assessment of relap/scdapsim/mod3.4 prediction capability...

Research ArticleAssessment of RELAP/SCDAPSIM/MOD3.4 PredictionCapability with Severe Fuel Damage Scoping Test

Noppawan Rattanadecho,1 Somboon Rassame,1 Kampanart Silva,2

Chris Allison,3 and Judith Hohorst3

1Department of Nuclear Engineering, Chulalongkorn University, 254 Phayathai Rd, Pathumwan, Bangkok 10330, Thailand2Thailand Institute of Nuclear Technology (Public Organization), 9/9 Moo 7, Saimoon, Ongkharak,Nakorn Nayok 26120, Thailand3Innovation Systems Software, 2585 Briar Creek Ln., Ammon, Idaho 83406, USA

Correspondence should be addressed to Somboon Rassame; [email protected]

Received 12 July 2017; Revised 4 September 2017; Accepted 14 September 2017; Published 25 October 2017

Academic Editor: Arkady Serikov

Copyright © 2017 Noppawan Rattanadecho et al. This is an open access article distributed under the Creative CommonsAttribution License, which permits unrestricted use, distribution, and reproduction in any medium, provided the original work isproperly cited.

The Power Burst Facility (PBF) was designed to provide experimental data to determine the thresholds for failure duringaccident conditions. Thus, the PBF benchmark using severe accidental analysis codes is essential to designing reactor for currentdirections. This assessment verified and validated that the RELAP/SCDAPSIM/MOD3.4 code can be used to assess the SevereFuel Damage Scoping Test (SFD-ST) performed in the PBF facility. This study compares the cladding temperatures and hydrogenproduction results calculated by the RELAP/SCDAPSIM/MOD3.4 code with experimental data and calculated results fromthe SCDAP/RELAP5/MOD3.2 and SCDAP/RELAP5/MOD3.3 codes. The interested parameters are cladding temperature andhydrogen production since the cladding temperature affects hydrogen production and consequently influences the accidentscenario. The calculated cladding temperatures and hydrogen production results from the RELAP/SCDAPSIM/MOD3.4 codeare in a good agreement with the experimental data and are generally more reasonable than the calculated results from theSCDAP/RELAP5/MOD3.2 and SCDAP/RELAP5/MOD3.3 codes. There are some discrepancies in the cladding temperature andhydrogen production results but they are expected.

1. Introduction

In 1979, a severe accident occurred at the Three Mile IslandUnit 2 (TMI-2) [1]. A failure in the nonnuclear part of theplant system triggered series of some automated responsesin the reactor coolant system (RCS) and the relief valve atthe top of the pressurizer failed to close when the pressurereturned to a proper value. After that, the operators wereunaware that cooling water was pouring out of the stuckopen valve. The lack of proper water flow allowed the reactorcore to become partially uncovered and severely damaged.Fukushima Daiichi nuclear accident occurred in 2011 due tothe occurrence of a huge tsunami after a massive earthquake[2]. The earthquake situation affected the electrical powersupply lines at the site and the tsunamis caused the massive

damage to the site infrastructure resulting in the loss ofthe cooling system. The above severe accidents significantlydestroyed the infrastructure and were potentially harmful tohumans and the environment. Consequently, it is necessaryto study and understand the progression of severe accidentsto prevent the occurrence of severe accidents ormitigate theirconsequences in case they are unavoidable.

The Power Burst Facility (PBF) was designed to testfuel samples under accident conditions and to learn theimplications of fuel failure to safety when operating powerreactors. The main purposes of the experiments performedin the PBF facility were to understand the fuel behavior andthe generation of hydrogen during a severe accident. ThePBF facility was designed to provide experimental data todefine the thresholds of failure during postulated accident

HindawiScience and Technology of Nuclear InstallationsVolume 2017, Article ID 7456380, 12 pageshttps://doi.org/10.1155/2017/7456380

https://doi.org/10.1155/2017/7456380

2 Science and Technology of Nuclear Installations

conditions and to understand the failure mechanisms [3].The Severe Fuel Damage Scoping Test (SFD-ST) was the firstlarge scale severe fuel damage experiment that was performedin PBF [4, 5]. The objectives of this experiment were tounderstand the fuel bundle dynamics, the related hydrogengeneration, and fission product behavior during transientperiod.

The RELAP/SCDAPSIM code has been developed byInnovative Systems Software (ISS) and has been used topredict the RCS behavior, thermal hydraulic, and core behav-ior under normal and accident conditions. The RELAP/SCDAPSIM code contains improved SCDAP models devel-oped from recent experimental programs, improved numer-ical analysis, and a three-dimension display. The assessmentof RELAP/SCDAPSIM code has been performed with someresearch reactor [6], PWR core [7], RBMK [8], CANDU[9, 10], QUENCH [11, 12], and CORA-17 [13].There are manybenchmark studies using RELAP/SCDAPSIM computer pro-gram to predict the results of several severe experimentsexcept the SFD-ST test program.

This study is aimed to study and assess the predictioncapability of RELAP/SCDAPSIM/MOD3.4 code on the PBFSFD-ST experiment. Section 2 provides the initial conditionsand the details that are used to run the code. Section 3 showsthe predicted results from the RELAP/SCDAPSIM/MOD3.4code and a comparison of the predicted data with theexperimental results and results from another two versionsof SCDAP/RELAP5 codes and a discussion of the results.Section 4 summarizes and gives the important conclusionsfrom this study.

2. Description of the RELAP/SCDAPSIM/MOD3.4 Code and the Experiment

This section provides the details of the code used in thisstudy and experiment used to predict and analyze the results.TheRELAP/SCDAPSIM/MOD3.4 code is used to analyze thethermal hydraulic response and core behavior of the SevereDamage Fuel Scoping Test (SFD-ST) conducted in the PowerBurst Facility (PBF) at the Idaho National Laboratory, USA.

2.1. RELAP/SCDAPSIM/MOD3.4. The RELAP/SCDAPSIMcode, designed to predict the behavior of reactor systemsduring normal and accident conditions, is being devel-oped as part of the international SCDAP Developmentand Training Program (SDTP) [14]. Innovative SystemsSoftware (ISS) is the administrator for the program. TheRELAP/SCDAPSIM/MOD3.4 code used by general usercommunity for production safety analysis runs wide range oftransients fast and more reliably than earlier versions of thecode.The purpose of the code is to predict core response andthe thermal hydraulic response of the reactor coolant system(RCS) during normal operation, design basis, and severeaccidents.The RELAP/SCDAPSIM code combines two codesRELAP5 and SCDAP into one code [15]. The RELAP5 modelis used to analyze the thermal hydraulic response of the RCS,control system behavior, and rector kinetics and the behaviorof some reactor components such as valves and pumps. The

SCDAP model is applied to estimate the behavior of the coreand vessel structures under normal and accident conditions.It is also capable of predicting the later stages of a severeaccident including themelting of fuel, debris andmolten poolformation, vessel interactions, and failed structural vessel.

The RELAP/SCDAPSIM/MOD3.4 code uses the publiclyavailable RELAP5/MOD3.3 and SCDAP/RELAP5/MOD3.2and 3.3 models developed by the US Nuclear RegulatoryCommission in combination with proprietary and someadvanced features such as (a) advanced programming andnumerical methods, (b) user options, (c) models developedby ISS, and other STDP members.

2.2. Code Nodalization for PBF SFD-ST Experiment [4].The PBF SFD-ST program was the first experiment in theseries of four severe fuel damage tests that was performedto gain some operating experiences for the following testprograms.Themain purposewas to understand the dynamicsof a fuel bundle, hydrogen production during the normaltransient phase, and the coolability of the bundle duringthe reflood phase. The results of the test proved that theinstrumentation and procedures used gave usable results andprovided informative guidance for the posttest examinationand analysis.

The PBF SFD-ST test bundle consisted of 32 fresh pres-surized water (PWR) design fuel rods with 0.9114m in lengthand arranged in a 6 × 6 array with the corner rods missingsurrounded by a shroud with an inner diameter of 127mm asshown in Figure 1. The shroud consisted of a zirconia insu-lator placed between Zircaloy layers. The test train assemblywas built by the Pacific Northwest Laboratory (PNL) andassembled at the Idaho National Laboratory (INL), formerlythe Idaho National Engineering Laboratory (INEL). Themain components were the inlet, outlet, insulated shroud,bundle, flow tube, and closure head. The used instrumentswere thermocouples, coolant pressure transducers, flow ratemeters, and rod-failure pressure switches. The measuredelevations were 0.35, 0.5, and 0.7m above the bottom of thefuel.The peak temperatures,measured by the thermocouples,were up to 2400K for the cladding and 1600K for the shroud.

The PBF SFD-ST test had two phases, the high temper-ature transient (prereflood phase) and reflood phase. Forthe high temperature transient phase, the initial reactorpower was 35 kW and the initial coolant flow through thebundle was approximately 0.016 l/s. The reactor power wasincreased from 35 kW to 93 kW (maximum power) in the205 minutes from the beginning of transient phase to theend of the experiment. At approximately 20 minutes after thestart of the experiment, the coolant flow rate was increasedto be 0.02 l/s. The termination of the transient phase was atapproximately 205 minutes at which time the reactor wasscrammed. After the occurrence of the reactor scram, thecoolant flow rate significantly increased 0.035 l/s to reduce thebundle temperature.This periodwas called the reflood phase.

The nodalization scheme of PBF SFD-ST code used forthe input code in the calculation is shown in Figure 2. Itshows the RELAP5 and SCDAP components in detail [15].The RELAP5 components are divided into two subparts, thebundle and bypass. In the RELAP5 nodalization of the test

Science and Technology of Nuclear Installations 3

80.01 mm

91.69mm

High density ：Ｌ／2 cylindersLow density ：Ｌ／2 fiber

Bundle coolant inlet lines

：Ｌ／2 insulation region7.87 mm thick

Instrument hardlines

Fission chambers

Outside shroud wall,127.0 mm ID, 1.52 mm thick

Inside shroud wall,115.82 mm ID, 1.52 mm thick

Zircaloy flow tube,140.97mm ID, 3.18 mm wall thickness

Inner wall of in pile tube,154.94mm ID

Fresh fuel rod

Zircaloy lead carrier

Zircaloy saddle

Zircaloy inner liner,0.76 mm thick

Bypass flow down

Bypass �ow up

Pressure regulator line

Figure 1: Cross section view of SFD-ST shroud and bundle.

SCDAP componentsRELAP5 components

Component 4(shroud)

Component 3Component 2Component 1

10

09

08

07

06

05

04

03

02

01

02

01

130 componentPipe for bypass

120 componentWater injection junction

110 componentSource volume for bypass

021 componentWater injection junction

010 componentSource volume for water injection

030 componentTest train

040 componentPipe outlet junction

050 componentCollector

060 componentCollector outlet junction

070 componentOutlet volume

01

10

09

08

07

06

05

04

03

02

01

10

09

08

07

06

05

04

03

02

01

10

09

08

07

06

05

04

03

02

01

10

09

08

07

06

05

04

03

02

Shro

ud

16 fr

esh

fuel

rods

12 fr

esh

fuel

rods

4 fre

sh fu

el ro

ds

Time-dependent volume

Single volume

Pipe

Junction orientationTime-dependent junctionSingle junction

InsulatingShroud

150 componentOutlet volume for bypass

140 componentOutlet junction

Figure 2: Nodalization of the PBF SFD-ST experiment.

bundle, source volume 10 injects water into the test train,pipe volume 30, through junction 021. Pipe volume 30 isdivided into 0.1m subvolumes. The outlet volume of pipe30 is connected to the collector, volume 50, by junction 40.

Volume 50 has two 0.5m in length subvolumes. The outletvolume 050 is connected to volume 070 by junction 060.In the bypass part, source volume 110 is connected to pipecomponent volume 130, to inject water into the bypass region


High temperature transient

Reactor scram at 205 minutes

Reflood

40 80 120 160 200 2400

Time (min)

0

20

40

60

80

100

Pow

er (k

W)

Figure 3: Two phases of the experiment, high temperature transientand reflood phase, in view of the bundle power-time profile.

Table 1: Examples of boundary conditions used in the input code.

Boundaryconditions Values

Bundle description 32 fresh rodsPower (kW) 35–90Pressure (MPa) 6.6–6.7Nominal inlet flowrate (L/s) 0.02

Cooldownprocedure

After reactor scram, 0.02 L/s refloodincreasing to ∼0.35 L/s

through junction 120. Volume 130 is 1m in length and isconnected to sink volume, 150 by junction 140.

Four SCDAP components were used to model the PBFSFD-ST test bundle. Components 1, 2, and 3 represent theinner ring of 4 fresh fuel rods, the middle ring of 12 fuel rods,and the outer ring of 16 fuel rods, and component 4 representsthe shroud. Each component has 10 axial nodes 0.1m inlength. The fuel rod material used in this calculation wasfreshUO2.The fuel pellet, outer cladding, and inner claddingradiuses were 4.13mm, 4.81mm, and 4.22mm, respectively.

2.3. Boundary Conditions. Figure 3 demonstrates two phasesof the experiment, high temperature transient phase (prere-flood phase), and reflood phase in view of the bundle power-time profile. Figure 4 and Table 1 give some examples ofboundary conditions for the PBF SFD-ST experiment usedin the input code. The key boundary conditions are bundlepressure, bundle inlet flow rate, power profile, and bundleinlet temperature. The power significantly increased from35 kW at the beginning of the experiment to a maximumpower of 93 kWwhen the reactor was scrammed.The reactorpower was reduced significantly after reactor scram at about205 minutes. The first 205 minutes of the experiment wascalled the high temperature transient. The reflood phasestarted after the reactor was scrammed.The coolant flow ratewas about 0.02 l/s at 525Kduring the transient phase from the

Table 2: Damage level states classified by the code.

Damage level codes Damage states0.0 Intact geometry0.1 Rupture due to ballooning0.2 Rubble (fragmented)0.4 Cohesive debris1.0 Molten pool

start of the experiment to 205 minutes. Afterwards, duringthe reflood phase the coolant flow rate was increased from0.02 l/s to 0.035 l/s to cool the hot test fuel. The bundle inlettemperature was a constant 525K during the transient phaseand was reduced to about 514K at 205 minutes, the time ofreactor scram. Figure 5 displays the flow rate and temperatureboundary conditions of the bypass line. The temperaturetrend in the bypass was similar to the bundle but the flow ratein the bypass was quite constant after 20 minutes, whereasthe flow rate in the bundle was dramatically increased atreactor scram. The boundary conditions of the bundle aregiven in source volume 010 and the boundary conditions forthe bypass are given in source volume 110.

2.4. Focused Parameters. Thepredictive capability of RELAP/SCDAPSIM/MOD3.4 is assessed by comparing the calcu-lated results with the PBF SFD-ST experimental results andcalculated results from other comparable codes such asthe SCDAP/RELAP/MOD3.2 and SCDAP/RELAP/MOD3.3code. The parameters being focused on in this paper are(1) the collapsed water level, (2) cladding temperatures, and(3) hydrogen production. It is known that the water level inreactor is a crucial parameter which affects the temperature ofthe cladding temperature. An accurate prediction of thewaterlevel in the core results in an acceptably accurate predictionof the core temperature.

Hydrogen is normally produced from the reaction ofZircalloy cladding with steam when the bundle temperaturereaches 1273K [17]. The production of hydrogen usuallyoccurs during a severe accident. Since the cladding tempera-ture is a key parameter in determining hydrogen productionrate, the accurate prediction of cladding temperature leadsto the good estimation of hydrogen production. Actually, thecladding or fuel temperature in a damage state is interested,but there are limitations in their measurements, since thethermocouples failed to measure at temperatures above2000K. Analysis the comparison provides the understandingof the core behavior and the phenomena and some limitationof codes.

Additionally, this paper shows some examples ofthe calculated temperature distribution and damagelevel states of the fuel rod and cladding by the RELAP/SCDAPSIM/MOD3.4 code. The five damage level states aredefined in Table 2.The “0.0” code represents intact geometry,that is, no change in the geometry of fuel bundle. The “0.1”code indicates rupture of the cladding due to ballooning hasoccurred. The “0.2” code indicates the formation of rubbledebris and fragmentation occurred.The “0.4” and “1.0” codesindicate the formation of cohesive debris and a molten pool,


Temperature

Flow rate

Pressure

510

515

520

525

Bund

le in

let t

empe

ratu

re (K

)

40 80 120 160 200 2400

Time (min)

0.00

0.01

0.02

0.03

0.04

0.05

Bund

le in

let fl

ow ra

te (L

/s)

6.6

6.7

6.8

Bund

le p

ress

ure (

MPa

)

Figure 4: Boundary conditions used in the input code.

Temperature

Flow rate

510

515

520

525By

pass

tem

pera

ture

(K)

50 100 150 2000

Time (min)

Bypa

ss in

let fl

ow ra

te (L

/s)

2.1

2.2

2.3

2.4

2.5

2.6

2.7

Figure 5: Boundary conditions of the bypass line used in the inputcode.

respectively. It is known that fuel and cladding temperatureshave a relationship to the occurring damage level states ofthe fuel rod and cladding. Besides, damage level states havedepended on excessive internal pressure of fuel rod [5].

3. Analysis Results

Thecalculated results from theRELAP/SCDAPSIM/MOD3.4code are compared to the measured data and the cal-culated results from the SCDAP/RELAP5/MOD3.2 andSCDAP/RELAP5/MOD3.3 code [4, 16]. Information pertain-ing to the total hydrogen released during the experimentthat is used to compare the RELAP/SCDAPSIM/MOD3.4calculated values to themeasured results taken from availablesources [18, 19]. The use of different publications for theinformation of hydrogen productions is due to the factthat this related information was published latterly after thecompletion of the Postirradiation Examination (PIE) of thedamaged bundle.

ExperimentalCalculated SCDAP/RELAP5/MOD3.3Calculated RELAP/SCDAPSIM/MOD3.4

0.5m elevation

2400 120 160 20040 80

Time (min)

0.0

0.2

0.4

0.6

0.8

1.0

Col

lapse

d liq

uid

wat

er le

vel (

m)

Figure 6: Comparison of experimental and calculated water levelsin the bundle.

3.1. Cladding Temperatures. This subsection describes theresults pertaining to the cladding temperatures. Figure 6shows measured the collapsed liquid water levels in thebundle during the experiment and those calculated bythe SCDAP/RELAP5/MOD3.3 and RELAP/SCDAPSIM/MOD3.4 code. The values for the water level are based onthe interface between the two-phase region at the top ofthe bundle and the single-phase region at the bottom. Theexperimental water level results for the reflood phase ofthe experiment are not included in this paper (only hightemperature transient phase). In contrast, the temperaturevalues calculated by the SCDAP/RELAP5/MOD3.3 andRELAP/SCDAPSIM/MOD3.4 code include both the hightemperature transient and reflood phases. As shown inFigure 6, the SCDAP/RELAP5/MOD3.3 and RELAP/SCDAPSIM/MOD3.4 codes overpredict the water levelduring the initial 40 minutes of the experiment. If the water


At 0.35m elevation from the bottomExperimentalCalculated SCDAP/RELAP5/MOD3.2Calculated SCDAP/RELAP5/MOD3.3Calculated RELAP/SCDAPSIM/MOD3.4

120 2000 16040 24080

Time (min)

500

1000

1500

2000

2500

3000

Tem

pera

ture

(K)

Figure 7: Comparison of measured and calculated cladding tem-peratures at 0.35m elevation.

level is measured correctly during those two periods, thecoolant flow rate during the first 40 minutes which is used asthe initial flow boundary conditions in RELAP/SCDAPSIMcalculation may not be sufficiently accurate. This may resultin the overestimation of water level during the initial phase.It is noted that the coolant flow rate during 40 minutes ofthe experiment was relatively low compared to the flowrate after 40 minutes of experiment. After 40 minutes,the values predicted by both codes are generally similarto the experimental values. During the reflood phase, theboth codes are able to predict well the water level but theexperimental water level is not measured due to the failure ofthe instrumentation. The calculated cladding temperaturesby the RELAP/SCDAPSIM/MOD3.4 code are compared tothe temperatures measured during the experiment and thecalculated temperatures from the SCDAP/RELAP5/MOD3.2and SCDAP/RELAP5/MOD3.3 codes at 0.35, 0.5, and0.7m elevations. All versions of the SCDAP/RELAP5 andRELAP/SCDAPSIM codes calculate a temperature at themidpoint of an axial node (each axial node in this assessmentcalculation is 0.1 in length). Therefore, to obtain a predictedtemperature at the measured elevations of 0.5 and 0.7, alinear interpolation is performed between the calculatedtemperature from the axial node below and above thedesired elevation. (For example, for the 0.5m elevation thelinear interpolation is performed between the calculatedtemperatures at the 0.45m and 0.55m elevations, and for the0.7m elevation the interpolation was between the calculatedtemperatures at the 0.65 and 0.75m elevations.) Figures 7, 8,and 9 show the comparisons of the cladding temperatures at0.35, 0.5, and 0.7m elevations, respectively.

As shown in Figure 7, for the cladding temperatures at0.35m elevation, dry-out occurred 100 minutes from the

At 0.5 m elevation from the bottomExperimentalCalculated SCDAP/RELAP5/MOD3.2Calculated SCDAP/RELAP5/MOD3.3Calculated RELAP/SCDAPSIM/MOD3.4

500

1000

1500

2000

2500

3000

Tem

pera

ture

(K)

120 2000 16040 24080

Time (min)


beginning of the transient. From 100 minutes to 200 minutesthe temperature continuously increases until the limitation ofinstrumentation is reached and the reactor is scrammed.Thecladding calculated temperature fluctuates slightly between100 and 140 minutes. The reactor was scrammed at 205minutes.

As shown in the Figure 8, the sudden changes ofmeasuredcladding temperature at 0.5m elevation are observed duringapproximately 20 to 30 minutes while the opposite trend ofmeasured water level changes around 0.5m elevation is alsoobserved during that period of time. This indicates that thecladding temperature is significantly affected by the liquidwater level.The calculated temperatures from the codes beingcompared in this study show a small amount of fluctuationbetween 40 and 60 minutes with the predicted trends closelymatching the measured temperature up to the time of reactorscram.

As shown in Figure 9, the predicted cladding temper-atures at 0.7m elevation from the three codes follow themeasured temperature reasonably well up to thermocouplefailure. The calculated temperatures fluctuate moderatelyfrom 0 to 30 minutes. From 30 minutes until the end of thetransient phase, the predicted temperature steadily increases.

After the end of the transient phase, the bundle isreflooded. During this phase, changes in the bundle geom-etry occurred. During the experiment, the instrumentationcould not measure the data during the reflood phase dueto the measureable limitation of the instruments, but theRELAP/SCDAPSIM/MOD3.4 code is capable of predictingbundle behaviors. During the reflood phase, the calculatedwater levels increase rapidly depending on the inlet flow.During the refloodphase starting at 205minutes, the claddingtemperature at all elevations increases rapidly for several


At 0.7m elevation from the bottomExperimentalCalculated SCDAP/RELAP5/MOD3.2Calculated SCDAP/RELAP5/MOD3.3Calculated RELAP/SCDAPSIM/MOD3.4

40 80 160 2402001200

Time (min)

500

1000

1500

2000

2500

3000

Tem

pera

ture

(K)


Experimental [12]Calculated SCDAP/RELAP5/MOD3.2Calculated SCDAP/RELAP5/MOD3.3Calculated RELAP/SCDAPSIM/MOD3.4

40 80 160 2400 120 200

Time (min)

0.0

0.4

0.8

1.2

1.6

Hyd

roge

n re

leas

e rat

e (g/

s)

Figure 10: Comparison of hydrogen release rate measured in theexperiment and predicted by the codes.

minutes before it dramatically decreases to about 500K by210 minutes.

From the Figures 7, 8, and 9, the calculated RELAP/SCDAPSIM/MOD3.4 cladding temperatures at all elevationsare in a good agreement with experiment and are more accu-rate than those calculated by the SCDAP/RELAP/MOD3.2and SCDAP/RELAP/MOD3.3 code. There are some uncer-tainties in the calculated values by the code. The results of

Experimental [4]Experimental [16]Calculated SCDAP/RELAP5/MOD3.2Calculated SCDAP/RELAP5/MOD3.3Calculated RELAP/SCDAPSIM/MOD3.4

0

100

200

300

400

500

Cum

ulat

ive h

ydro

gen

prod

uctio

n (g

)

40 80 120 160 200 2400

Time (min)

Figure 11: Comparison of cumulative hydrogen production mea-sured in the experiment and predicted by the codes.

the RELAP/SCDAPSIM/MOD3.4 code are generally morereliable due to a better prediction of the interface water leveland the improved fuel assembly behavior and in-vessel meltretention models.

3.2. Hydrogen Production. Figures 10 and 11 display thehydrogen production rate and hydrogen accumulated pro-duction, respectively. The three calculated results are ingood agreement with measured trend in prereflood phaseonly. During the reflood phase, it can be seen that thecalculated results were unreasonable with the measuredresults. The significant uncertainties of hydrogen productionmeasurement during the reflood phase need to be furtherinvestigated in the details or the oxidation and hydrogenproduction correlations and related models utilized in theRELAP/SCDAPSIM/MOD3.4 may need to have a furtherimprovement.

Table 3 shows accumulated hydrogen production ingrams predicted by the three versions of the code beingcompared in this paper and themeasured quantity. In Table 3,there are 2 columns for each version of the code and themea-sured data. One column gives the total hydrogen producedduring the prereflood phase and another provides the totalhydrogen released in the experiment. In the experiment, onlythe total quantity of hydrogen produced was measured. Themeasured total hydrogen produced from reference [4], onlinemeasurements, was varied from 235 to 515 g.The total hydro-gen based on reference [16]was 150±35 g.The total calculatedhydrogen production of RELAP/SCDAPSIM/MOD3.4 was155 g.

The calculated RELAP/SCDAPSIM/MOD3.4 results arein reasonably good agreement with experimental although


Table3:Com

paris

onof

measuredandcalculated

hydrogen

prod

uctio

n.

Experim

ent

Hydrogenprod

uctio

n(g)

Measured[4]

Measured[16]

SCDAP/RE

LAP/MOD3.2[4]

SCDAP/RE

LAP/MOD3.3[16]

RELA

P/SC

DAPS

IM/M

OD3.4

Prerefloo

dTo

tal

Prerefloo

dTo

tal

Prerefloo

dTo

tal

Prerefloo

dTo

tal

Prerefloo

dTo

tal

PBFSFD-ST

N/A

235–515

N/A

150±35

121

121

98125

145

155


0200.0400.0600.0800.010001200140016001800200022002400

10 20 30 400

Radial cross section of bundle from the center (mm)

0.2

0.4

0.6

0.8

Elev

atio

n (m

)

(a) At 80 minutes

0200.0400.0600.0800.010001200140016001800200022002400

30 4010 200


0.2

0.4

0.6

0.8

Elev

atio

n (m

)

(b) At 180 minutes

0200.0400.0600.0800.010001200140016001800200022002400

10 20 30 400


0.2

0.4

0.6

0.8

Elev

atio

n (m

)

(c) At 190 minutes

0200.0400.0600.0800.010001200140016001800200022002400

10 20 30 400


0.2

0.4

0.6

0.8

Elev

atio

n (m

)

(d) At 200 minutes

0200.0400.0600.0800.010001200140016001800200022002400

10 20 30 400


0.2

0.4

0.6

0.8

Elev

atio

n (m

)

(e) At 210 minutes

0200.0400.0600.0800.010001200140016001800200022002400

10 20 30 400


0.2

0.4

0.6

0.8

Elev

atio

n (m

)

(f) At 220 minutes

Figure 12: Temperature distribution of fuel bundle and shroud predicted by the RELAP/SCDAPSIM/MOD3.4 code.

there were some uncertainties. A more accurate predictionof the cladding temperatures led to a better prediction ofhydrogen production.

3.3. Temperature Distribution. Figure 12 shows the axial andradial distribution of temperature from the bundle center tothe outer shroud at 80, 180, 190, 200, 210, and 220 minutes

from the beginning of the experiment, represented by (a), (b),(c), (d), (e), and (f), respectively. The distance from centralbundle to the outer shroud was 0 to 48.635mm. The bundleis about 40mm in radius and it is divided into 3 componentsin same radius. The shroud was 8.63mm in radius. Thefigure shows the changing temperature distributions axiallyand radially at ten elevations in the bundle (measured from


−0.1

0.0

0.1

0.2

0.3

0.4

Dam

age l

evel

40 80 120 160 200 2400

Time (min)

At 0.35m elevationAt 0.55m elevationAt 0.75m elevation

At 0.95m elevation

Figure 13: Damage level states of fuel bundle predicted by theRELAP/SCDAPSIM/MOD3.4 code.

0.05 to 0.95m). In the figure, (a), (b), (c), and (d) show thebundle temperature distribution prior to reactor scram; (e)and (f) show the distribution for two periods after scram,the first one 5 minutes after scram and the other 10 minuteslater.

As shown in the Figure 12, maximum bundle tem-peratures are between 0.70 and 0.80m elevations due toaccumulated heat of lower elevations and delayed time ofcoolant (fed inlet coolant from lowest elevation first). Thereis gap between bundle and shroud, so the gap distance isabout 38 to 40mm.At 80minutes, themaximum temperatureis about 1130 K at between 0.70 and 0.80m elevations. At180, 190, and 200 minutes, before occurring reactor scram(205 minutes), the maximum temperatures are about 1850,1959, and 2163K at between 0.70 and 0.80m elevationsand temperatures at every elevation increase at the follow-ing time points. At 210 and 220 minutes, after occurringreactor scram, temperatures decrease due to the increasedinlet coolant flow rate (0.035 l/s). At between 0.1 and 0.2melevations, the temperatures are at a maximum value sinceaccumulated major melts possibly occurred at the bottomelevation.

3.4. Damage Level States, Temperature, and Time of CladdingRupture. Figure 13 shows damage level states of fuel bundleat 0.35, 0.55, 0.75, and 0.95m elevations based on theRELAP/SCDAPSIM/MOD3.4 code prediction. The maxi-mum damage level state for all elevations was “0.2,” definedas fuel bundle rubble. At 0.35m elevation, the fuel bundleruptured at reactor scram and immediately changed torubble. At 0.55m elevation, the fuel bundles retained intactgeometry during experiment. At 0.75m elevation, rupture

occurred between 80 and 205 minutes from the beginning ofthe experiment forming rubble debris after reactor scram. At0.95m elevation, rubble debris formed after reactor scram.The first rupture of fuel rod and cladding occurred at0.75m elevation at about 80 minutes where this point wasmaximum temperature of all elevation at this time. At 0.35,0.75, and 0.95m elevations, the last damage level state ofrubble and fragmented of fuel bundle occurred while thereare not changes in fuel bundle and cladding geometry only at0.55m elevation during the test period. This may be slightlydifferent with the obtained results from the radiograph offuel bundle geometry after the test performing with whichthe fragment rods are founded along the fuel bundle length[4].

Table 4 shows the results of temperature and time ofcladding rupture. The results in the INL reference indicatedthat fuel rod rupture occurred at about 78 to 87 minuteswith maximum cladding temperatures between 1000 and1200K [4]. The experimental results reported in another ref-erence stated that the measured time when cladding ruptureoccurred was between 90 and 104 minutes from the begin-ning of experiment at a rupture temperature between 1150and 1200K [5].The SCDAP/RELAP/MOD3.3 code predictedthe cladding rupture to occur between 100 and 106 min-utes and the predicted rupture temperature is about 1050K[16]. The RELAP/SCDAPSIM/MOD3.4 code estimates thecladding rupture beginning at 80 minutes and the rupturetemperature at 1130K. Therefore, the calculated results pre-dicted by the RELAP/SCDAPSIM/MOD3.4 code are quitereliable.

4. Conclusion

This paper shows the assessment of prediction capabilityof the RELAP/SCDAPSIM/MOD3.4 code (developed byInnovative Systems Software) focusing on cladding temper-atures and hydrogen production in the PBF SFD-ST exper-iment. The calculated fuel and cladding temperatures andhydrogen production by the RELAP/SCDAPSIM/MOD3.4code are compared with experimental results and thoseparameters calculated by other comparable codes such asthe SCDAP/RELAP/MOD3.2 and SCDAP/RELAP/MOD3.3codes.

The calculated cladding temperatures by the RELAP/SCDAPSIM/MOD3.4 code are generally in a good agreementwith the experiment results and more reasonable than thosepredicted by the SCDAP/RELAP5/MOD3.2 and 3.3 codes.The accurate predicted interface liquid level is attributed tothe acceptable predicated cladding temperatures. The bun-dle and shroud calculated temperature distribution showedthe RELAP/SCDAPSIM/MOD3.4 calculated temperaturesin detail. The calculated hydrogen production results arereasonable with experimental data. The damage level statesare sufficiently corresponded to the experimental results.There are some discrepancies in the predicted claddingtemperatures, damage level, and hydrogen production due touncertainties in the boundary conditions and limitation ofthe current codes capability.


Table4:Com

paris

onof

temperature

andtim

eofcladd

ingruptureb

ythec

odes

andin

thee

xperim

ent.

Experim

ent

Rupturetem

perature

Timeo

fcladd

ingrupture

Measured[5]

Measured[4]

SCDAP/RE

LAP/

MOD3.3[16]

RELA

P/SC

DAPS

IM/

MOD3.4

Measured[5]

Measured[4]

SCDAP/RE

LAP/

MOD3.3[16]

RELA

P/SC

DAPS

IM/

MOD3.4

(K)

(K)

(K)

(K)

(min)

(min)

(min)

(min)

PBFSFD-ST

1150–1200

1000–1200

1050

1130

90–104

78–87

100–

106

80


Conflicts of Interest

The authors declare that there are no conflicts of interestregarding the publication of this paper.

Acknowledgments

This research was supported by the Graduate School, Chula-longkorn University, under the program of Research Assis-tant Scholarship.

References

[1] Nuclear Regulatory Commission Special Inquiry Group, Vol-ume I Three Mile Island: A Report to the commissioners andto The Public, Nuclear Regulatory Commission Special InquiryGroup, 1979.

[2] Y. Amano, The Fukushima Daiichi Accident, InternationalAtomic Energy Agency, 2011.

[3] J. I. M. Damian, A. Weir, V. L. Putnam, and J. D. Bess, “PowerBurst Facility: U(18)O2-CaO-ZrO2 Fuel Rods in Water,” Tech.Rep. INL/EXT-09-15446, 2009.

[4] A. D. Knipe, S. A. Ploger, and D. J. Osetek, PBF Severe FuelDamage Scoping Test-Test Results Report, EGG Idaho Inc, IdahoFalls, Idaho, USA, 1986.

[5] L. J. Siefken, “Models for the Configuration and Integrity ofPartially Oxidized Fuel Rod Cladding at High Temperatures -Final Design Report,” Tech. Rep. INEEL/EXT-99-00107, 1999.

[6] A. R. Antariksawan, M. Q. Huda, T. Liu, J. Zmitkova, C. M.Allison, and J. K. Hohorst, “Validation of RELAP/SCDAPSIM/MOD3.4 for research reactor applications,” in Proceedings of the13th International Conference on Nuclear Engineering, Beijing,China, May 2005.

[7] C. M. Allison and J. K. Hohorst, “An assessment of effectivenessof core exit temperatureswith respect to PWRcore damage stateusing RELAP/SCDAPSIM/MOD3.4,” Nuclear Engineering andDesign, vol. 238, no. 7, pp. 1547–1560, 2008.

[8] A. Kaliatka and E. Uspuras, “Development and testing ofRBMK-1500 model for BDBA analysis employing RELAP/SCDAPSIM code,” Annals of Nuclear Energy, vol. 35, no. 6, pp.977–992, 2008.

[9] M. Mladin, D. Dupleac, and I. Prisecaru, “SCDAP/RELAP5application to CANDU6 fuel channel analysis under postulatedLLOCA/LOECC conditions,” Nuclear Engineering and Design,vol. 239, no. 2, pp. 353–364, 2009.

[10] D. Dupleac, M. Mladin, and I. Prisecaru, “Generic CANDU6 plant severe accident analysis employing SCAPSIM/RELAP5code,”Nuclear Engineering andDesign, vol. 239, no. 10, pp. 2093–2103, 2009.

[11] T. Kaliatka, A. Kaliatka, V. Vileiniskis, and E. Uspuras, “Mod-elling of QUENCH-03 and QUENCH-06 experiments usingRELAP/SCDAPSIMandASTEC codes,” Science andTechnologyof Nuclear Installations, vol. 2014, Article ID 849480, 13 pages,2014.

[12] H. Madokoro, K. Okamoto, and Y. Ishiwatari, “SCDAP modelimprovement with QUENCH-06 analysis,” in Proceedings ofthe 2014 22nd International Conference on Nuclear Engineering,ICONE 2014, Prague, Czech Republic, July 2014.

[13] H. Madokoro, “Assessment of RELAP/SCDAPSIM/MOD3.5against the BWR core degradation experiment CORA-17,” inProceedings of the 10th International Topical Meeting on Nuclear

Thermal-Hydraulics, Operation and Safety, Okinawa, Japan,2014.

[14] C.M. Allison and J. K. Hohorst, “Role of RELAP/SCDAPSIM innuclear safety,” Science and Technology of Nuclear Installations,vol. 2010, Article ID 425658, 17 pages, 2010.

[15] N. Rattanadecho, S. Rassame, K. Silva, C. Allison, and J.Hohorst, “Assessment of RELAP/SCDAPSIM/MOD3.4 Predic-tionCapabilitywith Severe FuelDamage ScopingTest: Focusingon Reactor Core Temperatures and Hydrogen Production,” inProceeding the 11th International Topical Meeting on NuclearReactor Thermal Hydraulics, Operation and Safety, Gyeongju,Korea, October 2016.

[16] L. J. Siefken, E. W. Coryell, E. A. Harvego, and J. K. Hohorst,Assessment of Modeling of Reactor Core Behavior During SevereAccidents, daho National Engineering and Environmental Lab-oratory, Idaho Falls, Idaho, USA, 2001.

[17] A. L. Camp, J. C. Cummings, M. P. Sherman et al., “Light WaterReactor Hydrogen Manual,” NUREG/CR-2726, 1983.

[18] B. A. Cook, P. A. Kalish, and D. M. Tow, Posttest Examinationof the Severe Fuel Damage Scoping Test Bundle Geometry, IdahoNational Engineering Laboratory, Idaho Falls, Idaho, USA,1984.

[19] P. Hofmann, “Metallographic Examination of the Severe FuelDamage Scoping Test (SFD-ST) Fuel Rod Bundle Cross Sec-tions,” NUREG/CR-5119, July 1988.

TribologyAdvances in

Hindawi Publishing Corporationhttp://www.hindawi.com Volume 2014

FuelsJournal of


Journal ofPetroleum Engineering


Industrial EngineeringJournal of


Power ElectronicsHindawi Publishing Corporationhttp://www.hindawi.com Volume 2014

Advances in

CombustionJournal of


Journal of


Renewable Energy

Submit your manuscripts athttps://www.hindawi.com


StructuresJournal of

International Journal of

RotatingMachinery


EnergyJournal of


Hindawi Publishing Corporation http://www.hindawi.com

Journal of

Volume 201Hindawi Publishing Corporation http://www.hindawi.com Volume 201

International Journal ofInternational Journal of


Nuclear InstallationsScience and Technology of


Solar EnergyJournal of


Wind EnergyJournal of


Nuclear EnergyInternational Journal of


High Energy PhysicsAdvances in

The Scientific World JournalHindawi Publishing Corporation http://www.hindawi.com Volume 2014

assessment of relap/scdapsim/mod3.4 prediction capability...

Documents