Research ArticleSimulation of a TRIGA Reactor Core BlockageUsing RELAP5 Code

Patriacutecia A L Reis12 Antonella L Costa12 Claubia Pereira12

Maria Auxiliadora F Veloso12 and Amir Z Mesquita3

1Departamento de Engenharia Nuclear Escola de Engenharia Universidade Federal de Minas Gerais6627 Antonio Carlos Avenue 31270-901 Belo Horizonte MG Brazil2Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares InovadoresCNPq Brazil3Centro de Desenvolvimento da Tecnologia NuclearComissao Nacional de Energia Nuclear Campus UFMG6627 Antonio Carlos Avenue 31270-901 Belo Horizonte MG Brazil

Correspondence should be addressed to Antonella L Costa antonellanuclearufmgbr

Received 17 October 2014 Revised 21 January 2015 Accepted 22 January 2015

Academic Editor Giorgio Galassi

Copyright copy 2015 Patrıcia A L Reis et al This is an open access article distributed under the Creative Commons AttributionLicense which permits unrestricted use distribution and reproduction in any medium provided the original work is properlycited

Cases of core coolant flow blockage transient have been simulated and analysed for the TRIGA IPR-R1 research reactor using theRELAP5-MOD33 code The transients are related to partial and to total obstruction of the core coolant channels The reactorbehaviour after the loss of flow was analysed as well as the changes in the coolant and fuel temperatures The behaviour of thethermal hydraulic parameters from the transient simulations was analysed For a partial blockage it was observed that the reactorreaches a new steady state operation with new values for the thermal hydraulic parameters The total core blockage brings thereactor to an abnormal operation causing increase in core temperature

1 Introduction

Operation of a research reactor is characterized by humanactions and interventions on a daily basis whether the reactoris at power operation or shutdown Even with preventivemeasures human actions or errors can still cause an initiatingevent [1] The safety analysis of research reactors includessimulations of selected cases classified by the InternationalAtomic Energy Agency since the simulations are performedusing nodalizations verified and validated by users of interna-tionally recognized codes [1] The thermal hydraulic analysisis considered as an essential aspect in the study of safetyof nuclear power and research reactors since it can pre-dict proper working conditions steady state and transientthereby ensuring the safe operation of a nuclear reactor[2] Among thermal hydraulic accidents are the loss of flowaccident (LOFA) and loss of coolant accident (LOCA)

The aim is to investigate the behaviour of a TRIGAresearch reactor after partial and total coolant flow blockagein the thermal hydraulic channels (THC) characterizing

a LOFA transient The transients were simulated using aRELAP5 model LOFA constitutes one of the most severeaccidents that may occur during a research reactor lifetime[3] The transient was simulated considering reactor poweroperation at 100 kW and 265 kW that are the main IPR-R1power operation levels

IPR-R1 General Description The IPR-R1 is a reactor typeTRIGA (Training Research Isotope General Atomic)Mark-I model manufactured by the General Atomic Com-pany and installed at Nuclear Energy Development Centre(CDTN) of Brazilian Nuclear Energy Commission (CNEN)in BeloHorizonte Brazil It is a light water graphite-reflectedopen-pool type research reactor Since 1970 IPR-R1 worksat 100 kW but it is ready to operate at power of 250 kWIt presents low power and low pressure for application inresearch training and radioisotopes productionThe reactoris located in a 6625mdeep pool with 192m of internal diam-eter and filled with demineralized light water A schematicreactor diagram is illustrated in Figure 1

Hindawi Publishing CorporationScience and Technology of Nuclear InstallationsVolume 2015 Article ID 354163 10 pageshttpdxdoiorg1011552015354163

2 Science and Technology of Nuclear Installations

Tower

Orifice platepressure transmitter

FlowmeterGM

(primary)Pump

Pump

GM (outlet)

Flowmeter

Heat exchanger

GM (area)

GM (pool)

Forcedcirculation

Naturalcirculation

Core

Reactortank

Conductivity sensorIon (3) and fission (1) chambersGM Geiger MullerControl rod position

Level meter

Temperature sensor

Level

Secondarycooling loop

Primarycooling loop

6417

mm

GM (control room)

Tsoil

Tout channel

Tin channel

Tin

Tout

Tin

Tout Tair

TinTout inlet and outlet temperature inrelation to the heat exchanger

TPT100

Figure 1 Schematic representation of the TRIGA IPR-R1 [17]

The water in the pool works as coolant moderator neu-tron reflector and it is able to assure an adequate biologicalradioactive shielding The reactor cooling occurs predom-inantly by natural convection with the circulation forcesgoverned by the water density differences The removal ofthe heat generated from the nuclear fissions is performedby pumping the pool water through a heat exchanger Thecore has a radial cylindrical configuration with six concentricrings (A B C D E and F)with 91 positions able to host eitherfuel rods or other components like control rods reflectorsand irradiator channel There are in the core 63 fuel elementsconstituted by a cylindrical metal cladding filled with ahomogeneous mixture of zirconium hydride and uranium20 enriched in 235U isotopeThese fuel elements have threeaxial sections an upper and a lower reflector (graphite) andthe central portion filled with fuel (U-ZrHx) [4] The radialreflectors elements are coveredwith aluminumandfilledwithgraphite having the same dimensions of the fuel elements

The control rods are composed by boron carbide andaluminum cladding In the center of the reactor there is analuminum tube (central thimble) for irradiation of experi-mental samples This tube is removable and when it is notin use the reactor pool water fills its volume Furthermorethe core has an annular graphite reflector with aluminumcladding Such annular reflector has a radial groove where arotary rack is assembled for insertion of the samples to irradi-ation In such rotary rack it is possible to place the samples in40 different positions around the core Moreover tangent to

annular reflector there is a pneumatic tubewhere the samplesalso can be inserted for irradiation

TRIGA reactors fuel has an important characteristic ofintrinsic safety with large prompt negative coefficient ofreactivity that effectively controls the reactor in cases of pos-itive reactivity insertions In this way any sudden reactivityaddition causes an increase in power which heats the fuel-moderator material (U-ZrHx) instantaneously causing thenumber of fissions to immediately decrease because of neu-tron energy spectrum hardening within the fuel pin

2 IPR-R1 Nodalization

21 Reactor Modelling The radial power distribution shownin Figure 2(a) was calculated in preceding works using theWIMSD4C and CITATION codes [5 6] and also experimen-tal data [7] The radial factor is defined as the ratio of theaverage linear power density in the element to the averagelinear power density in the core

RELAP5 nodalization for the IPR-R1 verified in previouswork (with 91 THC) [8] was utilized in this simulation Thecore model has 91 hydrodynamic channels (see Figure 2(b))The complete reactor nodalization in the RELAP5 is pre-sented in Figure 3 [8ndash10] To simulate the cross flow amongthe THCs a total of 186 single junctions were used intercon-necting selected volumes at different high positions of thenodalized channels These junctions have been added to themodel to observe the behaviour of the coolant inside the core

Science and Technology of Nuclear Installations 3

D7

D8

D9D10D11

D12

D13

D14

D15

D16

D17

D18 D1 D2

D3

D4

D5

D6

C2C1C12

C11

C10

C9

C8 C7 C6C5

C4

C3B2

B1

A1B6

B5B4

B3

068

F1 F2 000000

000

000

000

000000

000

000

000

000

000

000

065

000

000

064065

081

080

080

080

080

083

083

084

084

084

084

085

085

079

081

099099

099

099

099

100

101

107113

107

102

102

101

100

100

131

121

133

142

141

128

103109

153

141

142

119

131

132

130

155

128

131

103

081081082

081

084

084082

082

082

066

000

000

000

000

000000000000

000

000

000

000

000

F3F4

F5

F6E5

E4E3

E2E1E24E23

E22

E21

E20

E19

E18

E17

E16E15

E14 E13 E12E11

E10

E9

E8

E7

E6F7

F9

F8

F10

F11

F12F13

F14F15F16F17

F18F19

F21

F20

F22

F23

F24

F25

F26

F27

F28F29

F30

B1155 A1

Radial power factor

LocalizationFuel element aluminium

Graphite element

Fuel element stainless steel

Control rods C1 C7F16

Neutrons source

Central trap

Other types of components GraphiteFuel Element

91-THC

(a) (b)

Figure 2 IPR-R1mdashradial relative power distribution (a) model of the 91 THC in the RELAP5 code (b)

340 350 160 144

Pump Valve

HEX820

TDV TDV

720

SV 060

201 291

SV 100

10090807 0605

0403

02

01

TDV 500

SJ 059

Pool 050Pool 020

SJ 049

SJ 101SJ 099SJ 098SJ 097SJ 096SJ 095SJ 094SJ 093

SJ 092

SJ 091

132330

Core91 THC

63 heatstructures

201ndash291

Figure 3 IPR-R1mdashNodalization in the RELAP5-MOD33 code

after the THCblockage transientThis nodalization is capableof simulating situations of forced coolant recirculation andalso natural circulation

22 Heat Structures Modelling Each of the 63 fuel elementswas modeled separately and 63 heat structure componentsdivided in 21 axial levels were consideredThe heat structures

in RELAP5 permit the calculation of heat across the solidboundaries of the hydrodynamic volumes Radially the heatstructures were divided in 17 mesh points considering theactual radius of the fuel (heat source) gap and cladding Thefuel elements were modelled using cylindrical geometry andthe characteristics of thematerials were inserted to reproducethe heat transfer For each material specified corresponding

4 Science and Technology of Nuclear Installations

thermal property data were entered to define the thermalconductivity and volumetric heat capacity as functions oftemperature

The axial power distribution was calculated consideringa cosine profile predicted using the data from Figure 2 andtaking into account the power cutoff in the extremes of theelement due the presence of the graphite The point kineticsmodel was used in the current modeling which assumes thatchanges in reactivity uniformly affect the core flux resultingin no relative spatial variation over time Point kinetics is notindicated when a pronounced flux spike and radial neutronwave occur

3 THCs Blockage Transient

THCs blockage transient has been investigated using theRELAP5 code For the open pool research reactors configu-ration the probability of a blockage in the core upper plenumzone is higher than a blockage from the bottom [11] Thistransient situation may be caused for example by swelling ofthe fuel or fall of some material in the reactor pool leading toblockage of one or more channels [12] A fuel channel block-age event has different characteristics depending on the flowdirectionDownward cooling flow can lead to blockage due toobjects dropping into the pool Upward cooling flow can leadto blockage due to objects inside the primary cooling systempiping being dragged into the core by the action of the pump[1] In this work all the simulations considered the blockageat the core inlet

Two cases of transient were considered In the first casethe aim was to investigate a partial core blockage simulationin the second one it was considered a total core blockageBoth transient cases described respectively in more detailsin Sections 41 and 42 were initiated after the calculationreached steady state condition Two values of power opera-tion 100 kWand 265 kW typical IPR-R1 power operation val-ues were considered To perform the partial blockage simu-lation using the RELAP5 37 components of type valves at theinlet of TH channels in the rings A B C and D were used(Figure 4) The valves were closed at 4000 s To simulate thetotal core blockage the valve number 38 was closed at 4000 sstopping totally the coolant flow in the core

4 Simulation Results

Since each flow channel provides its own driving force it ispossible to consider flow channel independently in a steadystate situation [13] However to perform this specific type oftransient the interaction between all the obstructed channelsand all the adjacent channels must be considered RELAP5 isa one-dimensional code and each hydrodynamic cell has twofaces in the normal direction at the inlet and outlet Then tomore realistically analyse the core behaviour in the case of achannel blockage it is necessary to consider the coolant crossflow Therefore the RELAP5 cross flow junction model [14]has been used to represent the multidimensional phenomenaof the flow interconnecting the channels using componentsof type single junctions

41 Partial Core Coolant Flow Blockage The transient beginsat 4000 s The coolant flow rate in the valves reaches thetotal obstructed condition about 5 s after the beginning of theevent (Figure 5) Due the blockage at inlet of the THCs thecoolant flow changes direction and module to adequate theparameters for the new core operation conditions

Themass flow rate at the inlet of the channel B1 decreasesafter the blockage event reaching zero value (Figure 6) Themass flow rate at the midheight of this channel increasesand changes the flow direction Due the THC obstructionthe coolant flow is redistributed among adjacent channelsAfter the transient a new steady state condition is reached(Figure 7)

Figure 8(a) shows a detail of the nodalization wherethe junctions make possible the cross flow between THCsduring the transient The mass flow rate at the cross flowjunctions between the THCs C2 and D3 in four axial levelsare represented in Figure 8(b) It is possible to observe thatuntil 4000 s of calculation the mass flow rate in the junctionsis practically zero After the transient the coolant flowsalso among the channels characterizing the cross flow andreproducing the actual situation

After the beginning of the transient the coolant temper-ature increases in the obstructed channels at midheight anddue to the change of flow direction the outlet temperaturesdecrease A new steady state is reached As an exampleFigure 9 shows the coolant temperature in three axial levelsof the THC D4 at 100 kW and 265 kW of power operation

After the transient it is possible to observe void formationin the obstructed channels Figure 10 shows the void fractionat the outlet of the blocked channels B2 C3 and D4

411 Heat Transfer Analysis The heat transfer from the clad-ding surface to the water occurs locally in the regime of sub-cooled nucleate boiling so that the cladding surface temper-ature (119879sur) is the saturation temperature of the water plus thewall superheat given by McAdams correlations [15] There-fore the cladding surface temperature 119879sur can be foundusing the expression

119879sur = 119879sat + Δ119879sat (1)

where 119879sat is the water saturation temperature and it isequivalent to 11137∘C or 38454∘K at pressure of 15 bar [16]The Δ119879sat is the wall superheat temperature that is given by

Δ119879sat = 081 (11990210158401015840)0259

(2)

with temperature in ∘C and heat flux 11990210158401015840 in Wm2Specifically for the IPR-R1 TRIGA reactor conditions the

transition point between the single-phase convection regimeto subcooled nucleate boiling regime is approximately at60 kW of power operation [17]

Figure 11 shows the time evolution of the heat flux 11990210158401015840 andthe heat transfer coefficient ℎ

119904 in the heat structure B1 at

Science and Technology of Nuclear Installations 5

D7

D8D9D10D11

D12

D13

D14

D15

D16

D17D18 D1 D2

D3

D4

D6

D6

C2C1C12

C11

C10

C9

C8 C7 C6C5

C4

C3B2

B1

A1B6

B5B4

B3

068F1 F2 000

000

000

000

000

000000

000

000

000

000

000

000

065

000

000

064065

081

080

080

080

080

083

083

084

084

084

084085

085

079

081099099

099

099

099

100

101107113107

102

102

101

100

100131

121

133142

141

128

103 109

153141

142119

131

132130

155

128

131

103

081081082

081

084

084082

082

082066

000

000

000000

000000000000

000000

000

000

000

F3F4

F5

F6E5

E4E3

E2E1E24E23

E22

E21

E20

E19

E18

E17

E16E15

E14 E13 E12E11

E10

E9

E8

E7

E6F7

F9

F8

F10

F11

F12F13

F14F15F16F17

F18F19

F21

F20

F22

F23

F24

F25

F26

F27

F28F29

F30

A B

33

34

3518

19

17

1615

1413

12

1 1

1098

762

31

45

3637 20 21

27

26

25

22

23

24

28293031

32

(a)

Outlet

Inlet

F24 E19 D15 C10 B6 A1 B2 C4 D5 E7 F8

060

100

38

34 17 02 01 06 11 24

(b)

Figure 4 Core nodalization showing in details the positions of the valves used to perform the partial core blockage (a) central cross view ofthe core nodalization (b) The valve number 38 connected to the element number 100 is used to perform the total core blockage

midheight and at 100 kW of power operation for RELAP5calculation

The values of 11990210158401015840 were taken from RELAP5 steady stationcalculation and Δ119879sat and 119879sur were calculated using (1) and(2) The values for 100 kW and 265 kW are respectively

Δ119879sat(100 kW) = 081 (7292788)0259= 1472

∘C

Δ119879sat(265 kW) = 081 (1930676)0259= 1895

∘C(3)

Therefore

119879(sur100 kW) = 12609

∘C

119879(sur265 kW) = 13031

∘C(4)

These results means that the reactor regime is the subcoolednucleate boiling in which 119879sur gt 119879sat but 119879fluid lt 119879sat

Table 1 presents a comparison among experimental andcalculated values of the fuel thermal parameters The calcu-lated parameters using RELAP5 code are taken at midheight

6 Science and Technology of Nuclear Installations

01

03 10

0 4000 6000 8000 10000 1400012000 160002000

Mas

s flow

rate

(kg

s)

minus004

minus002

020

000

002

004

006

008

010

012

014

016

018

Time (s)

(a)

01

03

10

3990 3992 3994 3996 4010400840064004400240003998

Mas

s flow

rate

(kg

s)

minus004

minus002

020

000

002

004

006

008

010

012

014

016

018

Time (s)

(b)

Figure 5 Mass flow rate evolution in valves 1 3 and 10 at 265 kW of power operation tripped at 4000 s (a) The time window shows it indetails (b)

Inlet

Middle

0 2000 4000 6000 8000 100001000 3000 5000 7000 9000Time (s)

minus030

minus025

minus020

minus015

minus010

minus005

035

040

030

025

020

015

010

005

000

Mas

s flow

rate

(kg

s)

Figure 6 Inlet and midheight mass flow rate in the channel B1mdashbefore and after the blockage initiated at 4000 s (power of 265 kW)

(axial level 11) of B1 fuel element and are in good agreementwith the experimental data [17]

The convective heat transfer coefficient as it can be seenin Figure 11 increases after the blockage reaching a newsteady state value As it was verified from the results obtainedthe transient presented variation of the thermal hydraulicparameters that quickly reached new steady state values

42 Total Core Blockage In this situation all the thermalhydraulic channels were blocked using a component typevalve (valve 38 in the Figure 4) closed at 4000 s of calculationtherefore after the steady state conditions were reachedFigure 12 shows themass flow rate at 100 and 265 kWof poweroperation for the THC D4Themass flow rate after the tran-sient oscillates around zero value

After the beginning of the transient the coolant tempera-ture in the core increases reaching the saturation temperatureThe regime now is the saturated or bulk boiling with

Table 1 Experimental and calculated values of some fuel thermalparameters

Power(kW)

11990210158401015840

(Wm2)Δ119879sat(∘C)

Claddingtemperature (∘C)

Experimental 2650 1946130 1900 1304RELAP5 2650 1930676 1895 1303

the coolant temperature remaining essentially constant andequal to 119879sat As an example Figure 13 shows the coolanttemperature in three axial levels of the THC B1 at 100 kW and265 kW of power operation The same behaviour is observedfor all TH core channels

In addition Figure 14 shows the void fraction and thecoolant temperature time evolution in TH channel D4 at265 kW

Specifically to TRIGA reactors the limiting parameteris the fuel temperature The maximum fuel temperature isrelated to the dehydrogenation of uranium and zirconiumhydride and subsequent stress on the cladding The tem-perature at which the dehydrogenation occurs in UZrH

10

is approximately 550∘C This temperature should not beexceeded to avoid volumetric expansion of the UZrH anddeformation of the fuel element [4]

There is no significant increase in the fuel and claddingtemperatures after the total blockage (Figure 15) Even withthe loss of flow and the coolant reaching the saturationtemperature the fuel temperature remains below the safethreshold temperature

The loss of flow accidents is part of a category that involvesweak feedback effects when it occurs out of core However inthe total core blockage significant disturbance of the core wasobserved and also void formation was observed Thereforea more realistic analysis of this type of accident could beperformed considering cross section variations This is notpossible to do using a point kinetics model

Science and Technology of Nuclear Installations 7

A1

E5

B2

C3

D4

A1

E5E5

B2B

C3C

D4D4

A BC D E

F

Time (s)0 2000 4000 6000 8000 10000 12000 14000 16000

minus06

minus05

minus04

minus03

minus02

minus01

05

04

03

02

01

00

Mas

s flow

rate

(kg

s)

Figure 7 Outlet mass flow rate in the channels A1 B2 C3 D4 and E5 before and after the blockage (power of 265 kW)

Axial level 01

Axial level 03

Axial level 07

Axial level 11

Inlet C2

THC D3THC C2

Outlet C2

123456789

101112131415161718192021

123456789

101112131415161718192021

(a)

Time (s)0 2000 4000 6000 8000 10000

00

02

04

06

Axial level 11Axial level 07

Axial level 03Axial level 01

minus02

Mas

s flow

rate

(kg

s)

(b)

Figure 8 Details of the nodalization where the junctions make possible the cross flow between THCs during the transient (a) Mass flow rateat the cross flow junctions between the THCs C2 and D3 in four axial levels (b)

1 2 213 3 44 55 6 6 77 88 99 1010290 290

295 295

300300

305 305

310310

315 315

320 320

325325

330 330

OutletMiddle

Inlet

Outlet

Middle

Inlet

Time (times103 s) Time (times103 s)

Coo

lant

tem

pera

ture

(K)

Coo

lant

tem

pera

ture

(K)

P = 100 kW P = 265 kW

Figure 9 Coolant temperature at the THC D4 at 100 kW and 265 kW

8 Science and Technology of Nuclear Installations

B2 C3

D4D4

0 1 2 3 4 5 6 7 8 9 10 0 1 2 3

3210 4

4 5

5 6

6 7

7 8

8 9

9 10

10

00

4 4

0

4

6

68

12

16 16

16

20

2020

12

12

Void

frac

tion

(times10minus3)

Void

frac

tion

(times10minus3)

Void

frac

tion

(times10minus3)

Time (times103 s) Time (times103 s)

Time (times103 s)

Figure 10 Void fraction time evolution at outlet of the thermal hydraulic channels B2 C3 and D4 after the transient at 100 kW

100 20 30 40 50 60 70 80 90 10066

68

70

72

74

76

78

80

Time (times102 s)

Hea

t flux

(times103

Wm

2)

(a)

0 10 20 30 40 50 60 70 80 90 1000

3

6

9

12

15

18

21

24

27

30

Time (times102 s)

Hea

t tra

nsfe

r coe

ffici

ent(times102

Wm

2K)

(b)

Figure 11 Heat flux (a) and heat transfer coefficient (b) to the heat structure associated with the THC B1 at 100 kW of power operation

5 Conclusions

To perform the calculations a verified model in RELAP5MOD33 was considered The simulated transient character-izes a LOFA type accident Total and partial core coolant

blockage transient have been investigated for the IPR-R1research reactor In the simulation the channels were blockedusing components type valve in the inlet of the channels Inboth cases total and partial blockage the cross flowmodelingin the RELAP5 core nodalization is important in the analyses

Science and Technology of Nuclear Installations 9

Time (s)1000 2000 3000 4000 5000 60000

minus040

minus035

minus030

minus025

minus020

minus015

minus010

minus005

Mas

s flow

rate

(kg

s)

000

005

010

P = 100 kWP = 100 kW

Inlet

Outlet

(a)

Time (s)1000 2000 3000 4000 50000

00

05

10

Mas

s flow

rate

(kg

s)

P = 265 kW

Outlet

Middle = inlet

minus05

minus10

minus15

minus20

minus25

(b)

Figure 12 Mass flow rate at channel D4 after and before the total core blockage at 100 kW (a) and 265 kW (b) of power operation

Outlet

Middle

InletCoo

lant

tem

pera

ture

(K)

300

320

340

360

380

400

Time (s)1000 2000 3000 4000 5000 60000

P = 100 kW

(a)

Outlet

Middle

Inlet

Coo

lant

tem

pera

ture

(K)

300

320

340

360

380

400

Time (s)1000 2000 3000 4000 50000

P = 265 kW

(b)

Figure 13 Coolant temperature in the channel B2 at 100 kW (a) and 265 kW (b) of power operation

300

320

340

360

380

400

Coo

lant

tem

pera

ture

(K)

Time (s)1000 2000 3000 4000 50000

Outlet

MiddleInlet

(a)

Void

frac

tion

0

02

04

06

08

10

Time (s)1000 2000 3000 4000 50000

minus02

Inlet

Middle

Outlet

(b)

Figure 14 Coolant temperature (a) and void fraction (b) evolutions for the THC D4 at 100 kW

10 Science and Technology of Nuclear Installations

300

350

400

450

500

550

600

650

700

0 5 10 15 20 25 30 35 40 45 50Time (times102 s)

Tem

pera

ture

(K)

Fuel

Cladding

Figure 15 Fuel and cladding temperature evolutions for the B1 heatstructure at 265 kW of power operation

because it provides the coolant redistribution among adjacentchannels after the blockage as it is the expected behaviour

After the partial blockage transient an increase in thecoolant temperature of the blocked channels change in themass flow direction and modulus and void formation wereobserved In spite of this the reactor presented safe behaviourafter the transient reaching a new steady state However themodel is not capable of reproducing exactly the behaviour ofthe core due to the changes in the macroscopic cross sectionsthat are directly affected by coolant density and temperature

In the case of the total blockage the saturation coolanttemperature is reached into the core and the coolant boils infew minutes after the beginning of the transient One moretime in an actual situation the intrinsic safety characteristicof fuel shouldwork to control the reactor It was observed thata total core blockage disturbs significantly the core and thereactivity feedback effects should be considered A neutronicanalysis in this case could be necessary to investigate how theneutron flux is perturbed in the core after the total blockagetransient

Conflict of Interests

The authors declare that there is no conflict of interestsregarding the publication of this paper

Acknowledgments

The authors are grateful to CAPES CNPq CDTNCNENand FAPEMIG (all from Brazil) for the support

References

[1] IAEA Safety Analysis for Research Reactors Safety ReportsSeries No 55 International Atomic Energy Agency ViennaAustria 2008

[2] O Merroun A Almers T El Bardouni B El Bakkari andE Chakir ldquoAnalytical benchmarks for verification of thermal-hydraulic codes based on sub-channel approachrdquoNuclear Engi-neering and Design vol 239 no 4 pp 735ndash748 2009

[3] C S Y Suresh G Sateesh S K Das S P Veukateshanand M Rajan ldquoHeat transfer from a totally blocked fuelsubassembly of a liquid metal fast breeder reactor part Iexperimental investigation and part II numerical simulationrdquoNuclear Engineering and Design vol 235 pp 885ndash912 2005

[4] MA F VelosoAnalise Termofluidodinamica de Reatores Nucle-ares de Pesquisa Refrigerados a Agua em Regime de ConveccaoNatural [Doctor thesis] Universidade Estadual de CampinasCampinas Brazil 2004

[5] H M Dalle ldquoAvaliacao Neutronica do Reator TRIGA IPR-R1 ndashR1 com Configuracao de 63 Elementos Combustıveis e Barrade Regulacao em F16rdquo Restrict Document (NIndashEC3-0103)Centro de Desenvolvimento da Tecnologia Nuclear ComissaoNacional de Energia Nuclear Belo Horizonte Brazil 2003

[6] H M Dalle C Pereira and R G P Souza ldquoNeutroniccalculation to theTRIGA Ipr-R1 reactor using theWIMSD4 andCITATION codesrdquo Annals of Nuclear Energy vol 29 no 8 pp901ndash912 2002

[7] R M G P Souza ldquoResultados Experimentais da Calibracaodas Barras de Controle e do Excesso de Reatividade em CincoConfiguracoes do Nucleo do Reator IPR-R1rdquo Tech Rep NI-AT4-00199 CDTNCNEN Belo Horizonte Brasil 1999

[8] P A L Reis A L Costa C Pereira C A M Silva M A FVeloso and A Z Mesquita ldquoSensitivity analysis to a RELAP5nodalization developed for a typical TRIGA research reactorrdquoNuclear Engineering and Design vol 242 pp 300ndash306 2012

[9] A L Costa P A L Reis C Pereira M A F Veloso A ZMesquita and H V Soares ldquoThermal hydraulic analysis ofthe IPR-R1 TRIGA research reactor using a RELAP5 modelrdquoNuclear Engineering and Design vol 240 no 6 pp 1487ndash14942010

[10] P A L Reis A L Costa C Pereira et al ldquoAssessment of aRELAP5model for the IPR-R1 TRIGA research reactorrdquoAnnalsof Nuclear Energy vol 37 no 10 pp 1341ndash1350 2010

[11] M Adorni A Bousbia-Salah T Hamidouche B D MaroF Pierro and F DrsquoAuria ldquoAnalysis of partial and total flowblockage of a single fuel assembly of an MTR research reactorcorerdquo Annals of Nuclear Energy vol 32 no 15 pp 1679ndash16922005

[12] Q Lu S Qiu and G H Su ldquoDevelopment of a thermal-hydraulic analysis code for research reactors with plate fuelsrdquoAnnals of Nuclear Energy vol 36 no 4 pp 433ndash447 2009

[13] A Z Mesquita H C Rezende and R M G P Souza ldquoThermalpower calibrations of the IPR-R1 TRIGA nuclear reactorrdquo inProceedings of the 20th International Congress of MechanicalEngineering (COBEM rsquo09) Gramado Brazil November 2009

[14] RELAP5Mod33 Code Manuals Idaho National EngineeringLaboratory US NRC 2001

[15] W H McAdams Heat Transmission McGraw-Hill Book NewYork NY USA 3rd edition 1954

[16] W Wagner and A Kruse Properties of Water and SteammdashThe Industrial Standard IAPWS-IF97 for the ThermodynamicsProperties Springer Berlin Germany 1998

[17] A Z Mesquita A L Costa C Pereira M A F Veloso andP C A L Reis ldquoExperimental investigation of the onset ofsubcooled nucleate boiling in an open-pool nuclear researchreactorrdquo Journal of ASTM International vol 8 no 6 2011

TribologyAdvances in

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

International Journal of

AerospaceEngineeringHindawi Publishing Corporationhttpwwwhindawicom Volume 2014

FuelsJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Journal ofPetroleum Engineering

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Industrial EngineeringJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Power ElectronicsHindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Advances in

CombustionJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Journal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Renewable Energy

Submit your manuscripts athttpwwwhindawicom

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

StructuresJournal of

International Journal of

RotatingMachinery

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Hindawi Publishing Corporation httpwwwhindawicom

Journal ofEngineeringVolume 2014

Hindawi Publishing Corporation httpwwwhindawicom Volume 2014

International Journal ofPhotoenergy

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Nuclear InstallationsScience and Technology of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Solar EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Wind EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Nuclear EnergyInternational Journal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

High Energy PhysicsAdvances in

The Scientific World JournalHindawi Publishing Corporation httpwwwhindawicom Volume 2014

2 Science and Technology of Nuclear Installations

Tower

Orifice platepressure transmitter

FlowmeterGM

(primary)Pump

Pump

GM (outlet)

Flowmeter

Heat exchanger

GM (area)

GM (pool)

Forcedcirculation

Naturalcirculation

Core

Reactortank

Conductivity sensorIon (3) and fission (1) chambersGM Geiger MullerControl rod position

Level meter

Temperature sensor

Level

Secondarycooling loop

Primarycooling loop

6417

mm

GM (control room)

Tsoil

Tout channel

Tin channel

Tin

Tout

Tin

Tout Tair

TinTout inlet and outlet temperature inrelation to the heat exchanger

TPT100

Figure 1 Schematic representation of the TRIGA IPR-R1 [17]

The water in the pool works as coolant moderator neu-tron reflector and it is able to assure an adequate biologicalradioactive shielding The reactor cooling occurs predom-inantly by natural convection with the circulation forcesgoverned by the water density differences The removal ofthe heat generated from the nuclear fissions is performedby pumping the pool water through a heat exchanger Thecore has a radial cylindrical configuration with six concentricrings (A B C D E and F)with 91 positions able to host eitherfuel rods or other components like control rods reflectorsand irradiator channel There are in the core 63 fuel elementsconstituted by a cylindrical metal cladding filled with ahomogeneous mixture of zirconium hydride and uranium20 enriched in 235U isotopeThese fuel elements have threeaxial sections an upper and a lower reflector (graphite) andthe central portion filled with fuel (U-ZrHx) [4] The radialreflectors elements are coveredwith aluminumandfilledwithgraphite having the same dimensions of the fuel elements

The control rods are composed by boron carbide andaluminum cladding In the center of the reactor there is analuminum tube (central thimble) for irradiation of experi-mental samples This tube is removable and when it is notin use the reactor pool water fills its volume Furthermorethe core has an annular graphite reflector with aluminumcladding Such annular reflector has a radial groove where arotary rack is assembled for insertion of the samples to irradi-ation In such rotary rack it is possible to place the samples in40 different positions around the core Moreover tangent to

annular reflector there is a pneumatic tubewhere the samplesalso can be inserted for irradiation

TRIGA reactors fuel has an important characteristic ofintrinsic safety with large prompt negative coefficient ofreactivity that effectively controls the reactor in cases of pos-itive reactivity insertions In this way any sudden reactivityaddition causes an increase in power which heats the fuel-moderator material (U-ZrHx) instantaneously causing thenumber of fissions to immediately decrease because of neu-tron energy spectrum hardening within the fuel pin

2 IPR-R1 Nodalization

21 Reactor Modelling The radial power distribution shownin Figure 2(a) was calculated in preceding works using theWIMSD4C and CITATION codes [5 6] and also experimen-tal data [7] The radial factor is defined as the ratio of theaverage linear power density in the element to the averagelinear power density in the core

RELAP5 nodalization for the IPR-R1 verified in previouswork (with 91 THC) [8] was utilized in this simulation Thecore model has 91 hydrodynamic channels (see Figure 2(b))The complete reactor nodalization in the RELAP5 is pre-sented in Figure 3 [8ndash10] To simulate the cross flow amongthe THCs a total of 186 single junctions were used intercon-necting selected volumes at different high positions of thenodalized channels These junctions have been added to themodel to observe the behaviour of the coolant inside the core

Science and Technology of Nuclear Installations 3

D7

D8

D9D10D11

D12

D13

D14

D15

D16

D17

D18 D1 D2

D3

D4

D5

D6

C2C1C12

C11

C10

C9

C8 C7 C6C5

C4

C3B2

B1

A1B6

B5B4

B3

068

F1 F2 000000

000

000

000

000000

000

000

000

000

000

000

065

000

000

064065

081

080

080

080

080

083

083

084

084

084

084

085

085

079

081

099099

099

099

099

100

101

107113

107

102

102

101

100

100

131

121

133

142

141

128

103109

153

141

142

119

131

132

130

155

128

131

103

081081082

081

084

084082

082

082

066

000

000

000

000

000000000000

000

000

000

000

000

F3F4

F5

F6E5

E4E3

E2E1E24E23

E22

E21

E20

E19

E18

E17

E16E15

E14 E13 E12E11

E10

E9

E8

E7

E6F7

F9

F8

F10

F11

F12F13

F14F15F16F17

F18F19

F21

F20

F22

F23

F24

F25

F26

F27

F28F29

F30

B1155 A1

Radial power factor

LocalizationFuel element aluminium

Graphite element

Fuel element stainless steel

Control rods C1 C7F16

Neutrons source

Central trap

Other types of components GraphiteFuel Element

91-THC

(a) (b)

Figure 2 IPR-R1mdashradial relative power distribution (a) model of the 91 THC in the RELAP5 code (b)

340 350 160 144

Pump Valve

HEX820

TDV TDV

720

SV 060

201 291

SV 100

10090807 0605

0403

02

01

TDV 500

SJ 059

Pool 050Pool 020

SJ 049

SJ 101SJ 099SJ 098SJ 097SJ 096SJ 095SJ 094SJ 093

SJ 092

SJ 091

132330

Core91 THC

63 heatstructures

201ndash291

Figure 3 IPR-R1mdashNodalization in the RELAP5-MOD33 code

after the THCblockage transientThis nodalization is capableof simulating situations of forced coolant recirculation andalso natural circulation

22 Heat Structures Modelling Each of the 63 fuel elementswas modeled separately and 63 heat structure componentsdivided in 21 axial levels were consideredThe heat structures

in RELAP5 permit the calculation of heat across the solidboundaries of the hydrodynamic volumes Radially the heatstructures were divided in 17 mesh points considering theactual radius of the fuel (heat source) gap and cladding Thefuel elements were modelled using cylindrical geometry andthe characteristics of thematerials were inserted to reproducethe heat transfer For each material specified corresponding

4 Science and Technology of Nuclear Installations

thermal property data were entered to define the thermalconductivity and volumetric heat capacity as functions oftemperature

The axial power distribution was calculated consideringa cosine profile predicted using the data from Figure 2 andtaking into account the power cutoff in the extremes of theelement due the presence of the graphite The point kineticsmodel was used in the current modeling which assumes thatchanges in reactivity uniformly affect the core flux resultingin no relative spatial variation over time Point kinetics is notindicated when a pronounced flux spike and radial neutronwave occur

3 THCs Blockage Transient

THCs blockage transient has been investigated using theRELAP5 code For the open pool research reactors configu-ration the probability of a blockage in the core upper plenumzone is higher than a blockage from the bottom [11] Thistransient situation may be caused for example by swelling ofthe fuel or fall of some material in the reactor pool leading toblockage of one or more channels [12] A fuel channel block-age event has different characteristics depending on the flowdirectionDownward cooling flow can lead to blockage due toobjects dropping into the pool Upward cooling flow can leadto blockage due to objects inside the primary cooling systempiping being dragged into the core by the action of the pump[1] In this work all the simulations considered the blockageat the core inlet

Two cases of transient were considered In the first casethe aim was to investigate a partial core blockage simulationin the second one it was considered a total core blockageBoth transient cases described respectively in more detailsin Sections 41 and 42 were initiated after the calculationreached steady state condition Two values of power opera-tion 100 kWand 265 kW typical IPR-R1 power operation val-ues were considered To perform the partial blockage simu-lation using the RELAP5 37 components of type valves at theinlet of TH channels in the rings A B C and D were used(Figure 4) The valves were closed at 4000 s To simulate thetotal core blockage the valve number 38 was closed at 4000 sstopping totally the coolant flow in the core

4 Simulation Results

Since each flow channel provides its own driving force it ispossible to consider flow channel independently in a steadystate situation [13] However to perform this specific type oftransient the interaction between all the obstructed channelsand all the adjacent channels must be considered RELAP5 isa one-dimensional code and each hydrodynamic cell has twofaces in the normal direction at the inlet and outlet Then tomore realistically analyse the core behaviour in the case of achannel blockage it is necessary to consider the coolant crossflow Therefore the RELAP5 cross flow junction model [14]has been used to represent the multidimensional phenomenaof the flow interconnecting the channels using componentsof type single junctions

41 Partial Core Coolant Flow Blockage The transient beginsat 4000 s The coolant flow rate in the valves reaches thetotal obstructed condition about 5 s after the beginning of theevent (Figure 5) Due the blockage at inlet of the THCs thecoolant flow changes direction and module to adequate theparameters for the new core operation conditions

Themass flow rate at the inlet of the channel B1 decreasesafter the blockage event reaching zero value (Figure 6) Themass flow rate at the midheight of this channel increasesand changes the flow direction Due the THC obstructionthe coolant flow is redistributed among adjacent channelsAfter the transient a new steady state condition is reached(Figure 7)

Figure 8(a) shows a detail of the nodalization wherethe junctions make possible the cross flow between THCsduring the transient The mass flow rate at the cross flowjunctions between the THCs C2 and D3 in four axial levelsare represented in Figure 8(b) It is possible to observe thatuntil 4000 s of calculation the mass flow rate in the junctionsis practically zero After the transient the coolant flowsalso among the channels characterizing the cross flow andreproducing the actual situation

After the beginning of the transient the coolant temper-ature increases in the obstructed channels at midheight anddue to the change of flow direction the outlet temperaturesdecrease A new steady state is reached As an exampleFigure 9 shows the coolant temperature in three axial levelsof the THC D4 at 100 kW and 265 kW of power operation

After the transient it is possible to observe void formationin the obstructed channels Figure 10 shows the void fractionat the outlet of the blocked channels B2 C3 and D4

411 Heat Transfer Analysis The heat transfer from the clad-ding surface to the water occurs locally in the regime of sub-cooled nucleate boiling so that the cladding surface temper-ature (119879sur) is the saturation temperature of the water plus thewall superheat given by McAdams correlations [15] There-fore the cladding surface temperature 119879sur can be foundusing the expression

119879sur = 119879sat + Δ119879sat (1)

where 119879sat is the water saturation temperature and it isequivalent to 11137∘C or 38454∘K at pressure of 15 bar [16]The Δ119879sat is the wall superheat temperature that is given by

Δ119879sat = 081 (11990210158401015840)0259

(2)

with temperature in ∘C and heat flux 11990210158401015840 in Wm2Specifically for the IPR-R1 TRIGA reactor conditions the

transition point between the single-phase convection regimeto subcooled nucleate boiling regime is approximately at60 kW of power operation [17]

Figure 11 shows the time evolution of the heat flux 11990210158401015840 andthe heat transfer coefficient ℎ

119904 in the heat structure B1 at

Science and Technology of Nuclear Installations 5

D7

D8D9D10D11

D12

D13

D14

D15

D16

D17D18 D1 D2

D3

D4

D6

D6

C2C1C12

C11

C10

C9

C8 C7 C6C5

C4

C3B2

B1

A1B6

B5B4

B3

068F1 F2 000

000

000

000

000

000000

000

000

000

000

000

000

065

000

000

064065

081

080

080

080

080

083

083

084

084

084

084085

085

079

081099099

099

099

099

100

101107113107

102

102

101

100

100131

121

133142

141

128

103 109

153141

142119

131

132130

155

128

131

103

081081082

081

084

084082

082

082066

000

000

000000

000000000000

000000

000

000

000

F3F4

F5

F6E5

E4E3

E2E1E24E23

E22

E21

E20

E19

E18

E17

E16E15

E14 E13 E12E11

E10

E9

E8

E7

E6F7

F9

F8

F10

F11

F12F13

F14F15F16F17

F18F19

F21

F20

F22

F23

F24

F25

F26

F27

F28F29

F30

A B

33

34

3518

19

17

1615

1413

12

1 1

1098

762

31

45

3637 20 21

27

26

25

22

23

24

28293031

32

(a)

Outlet

Inlet

F24 E19 D15 C10 B6 A1 B2 C4 D5 E7 F8

060

100

38

34 17 02 01 06 11 24

(b)

Figure 4 Core nodalization showing in details the positions of the valves used to perform the partial core blockage (a) central cross view ofthe core nodalization (b) The valve number 38 connected to the element number 100 is used to perform the total core blockage

midheight and at 100 kW of power operation for RELAP5calculation

The values of 11990210158401015840 were taken from RELAP5 steady stationcalculation and Δ119879sat and 119879sur were calculated using (1) and(2) The values for 100 kW and 265 kW are respectively

Δ119879sat(100 kW) = 081 (7292788)0259= 1472

∘C

Δ119879sat(265 kW) = 081 (1930676)0259= 1895

∘C(3)

Therefore

119879(sur100 kW) = 12609

∘C

119879(sur265 kW) = 13031

∘C(4)

These results means that the reactor regime is the subcoolednucleate boiling in which 119879sur gt 119879sat but 119879fluid lt 119879sat

Table 1 presents a comparison among experimental andcalculated values of the fuel thermal parameters The calcu-lated parameters using RELAP5 code are taken at midheight

6 Science and Technology of Nuclear Installations

01

03 10

0 4000 6000 8000 10000 1400012000 160002000

Mas

s flow

rate

(kg

s)

minus004

minus002

020

000

002

004

006

008

010

012

014

016

018

Time (s)

(a)

01

03

10

3990 3992 3994 3996 4010400840064004400240003998

Mas

s flow

rate

(kg

s)

minus004

minus002

020

000

002

004

006

008

010

012

014

016

018

Time (s)

(b)

Figure 5 Mass flow rate evolution in valves 1 3 and 10 at 265 kW of power operation tripped at 4000 s (a) The time window shows it indetails (b)

Inlet

Middle

0 2000 4000 6000 8000 100001000 3000 5000 7000 9000Time (s)

minus030

minus025

minus020

minus015

minus010

minus005

035

040

030

025

020

015

010

005

000

Mas

s flow

rate

(kg

s)

Figure 6 Inlet and midheight mass flow rate in the channel B1mdashbefore and after the blockage initiated at 4000 s (power of 265 kW)

(axial level 11) of B1 fuel element and are in good agreementwith the experimental data [17]

The convective heat transfer coefficient as it can be seenin Figure 11 increases after the blockage reaching a newsteady state value As it was verified from the results obtainedthe transient presented variation of the thermal hydraulicparameters that quickly reached new steady state values

42 Total Core Blockage In this situation all the thermalhydraulic channels were blocked using a component typevalve (valve 38 in the Figure 4) closed at 4000 s of calculationtherefore after the steady state conditions were reachedFigure 12 shows themass flow rate at 100 and 265 kWof poweroperation for the THC D4Themass flow rate after the tran-sient oscillates around zero value

After the beginning of the transient the coolant tempera-ture in the core increases reaching the saturation temperatureThe regime now is the saturated or bulk boiling with

Table 1 Experimental and calculated values of some fuel thermalparameters

Power(kW)

11990210158401015840

(Wm2)Δ119879sat(∘C)

Claddingtemperature (∘C)

Experimental 2650 1946130 1900 1304RELAP5 2650 1930676 1895 1303

the coolant temperature remaining essentially constant andequal to 119879sat As an example Figure 13 shows the coolanttemperature in three axial levels of the THC B1 at 100 kW and265 kW of power operation The same behaviour is observedfor all TH core channels

In addition Figure 14 shows the void fraction and thecoolant temperature time evolution in TH channel D4 at265 kW

Specifically to TRIGA reactors the limiting parameteris the fuel temperature The maximum fuel temperature isrelated to the dehydrogenation of uranium and zirconiumhydride and subsequent stress on the cladding The tem-perature at which the dehydrogenation occurs in UZrH

10

is approximately 550∘C This temperature should not beexceeded to avoid volumetric expansion of the UZrH anddeformation of the fuel element [4]

There is no significant increase in the fuel and claddingtemperatures after the total blockage (Figure 15) Even withthe loss of flow and the coolant reaching the saturationtemperature the fuel temperature remains below the safethreshold temperature

The loss of flow accidents is part of a category that involvesweak feedback effects when it occurs out of core However inthe total core blockage significant disturbance of the core wasobserved and also void formation was observed Thereforea more realistic analysis of this type of accident could beperformed considering cross section variations This is notpossible to do using a point kinetics model

Science and Technology of Nuclear Installations 7

A1

E5

B2

C3

D4

A1

E5E5

B2B

C3C

D4D4

A BC D E

F

Time (s)0 2000 4000 6000 8000 10000 12000 14000 16000

minus06

minus05

minus04

minus03

minus02

minus01

05

04

03

02

01

00

Mas

s flow

rate

(kg

s)

Figure 7 Outlet mass flow rate in the channels A1 B2 C3 D4 and E5 before and after the blockage (power of 265 kW)

Axial level 01

Axial level 03

Axial level 07

Axial level 11

Inlet C2

THC D3THC C2

Outlet C2

123456789

101112131415161718192021

123456789

101112131415161718192021

(a)

Time (s)0 2000 4000 6000 8000 10000

00

02

04

06

Axial level 11Axial level 07

Axial level 03Axial level 01

minus02

Mas

s flow

rate

(kg

s)

(b)

Figure 8 Details of the nodalization where the junctions make possible the cross flow between THCs during the transient (a) Mass flow rateat the cross flow junctions between the THCs C2 and D3 in four axial levels (b)

1 2 213 3 44 55 6 6 77 88 99 1010290 290

295 295

300300

305 305

310310

315 315

320 320

325325

330 330

OutletMiddle

Inlet

Outlet

Middle

Inlet

Time (times103 s) Time (times103 s)

Coo

lant

tem

pera

ture

(K)

Coo

lant

tem

pera

ture

(K)

P = 100 kW P = 265 kW

Figure 9 Coolant temperature at the THC D4 at 100 kW and 265 kW

8 Science and Technology of Nuclear Installations

B2 C3

D4D4

0 1 2 3 4 5 6 7 8 9 10 0 1 2 3

3210 4

4 5

5 6

6 7

7 8

8 9

9 10

10

00

4 4

0

4

6

68

12

16 16

16

20

2020

12

12

Void

frac

tion

(times10minus3)

Void

frac

tion

(times10minus3)

Void

frac

tion

(times10minus3)

Time (times103 s) Time (times103 s)

Time (times103 s)

Figure 10 Void fraction time evolution at outlet of the thermal hydraulic channels B2 C3 and D4 after the transient at 100 kW

100 20 30 40 50 60 70 80 90 10066

68

70

72

74

76

78

80

Time (times102 s)

Hea

t flux

(times103

Wm

2)

(a)

0 10 20 30 40 50 60 70 80 90 1000

3

6

9

12

15

18

21

24

27

30

Time (times102 s)

Hea

t tra

nsfe

r coe

ffici

ent(times102

Wm

2K)

(b)

Figure 11 Heat flux (a) and heat transfer coefficient (b) to the heat structure associated with the THC B1 at 100 kW of power operation

5 Conclusions

To perform the calculations a verified model in RELAP5MOD33 was considered The simulated transient character-izes a LOFA type accident Total and partial core coolant

blockage transient have been investigated for the IPR-R1research reactor In the simulation the channels were blockedusing components type valve in the inlet of the channels Inboth cases total and partial blockage the cross flowmodelingin the RELAP5 core nodalization is important in the analyses

Science and Technology of Nuclear Installations 9

Time (s)1000 2000 3000 4000 5000 60000

minus040

minus035

minus030

minus025

minus020

minus015

minus010

minus005

Mas

s flow

rate

(kg

s)

000

005

010

P = 100 kWP = 100 kW

Inlet

Outlet

(a)

Time (s)1000 2000 3000 4000 50000

00

05

10

Mas

s flow

rate

(kg

s)

P = 265 kW

Outlet

Middle = inlet

minus05

minus10

minus15

minus20

minus25

(b)

Figure 12 Mass flow rate at channel D4 after and before the total core blockage at 100 kW (a) and 265 kW (b) of power operation

Outlet

Middle

InletCoo

lant

tem

pera

ture

(K)

300

320

340

360

380

400

Time (s)1000 2000 3000 4000 5000 60000

P = 100 kW

(a)

Outlet

Middle

Inlet

Coo

lant

tem

pera

ture

(K)

300

320

340

360

380

400

Time (s)1000 2000 3000 4000 50000

P = 265 kW

(b)

Figure 13 Coolant temperature in the channel B2 at 100 kW (a) and 265 kW (b) of power operation

300

320

340

360

380

400

Coo

lant

tem

pera

ture

(K)

Time (s)1000 2000 3000 4000 50000

Outlet

MiddleInlet

(a)

Void

frac

tion

0

02

04

06

08

10

Time (s)1000 2000 3000 4000 50000

minus02

Inlet

Middle

Outlet

(b)

Figure 14 Coolant temperature (a) and void fraction (b) evolutions for the THC D4 at 100 kW

10 Science and Technology of Nuclear Installations

300

350

400

450

500

550

600

650

700

0 5 10 15 20 25 30 35 40 45 50Time (times102 s)

Tem

pera

ture

(K)

Fuel

Cladding

Figure 15 Fuel and cladding temperature evolutions for the B1 heatstructure at 265 kW of power operation

because it provides the coolant redistribution among adjacentchannels after the blockage as it is the expected behaviour

After the partial blockage transient an increase in thecoolant temperature of the blocked channels change in themass flow direction and modulus and void formation wereobserved In spite of this the reactor presented safe behaviourafter the transient reaching a new steady state However themodel is not capable of reproducing exactly the behaviour ofthe core due to the changes in the macroscopic cross sectionsthat are directly affected by coolant density and temperature

In the case of the total blockage the saturation coolanttemperature is reached into the core and the coolant boils infew minutes after the beginning of the transient One moretime in an actual situation the intrinsic safety characteristicof fuel shouldwork to control the reactor It was observed thata total core blockage disturbs significantly the core and thereactivity feedback effects should be considered A neutronicanalysis in this case could be necessary to investigate how theneutron flux is perturbed in the core after the total blockagetransient

Conflict of Interests

The authors declare that there is no conflict of interestsregarding the publication of this paper

Acknowledgments

The authors are grateful to CAPES CNPq CDTNCNENand FAPEMIG (all from Brazil) for the support

References

[1] IAEA Safety Analysis for Research Reactors Safety ReportsSeries No 55 International Atomic Energy Agency ViennaAustria 2008

[2] O Merroun A Almers T El Bardouni B El Bakkari andE Chakir ldquoAnalytical benchmarks for verification of thermal-hydraulic codes based on sub-channel approachrdquoNuclear Engi-neering and Design vol 239 no 4 pp 735ndash748 2009

[3] C S Y Suresh G Sateesh S K Das S P Veukateshanand M Rajan ldquoHeat transfer from a totally blocked fuelsubassembly of a liquid metal fast breeder reactor part Iexperimental investigation and part II numerical simulationrdquoNuclear Engineering and Design vol 235 pp 885ndash912 2005

[4] MA F VelosoAnalise Termofluidodinamica de Reatores Nucle-ares de Pesquisa Refrigerados a Agua em Regime de ConveccaoNatural [Doctor thesis] Universidade Estadual de CampinasCampinas Brazil 2004

[5] H M Dalle ldquoAvaliacao Neutronica do Reator TRIGA IPR-R1 ndashR1 com Configuracao de 63 Elementos Combustıveis e Barrade Regulacao em F16rdquo Restrict Document (NIndashEC3-0103)Centro de Desenvolvimento da Tecnologia Nuclear ComissaoNacional de Energia Nuclear Belo Horizonte Brazil 2003

[6] H M Dalle C Pereira and R G P Souza ldquoNeutroniccalculation to theTRIGA Ipr-R1 reactor using theWIMSD4 andCITATION codesrdquo Annals of Nuclear Energy vol 29 no 8 pp901ndash912 2002

[7] R M G P Souza ldquoResultados Experimentais da Calibracaodas Barras de Controle e do Excesso de Reatividade em CincoConfiguracoes do Nucleo do Reator IPR-R1rdquo Tech Rep NI-AT4-00199 CDTNCNEN Belo Horizonte Brasil 1999

[8] P A L Reis A L Costa C Pereira C A M Silva M A FVeloso and A Z Mesquita ldquoSensitivity analysis to a RELAP5nodalization developed for a typical TRIGA research reactorrdquoNuclear Engineering and Design vol 242 pp 300ndash306 2012

[9] A L Costa P A L Reis C Pereira M A F Veloso A ZMesquita and H V Soares ldquoThermal hydraulic analysis ofthe IPR-R1 TRIGA research reactor using a RELAP5 modelrdquoNuclear Engineering and Design vol 240 no 6 pp 1487ndash14942010

[10] P A L Reis A L Costa C Pereira et al ldquoAssessment of aRELAP5model for the IPR-R1 TRIGA research reactorrdquoAnnalsof Nuclear Energy vol 37 no 10 pp 1341ndash1350 2010

[11] M Adorni A Bousbia-Salah T Hamidouche B D MaroF Pierro and F DrsquoAuria ldquoAnalysis of partial and total flowblockage of a single fuel assembly of an MTR research reactorcorerdquo Annals of Nuclear Energy vol 32 no 15 pp 1679ndash16922005

[12] Q Lu S Qiu and G H Su ldquoDevelopment of a thermal-hydraulic analysis code for research reactors with plate fuelsrdquoAnnals of Nuclear Energy vol 36 no 4 pp 433ndash447 2009

[13] A Z Mesquita H C Rezende and R M G P Souza ldquoThermalpower calibrations of the IPR-R1 TRIGA nuclear reactorrdquo inProceedings of the 20th International Congress of MechanicalEngineering (COBEM rsquo09) Gramado Brazil November 2009

[14] RELAP5Mod33 Code Manuals Idaho National EngineeringLaboratory US NRC 2001

[15] W H McAdams Heat Transmission McGraw-Hill Book NewYork NY USA 3rd edition 1954

[16] W Wagner and A Kruse Properties of Water and SteammdashThe Industrial Standard IAPWS-IF97 for the ThermodynamicsProperties Springer Berlin Germany 1998

[17] A Z Mesquita A L Costa C Pereira M A F Veloso andP C A L Reis ldquoExperimental investigation of the onset ofsubcooled nucleate boiling in an open-pool nuclear researchreactorrdquo Journal of ASTM International vol 8 no 6 2011

TribologyAdvances in

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

International Journal of

AerospaceEngineeringHindawi Publishing Corporationhttpwwwhindawicom Volume 2014

FuelsJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Journal ofPetroleum Engineering

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Industrial EngineeringJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Power ElectronicsHindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Advances in

CombustionJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Journal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Renewable Energy

Submit your manuscripts athttpwwwhindawicom

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

StructuresJournal of

International Journal of

RotatingMachinery

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Hindawi Publishing Corporation httpwwwhindawicom

Journal ofEngineeringVolume 2014

Hindawi Publishing Corporation httpwwwhindawicom Volume 2014

International Journal ofPhotoenergy

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Nuclear InstallationsScience and Technology of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Solar EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Wind EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Nuclear EnergyInternational Journal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

High Energy PhysicsAdvances in

The Scientific World JournalHindawi Publishing Corporation httpwwwhindawicom Volume 2014

Science and Technology of Nuclear Installations 3

D7

D8

D9D10D11

D12

D13

D14

D15

D16

D17

D18 D1 D2

D3

D4

D5

D6

C2C1C12

C11

C10

C9

C8 C7 C6C5

C4

C3B2

B1

A1B6

B5B4

B3

068

F1 F2 000000

000

000

000

000000

000

000

000

000

000

000

065

000

000

064065

081

080

080

080

080

083

083

084

084

084

084

085

085

079

081

099099

099

099

099

100

101

107113

107

102

102

101

100

100

131

121

133

142

141

128

103109

153

141

142

119

131

132

130

155

128

131

103

081081082

081

084

084082

082

082

066

000

000

000

000

000000000000

000

000

000

000

000

F3F4

F5

F6E5

E4E3

E2E1E24E23

E22

E21

E20

E19

E18

E17

E16E15

E14 E13 E12E11

E10

E9

E8

E7

E6F7

F9

F8

F10

F11

F12F13

F14F15F16F17

F18F19

F21

F20

F22

F23

F24

F25

F26

F27

F28F29

F30

B1155 A1

Radial power factor

LocalizationFuel element aluminium

Graphite element

Fuel element stainless steel

Control rods C1 C7F16

Neutrons source

Central trap

Other types of components GraphiteFuel Element

91-THC

(a) (b)

Figure 2 IPR-R1mdashradial relative power distribution (a) model of the 91 THC in the RELAP5 code (b)

340 350 160 144

Pump Valve

HEX820

TDV TDV

720

SV 060

201 291

SV 100

10090807 0605

0403

02

01

TDV 500

SJ 059

Pool 050Pool 020

SJ 049

SJ 101SJ 099SJ 098SJ 097SJ 096SJ 095SJ 094SJ 093

SJ 092

SJ 091

132330

Core91 THC

63 heatstructures

201ndash291

Figure 3 IPR-R1mdashNodalization in the RELAP5-MOD33 code

after the THCblockage transientThis nodalization is capableof simulating situations of forced coolant recirculation andalso natural circulation

22 Heat Structures Modelling Each of the 63 fuel elementswas modeled separately and 63 heat structure componentsdivided in 21 axial levels were consideredThe heat structures

in RELAP5 permit the calculation of heat across the solidboundaries of the hydrodynamic volumes Radially the heatstructures were divided in 17 mesh points considering theactual radius of the fuel (heat source) gap and cladding Thefuel elements were modelled using cylindrical geometry andthe characteristics of thematerials were inserted to reproducethe heat transfer For each material specified corresponding

4 Science and Technology of Nuclear Installations

thermal property data were entered to define the thermalconductivity and volumetric heat capacity as functions oftemperature

The axial power distribution was calculated consideringa cosine profile predicted using the data from Figure 2 andtaking into account the power cutoff in the extremes of theelement due the presence of the graphite The point kineticsmodel was used in the current modeling which assumes thatchanges in reactivity uniformly affect the core flux resultingin no relative spatial variation over time Point kinetics is notindicated when a pronounced flux spike and radial neutronwave occur

3 THCs Blockage Transient

THCs blockage transient has been investigated using theRELAP5 code For the open pool research reactors configu-ration the probability of a blockage in the core upper plenumzone is higher than a blockage from the bottom [11] Thistransient situation may be caused for example by swelling ofthe fuel or fall of some material in the reactor pool leading toblockage of one or more channels [12] A fuel channel block-age event has different characteristics depending on the flowdirectionDownward cooling flow can lead to blockage due toobjects dropping into the pool Upward cooling flow can leadto blockage due to objects inside the primary cooling systempiping being dragged into the core by the action of the pump[1] In this work all the simulations considered the blockageat the core inlet

Two cases of transient were considered In the first casethe aim was to investigate a partial core blockage simulationin the second one it was considered a total core blockageBoth transient cases described respectively in more detailsin Sections 41 and 42 were initiated after the calculationreached steady state condition Two values of power opera-tion 100 kWand 265 kW typical IPR-R1 power operation val-ues were considered To perform the partial blockage simu-lation using the RELAP5 37 components of type valves at theinlet of TH channels in the rings A B C and D were used(Figure 4) The valves were closed at 4000 s To simulate thetotal core blockage the valve number 38 was closed at 4000 sstopping totally the coolant flow in the core

4 Simulation Results

Since each flow channel provides its own driving force it ispossible to consider flow channel independently in a steadystate situation [13] However to perform this specific type oftransient the interaction between all the obstructed channelsand all the adjacent channels must be considered RELAP5 isa one-dimensional code and each hydrodynamic cell has twofaces in the normal direction at the inlet and outlet Then tomore realistically analyse the core behaviour in the case of achannel blockage it is necessary to consider the coolant crossflow Therefore the RELAP5 cross flow junction model [14]has been used to represent the multidimensional phenomenaof the flow interconnecting the channels using componentsof type single junctions

41 Partial Core Coolant Flow Blockage The transient beginsat 4000 s The coolant flow rate in the valves reaches thetotal obstructed condition about 5 s after the beginning of theevent (Figure 5) Due the blockage at inlet of the THCs thecoolant flow changes direction and module to adequate theparameters for the new core operation conditions

Themass flow rate at the inlet of the channel B1 decreasesafter the blockage event reaching zero value (Figure 6) Themass flow rate at the midheight of this channel increasesand changes the flow direction Due the THC obstructionthe coolant flow is redistributed among adjacent channelsAfter the transient a new steady state condition is reached(Figure 7)

Figure 8(a) shows a detail of the nodalization wherethe junctions make possible the cross flow between THCsduring the transient The mass flow rate at the cross flowjunctions between the THCs C2 and D3 in four axial levelsare represented in Figure 8(b) It is possible to observe thatuntil 4000 s of calculation the mass flow rate in the junctionsis practically zero After the transient the coolant flowsalso among the channels characterizing the cross flow andreproducing the actual situation

After the beginning of the transient the coolant temper-ature increases in the obstructed channels at midheight anddue to the change of flow direction the outlet temperaturesdecrease A new steady state is reached As an exampleFigure 9 shows the coolant temperature in three axial levelsof the THC D4 at 100 kW and 265 kW of power operation

After the transient it is possible to observe void formationin the obstructed channels Figure 10 shows the void fractionat the outlet of the blocked channels B2 C3 and D4

411 Heat Transfer Analysis The heat transfer from the clad-ding surface to the water occurs locally in the regime of sub-cooled nucleate boiling so that the cladding surface temper-ature (119879sur) is the saturation temperature of the water plus thewall superheat given by McAdams correlations [15] There-fore the cladding surface temperature 119879sur can be foundusing the expression

119879sur = 119879sat + Δ119879sat (1)

where 119879sat is the water saturation temperature and it isequivalent to 11137∘C or 38454∘K at pressure of 15 bar [16]The Δ119879sat is the wall superheat temperature that is given by

Δ119879sat = 081 (11990210158401015840)0259

(2)

with temperature in ∘C and heat flux 11990210158401015840 in Wm2Specifically for the IPR-R1 TRIGA reactor conditions the

transition point between the single-phase convection regimeto subcooled nucleate boiling regime is approximately at60 kW of power operation [17]

Figure 11 shows the time evolution of the heat flux 11990210158401015840 andthe heat transfer coefficient ℎ

119904 in the heat structure B1 at

Science and Technology of Nuclear Installations 5

D7

D8D9D10D11

D12

D13

D14

D15

D16

D17D18 D1 D2

D3

D4

D6

D6

C2C1C12

C11

C10

C9

C8 C7 C6C5

C4

C3B2

B1

A1B6

B5B4

B3

068F1 F2 000

000

000

000

000

000000

000

000

000

000

000

000

065

000

000

064065

081

080

080

080

080

083

083

084

084

084

084085

085

079

081099099

099

099

099

100

101107113107

102

102

101

100

100131

121

133142

141

128

103 109

153141

142119

131

132130

155

128

131

103

081081082

081

084

084082

082

082066

000

000

000000

000000000000

000000

000

000

000

F3F4

F5

F6E5

E4E3

E2E1E24E23

E22

E21

E20

E19

E18

E17

E16E15

E14 E13 E12E11

E10

E9

E8

E7

E6F7

F9

F8

F10

F11

F12F13

F14F15F16F17

F18F19

F21

F20

F22

F23

F24

F25

F26

F27

F28F29

F30

A B

33

34

3518

19

17

1615

1413

12

1 1

1098

762

31

45

3637 20 21

27

26

25

22

23

24

28293031

32

(a)

Outlet

Inlet

F24 E19 D15 C10 B6 A1 B2 C4 D5 E7 F8

060

100

38

34 17 02 01 06 11 24

(b)

Figure 4 Core nodalization showing in details the positions of the valves used to perform the partial core blockage (a) central cross view ofthe core nodalization (b) The valve number 38 connected to the element number 100 is used to perform the total core blockage

midheight and at 100 kW of power operation for RELAP5calculation

The values of 11990210158401015840 were taken from RELAP5 steady stationcalculation and Δ119879sat and 119879sur were calculated using (1) and(2) The values for 100 kW and 265 kW are respectively

Δ119879sat(100 kW) = 081 (7292788)0259= 1472

∘C

Δ119879sat(265 kW) = 081 (1930676)0259= 1895

∘C(3)

Therefore

119879(sur100 kW) = 12609

∘C

119879(sur265 kW) = 13031

∘C(4)

These results means that the reactor regime is the subcoolednucleate boiling in which 119879sur gt 119879sat but 119879fluid lt 119879sat

Table 1 presents a comparison among experimental andcalculated values of the fuel thermal parameters The calcu-lated parameters using RELAP5 code are taken at midheight

6 Science and Technology of Nuclear Installations

01

03 10

0 4000 6000 8000 10000 1400012000 160002000

Mas

s flow

rate

(kg

s)

minus004

minus002

020

000

002

004

006

008

010

012

014

016

018

Time (s)

(a)

01

03

10

3990 3992 3994 3996 4010400840064004400240003998

Mas

s flow

rate

(kg

s)

minus004

minus002

020

000

002

004

006

008

010

012

014

016

018

Time (s)

(b)

Figure 5 Mass flow rate evolution in valves 1 3 and 10 at 265 kW of power operation tripped at 4000 s (a) The time window shows it indetails (b)

Inlet

Middle

0 2000 4000 6000 8000 100001000 3000 5000 7000 9000Time (s)

minus030

minus025

minus020

minus015

minus010

minus005

035

040

030

025

020

015

010

005

000

Mas

s flow

rate

(kg

s)

Figure 6 Inlet and midheight mass flow rate in the channel B1mdashbefore and after the blockage initiated at 4000 s (power of 265 kW)

(axial level 11) of B1 fuel element and are in good agreementwith the experimental data [17]

The convective heat transfer coefficient as it can be seenin Figure 11 increases after the blockage reaching a newsteady state value As it was verified from the results obtainedthe transient presented variation of the thermal hydraulicparameters that quickly reached new steady state values

42 Total Core Blockage In this situation all the thermalhydraulic channels were blocked using a component typevalve (valve 38 in the Figure 4) closed at 4000 s of calculationtherefore after the steady state conditions were reachedFigure 12 shows themass flow rate at 100 and 265 kWof poweroperation for the THC D4Themass flow rate after the tran-sient oscillates around zero value

After the beginning of the transient the coolant tempera-ture in the core increases reaching the saturation temperatureThe regime now is the saturated or bulk boiling with

Table 1 Experimental and calculated values of some fuel thermalparameters

Power(kW)

11990210158401015840

(Wm2)Δ119879sat(∘C)

Claddingtemperature (∘C)

Experimental 2650 1946130 1900 1304RELAP5 2650 1930676 1895 1303

the coolant temperature remaining essentially constant andequal to 119879sat As an example Figure 13 shows the coolanttemperature in three axial levels of the THC B1 at 100 kW and265 kW of power operation The same behaviour is observedfor all TH core channels

In addition Figure 14 shows the void fraction and thecoolant temperature time evolution in TH channel D4 at265 kW

Specifically to TRIGA reactors the limiting parameteris the fuel temperature The maximum fuel temperature isrelated to the dehydrogenation of uranium and zirconiumhydride and subsequent stress on the cladding The tem-perature at which the dehydrogenation occurs in UZrH

10

is approximately 550∘C This temperature should not beexceeded to avoid volumetric expansion of the UZrH anddeformation of the fuel element [4]

There is no significant increase in the fuel and claddingtemperatures after the total blockage (Figure 15) Even withthe loss of flow and the coolant reaching the saturationtemperature the fuel temperature remains below the safethreshold temperature

The loss of flow accidents is part of a category that involvesweak feedback effects when it occurs out of core However inthe total core blockage significant disturbance of the core wasobserved and also void formation was observed Thereforea more realistic analysis of this type of accident could beperformed considering cross section variations This is notpossible to do using a point kinetics model

Science and Technology of Nuclear Installations 7

A1

E5

B2

C3

D4

A1

E5E5

B2B

C3C

D4D4

A BC D E

F

Time (s)0 2000 4000 6000 8000 10000 12000 14000 16000

minus06

minus05

minus04

minus03

minus02

minus01

05

04

03

02

01

00

Mas

s flow

rate

(kg

s)

Figure 7 Outlet mass flow rate in the channels A1 B2 C3 D4 and E5 before and after the blockage (power of 265 kW)

Axial level 01

Axial level 03

Axial level 07

Axial level 11

Inlet C2

THC D3THC C2

Outlet C2

123456789

101112131415161718192021

123456789

101112131415161718192021

(a)

Time (s)0 2000 4000 6000 8000 10000

00

02

04

06

Axial level 11Axial level 07

Axial level 03Axial level 01

minus02

Mas

s flow

rate

(kg

s)

(b)

Figure 8 Details of the nodalization where the junctions make possible the cross flow between THCs during the transient (a) Mass flow rateat the cross flow junctions between the THCs C2 and D3 in four axial levels (b)

1 2 213 3 44 55 6 6 77 88 99 1010290 290

295 295

300300

305 305

310310

315 315

320 320

325325

330 330

OutletMiddle

Inlet

Outlet

Middle

Inlet

Time (times103 s) Time (times103 s)

Coo

lant

tem

pera

ture

(K)

Coo

lant

tem

pera

ture

(K)

P = 100 kW P = 265 kW

Figure 9 Coolant temperature at the THC D4 at 100 kW and 265 kW

8 Science and Technology of Nuclear Installations

B2 C3

D4D4

0 1 2 3 4 5 6 7 8 9 10 0 1 2 3

3210 4

4 5

5 6

6 7

7 8

8 9

9 10

10

00

4 4

0

4

6

68

12

16 16

16

20

2020

12

12

Void

frac

tion

(times10minus3)

Void

frac

tion

(times10minus3)

Void

frac

tion

(times10minus3)

Time (times103 s) Time (times103 s)

Time (times103 s)

Figure 10 Void fraction time evolution at outlet of the thermal hydraulic channels B2 C3 and D4 after the transient at 100 kW

100 20 30 40 50 60 70 80 90 10066

68

70

72

74

76

78

80

Time (times102 s)

Hea

t flux

(times103

Wm

2)

(a)

0 10 20 30 40 50 60 70 80 90 1000

3

6

9

12

15

18

21

24

27

30

Time (times102 s)

Hea

t tra

nsfe

r coe

ffici

ent(times102

Wm

2K)

(b)

Figure 11 Heat flux (a) and heat transfer coefficient (b) to the heat structure associated with the THC B1 at 100 kW of power operation

5 Conclusions

To perform the calculations a verified model in RELAP5MOD33 was considered The simulated transient character-izes a LOFA type accident Total and partial core coolant

blockage transient have been investigated for the IPR-R1research reactor In the simulation the channels were blockedusing components type valve in the inlet of the channels Inboth cases total and partial blockage the cross flowmodelingin the RELAP5 core nodalization is important in the analyses

Science and Technology of Nuclear Installations 9

Time (s)1000 2000 3000 4000 5000 60000

minus040

minus035

minus030

minus025

minus020

minus015

minus010

minus005

Mas

s flow

rate

(kg

s)

000

005

010

P = 100 kWP = 100 kW

Inlet

Outlet

(a)

Time (s)1000 2000 3000 4000 50000

00

05

10

Mas

s flow

rate

(kg

s)

P = 265 kW

Outlet

Middle = inlet

minus05

minus10

minus15

minus20

minus25

(b)

Figure 12 Mass flow rate at channel D4 after and before the total core blockage at 100 kW (a) and 265 kW (b) of power operation

Outlet

Middle

InletCoo

lant

tem

pera

ture

(K)

300

320

340

360

380

400

Time (s)1000 2000 3000 4000 5000 60000

P = 100 kW

(a)

Outlet

Middle

Inlet

Coo

lant

tem

pera

ture

(K)

300

320

340

360

380

400

Time (s)1000 2000 3000 4000 50000

P = 265 kW

(b)

Figure 13 Coolant temperature in the channel B2 at 100 kW (a) and 265 kW (b) of power operation

300

320

340

360

380

400

Coo

lant

tem

pera

ture

(K)

Time (s)1000 2000 3000 4000 50000

Outlet

MiddleInlet

(a)

Void

frac

tion

0

02

04

06

08

10

Time (s)1000 2000 3000 4000 50000

minus02

Inlet

Middle

Outlet

(b)

Figure 14 Coolant temperature (a) and void fraction (b) evolutions for the THC D4 at 100 kW

10 Science and Technology of Nuclear Installations

300

350

400

450

500

550

600

650

700

0 5 10 15 20 25 30 35 40 45 50Time (times102 s)

Tem

pera

ture

(K)

Fuel

Cladding

Figure 15 Fuel and cladding temperature evolutions for the B1 heatstructure at 265 kW of power operation

because it provides the coolant redistribution among adjacentchannels after the blockage as it is the expected behaviour

After the partial blockage transient an increase in thecoolant temperature of the blocked channels change in themass flow direction and modulus and void formation wereobserved In spite of this the reactor presented safe behaviourafter the transient reaching a new steady state However themodel is not capable of reproducing exactly the behaviour ofthe core due to the changes in the macroscopic cross sectionsthat are directly affected by coolant density and temperature

In the case of the total blockage the saturation coolanttemperature is reached into the core and the coolant boils infew minutes after the beginning of the transient One moretime in an actual situation the intrinsic safety characteristicof fuel shouldwork to control the reactor It was observed thata total core blockage disturbs significantly the core and thereactivity feedback effects should be considered A neutronicanalysis in this case could be necessary to investigate how theneutron flux is perturbed in the core after the total blockagetransient

Conflict of Interests

The authors declare that there is no conflict of interestsregarding the publication of this paper

Acknowledgments

The authors are grateful to CAPES CNPq CDTNCNENand FAPEMIG (all from Brazil) for the support

References

[1] IAEA Safety Analysis for Research Reactors Safety ReportsSeries No 55 International Atomic Energy Agency ViennaAustria 2008

[2] O Merroun A Almers T El Bardouni B El Bakkari andE Chakir ldquoAnalytical benchmarks for verification of thermal-hydraulic codes based on sub-channel approachrdquoNuclear Engi-neering and Design vol 239 no 4 pp 735ndash748 2009

[3] C S Y Suresh G Sateesh S K Das S P Veukateshanand M Rajan ldquoHeat transfer from a totally blocked fuelsubassembly of a liquid metal fast breeder reactor part Iexperimental investigation and part II numerical simulationrdquoNuclear Engineering and Design vol 235 pp 885ndash912 2005

[4] MA F VelosoAnalise Termofluidodinamica de Reatores Nucle-ares de Pesquisa Refrigerados a Agua em Regime de ConveccaoNatural [Doctor thesis] Universidade Estadual de CampinasCampinas Brazil 2004

[5] H M Dalle ldquoAvaliacao Neutronica do Reator TRIGA IPR-R1 ndashR1 com Configuracao de 63 Elementos Combustıveis e Barrade Regulacao em F16rdquo Restrict Document (NIndashEC3-0103)Centro de Desenvolvimento da Tecnologia Nuclear ComissaoNacional de Energia Nuclear Belo Horizonte Brazil 2003

[6] H M Dalle C Pereira and R G P Souza ldquoNeutroniccalculation to theTRIGA Ipr-R1 reactor using theWIMSD4 andCITATION codesrdquo Annals of Nuclear Energy vol 29 no 8 pp901ndash912 2002

[7] R M G P Souza ldquoResultados Experimentais da Calibracaodas Barras de Controle e do Excesso de Reatividade em CincoConfiguracoes do Nucleo do Reator IPR-R1rdquo Tech Rep NI-AT4-00199 CDTNCNEN Belo Horizonte Brasil 1999

[8] P A L Reis A L Costa C Pereira C A M Silva M A FVeloso and A Z Mesquita ldquoSensitivity analysis to a RELAP5nodalization developed for a typical TRIGA research reactorrdquoNuclear Engineering and Design vol 242 pp 300ndash306 2012

[9] A L Costa P A L Reis C Pereira M A F Veloso A ZMesquita and H V Soares ldquoThermal hydraulic analysis ofthe IPR-R1 TRIGA research reactor using a RELAP5 modelrdquoNuclear Engineering and Design vol 240 no 6 pp 1487ndash14942010

[10] P A L Reis A L Costa C Pereira et al ldquoAssessment of aRELAP5model for the IPR-R1 TRIGA research reactorrdquoAnnalsof Nuclear Energy vol 37 no 10 pp 1341ndash1350 2010

[11] M Adorni A Bousbia-Salah T Hamidouche B D MaroF Pierro and F DrsquoAuria ldquoAnalysis of partial and total flowblockage of a single fuel assembly of an MTR research reactorcorerdquo Annals of Nuclear Energy vol 32 no 15 pp 1679ndash16922005

[12] Q Lu S Qiu and G H Su ldquoDevelopment of a thermal-hydraulic analysis code for research reactors with plate fuelsrdquoAnnals of Nuclear Energy vol 36 no 4 pp 433ndash447 2009

[13] A Z Mesquita H C Rezende and R M G P Souza ldquoThermalpower calibrations of the IPR-R1 TRIGA nuclear reactorrdquo inProceedings of the 20th International Congress of MechanicalEngineering (COBEM rsquo09) Gramado Brazil November 2009

[14] RELAP5Mod33 Code Manuals Idaho National EngineeringLaboratory US NRC 2001

[15] W H McAdams Heat Transmission McGraw-Hill Book NewYork NY USA 3rd edition 1954

[16] W Wagner and A Kruse Properties of Water and SteammdashThe Industrial Standard IAPWS-IF97 for the ThermodynamicsProperties Springer Berlin Germany 1998

[17] A Z Mesquita A L Costa C Pereira M A F Veloso andP C A L Reis ldquoExperimental investigation of the onset ofsubcooled nucleate boiling in an open-pool nuclear researchreactorrdquo Journal of ASTM International vol 8 no 6 2011

TribologyAdvances in

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

International Journal of

AerospaceEngineeringHindawi Publishing Corporationhttpwwwhindawicom Volume 2014

FuelsJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Journal ofPetroleum Engineering

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Industrial EngineeringJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Power ElectronicsHindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Advances in

CombustionJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Journal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Renewable Energy

Submit your manuscripts athttpwwwhindawicom

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

StructuresJournal of

International Journal of

RotatingMachinery

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Hindawi Publishing Corporation httpwwwhindawicom

Journal ofEngineeringVolume 2014

Hindawi Publishing Corporation httpwwwhindawicom Volume 2014

International Journal ofPhotoenergy

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Nuclear InstallationsScience and Technology of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Solar EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Wind EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Nuclear EnergyInternational Journal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

High Energy PhysicsAdvances in

The Scientific World JournalHindawi Publishing Corporation httpwwwhindawicom Volume 2014

4 Science and Technology of Nuclear Installations

thermal property data were entered to define the thermalconductivity and volumetric heat capacity as functions oftemperature

The axial power distribution was calculated consideringa cosine profile predicted using the data from Figure 2 andtaking into account the power cutoff in the extremes of theelement due the presence of the graphite The point kineticsmodel was used in the current modeling which assumes thatchanges in reactivity uniformly affect the core flux resultingin no relative spatial variation over time Point kinetics is notindicated when a pronounced flux spike and radial neutronwave occur

3 THCs Blockage Transient

THCs blockage transient has been investigated using theRELAP5 code For the open pool research reactors configu-ration the probability of a blockage in the core upper plenumzone is higher than a blockage from the bottom [11] Thistransient situation may be caused for example by swelling ofthe fuel or fall of some material in the reactor pool leading toblockage of one or more channels [12] A fuel channel block-age event has different characteristics depending on the flowdirectionDownward cooling flow can lead to blockage due toobjects dropping into the pool Upward cooling flow can leadto blockage due to objects inside the primary cooling systempiping being dragged into the core by the action of the pump[1] In this work all the simulations considered the blockageat the core inlet

Two cases of transient were considered In the first casethe aim was to investigate a partial core blockage simulationin the second one it was considered a total core blockageBoth transient cases described respectively in more detailsin Sections 41 and 42 were initiated after the calculationreached steady state condition Two values of power opera-tion 100 kWand 265 kW typical IPR-R1 power operation val-ues were considered To perform the partial blockage simu-lation using the RELAP5 37 components of type valves at theinlet of TH channels in the rings A B C and D were used(Figure 4) The valves were closed at 4000 s To simulate thetotal core blockage the valve number 38 was closed at 4000 sstopping totally the coolant flow in the core

4 Simulation Results

Since each flow channel provides its own driving force it ispossible to consider flow channel independently in a steadystate situation [13] However to perform this specific type oftransient the interaction between all the obstructed channelsand all the adjacent channels must be considered RELAP5 isa one-dimensional code and each hydrodynamic cell has twofaces in the normal direction at the inlet and outlet Then tomore realistically analyse the core behaviour in the case of achannel blockage it is necessary to consider the coolant crossflow Therefore the RELAP5 cross flow junction model [14]has been used to represent the multidimensional phenomenaof the flow interconnecting the channels using componentsof type single junctions

41 Partial Core Coolant Flow Blockage The transient beginsat 4000 s The coolant flow rate in the valves reaches thetotal obstructed condition about 5 s after the beginning of theevent (Figure 5) Due the blockage at inlet of the THCs thecoolant flow changes direction and module to adequate theparameters for the new core operation conditions

Themass flow rate at the inlet of the channel B1 decreasesafter the blockage event reaching zero value (Figure 6) Themass flow rate at the midheight of this channel increasesand changes the flow direction Due the THC obstructionthe coolant flow is redistributed among adjacent channelsAfter the transient a new steady state condition is reached(Figure 7)

Figure 8(a) shows a detail of the nodalization wherethe junctions make possible the cross flow between THCsduring the transient The mass flow rate at the cross flowjunctions between the THCs C2 and D3 in four axial levelsare represented in Figure 8(b) It is possible to observe thatuntil 4000 s of calculation the mass flow rate in the junctionsis practically zero After the transient the coolant flowsalso among the channels characterizing the cross flow andreproducing the actual situation

After the beginning of the transient the coolant temper-ature increases in the obstructed channels at midheight anddue to the change of flow direction the outlet temperaturesdecrease A new steady state is reached As an exampleFigure 9 shows the coolant temperature in three axial levelsof the THC D4 at 100 kW and 265 kW of power operation

After the transient it is possible to observe void formationin the obstructed channels Figure 10 shows the void fractionat the outlet of the blocked channels B2 C3 and D4

411 Heat Transfer Analysis The heat transfer from the clad-ding surface to the water occurs locally in the regime of sub-cooled nucleate boiling so that the cladding surface temper-ature (119879sur) is the saturation temperature of the water plus thewall superheat given by McAdams correlations [15] There-fore the cladding surface temperature 119879sur can be foundusing the expression

119879sur = 119879sat + Δ119879sat (1)

where 119879sat is the water saturation temperature and it isequivalent to 11137∘C or 38454∘K at pressure of 15 bar [16]The Δ119879sat is the wall superheat temperature that is given by

Δ119879sat = 081 (11990210158401015840)0259

(2)

with temperature in ∘C and heat flux 11990210158401015840 in Wm2Specifically for the IPR-R1 TRIGA reactor conditions the

transition point between the single-phase convection regimeto subcooled nucleate boiling regime is approximately at60 kW of power operation [17]

Figure 11 shows the time evolution of the heat flux 11990210158401015840 andthe heat transfer coefficient ℎ

119904 in the heat structure B1 at

Science and Technology of Nuclear Installations 5

D7

D8D9D10D11

D12

D13

D14

D15

D16

D17D18 D1 D2

D3

D4

D6

D6

C2C1C12

C11

C10

C9

C8 C7 C6C5

C4

C3B2

B1

A1B6

B5B4

B3

068F1 F2 000

000

000

000

000

000000

000

000

000

000

000

000

065

000

000

064065

081

080

080

080

080

083

083

084

084

084

084085

085

079

081099099

099

099

099

100

101107113107

102

102

101

100

100131

121

133142

141

128

103 109

153141

142119

131

132130

155

128

131

103

081081082

081

084

084082

082

082066

000

000

000000

000000000000

000000

000

000

000

F3F4

F5

F6E5

E4E3

E2E1E24E23

E22

E21

E20

E19

E18

E17

E16E15

E14 E13 E12E11

E10

E9

E8

E7

E6F7

F9

F8

F10

F11

F12F13

F14F15F16F17

F18F19

F21

F20

F22

F23

F24

F25

F26

F27

F28F29

F30

A B

33

34

3518

19

17

1615

1413

12

1 1

1098

762

31

45

3637 20 21

27

26

25

22

23

24

28293031

32

(a)

Outlet

Inlet

F24 E19 D15 C10 B6 A1 B2 C4 D5 E7 F8

060

100

38

34 17 02 01 06 11 24

(b)

Figure 4 Core nodalization showing in details the positions of the valves used to perform the partial core blockage (a) central cross view ofthe core nodalization (b) The valve number 38 connected to the element number 100 is used to perform the total core blockage

midheight and at 100 kW of power operation for RELAP5calculation

The values of 11990210158401015840 were taken from RELAP5 steady stationcalculation and Δ119879sat and 119879sur were calculated using (1) and(2) The values for 100 kW and 265 kW are respectively

Δ119879sat(100 kW) = 081 (7292788)0259= 1472

∘C

Δ119879sat(265 kW) = 081 (1930676)0259= 1895

∘C(3)

Therefore

119879(sur100 kW) = 12609

∘C

119879(sur265 kW) = 13031

∘C(4)

These results means that the reactor regime is the subcoolednucleate boiling in which 119879sur gt 119879sat but 119879fluid lt 119879sat

Table 1 presents a comparison among experimental andcalculated values of the fuel thermal parameters The calcu-lated parameters using RELAP5 code are taken at midheight

6 Science and Technology of Nuclear Installations

01

03 10

0 4000 6000 8000 10000 1400012000 160002000

Mas

s flow

rate

(kg

s)

minus004

minus002

020

000

002

004

006

008

010

012

014

016

018

Time (s)

(a)

01

03

10

3990 3992 3994 3996 4010400840064004400240003998

Mas

s flow

rate

(kg

s)

minus004

minus002

020

000

002

004

006

008

010

012

014

016

018

Time (s)

(b)

Figure 5 Mass flow rate evolution in valves 1 3 and 10 at 265 kW of power operation tripped at 4000 s (a) The time window shows it indetails (b)

Inlet

Middle

0 2000 4000 6000 8000 100001000 3000 5000 7000 9000Time (s)

minus030

minus025

minus020

minus015

minus010

minus005

035

040

030

025

020

015

010

005

000

Mas

s flow

rate

(kg

s)

Figure 6 Inlet and midheight mass flow rate in the channel B1mdashbefore and after the blockage initiated at 4000 s (power of 265 kW)

(axial level 11) of B1 fuel element and are in good agreementwith the experimental data [17]

The convective heat transfer coefficient as it can be seenin Figure 11 increases after the blockage reaching a newsteady state value As it was verified from the results obtainedthe transient presented variation of the thermal hydraulicparameters that quickly reached new steady state values

42 Total Core Blockage In this situation all the thermalhydraulic channels were blocked using a component typevalve (valve 38 in the Figure 4) closed at 4000 s of calculationtherefore after the steady state conditions were reachedFigure 12 shows themass flow rate at 100 and 265 kWof poweroperation for the THC D4Themass flow rate after the tran-sient oscillates around zero value

After the beginning of the transient the coolant tempera-ture in the core increases reaching the saturation temperatureThe regime now is the saturated or bulk boiling with

Table 1 Experimental and calculated values of some fuel thermalparameters

Power(kW)

11990210158401015840

(Wm2)Δ119879sat(∘C)

Claddingtemperature (∘C)

Experimental 2650 1946130 1900 1304RELAP5 2650 1930676 1895 1303

the coolant temperature remaining essentially constant andequal to 119879sat As an example Figure 13 shows the coolanttemperature in three axial levels of the THC B1 at 100 kW and265 kW of power operation The same behaviour is observedfor all TH core channels

In addition Figure 14 shows the void fraction and thecoolant temperature time evolution in TH channel D4 at265 kW

Specifically to TRIGA reactors the limiting parameteris the fuel temperature The maximum fuel temperature isrelated to the dehydrogenation of uranium and zirconiumhydride and subsequent stress on the cladding The tem-perature at which the dehydrogenation occurs in UZrH

10

is approximately 550∘C This temperature should not beexceeded to avoid volumetric expansion of the UZrH anddeformation of the fuel element [4]

There is no significant increase in the fuel and claddingtemperatures after the total blockage (Figure 15) Even withthe loss of flow and the coolant reaching the saturationtemperature the fuel temperature remains below the safethreshold temperature

The loss of flow accidents is part of a category that involvesweak feedback effects when it occurs out of core However inthe total core blockage significant disturbance of the core wasobserved and also void formation was observed Thereforea more realistic analysis of this type of accident could beperformed considering cross section variations This is notpossible to do using a point kinetics model

Science and Technology of Nuclear Installations 7

A1

E5

B2

C3

D4

A1

E5E5

B2B

C3C

D4D4

A BC D E

F

Time (s)0 2000 4000 6000 8000 10000 12000 14000 16000

minus06

minus05

minus04

minus03

minus02

minus01

05

04

03

02

01

00

Mas

s flow

rate

(kg

s)

Figure 7 Outlet mass flow rate in the channels A1 B2 C3 D4 and E5 before and after the blockage (power of 265 kW)

Axial level 01

Axial level 03

Axial level 07

Axial level 11

Inlet C2

THC D3THC C2

Outlet C2

123456789

101112131415161718192021

123456789

101112131415161718192021

(a)

Time (s)0 2000 4000 6000 8000 10000

00

02

04

06

Axial level 11Axial level 07

Axial level 03Axial level 01

minus02

Mas

s flow

rate

(kg

s)

(b)

Figure 8 Details of the nodalization where the junctions make possible the cross flow between THCs during the transient (a) Mass flow rateat the cross flow junctions between the THCs C2 and D3 in four axial levels (b)

1 2 213 3 44 55 6 6 77 88 99 1010290 290

295 295

300300

305 305

310310

315 315

320 320

325325

330 330

OutletMiddle

Inlet

Outlet

Middle

Inlet

Time (times103 s) Time (times103 s)

Coo

lant

tem

pera

ture

(K)

Coo

lant

tem

pera

ture

(K)

P = 100 kW P = 265 kW

Figure 9 Coolant temperature at the THC D4 at 100 kW and 265 kW

8 Science and Technology of Nuclear Installations

B2 C3

D4D4

0 1 2 3 4 5 6 7 8 9 10 0 1 2 3

3210 4

4 5

5 6

6 7

7 8

8 9

9 10

10

00

4 4

0

4

6

68

12

16 16

16

20

2020

12

12

Void

frac

tion

(times10minus3)

Void

frac

tion

(times10minus3)

Void

frac

tion

(times10minus3)

Time (times103 s) Time (times103 s)

Time (times103 s)

Figure 10 Void fraction time evolution at outlet of the thermal hydraulic channels B2 C3 and D4 after the transient at 100 kW

100 20 30 40 50 60 70 80 90 10066

68

70

72

74

76

78

80

Time (times102 s)

Hea

t flux

(times103

Wm

2)

(a)

0 10 20 30 40 50 60 70 80 90 1000

3

6

9

12

15

18

21

24

27

30

Time (times102 s)

Hea

t tra

nsfe

r coe

ffici

ent(times102

Wm

2K)

(b)

Figure 11 Heat flux (a) and heat transfer coefficient (b) to the heat structure associated with the THC B1 at 100 kW of power operation

5 Conclusions

To perform the calculations a verified model in RELAP5MOD33 was considered The simulated transient character-izes a LOFA type accident Total and partial core coolant

blockage transient have been investigated for the IPR-R1research reactor In the simulation the channels were blockedusing components type valve in the inlet of the channels Inboth cases total and partial blockage the cross flowmodelingin the RELAP5 core nodalization is important in the analyses

Science and Technology of Nuclear Installations 9

Time (s)1000 2000 3000 4000 5000 60000

minus040

minus035

minus030

minus025

minus020

minus015

minus010

minus005

Mas

s flow

rate

(kg

s)

000

005

010

P = 100 kWP = 100 kW

Inlet

Outlet

(a)

Time (s)1000 2000 3000 4000 50000

00

05

10

Mas

s flow

rate

(kg

s)

P = 265 kW

Outlet

Middle = inlet

minus05

minus10

minus15

minus20

minus25

(b)

Figure 12 Mass flow rate at channel D4 after and before the total core blockage at 100 kW (a) and 265 kW (b) of power operation

Outlet

Middle

InletCoo

lant

tem

pera

ture

(K)

300

320

340

360

380

400

Time (s)1000 2000 3000 4000 5000 60000

P = 100 kW

(a)

Outlet

Middle

Inlet

Coo

lant

tem

pera

ture

(K)

300

320

340

360

380

400

Time (s)1000 2000 3000 4000 50000

P = 265 kW

(b)

Figure 13 Coolant temperature in the channel B2 at 100 kW (a) and 265 kW (b) of power operation

300

320

340

360

380

400

Coo

lant

tem

pera

ture

(K)

Time (s)1000 2000 3000 4000 50000

Outlet

MiddleInlet

(a)

Void

frac

tion

0

02

04

06

08

10

Time (s)1000 2000 3000 4000 50000

minus02

Inlet

Middle

Outlet

(b)

Figure 14 Coolant temperature (a) and void fraction (b) evolutions for the THC D4 at 100 kW

10 Science and Technology of Nuclear Installations

300

350

400

450

500

550

600

650

700

0 5 10 15 20 25 30 35 40 45 50Time (times102 s)

Tem

pera

ture

(K)

Fuel

Cladding

Figure 15 Fuel and cladding temperature evolutions for the B1 heatstructure at 265 kW of power operation

because it provides the coolant redistribution among adjacentchannels after the blockage as it is the expected behaviour

After the partial blockage transient an increase in thecoolant temperature of the blocked channels change in themass flow direction and modulus and void formation wereobserved In spite of this the reactor presented safe behaviourafter the transient reaching a new steady state However themodel is not capable of reproducing exactly the behaviour ofthe core due to the changes in the macroscopic cross sectionsthat are directly affected by coolant density and temperature

In the case of the total blockage the saturation coolanttemperature is reached into the core and the coolant boils infew minutes after the beginning of the transient One moretime in an actual situation the intrinsic safety characteristicof fuel shouldwork to control the reactor It was observed thata total core blockage disturbs significantly the core and thereactivity feedback effects should be considered A neutronicanalysis in this case could be necessary to investigate how theneutron flux is perturbed in the core after the total blockagetransient

Conflict of Interests

The authors declare that there is no conflict of interestsregarding the publication of this paper

Acknowledgments

The authors are grateful to CAPES CNPq CDTNCNENand FAPEMIG (all from Brazil) for the support

References

[1] IAEA Safety Analysis for Research Reactors Safety ReportsSeries No 55 International Atomic Energy Agency ViennaAustria 2008

[2] O Merroun A Almers T El Bardouni B El Bakkari andE Chakir ldquoAnalytical benchmarks for verification of thermal-hydraulic codes based on sub-channel approachrdquoNuclear Engi-neering and Design vol 239 no 4 pp 735ndash748 2009

[3] C S Y Suresh G Sateesh S K Das S P Veukateshanand M Rajan ldquoHeat transfer from a totally blocked fuelsubassembly of a liquid metal fast breeder reactor part Iexperimental investigation and part II numerical simulationrdquoNuclear Engineering and Design vol 235 pp 885ndash912 2005

[4] MA F VelosoAnalise Termofluidodinamica de Reatores Nucle-ares de Pesquisa Refrigerados a Agua em Regime de ConveccaoNatural [Doctor thesis] Universidade Estadual de CampinasCampinas Brazil 2004

[5] H M Dalle ldquoAvaliacao Neutronica do Reator TRIGA IPR-R1 ndashR1 com Configuracao de 63 Elementos Combustıveis e Barrade Regulacao em F16rdquo Restrict Document (NIndashEC3-0103)Centro de Desenvolvimento da Tecnologia Nuclear ComissaoNacional de Energia Nuclear Belo Horizonte Brazil 2003

[6] H M Dalle C Pereira and R G P Souza ldquoNeutroniccalculation to theTRIGA Ipr-R1 reactor using theWIMSD4 andCITATION codesrdquo Annals of Nuclear Energy vol 29 no 8 pp901ndash912 2002

[7] R M G P Souza ldquoResultados Experimentais da Calibracaodas Barras de Controle e do Excesso de Reatividade em CincoConfiguracoes do Nucleo do Reator IPR-R1rdquo Tech Rep NI-AT4-00199 CDTNCNEN Belo Horizonte Brasil 1999

[8] P A L Reis A L Costa C Pereira C A M Silva M A FVeloso and A Z Mesquita ldquoSensitivity analysis to a RELAP5nodalization developed for a typical TRIGA research reactorrdquoNuclear Engineering and Design vol 242 pp 300ndash306 2012

[9] A L Costa P A L Reis C Pereira M A F Veloso A ZMesquita and H V Soares ldquoThermal hydraulic analysis ofthe IPR-R1 TRIGA research reactor using a RELAP5 modelrdquoNuclear Engineering and Design vol 240 no 6 pp 1487ndash14942010

[10] P A L Reis A L Costa C Pereira et al ldquoAssessment of aRELAP5model for the IPR-R1 TRIGA research reactorrdquoAnnalsof Nuclear Energy vol 37 no 10 pp 1341ndash1350 2010

[11] M Adorni A Bousbia-Salah T Hamidouche B D MaroF Pierro and F DrsquoAuria ldquoAnalysis of partial and total flowblockage of a single fuel assembly of an MTR research reactorcorerdquo Annals of Nuclear Energy vol 32 no 15 pp 1679ndash16922005

[12] Q Lu S Qiu and G H Su ldquoDevelopment of a thermal-hydraulic analysis code for research reactors with plate fuelsrdquoAnnals of Nuclear Energy vol 36 no 4 pp 433ndash447 2009

[13] A Z Mesquita H C Rezende and R M G P Souza ldquoThermalpower calibrations of the IPR-R1 TRIGA nuclear reactorrdquo inProceedings of the 20th International Congress of MechanicalEngineering (COBEM rsquo09) Gramado Brazil November 2009

[14] RELAP5Mod33 Code Manuals Idaho National EngineeringLaboratory US NRC 2001

[15] W H McAdams Heat Transmission McGraw-Hill Book NewYork NY USA 3rd edition 1954

[16] W Wagner and A Kruse Properties of Water and SteammdashThe Industrial Standard IAPWS-IF97 for the ThermodynamicsProperties Springer Berlin Germany 1998

[17] A Z Mesquita A L Costa C Pereira M A F Veloso andP C A L Reis ldquoExperimental investigation of the onset ofsubcooled nucleate boiling in an open-pool nuclear researchreactorrdquo Journal of ASTM International vol 8 no 6 2011

TribologyAdvances in

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

International Journal of

AerospaceEngineeringHindawi Publishing Corporationhttpwwwhindawicom Volume 2014

FuelsJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Journal ofPetroleum Engineering

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Industrial EngineeringJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Power ElectronicsHindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Advances in

CombustionJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Journal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Renewable Energy

Submit your manuscripts athttpwwwhindawicom

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

StructuresJournal of

International Journal of

RotatingMachinery

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Hindawi Publishing Corporation httpwwwhindawicom

Journal ofEngineeringVolume 2014

Hindawi Publishing Corporation httpwwwhindawicom Volume 2014

International Journal ofPhotoenergy

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Nuclear InstallationsScience and Technology of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Solar EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Wind EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Nuclear EnergyInternational Journal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

High Energy PhysicsAdvances in

The Scientific World JournalHindawi Publishing Corporation httpwwwhindawicom Volume 2014

Science and Technology of Nuclear Installations 5

D7

D8D9D10D11

D12

D13

D14

D15

D16

D17D18 D1 D2

D3

D4

D6

D6

C2C1C12

C11

C10

C9

C8 C7 C6C5

C4

C3B2

B1

A1B6

B5B4

B3

068F1 F2 000

000

000

000

000

000000

000

000

000

000

000

000

065

000

000

064065

081

080

080

080

080

083

083

084

084

084

084085

085

079

081099099

099

099

099

100

101107113107

102

102

101

100

100131

121

133142

141

128

103 109

153141

142119

131

132130

155

128

131

103

081081082

081

084

084082

082

082066

000

000

000000

000000000000

000000

000

000

000

F3F4

F5

F6E5

E4E3

E2E1E24E23

E22

E21

E20

E19

E18

E17

E16E15

E14 E13 E12E11

E10

E9

E8

E7

E6F7

F9

F8

F10

F11

F12F13

F14F15F16F17

F18F19

F21

F20

F22

F23

F24

F25

F26

F27

F28F29

F30

A B

33

34

3518

19

17

1615

1413

12

1 1

1098

762

31

45

3637 20 21

27

26

25

22

23

24

28293031

32

(a)

Outlet

Inlet

F24 E19 D15 C10 B6 A1 B2 C4 D5 E7 F8

060

100

38

34 17 02 01 06 11 24

(b)

Figure 4 Core nodalization showing in details the positions of the valves used to perform the partial core blockage (a) central cross view ofthe core nodalization (b) The valve number 38 connected to the element number 100 is used to perform the total core blockage

midheight and at 100 kW of power operation for RELAP5calculation

The values of 11990210158401015840 were taken from RELAP5 steady stationcalculation and Δ119879sat and 119879sur were calculated using (1) and(2) The values for 100 kW and 265 kW are respectively

Δ119879sat(100 kW) = 081 (7292788)0259= 1472

∘C

Δ119879sat(265 kW) = 081 (1930676)0259= 1895

∘C(3)

Therefore

119879(sur100 kW) = 12609

∘C

119879(sur265 kW) = 13031

∘C(4)

These results means that the reactor regime is the subcoolednucleate boiling in which 119879sur gt 119879sat but 119879fluid lt 119879sat

Table 1 presents a comparison among experimental andcalculated values of the fuel thermal parameters The calcu-lated parameters using RELAP5 code are taken at midheight

6 Science and Technology of Nuclear Installations

01

03 10

0 4000 6000 8000 10000 1400012000 160002000

Mas

s flow

rate

(kg

s)

minus004

minus002

020

000

002

004

006

008

010

012

014

016

018

Time (s)

(a)

01

03

10

3990 3992 3994 3996 4010400840064004400240003998

Mas

s flow

rate

(kg

s)

minus004

minus002

020

000

002

004

006

008

010

012

014

016

018

Time (s)

(b)

Figure 5 Mass flow rate evolution in valves 1 3 and 10 at 265 kW of power operation tripped at 4000 s (a) The time window shows it indetails (b)

Inlet

Middle

0 2000 4000 6000 8000 100001000 3000 5000 7000 9000Time (s)

minus030

minus025

minus020

minus015

minus010

minus005

035

040

030

025

020

015

010

005

000

Mas

s flow

rate

(kg

s)

Figure 6 Inlet and midheight mass flow rate in the channel B1mdashbefore and after the blockage initiated at 4000 s (power of 265 kW)

(axial level 11) of B1 fuel element and are in good agreementwith the experimental data [17]

The convective heat transfer coefficient as it can be seenin Figure 11 increases after the blockage reaching a newsteady state value As it was verified from the results obtainedthe transient presented variation of the thermal hydraulicparameters that quickly reached new steady state values

42 Total Core Blockage In this situation all the thermalhydraulic channels were blocked using a component typevalve (valve 38 in the Figure 4) closed at 4000 s of calculationtherefore after the steady state conditions were reachedFigure 12 shows themass flow rate at 100 and 265 kWof poweroperation for the THC D4Themass flow rate after the tran-sient oscillates around zero value

After the beginning of the transient the coolant tempera-ture in the core increases reaching the saturation temperatureThe regime now is the saturated or bulk boiling with

Table 1 Experimental and calculated values of some fuel thermalparameters

Power(kW)

11990210158401015840

(Wm2)Δ119879sat(∘C)

Claddingtemperature (∘C)

Experimental 2650 1946130 1900 1304RELAP5 2650 1930676 1895 1303

the coolant temperature remaining essentially constant andequal to 119879sat As an example Figure 13 shows the coolanttemperature in three axial levels of the THC B1 at 100 kW and265 kW of power operation The same behaviour is observedfor all TH core channels

In addition Figure 14 shows the void fraction and thecoolant temperature time evolution in TH channel D4 at265 kW

Specifically to TRIGA reactors the limiting parameteris the fuel temperature The maximum fuel temperature isrelated to the dehydrogenation of uranium and zirconiumhydride and subsequent stress on the cladding The tem-perature at which the dehydrogenation occurs in UZrH

10

is approximately 550∘C This temperature should not beexceeded to avoid volumetric expansion of the UZrH anddeformation of the fuel element [4]

There is no significant increase in the fuel and claddingtemperatures after the total blockage (Figure 15) Even withthe loss of flow and the coolant reaching the saturationtemperature the fuel temperature remains below the safethreshold temperature

The loss of flow accidents is part of a category that involvesweak feedback effects when it occurs out of core However inthe total core blockage significant disturbance of the core wasobserved and also void formation was observed Thereforea more realistic analysis of this type of accident could beperformed considering cross section variations This is notpossible to do using a point kinetics model

Science and Technology of Nuclear Installations 7

A1

E5

B2

C3

D4

A1

E5E5

B2B

C3C

D4D4

A BC D E

F

Time (s)0 2000 4000 6000 8000 10000 12000 14000 16000

minus06

minus05

minus04

minus03

minus02

minus01

05

04

03

02

01

00

Mas

s flow

rate

(kg

s)

Figure 7 Outlet mass flow rate in the channels A1 B2 C3 D4 and E5 before and after the blockage (power of 265 kW)

Axial level 01

Axial level 03

Axial level 07

Axial level 11

Inlet C2

THC D3THC C2

Outlet C2

123456789

101112131415161718192021

123456789

101112131415161718192021

(a)

Time (s)0 2000 4000 6000 8000 10000

00

02

04

06

Axial level 11Axial level 07

Axial level 03Axial level 01

minus02

Mas

s flow

rate

(kg

s)

(b)

Figure 8 Details of the nodalization where the junctions make possible the cross flow between THCs during the transient (a) Mass flow rateat the cross flow junctions between the THCs C2 and D3 in four axial levels (b)

1 2 213 3 44 55 6 6 77 88 99 1010290 290

295 295

300300

305 305

310310

315 315

320 320

325325

330 330

OutletMiddle

Inlet

Outlet

Middle

Inlet

Time (times103 s) Time (times103 s)

Coo

lant

tem

pera

ture

(K)

Coo

lant

tem

pera

ture

(K)

P = 100 kW P = 265 kW

Figure 9 Coolant temperature at the THC D4 at 100 kW and 265 kW

8 Science and Technology of Nuclear Installations

B2 C3

D4D4

0 1 2 3 4 5 6 7 8 9 10 0 1 2 3

3210 4

4 5

5 6

6 7

7 8

8 9

9 10

10

00

4 4

0

4

6

68

12

16 16

16

20

2020

12

12

Void

frac

tion

(times10minus3)

Void

frac

tion

(times10minus3)

Void

frac

tion

(times10minus3)

Time (times103 s) Time (times103 s)

Time (times103 s)

Figure 10 Void fraction time evolution at outlet of the thermal hydraulic channels B2 C3 and D4 after the transient at 100 kW

100 20 30 40 50 60 70 80 90 10066

68

70

72

74

76

78

80

Time (times102 s)

Hea

t flux

(times103

Wm

2)

(a)

0 10 20 30 40 50 60 70 80 90 1000

3

6

9

12

15

18

21

24

27

30

Time (times102 s)

Hea

t tra

nsfe

r coe

ffici

ent(times102

Wm

2K)

(b)

Figure 11 Heat flux (a) and heat transfer coefficient (b) to the heat structure associated with the THC B1 at 100 kW of power operation

5 Conclusions

To perform the calculations a verified model in RELAP5MOD33 was considered The simulated transient character-izes a LOFA type accident Total and partial core coolant

blockage transient have been investigated for the IPR-R1research reactor In the simulation the channels were blockedusing components type valve in the inlet of the channels Inboth cases total and partial blockage the cross flowmodelingin the RELAP5 core nodalization is important in the analyses

Science and Technology of Nuclear Installations 9

Time (s)1000 2000 3000 4000 5000 60000

minus040

minus035

minus030

minus025

minus020

minus015

minus010

minus005

Mas

s flow

rate

(kg

s)

000

005

010

P = 100 kWP = 100 kW

Inlet

Outlet

(a)

Time (s)1000 2000 3000 4000 50000

00

05

10

Mas

s flow

rate

(kg

s)

P = 265 kW

Outlet

Middle = inlet

minus05

minus10

minus15

minus20

minus25

(b)

Figure 12 Mass flow rate at channel D4 after and before the total core blockage at 100 kW (a) and 265 kW (b) of power operation

Outlet

Middle

InletCoo

lant

tem

pera

ture

(K)

300

320

340

360

380

400

Time (s)1000 2000 3000 4000 5000 60000

P = 100 kW

(a)

Outlet

Middle

Inlet

Coo

lant

tem

pera

ture

(K)

300

320

340

360

380

400

Time (s)1000 2000 3000 4000 50000

P = 265 kW

(b)

Figure 13 Coolant temperature in the channel B2 at 100 kW (a) and 265 kW (b) of power operation

300

320

340

360

380

400

Coo

lant

tem

pera

ture

(K)

Time (s)1000 2000 3000 4000 50000

Outlet

MiddleInlet

(a)

Void

frac

tion

0

02

04

06

08

10

Time (s)1000 2000 3000 4000 50000

minus02

Inlet

Middle

Outlet

(b)

Figure 14 Coolant temperature (a) and void fraction (b) evolutions for the THC D4 at 100 kW

10 Science and Technology of Nuclear Installations

300

350

400

450

500

550

600

650

700

0 5 10 15 20 25 30 35 40 45 50Time (times102 s)

Tem

pera

ture

(K)

Fuel

Cladding

Figure 15 Fuel and cladding temperature evolutions for the B1 heatstructure at 265 kW of power operation

because it provides the coolant redistribution among adjacentchannels after the blockage as it is the expected behaviour

After the partial blockage transient an increase in thecoolant temperature of the blocked channels change in themass flow direction and modulus and void formation wereobserved In spite of this the reactor presented safe behaviourafter the transient reaching a new steady state However themodel is not capable of reproducing exactly the behaviour ofthe core due to the changes in the macroscopic cross sectionsthat are directly affected by coolant density and temperature

In the case of the total blockage the saturation coolanttemperature is reached into the core and the coolant boils infew minutes after the beginning of the transient One moretime in an actual situation the intrinsic safety characteristicof fuel shouldwork to control the reactor It was observed thata total core blockage disturbs significantly the core and thereactivity feedback effects should be considered A neutronicanalysis in this case could be necessary to investigate how theneutron flux is perturbed in the core after the total blockagetransient

Conflict of Interests

The authors declare that there is no conflict of interestsregarding the publication of this paper

Acknowledgments

The authors are grateful to CAPES CNPq CDTNCNENand FAPEMIG (all from Brazil) for the support

References

[1] IAEA Safety Analysis for Research Reactors Safety ReportsSeries No 55 International Atomic Energy Agency ViennaAustria 2008

[2] O Merroun A Almers T El Bardouni B El Bakkari andE Chakir ldquoAnalytical benchmarks for verification of thermal-hydraulic codes based on sub-channel approachrdquoNuclear Engi-neering and Design vol 239 no 4 pp 735ndash748 2009

[3] C S Y Suresh G Sateesh S K Das S P Veukateshanand M Rajan ldquoHeat transfer from a totally blocked fuelsubassembly of a liquid metal fast breeder reactor part Iexperimental investigation and part II numerical simulationrdquoNuclear Engineering and Design vol 235 pp 885ndash912 2005

[4] MA F VelosoAnalise Termofluidodinamica de Reatores Nucle-ares de Pesquisa Refrigerados a Agua em Regime de ConveccaoNatural [Doctor thesis] Universidade Estadual de CampinasCampinas Brazil 2004

[5] H M Dalle ldquoAvaliacao Neutronica do Reator TRIGA IPR-R1 ndashR1 com Configuracao de 63 Elementos Combustıveis e Barrade Regulacao em F16rdquo Restrict Document (NIndashEC3-0103)Centro de Desenvolvimento da Tecnologia Nuclear ComissaoNacional de Energia Nuclear Belo Horizonte Brazil 2003

[6] H M Dalle C Pereira and R G P Souza ldquoNeutroniccalculation to theTRIGA Ipr-R1 reactor using theWIMSD4 andCITATION codesrdquo Annals of Nuclear Energy vol 29 no 8 pp901ndash912 2002

[7] R M G P Souza ldquoResultados Experimentais da Calibracaodas Barras de Controle e do Excesso de Reatividade em CincoConfiguracoes do Nucleo do Reator IPR-R1rdquo Tech Rep NI-AT4-00199 CDTNCNEN Belo Horizonte Brasil 1999

[8] P A L Reis A L Costa C Pereira C A M Silva M A FVeloso and A Z Mesquita ldquoSensitivity analysis to a RELAP5nodalization developed for a typical TRIGA research reactorrdquoNuclear Engineering and Design vol 242 pp 300ndash306 2012

[9] A L Costa P A L Reis C Pereira M A F Veloso A ZMesquita and H V Soares ldquoThermal hydraulic analysis ofthe IPR-R1 TRIGA research reactor using a RELAP5 modelrdquoNuclear Engineering and Design vol 240 no 6 pp 1487ndash14942010

[10] P A L Reis A L Costa C Pereira et al ldquoAssessment of aRELAP5model for the IPR-R1 TRIGA research reactorrdquoAnnalsof Nuclear Energy vol 37 no 10 pp 1341ndash1350 2010

[11] M Adorni A Bousbia-Salah T Hamidouche B D MaroF Pierro and F DrsquoAuria ldquoAnalysis of partial and total flowblockage of a single fuel assembly of an MTR research reactorcorerdquo Annals of Nuclear Energy vol 32 no 15 pp 1679ndash16922005

[12] Q Lu S Qiu and G H Su ldquoDevelopment of a thermal-hydraulic analysis code for research reactors with plate fuelsrdquoAnnals of Nuclear Energy vol 36 no 4 pp 433ndash447 2009

[13] A Z Mesquita H C Rezende and R M G P Souza ldquoThermalpower calibrations of the IPR-R1 TRIGA nuclear reactorrdquo inProceedings of the 20th International Congress of MechanicalEngineering (COBEM rsquo09) Gramado Brazil November 2009

[14] RELAP5Mod33 Code Manuals Idaho National EngineeringLaboratory US NRC 2001

[15] W H McAdams Heat Transmission McGraw-Hill Book NewYork NY USA 3rd edition 1954

[16] W Wagner and A Kruse Properties of Water and SteammdashThe Industrial Standard IAPWS-IF97 for the ThermodynamicsProperties Springer Berlin Germany 1998

[17] A Z Mesquita A L Costa C Pereira M A F Veloso andP C A L Reis ldquoExperimental investigation of the onset ofsubcooled nucleate boiling in an open-pool nuclear researchreactorrdquo Journal of ASTM International vol 8 no 6 2011

TribologyAdvances in

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

International Journal of

AerospaceEngineeringHindawi Publishing Corporationhttpwwwhindawicom Volume 2014

FuelsJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Journal ofPetroleum Engineering

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Industrial EngineeringJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Power ElectronicsHindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Advances in

CombustionJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Journal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Renewable Energy

Submit your manuscripts athttpwwwhindawicom

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

StructuresJournal of

International Journal of

RotatingMachinery

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Hindawi Publishing Corporation httpwwwhindawicom

Journal ofEngineeringVolume 2014

Hindawi Publishing Corporation httpwwwhindawicom Volume 2014

International Journal ofPhotoenergy

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Nuclear InstallationsScience and Technology of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Solar EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Wind EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Nuclear EnergyInternational Journal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

High Energy PhysicsAdvances in

The Scientific World JournalHindawi Publishing Corporation httpwwwhindawicom Volume 2014

6 Science and Technology of Nuclear Installations

01

03 10

0 4000 6000 8000 10000 1400012000 160002000

Mas

s flow

rate

(kg

s)

minus004

minus002

020

000

002

004

006

008

010

012

014

016

018

Time (s)

(a)

01

03

10

3990 3992 3994 3996 4010400840064004400240003998

Mas

s flow

rate

(kg

s)

minus004

minus002

020

000

002

004

006

008

010

012

014

016

018

Time (s)

(b)

Figure 5 Mass flow rate evolution in valves 1 3 and 10 at 265 kW of power operation tripped at 4000 s (a) The time window shows it indetails (b)

Inlet

Middle

0 2000 4000 6000 8000 100001000 3000 5000 7000 9000Time (s)

minus030

minus025

minus020

minus015

minus010

minus005

035

040

030

025

020

015

010

005

000

Mas

s flow

rate

(kg

s)

Figure 6 Inlet and midheight mass flow rate in the channel B1mdashbefore and after the blockage initiated at 4000 s (power of 265 kW)

(axial level 11) of B1 fuel element and are in good agreementwith the experimental data [17]

The convective heat transfer coefficient as it can be seenin Figure 11 increases after the blockage reaching a newsteady state value As it was verified from the results obtainedthe transient presented variation of the thermal hydraulicparameters that quickly reached new steady state values

42 Total Core Blockage In this situation all the thermalhydraulic channels were blocked using a component typevalve (valve 38 in the Figure 4) closed at 4000 s of calculationtherefore after the steady state conditions were reachedFigure 12 shows themass flow rate at 100 and 265 kWof poweroperation for the THC D4Themass flow rate after the tran-sient oscillates around zero value

After the beginning of the transient the coolant tempera-ture in the core increases reaching the saturation temperatureThe regime now is the saturated or bulk boiling with

Table 1 Experimental and calculated values of some fuel thermalparameters

Power(kW)

11990210158401015840

(Wm2)Δ119879sat(∘C)

Claddingtemperature (∘C)

Experimental 2650 1946130 1900 1304RELAP5 2650 1930676 1895 1303

the coolant temperature remaining essentially constant andequal to 119879sat As an example Figure 13 shows the coolanttemperature in three axial levels of the THC B1 at 100 kW and265 kW of power operation The same behaviour is observedfor all TH core channels

In addition Figure 14 shows the void fraction and thecoolant temperature time evolution in TH channel D4 at265 kW

Specifically to TRIGA reactors the limiting parameteris the fuel temperature The maximum fuel temperature isrelated to the dehydrogenation of uranium and zirconiumhydride and subsequent stress on the cladding The tem-perature at which the dehydrogenation occurs in UZrH

10

is approximately 550∘C This temperature should not beexceeded to avoid volumetric expansion of the UZrH anddeformation of the fuel element [4]

There is no significant increase in the fuel and claddingtemperatures after the total blockage (Figure 15) Even withthe loss of flow and the coolant reaching the saturationtemperature the fuel temperature remains below the safethreshold temperature

The loss of flow accidents is part of a category that involvesweak feedback effects when it occurs out of core However inthe total core blockage significant disturbance of the core wasobserved and also void formation was observed Thereforea more realistic analysis of this type of accident could beperformed considering cross section variations This is notpossible to do using a point kinetics model

Science and Technology of Nuclear Installations 7

A1

E5

B2

C3

D4

A1

E5E5

B2B

C3C

D4D4

A BC D E

F

Time (s)0 2000 4000 6000 8000 10000 12000 14000 16000

minus06

minus05

minus04

minus03

minus02

minus01

05

04

03

02

01

00

Mas

s flow

rate

(kg

s)

Figure 7 Outlet mass flow rate in the channels A1 B2 C3 D4 and E5 before and after the blockage (power of 265 kW)

Axial level 01

Axial level 03

Axial level 07

Axial level 11

Inlet C2

THC D3THC C2

Outlet C2

123456789

101112131415161718192021

123456789

101112131415161718192021

(a)

Time (s)0 2000 4000 6000 8000 10000

00

02

04

06

Axial level 11Axial level 07

Axial level 03Axial level 01

minus02

Mas

s flow

rate

(kg

s)

(b)

Figure 8 Details of the nodalization where the junctions make possible the cross flow between THCs during the transient (a) Mass flow rateat the cross flow junctions between the THCs C2 and D3 in four axial levels (b)

1 2 213 3 44 55 6 6 77 88 99 1010290 290

295 295

300300

305 305

310310

315 315

320 320

325325

330 330

OutletMiddle

Inlet

Outlet

Middle

Inlet

Time (times103 s) Time (times103 s)

Coo

lant

tem

pera

ture

(K)

Coo

lant

tem

pera

ture

(K)

P = 100 kW P = 265 kW

Figure 9 Coolant temperature at the THC D4 at 100 kW and 265 kW

8 Science and Technology of Nuclear Installations

B2 C3

D4D4

0 1 2 3 4 5 6 7 8 9 10 0 1 2 3

3210 4

4 5

5 6

6 7

7 8

8 9

9 10

10

00

4 4

0

4

6

68

12

16 16

16

20

2020

12

12

Void

frac

tion

(times10minus3)

Void

frac

tion

(times10minus3)

Void

frac

tion

(times10minus3)

Time (times103 s) Time (times103 s)

Time (times103 s)

Figure 10 Void fraction time evolution at outlet of the thermal hydraulic channels B2 C3 and D4 after the transient at 100 kW

100 20 30 40 50 60 70 80 90 10066

68

70

72

74

76

78

80

Time (times102 s)

Hea

t flux

(times103

Wm

2)

(a)

0 10 20 30 40 50 60 70 80 90 1000

3

6

9

12

15

18

21

24

27

30

Time (times102 s)

Hea

t tra

nsfe

r coe

ffici

ent(times102

Wm

2K)

(b)

Figure 11 Heat flux (a) and heat transfer coefficient (b) to the heat structure associated with the THC B1 at 100 kW of power operation

5 Conclusions

To perform the calculations a verified model in RELAP5MOD33 was considered The simulated transient character-izes a LOFA type accident Total and partial core coolant

blockage transient have been investigated for the IPR-R1research reactor In the simulation the channels were blockedusing components type valve in the inlet of the channels Inboth cases total and partial blockage the cross flowmodelingin the RELAP5 core nodalization is important in the analyses

Science and Technology of Nuclear Installations 9

Time (s)1000 2000 3000 4000 5000 60000

minus040

minus035

minus030

minus025

minus020

minus015

minus010

minus005

Mas

s flow

rate

(kg

s)

000

005

010

P = 100 kWP = 100 kW

Inlet

Outlet

(a)

Time (s)1000 2000 3000 4000 50000

00

05

10

Mas

s flow

rate

(kg

s)

P = 265 kW

Outlet

Middle = inlet

minus05

minus10

minus15

minus20

minus25

(b)

Figure 12 Mass flow rate at channel D4 after and before the total core blockage at 100 kW (a) and 265 kW (b) of power operation

Outlet

Middle

InletCoo

lant

tem

pera

ture

(K)

300

320

340

360

380

400

Time (s)1000 2000 3000 4000 5000 60000

P = 100 kW

(a)

Outlet

Middle

Inlet

Coo

lant

tem

pera

ture

(K)

300

320

340

360

380

400

Time (s)1000 2000 3000 4000 50000

P = 265 kW

(b)

Figure 13 Coolant temperature in the channel B2 at 100 kW (a) and 265 kW (b) of power operation

300

320

340

360

380

400

Coo

lant

tem

pera

ture

(K)

Time (s)1000 2000 3000 4000 50000

Outlet

MiddleInlet

(a)

Void

frac

tion

0

02

04

06

08

10

Time (s)1000 2000 3000 4000 50000

minus02

Inlet

Middle

Outlet

(b)

Figure 14 Coolant temperature (a) and void fraction (b) evolutions for the THC D4 at 100 kW

10 Science and Technology of Nuclear Installations

300

350

400

450

500

550

600

650

700

0 5 10 15 20 25 30 35 40 45 50Time (times102 s)

Tem

pera

ture

(K)

Fuel

Cladding

Figure 15 Fuel and cladding temperature evolutions for the B1 heatstructure at 265 kW of power operation

because it provides the coolant redistribution among adjacentchannels after the blockage as it is the expected behaviour

After the partial blockage transient an increase in thecoolant temperature of the blocked channels change in themass flow direction and modulus and void formation wereobserved In spite of this the reactor presented safe behaviourafter the transient reaching a new steady state However themodel is not capable of reproducing exactly the behaviour ofthe core due to the changes in the macroscopic cross sectionsthat are directly affected by coolant density and temperature

In the case of the total blockage the saturation coolanttemperature is reached into the core and the coolant boils infew minutes after the beginning of the transient One moretime in an actual situation the intrinsic safety characteristicof fuel shouldwork to control the reactor It was observed thata total core blockage disturbs significantly the core and thereactivity feedback effects should be considered A neutronicanalysis in this case could be necessary to investigate how theneutron flux is perturbed in the core after the total blockagetransient

Conflict of Interests

The authors declare that there is no conflict of interestsregarding the publication of this paper

Acknowledgments

The authors are grateful to CAPES CNPq CDTNCNENand FAPEMIG (all from Brazil) for the support

References

[1] IAEA Safety Analysis for Research Reactors Safety ReportsSeries No 55 International Atomic Energy Agency ViennaAustria 2008

[2] O Merroun A Almers T El Bardouni B El Bakkari andE Chakir ldquoAnalytical benchmarks for verification of thermal-hydraulic codes based on sub-channel approachrdquoNuclear Engi-neering and Design vol 239 no 4 pp 735ndash748 2009

[3] C S Y Suresh G Sateesh S K Das S P Veukateshanand M Rajan ldquoHeat transfer from a totally blocked fuelsubassembly of a liquid metal fast breeder reactor part Iexperimental investigation and part II numerical simulationrdquoNuclear Engineering and Design vol 235 pp 885ndash912 2005

[4] MA F VelosoAnalise Termofluidodinamica de Reatores Nucle-ares de Pesquisa Refrigerados a Agua em Regime de ConveccaoNatural [Doctor thesis] Universidade Estadual de CampinasCampinas Brazil 2004

[5] H M Dalle ldquoAvaliacao Neutronica do Reator TRIGA IPR-R1 ndashR1 com Configuracao de 63 Elementos Combustıveis e Barrade Regulacao em F16rdquo Restrict Document (NIndashEC3-0103)Centro de Desenvolvimento da Tecnologia Nuclear ComissaoNacional de Energia Nuclear Belo Horizonte Brazil 2003

[6] H M Dalle C Pereira and R G P Souza ldquoNeutroniccalculation to theTRIGA Ipr-R1 reactor using theWIMSD4 andCITATION codesrdquo Annals of Nuclear Energy vol 29 no 8 pp901ndash912 2002

[7] R M G P Souza ldquoResultados Experimentais da Calibracaodas Barras de Controle e do Excesso de Reatividade em CincoConfiguracoes do Nucleo do Reator IPR-R1rdquo Tech Rep NI-AT4-00199 CDTNCNEN Belo Horizonte Brasil 1999

[8] P A L Reis A L Costa C Pereira C A M Silva M A FVeloso and A Z Mesquita ldquoSensitivity analysis to a RELAP5nodalization developed for a typical TRIGA research reactorrdquoNuclear Engineering and Design vol 242 pp 300ndash306 2012

[9] A L Costa P A L Reis C Pereira M A F Veloso A ZMesquita and H V Soares ldquoThermal hydraulic analysis ofthe IPR-R1 TRIGA research reactor using a RELAP5 modelrdquoNuclear Engineering and Design vol 240 no 6 pp 1487ndash14942010

[10] P A L Reis A L Costa C Pereira et al ldquoAssessment of aRELAP5model for the IPR-R1 TRIGA research reactorrdquoAnnalsof Nuclear Energy vol 37 no 10 pp 1341ndash1350 2010

[11] M Adorni A Bousbia-Salah T Hamidouche B D MaroF Pierro and F DrsquoAuria ldquoAnalysis of partial and total flowblockage of a single fuel assembly of an MTR research reactorcorerdquo Annals of Nuclear Energy vol 32 no 15 pp 1679ndash16922005

[12] Q Lu S Qiu and G H Su ldquoDevelopment of a thermal-hydraulic analysis code for research reactors with plate fuelsrdquoAnnals of Nuclear Energy vol 36 no 4 pp 433ndash447 2009

[13] A Z Mesquita H C Rezende and R M G P Souza ldquoThermalpower calibrations of the IPR-R1 TRIGA nuclear reactorrdquo inProceedings of the 20th International Congress of MechanicalEngineering (COBEM rsquo09) Gramado Brazil November 2009

[14] RELAP5Mod33 Code Manuals Idaho National EngineeringLaboratory US NRC 2001

[15] W H McAdams Heat Transmission McGraw-Hill Book NewYork NY USA 3rd edition 1954

[16] W Wagner and A Kruse Properties of Water and SteammdashThe Industrial Standard IAPWS-IF97 for the ThermodynamicsProperties Springer Berlin Germany 1998

[17] A Z Mesquita A L Costa C Pereira M A F Veloso andP C A L Reis ldquoExperimental investigation of the onset ofsubcooled nucleate boiling in an open-pool nuclear researchreactorrdquo Journal of ASTM International vol 8 no 6 2011

TribologyAdvances in

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

International Journal of

AerospaceEngineeringHindawi Publishing Corporationhttpwwwhindawicom Volume 2014

FuelsJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Journal ofPetroleum Engineering

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Industrial EngineeringJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Power ElectronicsHindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Advances in

CombustionJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Journal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Renewable Energy

Submit your manuscripts athttpwwwhindawicom

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

StructuresJournal of

International Journal of

RotatingMachinery

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Hindawi Publishing Corporation httpwwwhindawicom

Journal ofEngineeringVolume 2014

Hindawi Publishing Corporation httpwwwhindawicom Volume 2014

International Journal ofPhotoenergy

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Nuclear InstallationsScience and Technology of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Solar EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Wind EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Nuclear EnergyInternational Journal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

High Energy PhysicsAdvances in

The Scientific World JournalHindawi Publishing Corporation httpwwwhindawicom Volume 2014

Science and Technology of Nuclear Installations 7

A1

E5

B2

C3

D4

A1

E5E5

B2B

C3C

D4D4

A BC D E

F

Time (s)0 2000 4000 6000 8000 10000 12000 14000 16000

minus06

minus05

minus04

minus03

minus02

minus01

05

04

03

02

01

00

Mas

s flow

rate

(kg

s)

Figure 7 Outlet mass flow rate in the channels A1 B2 C3 D4 and E5 before and after the blockage (power of 265 kW)

Axial level 01

Axial level 03

Axial level 07

Axial level 11

Inlet C2

THC D3THC C2

Outlet C2

123456789

101112131415161718192021

123456789

101112131415161718192021

(a)

Time (s)0 2000 4000 6000 8000 10000

00

02

04

06

Axial level 11Axial level 07

Axial level 03Axial level 01

minus02

Mas

s flow

rate

(kg

s)

(b)

Figure 8 Details of the nodalization where the junctions make possible the cross flow between THCs during the transient (a) Mass flow rateat the cross flow junctions between the THCs C2 and D3 in four axial levels (b)

1 2 213 3 44 55 6 6 77 88 99 1010290 290

295 295

300300

305 305

310310

315 315

320 320

325325

330 330

OutletMiddle

Inlet

Outlet

Middle

Inlet

Time (times103 s) Time (times103 s)

Coo

lant

tem

pera

ture

(K)

Coo

lant

tem

pera

ture

(K)

P = 100 kW P = 265 kW

Figure 9 Coolant temperature at the THC D4 at 100 kW and 265 kW

8 Science and Technology of Nuclear Installations

B2 C3

D4D4

0 1 2 3 4 5 6 7 8 9 10 0 1 2 3

3210 4

4 5

5 6

6 7

7 8

8 9

9 10

10

00

4 4

0

4

6

68

12

16 16

16

20

2020

12

12

Void

frac

tion

(times10minus3)

Void

frac

tion

(times10minus3)

Void

frac

tion

(times10minus3)

Time (times103 s) Time (times103 s)

Time (times103 s)

Figure 10 Void fraction time evolution at outlet of the thermal hydraulic channels B2 C3 and D4 after the transient at 100 kW

100 20 30 40 50 60 70 80 90 10066

68

70

72

74

76

78

80

Time (times102 s)

Hea

t flux

(times103

Wm

2)

(a)

0 10 20 30 40 50 60 70 80 90 1000

3

6

9

12

15

18

21

24

27

30

Time (times102 s)

Hea

t tra

nsfe

r coe

ffici

ent(times102

Wm

2K)

(b)

Figure 11 Heat flux (a) and heat transfer coefficient (b) to the heat structure associated with the THC B1 at 100 kW of power operation

5 Conclusions

To perform the calculations a verified model in RELAP5MOD33 was considered The simulated transient character-izes a LOFA type accident Total and partial core coolant

blockage transient have been investigated for the IPR-R1research reactor In the simulation the channels were blockedusing components type valve in the inlet of the channels Inboth cases total and partial blockage the cross flowmodelingin the RELAP5 core nodalization is important in the analyses

Science and Technology of Nuclear Installations 9

Time (s)1000 2000 3000 4000 5000 60000

minus040

minus035

minus030

minus025

minus020

minus015

minus010

minus005

Mas

s flow

rate

(kg

s)

000

005

010

P = 100 kWP = 100 kW

Inlet

Outlet

(a)

Time (s)1000 2000 3000 4000 50000

00

05

10

Mas

s flow

rate

(kg

s)

P = 265 kW

Outlet

Middle = inlet

minus05

minus10

minus15

minus20

minus25

(b)

Figure 12 Mass flow rate at channel D4 after and before the total core blockage at 100 kW (a) and 265 kW (b) of power operation

Outlet

Middle

InletCoo

lant

tem

pera

ture

(K)

300

320

340

360

380

400

Time (s)1000 2000 3000 4000 5000 60000

P = 100 kW

(a)

Outlet

Middle

Inlet

Coo

lant

tem

pera

ture

(K)

300

320

340

360

380

400

Time (s)1000 2000 3000 4000 50000

P = 265 kW

(b)

Figure 13 Coolant temperature in the channel B2 at 100 kW (a) and 265 kW (b) of power operation

300

320

340

360

380

400

Coo

lant

tem

pera

ture

(K)

Time (s)1000 2000 3000 4000 50000

Outlet

MiddleInlet

(a)

Void

frac

tion

0

02

04

06

08

10

Time (s)1000 2000 3000 4000 50000

minus02

Inlet

Middle

Outlet

(b)

Figure 14 Coolant temperature (a) and void fraction (b) evolutions for the THC D4 at 100 kW

10 Science and Technology of Nuclear Installations

300

350

400

450

500

550

600

650

700

0 5 10 15 20 25 30 35 40 45 50Time (times102 s)

Tem

pera

ture

(K)

Fuel

Cladding

Figure 15 Fuel and cladding temperature evolutions for the B1 heatstructure at 265 kW of power operation

because it provides the coolant redistribution among adjacentchannels after the blockage as it is the expected behaviour

After the partial blockage transient an increase in thecoolant temperature of the blocked channels change in themass flow direction and modulus and void formation wereobserved In spite of this the reactor presented safe behaviourafter the transient reaching a new steady state However themodel is not capable of reproducing exactly the behaviour ofthe core due to the changes in the macroscopic cross sectionsthat are directly affected by coolant density and temperature

In the case of the total blockage the saturation coolanttemperature is reached into the core and the coolant boils infew minutes after the beginning of the transient One moretime in an actual situation the intrinsic safety characteristicof fuel shouldwork to control the reactor It was observed thata total core blockage disturbs significantly the core and thereactivity feedback effects should be considered A neutronicanalysis in this case could be necessary to investigate how theneutron flux is perturbed in the core after the total blockagetransient

Conflict of Interests

The authors declare that there is no conflict of interestsregarding the publication of this paper

Acknowledgments

The authors are grateful to CAPES CNPq CDTNCNENand FAPEMIG (all from Brazil) for the support

References

[1] IAEA Safety Analysis for Research Reactors Safety ReportsSeries No 55 International Atomic Energy Agency ViennaAustria 2008

[2] O Merroun A Almers T El Bardouni B El Bakkari andE Chakir ldquoAnalytical benchmarks for verification of thermal-hydraulic codes based on sub-channel approachrdquoNuclear Engi-neering and Design vol 239 no 4 pp 735ndash748 2009

[3] C S Y Suresh G Sateesh S K Das S P Veukateshanand M Rajan ldquoHeat transfer from a totally blocked fuelsubassembly of a liquid metal fast breeder reactor part Iexperimental investigation and part II numerical simulationrdquoNuclear Engineering and Design vol 235 pp 885ndash912 2005

[4] MA F VelosoAnalise Termofluidodinamica de Reatores Nucle-ares de Pesquisa Refrigerados a Agua em Regime de ConveccaoNatural [Doctor thesis] Universidade Estadual de CampinasCampinas Brazil 2004

[5] H M Dalle ldquoAvaliacao Neutronica do Reator TRIGA IPR-R1 ndashR1 com Configuracao de 63 Elementos Combustıveis e Barrade Regulacao em F16rdquo Restrict Document (NIndashEC3-0103)Centro de Desenvolvimento da Tecnologia Nuclear ComissaoNacional de Energia Nuclear Belo Horizonte Brazil 2003

[6] H M Dalle C Pereira and R G P Souza ldquoNeutroniccalculation to theTRIGA Ipr-R1 reactor using theWIMSD4 andCITATION codesrdquo Annals of Nuclear Energy vol 29 no 8 pp901ndash912 2002

[7] R M G P Souza ldquoResultados Experimentais da Calibracaodas Barras de Controle e do Excesso de Reatividade em CincoConfiguracoes do Nucleo do Reator IPR-R1rdquo Tech Rep NI-AT4-00199 CDTNCNEN Belo Horizonte Brasil 1999

[8] P A L Reis A L Costa C Pereira C A M Silva M A FVeloso and A Z Mesquita ldquoSensitivity analysis to a RELAP5nodalization developed for a typical TRIGA research reactorrdquoNuclear Engineering and Design vol 242 pp 300ndash306 2012

[9] A L Costa P A L Reis C Pereira M A F Veloso A ZMesquita and H V Soares ldquoThermal hydraulic analysis ofthe IPR-R1 TRIGA research reactor using a RELAP5 modelrdquoNuclear Engineering and Design vol 240 no 6 pp 1487ndash14942010

[10] P A L Reis A L Costa C Pereira et al ldquoAssessment of aRELAP5model for the IPR-R1 TRIGA research reactorrdquoAnnalsof Nuclear Energy vol 37 no 10 pp 1341ndash1350 2010

[11] M Adorni A Bousbia-Salah T Hamidouche B D MaroF Pierro and F DrsquoAuria ldquoAnalysis of partial and total flowblockage of a single fuel assembly of an MTR research reactorcorerdquo Annals of Nuclear Energy vol 32 no 15 pp 1679ndash16922005

[12] Q Lu S Qiu and G H Su ldquoDevelopment of a thermal-hydraulic analysis code for research reactors with plate fuelsrdquoAnnals of Nuclear Energy vol 36 no 4 pp 433ndash447 2009

[13] A Z Mesquita H C Rezende and R M G P Souza ldquoThermalpower calibrations of the IPR-R1 TRIGA nuclear reactorrdquo inProceedings of the 20th International Congress of MechanicalEngineering (COBEM rsquo09) Gramado Brazil November 2009

[14] RELAP5Mod33 Code Manuals Idaho National EngineeringLaboratory US NRC 2001

[15] W H McAdams Heat Transmission McGraw-Hill Book NewYork NY USA 3rd edition 1954

[16] W Wagner and A Kruse Properties of Water and SteammdashThe Industrial Standard IAPWS-IF97 for the ThermodynamicsProperties Springer Berlin Germany 1998

[17] A Z Mesquita A L Costa C Pereira M A F Veloso andP C A L Reis ldquoExperimental investigation of the onset ofsubcooled nucleate boiling in an open-pool nuclear researchreactorrdquo Journal of ASTM International vol 8 no 6 2011

TribologyAdvances in

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

International Journal of

AerospaceEngineeringHindawi Publishing Corporationhttpwwwhindawicom Volume 2014

FuelsJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Journal ofPetroleum Engineering

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Industrial EngineeringJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Power ElectronicsHindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Advances in

CombustionJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Journal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Renewable Energy

Submit your manuscripts athttpwwwhindawicom

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

StructuresJournal of

International Journal of

RotatingMachinery

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Hindawi Publishing Corporation httpwwwhindawicom

Journal ofEngineeringVolume 2014

Hindawi Publishing Corporation httpwwwhindawicom Volume 2014

International Journal ofPhotoenergy

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Nuclear InstallationsScience and Technology of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Solar EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Wind EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Nuclear EnergyInternational Journal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

High Energy PhysicsAdvances in

The Scientific World JournalHindawi Publishing Corporation httpwwwhindawicom Volume 2014

8 Science and Technology of Nuclear Installations

B2 C3

D4D4

0 1 2 3 4 5 6 7 8 9 10 0 1 2 3

3210 4

4 5

5 6

6 7

7 8

8 9

9 10

10

00

4 4

0

4

6

68

12

16 16

16

20

2020

12

12

Void

frac

tion

(times10minus3)

Void

frac

tion

(times10minus3)

Void

frac

tion

(times10minus3)

Time (times103 s) Time (times103 s)

Time (times103 s)

Figure 10 Void fraction time evolution at outlet of the thermal hydraulic channels B2 C3 and D4 after the transient at 100 kW

100 20 30 40 50 60 70 80 90 10066

68

70

72

74

76

78

80

Time (times102 s)

Hea

t flux

(times103

Wm

2)

(a)

0 10 20 30 40 50 60 70 80 90 1000

3

6

9

12

15

18

21

24

27

30

Time (times102 s)

Hea

t tra

nsfe

r coe

ffici

ent(times102

Wm

2K)

(b)

Figure 11 Heat flux (a) and heat transfer coefficient (b) to the heat structure associated with the THC B1 at 100 kW of power operation

5 Conclusions

To perform the calculations a verified model in RELAP5MOD33 was considered The simulated transient character-izes a LOFA type accident Total and partial core coolant

blockage transient have been investigated for the IPR-R1research reactor In the simulation the channels were blockedusing components type valve in the inlet of the channels Inboth cases total and partial blockage the cross flowmodelingin the RELAP5 core nodalization is important in the analyses

Science and Technology of Nuclear Installations 9

Time (s)1000 2000 3000 4000 5000 60000

minus040

minus035

minus030

minus025

minus020

minus015

minus010

minus005

Mas

s flow

rate

(kg

s)

000

005

010

P = 100 kWP = 100 kW

Inlet

Outlet

(a)

Time (s)1000 2000 3000 4000 50000

00

05

10

Mas

s flow

rate

(kg

s)

P = 265 kW

Outlet

Middle = inlet

minus05

minus10

minus15

minus20

minus25

(b)

Figure 12 Mass flow rate at channel D4 after and before the total core blockage at 100 kW (a) and 265 kW (b) of power operation

Outlet

Middle

InletCoo

lant

tem

pera

ture

(K)

300

320

340

360

380

400

Time (s)1000 2000 3000 4000 5000 60000

P = 100 kW

(a)

Outlet

Middle

Inlet

Coo

lant

tem

pera

ture

(K)

300

320

340

360

380

400

Time (s)1000 2000 3000 4000 50000

P = 265 kW

(b)

Figure 13 Coolant temperature in the channel B2 at 100 kW (a) and 265 kW (b) of power operation

300

320

340

360

380

400

Coo

lant

tem

pera

ture

(K)

Time (s)1000 2000 3000 4000 50000

Outlet

MiddleInlet

(a)

Void

frac

tion

0

02

04

06

08

10

Time (s)1000 2000 3000 4000 50000

minus02

Inlet

Middle

Outlet

(b)

Figure 14 Coolant temperature (a) and void fraction (b) evolutions for the THC D4 at 100 kW

10 Science and Technology of Nuclear Installations

300

350

400

450

500

550

600

650

700

0 5 10 15 20 25 30 35 40 45 50Time (times102 s)

Tem

pera

ture

(K)

Fuel

Cladding

Figure 15 Fuel and cladding temperature evolutions for the B1 heatstructure at 265 kW of power operation

because it provides the coolant redistribution among adjacentchannels after the blockage as it is the expected behaviour

After the partial blockage transient an increase in thecoolant temperature of the blocked channels change in themass flow direction and modulus and void formation wereobserved In spite of this the reactor presented safe behaviourafter the transient reaching a new steady state However themodel is not capable of reproducing exactly the behaviour ofthe core due to the changes in the macroscopic cross sectionsthat are directly affected by coolant density and temperature

In the case of the total blockage the saturation coolanttemperature is reached into the core and the coolant boils infew minutes after the beginning of the transient One moretime in an actual situation the intrinsic safety characteristicof fuel shouldwork to control the reactor It was observed thata total core blockage disturbs significantly the core and thereactivity feedback effects should be considered A neutronicanalysis in this case could be necessary to investigate how theneutron flux is perturbed in the core after the total blockagetransient

Conflict of Interests

The authors declare that there is no conflict of interestsregarding the publication of this paper

Acknowledgments

The authors are grateful to CAPES CNPq CDTNCNENand FAPEMIG (all from Brazil) for the support

References

[1] IAEA Safety Analysis for Research Reactors Safety ReportsSeries No 55 International Atomic Energy Agency ViennaAustria 2008

[2] O Merroun A Almers T El Bardouni B El Bakkari andE Chakir ldquoAnalytical benchmarks for verification of thermal-hydraulic codes based on sub-channel approachrdquoNuclear Engi-neering and Design vol 239 no 4 pp 735ndash748 2009

[3] C S Y Suresh G Sateesh S K Das S P Veukateshanand M Rajan ldquoHeat transfer from a totally blocked fuelsubassembly of a liquid metal fast breeder reactor part Iexperimental investigation and part II numerical simulationrdquoNuclear Engineering and Design vol 235 pp 885ndash912 2005

[4] MA F VelosoAnalise Termofluidodinamica de Reatores Nucle-ares de Pesquisa Refrigerados a Agua em Regime de ConveccaoNatural [Doctor thesis] Universidade Estadual de CampinasCampinas Brazil 2004

[5] H M Dalle ldquoAvaliacao Neutronica do Reator TRIGA IPR-R1 ndashR1 com Configuracao de 63 Elementos Combustıveis e Barrade Regulacao em F16rdquo Restrict Document (NIndashEC3-0103)Centro de Desenvolvimento da Tecnologia Nuclear ComissaoNacional de Energia Nuclear Belo Horizonte Brazil 2003

[6] H M Dalle C Pereira and R G P Souza ldquoNeutroniccalculation to theTRIGA Ipr-R1 reactor using theWIMSD4 andCITATION codesrdquo Annals of Nuclear Energy vol 29 no 8 pp901ndash912 2002

[7] R M G P Souza ldquoResultados Experimentais da Calibracaodas Barras de Controle e do Excesso de Reatividade em CincoConfiguracoes do Nucleo do Reator IPR-R1rdquo Tech Rep NI-AT4-00199 CDTNCNEN Belo Horizonte Brasil 1999

[8] P A L Reis A L Costa C Pereira C A M Silva M A FVeloso and A Z Mesquita ldquoSensitivity analysis to a RELAP5nodalization developed for a typical TRIGA research reactorrdquoNuclear Engineering and Design vol 242 pp 300ndash306 2012

[9] A L Costa P A L Reis C Pereira M A F Veloso A ZMesquita and H V Soares ldquoThermal hydraulic analysis ofthe IPR-R1 TRIGA research reactor using a RELAP5 modelrdquoNuclear Engineering and Design vol 240 no 6 pp 1487ndash14942010

[10] P A L Reis A L Costa C Pereira et al ldquoAssessment of aRELAP5model for the IPR-R1 TRIGA research reactorrdquoAnnalsof Nuclear Energy vol 37 no 10 pp 1341ndash1350 2010

[11] M Adorni A Bousbia-Salah T Hamidouche B D MaroF Pierro and F DrsquoAuria ldquoAnalysis of partial and total flowblockage of a single fuel assembly of an MTR research reactorcorerdquo Annals of Nuclear Energy vol 32 no 15 pp 1679ndash16922005

[12] Q Lu S Qiu and G H Su ldquoDevelopment of a thermal-hydraulic analysis code for research reactors with plate fuelsrdquoAnnals of Nuclear Energy vol 36 no 4 pp 433ndash447 2009

[13] A Z Mesquita H C Rezende and R M G P Souza ldquoThermalpower calibrations of the IPR-R1 TRIGA nuclear reactorrdquo inProceedings of the 20th International Congress of MechanicalEngineering (COBEM rsquo09) Gramado Brazil November 2009

[14] RELAP5Mod33 Code Manuals Idaho National EngineeringLaboratory US NRC 2001

[15] W H McAdams Heat Transmission McGraw-Hill Book NewYork NY USA 3rd edition 1954

[16] W Wagner and A Kruse Properties of Water and SteammdashThe Industrial Standard IAPWS-IF97 for the ThermodynamicsProperties Springer Berlin Germany 1998

[17] A Z Mesquita A L Costa C Pereira M A F Veloso andP C A L Reis ldquoExperimental investigation of the onset ofsubcooled nucleate boiling in an open-pool nuclear researchreactorrdquo Journal of ASTM International vol 8 no 6 2011

TribologyAdvances in

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

International Journal of

AerospaceEngineeringHindawi Publishing Corporationhttpwwwhindawicom Volume 2014

FuelsJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Journal ofPetroleum Engineering

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Industrial EngineeringJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Power ElectronicsHindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Advances in

CombustionJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Journal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Renewable Energy

Submit your manuscripts athttpwwwhindawicom

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

StructuresJournal of

International Journal of

RotatingMachinery

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Hindawi Publishing Corporation httpwwwhindawicom

Journal ofEngineeringVolume 2014

Hindawi Publishing Corporation httpwwwhindawicom Volume 2014

International Journal ofPhotoenergy

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Nuclear InstallationsScience and Technology of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Solar EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Wind EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Nuclear EnergyInternational Journal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

High Energy PhysicsAdvances in

The Scientific World JournalHindawi Publishing Corporation httpwwwhindawicom Volume 2014

Science and Technology of Nuclear Installations 9

Time (s)1000 2000 3000 4000 5000 60000

minus040

minus035

minus030

minus025

minus020

minus015

minus010

minus005

Mas

s flow

rate

(kg

s)

000

005

010

P = 100 kWP = 100 kW

Inlet

Outlet

(a)

Time (s)1000 2000 3000 4000 50000

00

05

10

Mas

s flow

rate

(kg

s)

P = 265 kW

Outlet

Middle = inlet

minus05

minus10

minus15

minus20

minus25

(b)

Figure 12 Mass flow rate at channel D4 after and before the total core blockage at 100 kW (a) and 265 kW (b) of power operation

Outlet

Middle

InletCoo

lant

tem

pera

ture

(K)

300

320

340

360

380

400

Time (s)1000 2000 3000 4000 5000 60000

P = 100 kW

(a)

Outlet

Middle

Inlet

Coo

lant

tem

pera

ture

(K)

300

320

340

360

380

400

Time (s)1000 2000 3000 4000 50000

P = 265 kW

(b)

Figure 13 Coolant temperature in the channel B2 at 100 kW (a) and 265 kW (b) of power operation

300

320

340

360

380

400

Coo

lant

tem

pera

ture

(K)

Time (s)1000 2000 3000 4000 50000

Outlet

MiddleInlet

(a)

Void

frac

tion

0

02

04

06

08

10

Time (s)1000 2000 3000 4000 50000

minus02

Inlet

Middle

Outlet

(b)

Figure 14 Coolant temperature (a) and void fraction (b) evolutions for the THC D4 at 100 kW

10 Science and Technology of Nuclear Installations

300

350

400

450

500

550

600

650

700

0 5 10 15 20 25 30 35 40 45 50Time (times102 s)

Tem

pera

ture

(K)

Fuel

Cladding

Figure 15 Fuel and cladding temperature evolutions for the B1 heatstructure at 265 kW of power operation

because it provides the coolant redistribution among adjacentchannels after the blockage as it is the expected behaviour

After the partial blockage transient an increase in thecoolant temperature of the blocked channels change in themass flow direction and modulus and void formation wereobserved In spite of this the reactor presented safe behaviourafter the transient reaching a new steady state However themodel is not capable of reproducing exactly the behaviour ofthe core due to the changes in the macroscopic cross sectionsthat are directly affected by coolant density and temperature

In the case of the total blockage the saturation coolanttemperature is reached into the core and the coolant boils infew minutes after the beginning of the transient One moretime in an actual situation the intrinsic safety characteristicof fuel shouldwork to control the reactor It was observed thata total core blockage disturbs significantly the core and thereactivity feedback effects should be considered A neutronicanalysis in this case could be necessary to investigate how theneutron flux is perturbed in the core after the total blockagetransient

Conflict of Interests

The authors declare that there is no conflict of interestsregarding the publication of this paper

Acknowledgments

The authors are grateful to CAPES CNPq CDTNCNENand FAPEMIG (all from Brazil) for the support

References

[1] IAEA Safety Analysis for Research Reactors Safety ReportsSeries No 55 International Atomic Energy Agency ViennaAustria 2008

[2] O Merroun A Almers T El Bardouni B El Bakkari andE Chakir ldquoAnalytical benchmarks for verification of thermal-hydraulic codes based on sub-channel approachrdquoNuclear Engi-neering and Design vol 239 no 4 pp 735ndash748 2009

[3] C S Y Suresh G Sateesh S K Das S P Veukateshanand M Rajan ldquoHeat transfer from a totally blocked fuelsubassembly of a liquid metal fast breeder reactor part Iexperimental investigation and part II numerical simulationrdquoNuclear Engineering and Design vol 235 pp 885ndash912 2005

[4] MA F VelosoAnalise Termofluidodinamica de Reatores Nucle-ares de Pesquisa Refrigerados a Agua em Regime de ConveccaoNatural [Doctor thesis] Universidade Estadual de CampinasCampinas Brazil 2004

[5] H M Dalle ldquoAvaliacao Neutronica do Reator TRIGA IPR-R1 ndashR1 com Configuracao de 63 Elementos Combustıveis e Barrade Regulacao em F16rdquo Restrict Document (NIndashEC3-0103)Centro de Desenvolvimento da Tecnologia Nuclear ComissaoNacional de Energia Nuclear Belo Horizonte Brazil 2003

[6] H M Dalle C Pereira and R G P Souza ldquoNeutroniccalculation to theTRIGA Ipr-R1 reactor using theWIMSD4 andCITATION codesrdquo Annals of Nuclear Energy vol 29 no 8 pp901ndash912 2002

[7] R M G P Souza ldquoResultados Experimentais da Calibracaodas Barras de Controle e do Excesso de Reatividade em CincoConfiguracoes do Nucleo do Reator IPR-R1rdquo Tech Rep NI-AT4-00199 CDTNCNEN Belo Horizonte Brasil 1999

[8] P A L Reis A L Costa C Pereira C A M Silva M A FVeloso and A Z Mesquita ldquoSensitivity analysis to a RELAP5nodalization developed for a typical TRIGA research reactorrdquoNuclear Engineering and Design vol 242 pp 300ndash306 2012

[9] A L Costa P A L Reis C Pereira M A F Veloso A ZMesquita and H V Soares ldquoThermal hydraulic analysis ofthe IPR-R1 TRIGA research reactor using a RELAP5 modelrdquoNuclear Engineering and Design vol 240 no 6 pp 1487ndash14942010

[10] P A L Reis A L Costa C Pereira et al ldquoAssessment of aRELAP5model for the IPR-R1 TRIGA research reactorrdquoAnnalsof Nuclear Energy vol 37 no 10 pp 1341ndash1350 2010

[11] M Adorni A Bousbia-Salah T Hamidouche B D MaroF Pierro and F DrsquoAuria ldquoAnalysis of partial and total flowblockage of a single fuel assembly of an MTR research reactorcorerdquo Annals of Nuclear Energy vol 32 no 15 pp 1679ndash16922005

[12] Q Lu S Qiu and G H Su ldquoDevelopment of a thermal-hydraulic analysis code for research reactors with plate fuelsrdquoAnnals of Nuclear Energy vol 36 no 4 pp 433ndash447 2009

[13] A Z Mesquita H C Rezende and R M G P Souza ldquoThermalpower calibrations of the IPR-R1 TRIGA nuclear reactorrdquo inProceedings of the 20th International Congress of MechanicalEngineering (COBEM rsquo09) Gramado Brazil November 2009

[14] RELAP5Mod33 Code Manuals Idaho National EngineeringLaboratory US NRC 2001

[15] W H McAdams Heat Transmission McGraw-Hill Book NewYork NY USA 3rd edition 1954

[16] W Wagner and A Kruse Properties of Water and SteammdashThe Industrial Standard IAPWS-IF97 for the ThermodynamicsProperties Springer Berlin Germany 1998

[17] A Z Mesquita A L Costa C Pereira M A F Veloso andP C A L Reis ldquoExperimental investigation of the onset ofsubcooled nucleate boiling in an open-pool nuclear researchreactorrdquo Journal of ASTM International vol 8 no 6 2011

TribologyAdvances in

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

International Journal of

AerospaceEngineeringHindawi Publishing Corporationhttpwwwhindawicom Volume 2014

FuelsJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Journal ofPetroleum Engineering

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Industrial EngineeringJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Power ElectronicsHindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Advances in

CombustionJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Journal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Renewable Energy

Submit your manuscripts athttpwwwhindawicom

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

StructuresJournal of

International Journal of

RotatingMachinery

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Hindawi Publishing Corporation httpwwwhindawicom

Journal ofEngineeringVolume 2014

Hindawi Publishing Corporation httpwwwhindawicom Volume 2014

International Journal ofPhotoenergy

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Nuclear InstallationsScience and Technology of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Solar EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Wind EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Nuclear EnergyInternational Journal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

High Energy PhysicsAdvances in

The Scientific World JournalHindawi Publishing Corporation httpwwwhindawicom Volume 2014

10 Science and Technology of Nuclear Installations

300

350

400

450

500

550

600

650

700

0 5 10 15 20 25 30 35 40 45 50Time (times102 s)

Tem

pera

ture

(K)

Fuel

Cladding

Figure 15 Fuel and cladding temperature evolutions for the B1 heatstructure at 265 kW of power operation

because it provides the coolant redistribution among adjacentchannels after the blockage as it is the expected behaviour

After the partial blockage transient an increase in thecoolant temperature of the blocked channels change in themass flow direction and modulus and void formation wereobserved In spite of this the reactor presented safe behaviourafter the transient reaching a new steady state However themodel is not capable of reproducing exactly the behaviour ofthe core due to the changes in the macroscopic cross sectionsthat are directly affected by coolant density and temperature

In the case of the total blockage the saturation coolanttemperature is reached into the core and the coolant boils infew minutes after the beginning of the transient One moretime in an actual situation the intrinsic safety characteristicof fuel shouldwork to control the reactor It was observed thata total core blockage disturbs significantly the core and thereactivity feedback effects should be considered A neutronicanalysis in this case could be necessary to investigate how theneutron flux is perturbed in the core after the total blockagetransient

Conflict of Interests

The authors declare that there is no conflict of interestsregarding the publication of this paper

Acknowledgments

The authors are grateful to CAPES CNPq CDTNCNENand FAPEMIG (all from Brazil) for the support

References

[1] IAEA Safety Analysis for Research Reactors Safety ReportsSeries No 55 International Atomic Energy Agency ViennaAustria 2008

[2] O Merroun A Almers T El Bardouni B El Bakkari andE Chakir ldquoAnalytical benchmarks for verification of thermal-hydraulic codes based on sub-channel approachrdquoNuclear Engi-neering and Design vol 239 no 4 pp 735ndash748 2009

[3] C S Y Suresh G Sateesh S K Das S P Veukateshanand M Rajan ldquoHeat transfer from a totally blocked fuelsubassembly of a liquid metal fast breeder reactor part Iexperimental investigation and part II numerical simulationrdquoNuclear Engineering and Design vol 235 pp 885ndash912 2005

[4] MA F VelosoAnalise Termofluidodinamica de Reatores Nucle-ares de Pesquisa Refrigerados a Agua em Regime de ConveccaoNatural [Doctor thesis] Universidade Estadual de CampinasCampinas Brazil 2004

[5] H M Dalle ldquoAvaliacao Neutronica do Reator TRIGA IPR-R1 ndashR1 com Configuracao de 63 Elementos Combustıveis e Barrade Regulacao em F16rdquo Restrict Document (NIndashEC3-0103)Centro de Desenvolvimento da Tecnologia Nuclear ComissaoNacional de Energia Nuclear Belo Horizonte Brazil 2003

[6] H M Dalle C Pereira and R G P Souza ldquoNeutroniccalculation to theTRIGA Ipr-R1 reactor using theWIMSD4 andCITATION codesrdquo Annals of Nuclear Energy vol 29 no 8 pp901ndash912 2002

[7] R M G P Souza ldquoResultados Experimentais da Calibracaodas Barras de Controle e do Excesso de Reatividade em CincoConfiguracoes do Nucleo do Reator IPR-R1rdquo Tech Rep NI-AT4-00199 CDTNCNEN Belo Horizonte Brasil 1999

[8] P A L Reis A L Costa C Pereira C A M Silva M A FVeloso and A Z Mesquita ldquoSensitivity analysis to a RELAP5nodalization developed for a typical TRIGA research reactorrdquoNuclear Engineering and Design vol 242 pp 300ndash306 2012

[9] A L Costa P A L Reis C Pereira M A F Veloso A ZMesquita and H V Soares ldquoThermal hydraulic analysis ofthe IPR-R1 TRIGA research reactor using a RELAP5 modelrdquoNuclear Engineering and Design vol 240 no 6 pp 1487ndash14942010

[10] P A L Reis A L Costa C Pereira et al ldquoAssessment of aRELAP5model for the IPR-R1 TRIGA research reactorrdquoAnnalsof Nuclear Energy vol 37 no 10 pp 1341ndash1350 2010

[11] M Adorni A Bousbia-Salah T Hamidouche B D MaroF Pierro and F DrsquoAuria ldquoAnalysis of partial and total flowblockage of a single fuel assembly of an MTR research reactorcorerdquo Annals of Nuclear Energy vol 32 no 15 pp 1679ndash16922005

[12] Q Lu S Qiu and G H Su ldquoDevelopment of a thermal-hydraulic analysis code for research reactors with plate fuelsrdquoAnnals of Nuclear Energy vol 36 no 4 pp 433ndash447 2009

[13] A Z Mesquita H C Rezende and R M G P Souza ldquoThermalpower calibrations of the IPR-R1 TRIGA nuclear reactorrdquo inProceedings of the 20th International Congress of MechanicalEngineering (COBEM rsquo09) Gramado Brazil November 2009

[14] RELAP5Mod33 Code Manuals Idaho National EngineeringLaboratory US NRC 2001

[15] W H McAdams Heat Transmission McGraw-Hill Book NewYork NY USA 3rd edition 1954

[16] W Wagner and A Kruse Properties of Water and SteammdashThe Industrial Standard IAPWS-IF97 for the ThermodynamicsProperties Springer Berlin Germany 1998

[17] A Z Mesquita A L Costa C Pereira M A F Veloso andP C A L Reis ldquoExperimental investigation of the onset ofsubcooled nucleate boiling in an open-pool nuclear researchreactorrdquo Journal of ASTM International vol 8 no 6 2011

TribologyAdvances in

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

International Journal of

AerospaceEngineeringHindawi Publishing Corporationhttpwwwhindawicom Volume 2014

FuelsJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Journal ofPetroleum Engineering

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Industrial EngineeringJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Power ElectronicsHindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Advances in

CombustionJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Journal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Renewable Energy

Submit your manuscripts athttpwwwhindawicom

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

StructuresJournal of

International Journal of

RotatingMachinery

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Hindawi Publishing Corporation httpwwwhindawicom

Journal ofEngineeringVolume 2014

Hindawi Publishing Corporation httpwwwhindawicom Volume 2014

International Journal ofPhotoenergy

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Nuclear InstallationsScience and Technology of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Solar EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Wind EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Nuclear EnergyInternational Journal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

High Energy PhysicsAdvances in

The Scientific World JournalHindawi Publishing Corporation httpwwwhindawicom Volume 2014

TribologyAdvances in

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

International Journal of

AerospaceEngineeringHindawi Publishing Corporationhttpwwwhindawicom Volume 2014

FuelsJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Journal ofPetroleum Engineering

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Industrial EngineeringJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Power ElectronicsHindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Advances in

CombustionJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Journal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Renewable Energy

Submit your manuscripts athttpwwwhindawicom

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

StructuresJournal of

International Journal of

RotatingMachinery

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Hindawi Publishing Corporation httpwwwhindawicom

Journal ofEngineeringVolume 2014

Hindawi Publishing Corporation httpwwwhindawicom Volume 2014

International Journal ofPhotoenergy

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Nuclear InstallationsScience and Technology of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Solar EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Wind EnergyJournal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

Nuclear EnergyInternational Journal of

Hindawi Publishing Corporationhttpwwwhindawicom Volume 2014

High Energy PhysicsAdvances in

The Scientific World JournalHindawi Publishing Corporation httpwwwhindawicom Volume 2014