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Nuclear Engineering and Design 240 (2010) 1699–1706

Contents lists available at ScienceDirect

Nuclear Engineering and Design

journa l homepage: www.e lsev ier .com/ locate /nucengdes

pent fuel transport cask thermal evaluation under normal and accidentonditions

. Pugliese, R. Lo Frano ∗, G. Forasassiepartment of Mechanical, Nuclear and Production Engineering, University of Pisa, Via Diotisalvi, n◦2-56126 Pisa, Italy

r t i c l e i n f o

rticle history:eceived 7 October 2009eceived in revised form 11 February 2010ccepted 23 February 2010

a b s t r a c t

The casks used for transport of nuclear materials, especially the spent fuel element (SPE), must bedesigned according to rigorous acceptance criteria and standards requirements, e.g. the InternationalAtomic Energy Agency ones, in order to provide protection to people and environment against radiationexposure particularly in a severe accident scenario.

The aim of this work was the evaluation of the integrity of a spent fuel cask under both normal andaccident scenarios transport conditions, such as impact and rigorous fire events, in according to the IAEAaccident test requirements. The thermal behaviour and the temperatures distribution of a Light WaterReactor (LWR) spent fuel transport cask are presented in this paper, especially with reference to theItalian cask designed by AGN, which was characterized by a cylindrical body, with water or air inside theinternal cavity, and two lateral shock absorbers.

Using the finite element code ANSYS a series of thermal analyses (steady-state and transient thermalanalyses) were carried out in order to obtain the maximum fuel temperature and the temperatures fieldin the body of the cask, both in normal and in accidents scenario, considering all the heat transfer modesbetween the cask and the external environment (fire in the test or air in the normal conditions) as wellas inside the cask itself.

In order to follow the standards requirements, the thermal analyses in accidents scenarios were alsoperformed adopting a deformed shape of the shock absorbers to simulate the mechanical effects of aprevious IAEA 9 m drop test event. Impact tests on scale models of the shock absorbers have already beenconducted in the past at the Department of Mechanical, Nuclear and Production Engineering, Universityof Pisa, in the ‘80s. The obtained results, used for possible new licensing approval purposes by the Italian

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competent Authority of tare discussed.

. Introduction

The perspective of a worldwide “nuclear renaissance” for elec-ricity production, is also related to the nuclear spent fuel issues inerms of both short- and long-term management. Nevertheless theroblem of how to manage spent fuel in the interim, before a per-anent solution is agreed upon, is widely discussed in all countries

hat use nuclear power for the reason that it is important to under-tand all the issues involved in spent fuel storage and transportationo make the best choices possible. Moreover nuclear power genera-ion is recommended as a promising way to contribute to long-term

nergy supply and as a measure against global warming.

In this context, waiting for the future deployment of the todaynder development Generation IV reactors projects, as well as of alosed fuel cycle, the spent nuclear fuel (SNF) represents an open

∗ Corresponding author. Tel.: +39 050 2218093; fax: +39 050 2218065.E-mail address: rosa.lofrano@ing.unipi.it (R.L. Frano).

029-5493/$ – see front matter © 2010 Elsevier B.V. All rights reserved.oi:10.1016/j.nucengdes.2010.02.033

sk for PWR spent fuel cask transport by the Italian competent Authority,

© 2010 Elsevier B.V. All rights reserved.

issue to be explored thoroughly the used uranium fuel is widelyreferred in literature to as “spent fuel”.

Spent nuclear fuel is the nuclear fuel that has been irradiatedin a nuclear reactor during the operation of nuclear power plants(NPP) to the point where it is no longer useful in sustaining a nuclearreaction. Historically, the generated spent nuclear fuel is designatedas useful recyclable energy resource. SNF shall be properly storeduntil the reprocessing, that allows to chemically separate the valu-able material, such as uranium or plutonium, from the waste, wasdone. Moreover intermediate storage of spent fuel is important asa means for contributing to the flexible operation of the overallnuclear fuel cycle.

In fact the design of the packaging system is strictly related tothe need to protect the population and environment from exposure

to the radiation emitted by the radioactive materials contained inthe spent fuel. Therefore a cask packaging system must ensure that:

• The spent fuel remains contained even under severe accidentconditions;

1700 G. Pugliese et al. / Nuclear Engineering and Design 240 (2010) 1699–1706

Nomenclature

AGN Agip nuclearFEM finite element methodIAEA International Atomic Energy AgencyNPP nuclear power plantSNF spent nuclear fuelSPE spent fuel element

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PWR pressurized water reactorTE environmental temperature

The radiation levels at the surface of the container during normaltransport and under accident conditions are low;The transported spent fuel cannot accidentally undergo a nuclearfission reaction.

The use of a spent fuel packaging system, designed accordingo rigorous acceptance criteria and standards requirements, e.g.he International Atomic Energy Agency (IAEA) ones (Internationaltomic Energy Agency, 2005), is also required to allow the trans-ortation of SNF inside and away from the NPP site.

This paper is intended to evaluate the long-term integrity ofetal canister and actual materials (ensuring their safety func-

ions), which characterize the spent fuel casks, as well as to assesshe potential damage, if any, to the various components that mayomprise the cask itself, under both normal transport condition andccident scenarios (Rains, 1999; Lee et al., 2004), such as impact andigorous fire events (Chun and Ryu, 2000) in according with theentioned IAEA cask transport packaging test requirements and

ecommendations (that also serves as the world’s principal inter-overnmental forum for scientific and technical cooperation in theuclear field).

Moreover the thermal effects on the deformed shape of thepent nuclear fuel packaging subsequent to a free drop (or impact)ere also investigated and analyzed.

. Description of a cask packaging system

International requirements regulate commercial nuclear powereactors, non-power application of radioisotope research, fuel cycleacilities as well as the design, manufacture, use and rules governingnd instructions of the packaging for a low/high-level radioac-ive nuclear materials. The basic criteria for packages for shippingigh-level nuclear materials and spent fuel originated in 1946 andere based on the National Academy of Sciences recommenda-

ions, which represented a guidance for manufacture of the earlyhipping casks for spent fuel, and have been adopted by the Inter-ational Atomic Energy Agency. In according with this regulatoryequirement, nuclear power plants store spent fuel in enclosed typ-cal container, referred to as a cask (cooling pools and, in some cases,n dry storage casks), to await shipment to a temporary storage orermanent disposal facility.

Nuclear waste containers or packaging systems are an impor-ant factor in the design, cost and safe operation of nuclear wasteepositories (Kar et al., 2008). Spent nuclear fuel refers to uranium-earing fuel elements that have been used in nuclear reactors andhat are no longer producing enough energy to sustain a nucleareaction.

The spent fuel assemblies are, however, able to generate sig-

ificant amounts of radiation and heat; for this reason spent fuelust be shipped in containers or casks that shield and contain the

adioactivity and dissipate the heat.In NPP spent fuel is stored in enclosed cooling pools and, in

ome cases, in dry storage casks to await shipment to a tempo-

Fig. 1. Spent fuel packaging scheme.

rary storage, permanent disposal facility or reprocessing facility. Anessential component for any safe transportation is a robust spentfuel container, or “cask”, an example of which is shown in Fig. 1(U.S. Nuclear Regulatory Commission, 2003).

Prior to transportation, SNF is handled from its temporary dis-posal pool and accommodated in a canister of stainless steel. Thecanister (weld-sealed or open) is put into a cask. The sealing per-formance (if needed) of the canister has to be sustained during theinterim storage period.

The cask (Fig. 1) is typically constituted of a steel cylinder bodywith either welded or bolted closure ends. The steel cylinder pro-vides a leak-tight containment of the spent fuel. Inside the caskcavity, the fuel basket locates and supports the tubes of corrosionresistant material, being adapted to contain the spent nuclear fuelrod assemblies, in fixed positions.

The basket (horizontally or vertically oriented depending onthe cask design) is surrounded by the inner steel shell/cylinder,the additional concrete, the outer steel shell in order to provideadequate radiation shielding to workers and members of the public.

Moreover pressurized helium gas is, generally, used inside thecask to promote heat removal from the fuel assemblies to the caskwall, while air on the outside removes the heat.

Due to the radioactive characteristics of stored spent fuel, thecask shall provide enough shielding to reduce external radiation tolow carried out acceptable levels, according to the radiation stan-dards during transport (in general the transport must be safely inlarge, heavy containers that shield the public from exposure).

In addition one of the most important objective for a spentnuclear transportation cask design is to remove decay heat fromthe fuel array and maintain the peak cladding temperature belowthe design limits (Lee et al., 2009). Due to the fact that this limittemperature is usually not accessible to measurement, the designlimit of cask depends and must be below the materials’ meltingpoint.

Standards requirements cover both normal transport conditionand accidents scenarios such as impact tests and severe fire event.The spent fuel storage and transport cask must therefore withstandvarious accident conditions involving impact, puncture, fire, andsubmersion (Lee et al., 2004). To reduce the effects of the impactthe SNF casks are normally encased in energy absorbing structuresor protected by suitable impact limiters.

International Atomic Energy Agency (IAEA) cask transport pack-aging test requirements include a 9 m (30 feet) drop onto a flatunyielding surface and subsequent full exposure to an engulfing fire

for 30 min (fire test) or to an environment at 800 ◦C temperaturefor a numerical simulation or for a furnace test.

This paper deals with a cask, designed in Italy by AGN (Fig. 2)consisting of a cylindrical body, horizontally oriented during the

G. Pugliese et al. / Nuclear Engineering and Design 240 (2010) 1699–1706 1701

-1 cas

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Fig. 2. AGN

ransportation, with water or air in the internal cavity of the cask,.e. the primary cooling system, in which the fuel element basketsre inserted. This cask had been certified after the mentioned IAEAegulation for LWR spent fuel transport in the ‘80s (Aquaro andorasassi, 1983).

The AGN-1 cask configuration, like shown in Fig. 2, is a multibar-

ier confinement and containment system, having 54 t of weight,m length and a maximum diameter of 1.5 m. It is characterizedy a main body of carbon steel, internally and externally coatedith a stainless steel liner. Two shock absorbers (or impact lim-

Fig. 3. AGN-1 FEM models without air/water (a), and with air/w

k scheme.

iters) of stainless steel with elliptical shape, situated at the sidesof the body allow to reduce the mechanical effects of possibleimpacts, in the event of an accident. In fact these limiters mightcrush, absorbing impact forces and protecting the container and itscargo.

The external cask surface is also characterized by an extensive

cooling fins distribution which allow to increase the heat transfercapability and to enlarge the heat exchanger surfaces (secondarycooling system). Moreover a liquid shielding barrier is located inthe gap between the cask body and the cooling fins.

ater in not deformed (b) and deformed (c) cask shapes.

1 ring and Design 240 (2010) 1699–1706

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Fig. 4. AGN-1 detailed cask scheme.

702 G. Pugliese et al. / Nuclear Enginee

The fuel elements are stored inside stainless steel sealed basketsontaining water. Afterwards the fuel baskets, containing the SNFod assemblies, are correctly inserted inside the cask internal cavityn fixed positions, passing through appropriate canister channels.

In this investigation, the thermal behaviour and the distributionf the temperature of the AGN-1 PWR spent fuel transport cask,valuated both in mentioned (normal and accident) conditionsccording to the IAEA accident test requirements, are presented.

. Thermal analysis methodology

From the time the spent nuclear fuel elements are loaded intohe cask until the end of the temporary storage, the container

aterial is subjected to irradiation mainly due to �s, temperaturend pressure excursions, potential pyrophoric hazard and corrosivenvironments, etc.; an adequate design of the packaging is required.herefore complex physical phenomena should be considered toccurately describe the temporary storage and transportation con-itions.

In this paper the structural effects and thermal performances ofhe AGN-1 cask, under the normal and hypothetical accident condi-ions of transport, were evaluated adopting a numerical approachn order to verify the cask capability to withstand the consideredransport conditions without damages on the considered packag-ng system. In addition it was verified the decay heat removal wasnsured, so that the temperatures of the spent fuel and of cask com-onents remained within the allowable limits (e.g. the temperaturef cladding oxidation).

.1. AGN-1 FEM model description

In the present paper a deterministic approach was used to eval-ate the thermal analysis effects, performed for normal, off-normal,nd accident conditions. To attain the intent both steady-state andransient thermal analyses, considering also the maximum spentuel decay heat load approved for this cask design, were carriedut. The finite element code ANSYS was used in order to determinehe temperature distribution which might arise in the event of fire.

The first task in the adopted methodological approach was,herefore, the setting up and the implementation of a suitable 3-Dnite element model, representing, in as much detailed as possible,he real and complex geometry of the AGN-1 packaging system.

The set up AGN-1 cask model was characterized by double sym-etry (transversal and longitudinal one). In Fig. 3(a) it was clearly

epresented the FEM model of AGN-1 cask with all the previouslyescribed internal and external structures, such as the fuel bas-et, the closure lid, etc., without the air or water inventory. In thehown model also the basket components, consisting of the basketlements, the connecting components (i.e. separation grids), theandling elements, etc. were implemented. Moreover in Fig. 3(b)nd (c) it was shown the same model, implemented for both noteformed and deformed shock absorbers shapes/characteristics,ith the air or water inventory.

In order to achieve a reduction of the computational time it wasssumed to neglect the geometrical differences between the clo-ure lid and the bottom of the cask cavity, as it is represented inig. 4.

Solid and thick shell thermal elements type were chosen inrder to simulate all heat transfer mechanisms inside the cask and

etween the cask itself and the environment. The fuel elements, 28ruciform specimen inside each basket, were assumed as thin cylin-rical elements, like represented in Fig. 5. The AGN-1 FEM modellements number, equal about to 135,000, was found appropriateo achieve the needed accuracy required by the analyses.

Fig. 5. Spent fuel elements components inside the AGN-1 cask FEM model.

The carried out thermal analyses were considered as a supportfor the new licensing approval by the Italian competent Authority ofthe cask for PWR spent fuel transport. To match this intent, severalimportant features were taken into account to attain thermal caskperformances of the cask, like the following:

• The package geometry and materials;• The structural and mechanical features that may affect heat trans-

fer, such as cooling fins, insulating materials, surface conditions ofthe package components, and gaps or physical contacts betweeninternal components.

According to the storage concept, the heat decay of the spentfuels is removed through canister walls by dry/wet cooling; long-term integrity of the cask components must be therefore ensuredand, in particular, the cask must have sufficient heat removal per-formance.

The simulated thermal behaviour of cask was studied assum-ing that all the AGN-1 cask components, such as the canister orthe basket, etc., were exposed to various temperature conditionsdepending on decay heat during the storage period (Huang, 1996).Moreover both the wet and dry fuel storage, inside the cask cav-ity, were analyzed; in the case of dry storage design the dissipationof the residual heat, generated by the stored spent fuel elementsthemselves, was one of the main safety thermal issues for licens-

ing. Actually, in the dry or wet storage systems, the cask mightact as the final barrier to encapsulate spent fuels and radioactivematerials.

In what follows steady-state and transient analyses, accordingto the IAEA requirements, are reported and analysed, highlighting

ring and Design 240 (2010) 1699–1706 1703

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Table 1Thermal analyses results (TE = 38 ◦C) with not deformed cask shape.

Item Maximum temperature [◦C]

Wet storage Dry storage

Impact limiter 77 77Overpack (outside) 79 77

deformed shape of the shock absorbers, in order to simulate themechanical effects of the previous mentioned IAEA 9 m drop testevent and a 30 min fire exposure (event characterized by a firetemperature equal to 800 ◦C with an emissivity of 0.9 and a caskadsorbivity equal to 0.8).

G. Pugliese et al. / Nuclear Enginee

he cooling down conditions of the cask after the fire event for aime long enough to achieve again the steady-state condition.

Afterwards the methodological approach may be summarizedn the following several important steps:

Appropriate finite element modelling of packaging componentswith their real geometry and temperature-dependent mate-rial characteristics (models previously shown in Fig. 3(b) and(c) representing the pre- and post-impact test configurations,respectively indicative of the normal and accident conditiontransportations);Assumptions in representing heat sources, heat transfer pathsand modes and thermal properties for the package materials;Setting up of suitable expressions for the conductive, the convec-tive and the radiation heat transfer modes among the packagecomponents as well as the thermal boundary conditions;Evaluation of structural and thermal effects under normal condi-tions and accident conditions transportation (without and withdeformed shock absorber shape);

The heat transfer process in the cask is very complicated as aonsequence of the inherent complex cask geometry and of thenduced convection processes, which is dependent on the spent fuelecay heat, the thermal boundary condition, the orientation of theask, etc. (Heng et al., 2002). Therefore to determine in a detailednd realistic way, the cask thermal performance, all heat trans-er modes (conductive, convective and radiation ones) and thermaloads were taken into account, in the carried out thermal analyses,y means of adequate values and hypotheses, considering that:

. The conductive coefficient values were consistent with the caskmaterial properties and temperature;

. The convective coefficient values were considered negligible forair or water, in the hypothesis of the absence of fluid turbulenceinside the cask, while those ones representative of heat trans-fer mode between the impact limiters-cask or the cooling finssurfaces and the external environment were assumed in accord-ing respectively with Eqs. (1) and (2) (experimental correlation)(Agip Nucleare, 1982):

hf 1 = 1.23 ∗ (�t)1/3 [W/m2◦C] (1)

hf 2 = 0.835 ∗ (�t)1/3 [W/m2◦C] (2)

. The fluid region was characterized by conduction and radiationmodes;

. The radiation heat transfer mode was assumed to be character-ized by an emissivity values equal to 0.5 and 0.8, respectively fornormal and accident transport conditions, for the cask compo-nents;

. The heat flux values were conservatively determined assumingthat the solar irradiation on the external cask surface (in agree-ment with IAEA rules) was maximum, for a time period equal to12 h.

. The residual heat power decay of each fuel element, at the max-imum burn up level, was assumed equal to 36.5 W.

.2. Evaluation of thermal effects on the cask under normalransport conditions

The implemented packaging model was firstly analysed assum-ng a steady-state thermal condition. In particular, the structural

ntegrity for normal cask transportation was evaluated consideringhe internal heat loads as well as the thermal boundary conditionsue to environment temperature.

The numerical simulations were performed in order to evaluatehe effects of a normal transport of the cask exposed to environ-

Gasket 77 78Overpack (inside) 79 79Canister 81 88Fuel rod 81 88

mental temperatures (TE) of −10 ◦C, 0 ◦C and 38 ◦C, even if in thisstudy only the normal (38 ◦C) and accident conditions are analyzedand discussed (some results of which are summarized in Table 1).

Moreover, in Figs. 6 and 7 the obtained cask temperature dis-tributions, considering an environmental temperature of 38 ◦C, fordry and wet transportations, are shown. The obtained results high-lighted that the maximum values of temperature are located incorrespondence of the fuel elements basket and of the cooling fins.

Moreover it was observed that, even if the results seemed quitesimilar, the wet transport condition allowed to obtain lower tem-perature values. The shown results highlighted, in fact, that the drytransport condition in respect to the wet one resulted in a moreoverheated basket components with fuel temperature values equalto about 88 ◦C.

It is important to note that the temperature propagation insidethe cask, between the fuel basket and the outer steel shell, seemedto be mainly influenced by the convection mode. Anyway in allthe considered transport cases the safety of cask components isensured.

3.3. Evaluation of thermal effects on the cask under accidenttransport conditions

To determine the potential damage to the various componentsthat might compromise the cask integrity deformed shock absorberconfigurations (after drop test) were implemented and assumed inthe numerical simulations.

In order to follow the standards requirements, preliminary ther-mal analyses in accidents scenarios were performed adopting a

Fig. 6. AGN-1 temperature distributions for normal (38 ◦C) dry transport.

1704 G. Pugliese et al. / Nuclear Engineering and Design 240 (2010) 1699–1706

utions for normal (38 ◦C) wet transport.

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Fig. 7. AGN-1 temperature distrib

The prescribed hypothetical sequence of accidents, fully engulf-ng pool fire lasting 30 min with flame temperature no less than00 ◦C is designed to be more severe than 99% of all transporta-ion accidents, and is assumed to conservatively bound all credibleccident scenarios in the transport of spent nuclear fuel. More-ver a total leakage of the liquid shielding barrier contained in theap between the cask body and the cooling fins is assumed as aonsequence of the impact.

The deformed cask shape adopted to perform the accident anal-sis is obtained from the actual carried out impact tests performedn the AGN-1 cask. These tests on scale models of the shockbsorbers have already been conducted in the past (‘80s) at theniversity of Pisa (Fig. 8) (Aquaro and Forasassi, 1983).

Therefore the transient thermal analyses were carried out onn appropriate deformed cask FEM model (for both dry and wetonditions), reproducing the maximum registered shock absorberseformation in order to evaluate the thermal stresses and tem-erature behaviours, which can occur either during or after there.

Furthermore solar heat flux was also considered in the per-ormed analyses, applied with 400 W/m2, 800 W/m2 and 200 W/m2

or 12 h per day, respectively, for curved, horizontal and vertical

urfaces (IAEA Safety Series TS-R-1, 2005).

It is worthy to note that the cask steady-state temperature field,reliminarily obtained assuming a deformed shock absorbers shapet an environmental temperature of 38 ◦C, was used as input initial

Fig. 8. Vertical and lateral AGN-1 drop test.

Fig. 9. AGN-1 steady-state temperature field at an environmental temperature of38 ◦C with cask deformed shape.

condition in the transient thermal analyses to evaluate the tem-perature distribution inside and along the cask in the fire accident

events condition (Fig. 9).

Moreover in Tables 2 and 3, the temperature values, up to 0.5 hlater the occurrence of the fire event was also summarized.

A further evaluation of the cask cooling down condition, in termsof temperature distributions, up to 3 h later the occurrence of the

Table 2Thermal analyses results (TE = 800 ◦C) with not deformed cask shape.

Item Maximum temperature [◦C]

Wet storage Dry storage

Impact limiter 701 701Overpack (outside) 630 630Gasket 124 124Overpack (inside) 104 106Canister 90 92Fuel rod 81 88

G. Pugliese et al. / Nuclear Engineering and Design 240 (2010) 1699–1706 1705

Table 3Thermal analyses results (TE = 800 ◦C) with deformed cask shape.

Item Maximum temperature [◦C]

Wet storage Dry storage

Impact limiter 701 702Overpack (outside) 631 631Gasket 128 129

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Overpack (inside) 108 97Canister 95 90Fuel rod 84 89

re event was also carried out. The obtained results are representedn the following figures, only for the dry transport condition. Inig. 10, the temperature behaviour after half an hour of fire expo-ure in agreement with the IAEA rules is shown. Although thexposure for 30 min at 800 ◦C the temperature distribution insidehe cask was evaluated to be about 8 times lower than the external

nes.

The maximum temperature value experienced by the fuel bas-et resulted to be lower than the one at which SNF cladding lossts integrity (593–740 ◦C) and, hence, not enough to determine

ig. 10. AGN-1 temperature distributions in the cask (a) and in the fuel rods (b) forccident condition after half an hour of fire exposure.

Fig. 11. AGN-1 temperature distribution post-accident condition.

severe structural damages. Moreover the temperature reached bythe closure lid in correspondence of the sealing during the accidentsequence was calculated and resulted equal to about 176 ◦C andtherefore no sufficient to attain a reduction of bearing and sealingcapability of the seals themselves due to their characteristics.

Fig. 11 represents the temperature distribution calculated 3 hafter the accident event. It is possible to observe an overall reduc-tion of the temperature values on the external cask surfaces due tothe cooling condition operated by the external air. Otherwise theinner packaging system components suffer a lower increase of thetemperature; in fact the high thermal inertia does not allow to dis-sipate, in the considered lasting time, the heat absorbed during thefire exposure.

Reviewing the obtained temperature values, it possible toremark that the cask packaging system and all its internal com-ponents resulted to be safely maintained within their minimumand maximum temperature criteria, for normal, off-normal, andaccident conditions, and support the performance of the intendedsafety function.

3.4. Validation of thermal analyses

The need of a numerical approach validation was felt necessaryin order to take into account the uncertainties that unavoidablyaffects the analyses, due to the assumed input parameters and theapproximations introduced in setting up FEM models.

To attain the intent, the same AGN cask structure was alsoimplemented with FLUENT code (ANSYS FLUENT®, 2003) at theUniversity of Pisa, preserving the material properties, boundaryconditions and the heat transfer modes. FLUENT code was chosenbecause it is generally used to perform thermal-fluid flow analysisand solve problems with a variety turbulence, radiation, and heattransfer models, including flows in complex geometries.

The validation analysis was hence carried out with theintent to point out and demonstrate the reliability and perfor-mance/accuracy of the adopted ANSYS® code (ANSYS® Software,2007) to correctly represent the thermal heat transfer modes.

The comparison among the obtained results in terms of temper-ature, in the case of cask in the fire accident events condition, up to

0.5 h later the occurrence of the fire event, highlighted a discrep-ancy less than 3% (as visible in Fig. 12) demonstrating the accuracyof modelling and the insensitivity of the model in respect to thecode used to perform the transient thermal analyses.

1706 G. Pugliese et al. / Nuclear Engineering and Design 240 (2010) 1699–1706

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the cask from the multi-canister overpack-handling machine with air cushion”,SNF-5276 Rev. 0, Site-Wide Nuclear Safety Project. Engineering Report.

Fig. 12. Comparison between temperature dist

Considering that in the analyzed thermal scenarios the decayeat removal systems were passive, the maximum carried outemperature values were lower than the allowable limits for caskackaging system components (i.e. “the fuel cladding temperaturehould also generally be maintained below 570 ◦C for short-termff-normal, short-term accident, and fuel transfer operations (e.g., vac-um drying of the cask or dry transfer” (U.S. Nuclear Regulatoryommission, 2000)) ensuring, therefore, the integrity of the con-idered packaging system.

. Conclusion

Numerical thermal analyses of a packaging system were carried,ith reference to the Italian AGN-1 PWR spent nuclear fuel one in

rder to support the requirements of the a possible new licensingy the Competent Authority.

Adopting a numerical approach the relevant heat transfer mech-nisms between the spent nuclear elements inside the body and thexternal environment (fire in the test or air at 38 ◦C in the normalonditions) were analysed. Therefore a series of steady-state andransient analyses were performed taking into account all the heatransfer modes between the cask and the environment and insidehe cask itself were taken into account.

In order to follow the standards requirements, the thermal anal-ses in accidents scenarios were performed adopting a deformedhape of the shock absorbers to simulate the mechanical effects ofprevious IAEA 9 m drop test. The obtained results for the normal

ransport condition case highlighted that the maximum values ofemperature, located in correspondence of the fuel elements bas-et and of the cooling fins, are not enough to determine structuralamage of the cask.

As for the accident scenario it was observed that the high ther-al inertia of the packaging system does not allow to reach relevant

emperatures in the AGN-1 cask cavity also considering 30 min ofre exposure. The post-accident analyses results, after 3 h of cool-

ng down condition in air at 38 ◦C, highlighted that only the outerarts of the cask have dissipated the heat absorbed during the firexposure, due to the cask thermal inertia in the heat transfer pro-esses.

In addition it was also demonstrated the capability of ANSYSrogram to manage heat transfer processes and loads coupled toomplex geometry and various materials, in order to attain use-ul information to support the structural analysis, as well as theeliability of adopted ANSYS code.

on up to 0.5 h after the beginning of fire event.

Finally, the performed thermal analyses on AGN-1 cask systemindicated that thermal damages are avoided because the temper-ature experienced by the internal cask components, during thesimulated engulfing fire event, were about 120 ◦C for the gasketand about 90 ◦C for the fuel canister and rods and well below thecladding limit temperature, which is equal to about 570 ◦C. Conse-quently the AGN-1 cask design provided reasonable assurance ofadequate protection to people and environment against radiationexposure and possible leakage during transport.

Acknowledgments

The authors would thank Dr. N Forgione (University of Pisa) tokindly supply Fluent thermal analysis results, which were used inthe present work to validate the cask packaging system thermalanalysis methodological approach.

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