes at the break. The allowable loads connected with the actual temperatures of the fuel rods thusllowing the connection with the ‘dynamic’ transient calculationocumented in the previous paragraphs.
.1.5. Area ‘Neutron Kinetics, 3D Transient Neutron Flux’
The key result from the area ‘Neutron Kinetics, 3D Transient
eutron Flux’, obtained by the 3D coupled code Relap5/3D-estle, has already been given in Fig. 20a, initial 74 s of the
ransient. Besides, it is shown that the fission power producedn the affected channel decreases following the blockage event:
ig. 25. FC-BLOCKAGE scenario, area ‘Neutron Kinetics, 3D Transient Neu-ron Flux’, Relap5/3D-Nestle results for non-dimensional fission power: (a)elected axial regions of the affected fuel channel; (b) axial distribution atifferent times.
FUod
nd Design 238 (2008) 1026–1061 1051
his constitutes an input for the analysis performed within theramework of the present paper.
The decrease in power production is of the order of 10%f the initial power and occurs mostly in the initial 5 s of theransient, caused by the Doppler effect. Compared with the casef constant power, the consideration of the actual power bringso a delay of about 25 s in the prediction of the occurrence forressure tube rupture.
Additional results related to the application of the coupled 3Deutron kinetics code to the FC-BLOCKAGE scenario can beound in the companion paper by D’Auria et al. (2008d), includ-ng the background for the derivation of the mentioned resulte.g., cross-section derivation, power in neighbouring channels,tc.). The power change in non-dimensional form associated toach individual axial node of the affected channel is reportedn Fig. 25 together with the axial power distribution at differentimes. As expected, in some regions, e.g., bottom regions, the
ig. 26. FC-BLOCKAGE scenario, area ‘Structural Mechanics’, Katran-Stack results for pressure tube quantities, related to a tube close to the broken
ne, along the axis, at the time 23 s after the rupture: (a) bend stresses; (b)eformation of tube surface.
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052 F. D’Auria et al. / Nuclear Enginee
The neutron kinetics response associated with geometryhange are not considered in the present study. In the course ofn accident, these are expected to occur on the first time, whenhe threshold for clad damage (collapse in this case) is over-assed. Thereafter (i.e. with the predicted event progression),ossibly, rods bowing, contact(s) with pressure tube, violent oxi-ization connected with the H2 production and, possibly cladnd/or rods disruption due to thermal and hydraulic loads occur.ach of these event can be associated with change in neutron
roperties, i.e. ‘λ-matrices’ or macroscopic cross-sections, andreate an additional feedback upon fission power. The availableoupled 3D neutron kinetics thermal-hydraulic computationalools, namely Korsar-Bars, are ready to account for geometry
ctoa
ig. 27. FC-BLOCKAGE scenario, area ‘Structural Mechanics’, Ansys results: (a) sb) sensitivity of maximum principal stress to pressure tube temperature increase rate
nd Design 238 (2008) 1026–1061
hanges (D’Auria et al., 2005). However, such feedback wasot considered within the present framework.
.1.6. Area ‘Structural Mechanics’
Results for quantities related to the area ‘Structural Mechan-cs’ obtained by the application of the U-Stack and Ansys codesre given in Figs. 26 and 27. The use of those codes implieseglecting (case of Ansys) the consideration of the tube rupture
riterion given in Fig. 9 (see also Fig. 21b for its application) andhe use of more sophisticate criteria (case of Katran-U Stack)r of the material properties (case of Ansys). In the case ofpplication of these codes, a relevant the input is constituted by
ensitivity of maximum principal stress to fluence and temperature of graphite;; (c) plane distribution of maximum principal stress before crack propagation.
ring and Design 238 (2008) 1026–1061 1053
tpttsoi
dtpdebcfTgp
t
(
(
(
sij
tFibawt(F
tohc
FTc
fgdt
ggt(
6
ehafinement’.
The amount of H2 produced and transported inside the pri-mary system (MCC in the figure) and in the confinement, reactorcavity and accident localisation system (RC and ALS in the fig-
Table 6Refp calculation of the fission product inventory in one Smolensk-3 RBMK NPPfuel channel with average burn-up of approximately of 7500 MWd/t
he temperatures predicted by the thermal-hydraulic code, e.g.,ressure tube and graphite, and the output are the stresses andhe strains for the system under consideration as a function ofime, i.e. the zirconium tube, the graphite ring and the graphitetack. The rupture location and the occurring time is also anutput from the analysis (and may differ from the result shownn Fig. 21b).
The Katran-U Stack (Katran can be classified as one-imensional code) coupled code is capable of calculating, amonghe other things, the deformation and the shape change ofressure tube along the axis as a function of time. The axialistribution of bend stresses and of deformation after the breakvent for a tube close to the broken one, is given in Fig. 26. Thereak occurrence causes a bend deformation (in the reportedase, maximum at around 5.5 m from the bottom of active fuel)or about 14 cm, Fig. 26b (see also the right sketch in Fig. 6a).his implies the displacement and/or the break of at least tworaphite blocks, to accommodate for the movement of the metalressure tube (see chapter 6).
The use of the Ansys code implies (in addition to what men-ioned above):
a) The need to model with suitable level of detail (i.e. by finiteelements) each of the three components of the system underconsideration. Convergence should be demonstrated whendecreasing the mesh size.
b) The knowledge from the user or the user assumption inrelation to parameters like friction between adjacent com-ponents (e.g., ‘zirconium tube versus graphite rings’, or‘graphite brick versus graphite brick’), gap thicknesses andlocal material properties, both affected by the life in the coreand by the fluence.
c) The need to perform a variety of sensitivity studies. Theseincluded: ‘graphite ring-graphite brick’ thickness, fluenceand related hardening for graphite and pressure tube, averagegraphite temperature, pressure tube heating rate and numberof meshes.
Sample Ansys results, given in Fig. 27, relate to sensitivitytudies (Fig. 27a and b) and to the spatial distribution of max-mum principal stress in a horizontal plane, Fig. 27c at a timeust before the crack propagation start.
As expected, the rate of temperature increase of the pressureube has a large influence upon the time of the break occurrence,ig. 27a (Relap5 calculated values for such parameter are given
n Table 5). The smaller is the heating rate, the greater is the delayefore the onset of contact. In the case of gap equal to 0.7 mmnd fluence equal to 22 × 1021 n/cm2, the contact always occurshen the pressure tube temperature reaches about 1050 K, so
he mentioned delay is inversely proportional to the heating rateconsistently with the information provided by the diagram inig. 9).
One effect of irradiation on the graphite is the decrease of
he thermal conductivity that leads to an increase of the graphiteperating temperature. The effect of such increase on the timeistory of the maximum principal stress, assuming an initial gapompletely closed and two different fluence values can be seen
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ig. 28. FC-BLOCKAGE scenario, area ‘Fission Products, Generation andransport’, Melcor result for H2 transport in the primary system and in theonfinement.
rom Fig. 27b. As the graphite temperature increases, the actualap to be filled by the creeping pressure tube becomes largerue to the increased graphite thermal deformation. The effect ishe same as having a larger initial cold gap.
The stress intensification phenomenon at the edge of theraphite ring causes in the location of the contact with theraphite brick the most probable point for the propagation ofhe crack, the inside the graphite brick, as evidenced in Fig. 27cred ellipse in the bottom diagram).
.1.7. Area ‘Fission Products, Generation and Transport’
Within the safety technology area ‘Fission Products, Gen-ration and Transport’ the H2 transport is considered at firstereafter, even though this is a possible (typical) output of thereas ‘System Thermal-Hydraulics, Primary System and Con-
re, respectively) can be seen in Fig. 28, as calculated by theelcor code. The following should be observed:
Time 100 s is the start of the FC-BLOCKAGE event and,as expected, H2 production starts immediately after theevent.
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ig. 29. FC-BLOCKAGE scenario, area ‘Fission Products, Generation and Transporf xenon; (b) deposition of iodine mass at 3 h after the rupture and (c) release of Cesi
nd Design 238 (2008) 1026–1061
The total amount of H2 produced and transported is muchlower than the amount theoretically producible in one fuel
channel, i.e. of the order of 1%, see Table 1.The break occurrence time is predicted at about 50 s insteadof 74 s as given in Fig. 21. This is due to the considerationof constant full power in the period before the pressure tube
t’, Cocosys results for fission products transport in the confinement: (a) releaseum aerosol in the form CSOH.
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FC-BLOCKAGE, like the fuel channel Loss of Coolant Acci-dent, or the group distribution header blockage (e.g., FC-LOCAand GDH-BLOCKAGE, see the list in Table 2), or a variety ofAnticipated Transient Without Scram (ATWS) and/or Reactivity
ig. 30. Sketch dealing with the MPTR issue. One broken channel may cauropagation are given.
rupture (e.g., the power decrease calculated by the coupled3D neutron kinetics analysis was not taken into account inthe present study).
The amount of fission products leaving the rupture andransported inside various zones of the confinement ando the environment (confinement zones and flow-path tohe environment given in Fig. 5, more details available by’Auria et al., 2008c) can be seen in Fig. 29 as cal-
ulated by the Cocosys code. The following should bebserved:
The total amount of fission products ‘stored’ in one fuel chan-nel is given in Table 6 as calculated by the Refp code. Basedon the performed investigation, all radioactivity in the gap(third and fifth columns in Table 6) is released to the maincooling circuit and primarily to the confinement at the timeof the rupture opening.Time 0 s in the Fig. 29a and c is the opening time of thepressure tube rupture.Releases to the environment occur more than 1 h after theFC-BLOCKAGE event start.The amount of releases to the environment constitutes a frac-tion for around 4–8 orders of magnitude lower than the overallrelease in terms of mass of the isotopes (more details, includ-ing a comprehensive list of released isotopes is available fromD’Auria et al., 2005).The release of fission products is assumed to be causedonly by thermal damage of the fuel. Releases caused byhydraulic loads including the impact of already thermal dam-aged pieces of fuel with solid surfaces (e.g., graphite bricks) isneglected.Based on the post-accident examination of the events inChernobyl-1 (1982) and Leningrad-3 (1992) (NIKIET, 1983,
1992) and the supporting analyses performed within thepresent framework, it can be estimated that about 50 kg offuel looses its integrity and its design configuration followingthe FC-BLOCKAGE event.
Fp
lure of adjacent channels. In the right side, the worst conditions for rupture
. The MPTR related results
The consequences of the FC-BLOCKAGE event for theffected blocked channel and for the confinement and the envi-onment have been evaluated and discussed in the previoushapter. The problem here is to estimate the possibility that theupture of one channel (unavoidable for the considered event)riggers a domino effect including the rupture of other channelsi.e. dealing with the MPTR issue).
The MPTR scenario can be triggered by events other than the
ig. 31. FC-BLOCKAGE scenario, area ‘MPTR’: directions of spatial decom-osition for graphite block motion considered by U Stack code.
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nitiated Accidents (RIA), other than the notorious Chernobyl-event (1986). Although the Probabilistic Safety Assessment
PSA) has not been part of the present investigation, as alreadyentioned, and although MPTR analyses have been conducted
or transients other than the individual fuel channel blockage,.g., Novoselsky and Filinov (2000), the present investigationhows that the only “realistic” origin of MPTR to be consideredn Safety Analysis Report is the FC-BLOCKAGE event.
Two independent studies have been conducted within theresent framework, using the U-Stack code and a MPTR pro-edure detailed in the following. For both cases, the sketch inig. 30 applies, with assumptions adopted for the procedure
iven in the right side. The results discussed in chapter 5 con-tituted the basis for deriving the conclusion about the MPTRssue.
ANa
ig. 32. FC-BLOCKAGE scenario, area ‘MPTR’, U Stack results related to the Levent; (b) ‘stabilized’ situation at the end of the analysis.
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. Results from U Stack
Considering the ‘initial failure location’ as shown in Fig. 30,he U Stack uses the ‘spatial decomposition method’ to calcu-ate the motion of graphite blocks along the axes X, Y and Z asepicted in Fig. 31 and the rotation around the axes X, Y and.
The key results from the application of the U-Stack codere given in Fig. 32 in a pictorial form and are supported by theiagrams in Figs. 22 and 26. Furthermore, nodalization sketchesn Figs. 6 and 18 and related description should be consideredtogether with sketches in Figs. 30 and 31 above mentioned).
comprehensive documentation of the results can be found inIKIET (2004) (see also D’Auria et al., 2005). The following
spects are relevant:
ningrad-3 (1992) NPP event: (a) tube deformation process after the blockage
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The blockage event causes pressure tube overheating andcreep deformation as shown in the upper right part of Fig. 32related to the affected tube.The deformation of the tube can be observed in the upper leftpart of Fig. 32 where the profile of the tube is reported atdifferent times before the break.The time of break occurrence and the axial location of thebreak can also be deducted from the upper left diagram inFig. 32.After the break opening, graphite gaps are pressurized and afull map of the pressure in the gaps is calculated by the code,e.g., Fig. 22.The strongly asymmetric radial and axial pressure distribu-tion, with pressure peaks in the break region, constitutes themain source of load for neighbouring fuel channels with addi-tional contribution given by dynamic loads in the vicinity ofthe break.Hydraulic loads produce deformation and stresses in individ-ual pressure tubes and graphite stacks as depicted in Fig. 26in relation to pressure tubes.Actual deformation and stresses for pressure tube and graphitebricks are compared with (limit) allowable values.The final result is given in the bottom part of the Fig. 32:five graphite bricks are predicted to be damaged and severalbricks are displaced from their original position. A few chan-nels around the broken one are calculated to have undergoneplastic deformation.
Definitely, the rupture causes displacement of stacks neigh-ouring the broken channel, but no propagation of the rupture isalculated. This is in accordance with the evidence from the NPPvent (NIKIET, 1992) and from recent experiments performedn the TKR full size (related to a group of individual channels)xperimental facility, e.g., Medvedeva et al. (2004b).
.1. Results from the procedure
The proposed MPTR procedure basically reflects the archi-ecture of the U-Stack code, as described into detail by D’Auriat al. (2005). The procedure is based on the adoption of differentodes and different assumptions. The following three elementsf the procedure are distinguished:
I. Calculation of admissible loads: structural analyses areperformed with the aim of determining the maximumadmissible loads on the columns surrounding the initiatingfailure, depending on the failure location over the reac-tor core. An elasto-static model, based on the Ansys finiteelement code, was developed to perform such analyses.
II. Calculation of applicable (or actual) loads: thermal-hydraulic analyses are performed by means of the Relap5code, with the aim of estimating the actual loads acting onthe columns surrounding a failed channel.
II. Comparison between admissible and applicable loads: theresults of the previous two steps are compared so as to iden-tify those initiating failure locations, which could determinea rupture propagation.
ec
ig. 33. FC-BLOCKAGE scenario, area ‘MPTR’: characterization of breakocation inside the cross-section map of the RBMK core needed for the MPTRrocedure (the blue triangle shows ‘cold fuel elements).
The key approximations adopted in the procedure are: (a)o feedback is considered between structural mechanics andhermal-hydraulics calculations; (b) no consideration is giveno plasticity in any place; (c) no explicit dynamic analysiss performed involving the consideration of vibrations, pres-ure waves and fluid-structure interaction (see the third stepelow).
The first step for applying the procedure consists in the geo-etric characterization of the break location that includes (the
reak area has minor influence as already discussed):
The position of the affected fuel channel in the horizontalcross-section map of the RBMK core (Fig. 33).The axial location of the break, calculated according to thediscussion provided in relation to Fig. 13.The azimuthal location of the break, corner-wise or side-wise,right side of Fig. 30. Only side-wise breaks are considered,though the probability of occurrence of side-wise break islow: once a break is predicted at a given axial location thereis a probability around 10−2 that the azimuthal break axiscoincide with the centred orthogonal axis of neighbouringchannels. In this situation, a lower number of stacks constitutean active constraint for the propagation of the rupture, assum-ing no friction among sliding graphite bricks. The side-wisebreak position constitutes a conservative assumption.The distance from the tank of the affected row of channels andthe number of ‘cold’ channels between the broken channel andthe tanks wall, see the sketch in Fig. 33. In facts the Youngmodule of ‘cold’ channels is about 1.2 times larger than theYoung module of ‘hot’ channels.
The second step for applying the procedure addresses thelement ‘I.’ above. A map of admissible loads for each fuelhannel of the concerned RBMK core is created according to: (a)
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he number of stacks that separate the concerned stack from theank, also including the number of ‘cold’ stacks, (b) the distancef the last stack from the tank and (c) the axial elevation of thereak.
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ig. 34. FC-BLOCKAGE scenario, area ‘MPTR’, results from the special procedure rtep actual load; (c) fourth step channels prone to rupture propagation.
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The third step for applying the procedure addresses the ele-ent ‘II.’ above. The differential pressure distribution acting
long the axis of the closest stack is calculated as a function ofime following the break occurrence.
elated to the Smolensk-3 calculation: (a) second step, allowable loads; (b) third
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The fourth step for applying the procedure addresses the ele-ent ‘III.’ above and allows the ‘straightforward’ identification
n the map of the fuel channels that are prone to the propagationf the rupture.
The results from the second to the fourth step are given inig. 34. The following should be noted:
The red colour in Fig. 34a identifies the weakest fuel channels.Various maps have been derived in D’Auria et al. (2005),depending upon the load axial distribution, i.e. concentrated,uniform and peaked, and upon the consideration or not of thetank as resistant element. The data shown in the figure relateto the peaked load and to the presence of the tank.The diagram in Fig. 34b is the result of ‘static’ pressure dif-ference acting upon the opposite sides of the concerned fuelstack. The consideration along the axis of time evolutions likethose shown in Fig. 22 (in the present case obtained by Relap5,see also D’Auria et al., 2008c) allows the achievement of thediagram in Fig. 34b. The integral under the curve is the loadthat has the ‘peaked shape’ already considered for the resultsin Fig. 34a. The dynamic loads expected at the beginning ofthe transient (e.g., pressure wave and jet forces acting dur-ing a few seconds after the break occurrence) are added tothe loads in the form of ‘static’ pressure difference. The sumconstitutes the ‘actual’ load considered for the fourth step.The diagram in Fig. 34c shows the fuel channels that are proneto the propagation of the rupture: for those channels, the actualload is larger than the ‘resistant’ force and the channels mayget plastic condition as a consequence of a rupture occurringin the neighbouring position.
The obtained results show that about 2% of the RBMK corehannels, in the case of the Smolensk-3 NPP, may reach limit-ng condition, i.e. beginning of plastic deformation. However,wo key conservative assumptions have been considered: (1)ide-wise break configuration when calculating the ‘actual’ loadeven in such condition the ‘actual’ load is slightly larger thanhe ‘admissible’ load); (2) no plasticity when calculating theadmissible’ loads.
It is concluded that the MPTR issue is irrelevant for the safetyf the RBMK in the case of the FC-BLOCKAGE scenario, inccordance with the conclusions achieved in section 6.1 by thepplication of the U-Stack code.
.2. Conclusions
Results from best-estimate, coupled 3D structural mechanics,FD, thermal-hydraulics and neutron kinetics calculations forBMK core performance following the blockage of one channelre discussed in the paper. The study requested also the usef ‘severe-accident’ computational tools for the prediction ofydrogen and fission products transport inside the confinementnd to the environment.
The analyses are not supported by uncertainty evaluationnd should not be considered to the level of a licensing study.he main achieved purposes were to demonstrate that no unac-eptable situation is predicted during the considered accident
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volutions (e.g., radioactivity release to the environment belowhe threshold of acceptability) and to demonstrate the availabilitynd the suitability of sophisticate coupled computational tools.
The adopted coupled thermal-hydraulic neutron kineticsodes, namely Relap5/3D-Nestle and Korsar-Bars, were coupledith structural mechanics codes, e.g., Ansys and Katran-Stack and with the codes suitable for predicting fission
roduct transport, e.g., Melcor and Cocosys, respectively.The main conclusions from the analysis of the FC-
LOCKAGE scenario can be summarized as follows:
The scenario puts an enormous challenge to the codes: allkey technological areas relevant to the deterministic reactorsafety are involved. About 40 phenomena have been identifiedas characterizing the scenario and related computational toolshave been evaluated.The key output data from the analysis are (the Smolensk-3RBMK NPP is considered):• Affected bundle power decreases soon after the blockage
event owing to Doppler effect (negative) typically largerthan the positive reactivity effect induced by the channelvoiding.
• One fuel bundle destroyed (composed of two parts).• About 50 kg of irradiated fuel loose their geometric and
structural integrity and (potentially) escape the pressureboundary of the Main Cooling Circuit, but are collectedinside the reactor cavity.
• No challenge put to the structural integrity of the reactorcavity or to any other region of the confinement.
• No danger predicted that is originated by H2 deflagration.• The amount of radioactivity release to the environment,
starting about one hour after the event occurrence, is withinthe regulatory allowed limits.
• A few graphite bricks belonging to the affected fuel channelare damaged with bricks of surrounding column remainingintact.
The possibility for the occurrence of the multiple pressuretube rupture (MPTR) was excluded.
The qualification level of the U-Stack code in addressing thePTR issue as well as of an independent procedure based on
he use of different computational tools has been demonstratednd brought to the conclusion at the last bullet above.
It seems to be expedient to consider the offer related to aonitoring system to prevent the pressure tube rupture of the
ffected channel, see D’Auria et al. (2008e).
cknowledgements
The present paper is devoted to the memory of the eminentussian researcher and technologist Dr. Yuri Cherkashov whoassed away in May 2006. He contributed to the crucial effort ofesigning the fuel channel of the RBMK and was decorated and
ranted a State award for his services. Around fifty researcherst NIKIET and University of Pisa took part in the EC (Euro-ean Commission) Project activities that were at the origin ofhe present one plus five companion papers in this journal issue.
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ost of their names appear as co-authors of the papers or ofhe references. Their contribution is gratefully acknowledged. Aumber of persons provided a managerial support to the activi-ies; among them, we wish to recall V. Shandra and C. Sollima.he work would not have been possible without the contribu-
ion and the willingness of the Russian Beneficiary Institutionosenergoatom to cooperate and to supervise the activities. Spe-ial thanks are due to Dr. E. Hicken and Dr. R.B. Duffey whoook the charge of evaluating all this material and to Profs. M.
azzini and G. Petrangeli for their continuous supervision ofhe activities. Neither the EC nor any person acting on behalf ofhe Commission is responsible for the use which might be madef the information in the paper and the views expressed are theole responsibility of the authors and not necessarily reflect theiews of EC.
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