safety analysis of mtr type research reactor during postulated beam tube break inducing positive...

Annals of Nuclear Energy xxx (2014) xxx–xxx

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Annals of Nuclear Energy

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Safety analysis of MTR type research reactor during postulated beamtube break inducing positive reactivity

http://dx.doi.org/10.1016/j.anucene.2014.08.0180306-4549/� 2014 Elsevier Ltd. All rights reserved.

E-mail address: [email protected]

Please cite this article in press as: Khakim, A. Safety analysis of MTR type research reactor during postulated beam tube break inducing positive reaAnn. Nucl. Energy (2014), http://dx.doi.org/10.1016/j.anucene.2014.08.018

Azizul KhakimNuclear Energy Regulatory Agency (BAPETEN), Jl. Gajah Mada 8, Jakarta, 10120, Indonesia

a r t i c l e i n f o

Article history:Received 30 November 2013Accepted 12 August 2014Available online xxxx

Keywords:Beam tubePositive reactivityBreakRadial and tangentialVoidedFlooded

a b s t r a c t

MTR type research reactors are commonly equipped with beam tubes which functions as experimentalfacilities out off the core. When used for irradiation, the tubes are filled with air, in contrary they areflooded with water when idle. Should the break occur during irradiation, the water comes in and theair goes out. Beam tubes break should be viewed as potential accident to occur during reactor life time,caused, for instance by an earth quake. The break was assumed to occur inside the reactor pool and nowater leaked out of the pool as the tube isolation valves worked properly. When being used for irradia-tion, the change from voided to flooded condition due to the break, inserts positive reactivity as moreneutrons are moderated and reflected back to the core by the water. Functioning as moderator, water alsoreflects neutrons back to the core. More neutrons in the beam tube slowed down to a lower energy. Forthe above reason, safety analysis needs to be performed for beam tube break and its influence on reactordynamics. The radial beam tubes gave higher positive reactivity feedback than that of tangential ones.The calculations on reactivity induced by beam tubes break were performed with MCNP5 code. It wastreated as reactivity initiated accident analyzed using PARET/ANL code. Beam tube S5 gave the highestreactivity feedback of 0.062%dk/k and was treated as the initiating event. The maximum fuel and cladtemperature reached 175.28, and 136.75 �C, respectively. All safety parameters were confirmed underacceptable limits.

� 2014 Elsevier Ltd. All rights reserved.

1. Introduction

Material Testing Reactor (MTR) type research reactors usuallyprovide in-core and off-core irradiation facilities. The in-core irra-diation facilities include Irradiation Position (IP) and Central Irradi-ation Position (CIP). In addition, the reactor also has off-coreirradiation facilities such as beam tubes. Indonesian MTR typeresearch reactor has six beam tubes, four of which are radial andthe other two are tangential tubes. To obtain the desired neutronflux, the tubes are voided by draining water out of the tubes.Therefore, the tubes are filled with air when used for irradiation.

An accident due to beam tubes leakage needs to be taken intoaccount in the safety analysis, as the leakage leads the air to comeout and the water to come in the tubes which in turn induces posi-tive reactivity, although the facilities are located out of the core.That scenario is when the break occurs inside the reactor pooland the isolation valve remained intake. Thus, the leakage doesnot lead to LOCA (Loss of Coolant Accident). The insertion rate ofpositive reactivity is proportional to the leakage rate of air. Thetube break is assumed to occur in a connecting welding inside

the pool due to external event such as earth quake. Analysis ofbeam tube rupture leading to LOCA has been performed at 5 MWMTR type Tehran Research Reactor (TRR) (Hedayat et al., 2007).

Entering water into the beam tubes gives additional moderationeffect by water. In addition, it reduces the neutron leakage andmore neutrons are reflected back into the core. The increase inmoderation effect by water inside the tube leads to the increasein low energy neutron populations.

The reactivity feedbacks from each beam tube are calculatedusing Monte Carlo method MCNP5 (X-5 Monte Carlo Team, 2003)code employing continuous neutron energy of ENDF/B-VI nucleardata library. MCNP5 code has been validated with experimentaldata of first criticality of first core MPR GAS (Multi Purpose ReactorG.A. Siwabessy). The first core was composed of fuel plates made offresh U3O8Al. The first criticality was attained consisting of 9 fuelelements and 6 control elements, where 5 control elements werefully withdrawn and one control element of C-8 (regulating rod)at a position of 475 mm. First core MPR GAS comprised of 12 fuelelements and 6 control elements. The validation results showedthat MCNP5 agreed very well with the experiment, where themaximum relative deviation from the experimental data was only0.43% (Khakim Azizul, 2012).

ctivity.

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2 A. Khakim / Annals of Nuclear Energy xxx (2014) xxx–xxx

On the other hand, the effect of positive reactivity insertion onthe reactor power dynamics was simulated with PARET/ANL code.The code has been validated with experiment of SPERT-I and SPERTII for light water and heavy water systems of plate fuel type(Woodruff, 1989). In addition, PARET/ANL code has also been com-pared with RELAP5/MOD3 on 10 MW IAEA research reactor (IAEA,1992) for four transient cases, i.e., fast Loss of Flow (LOF) transient,slow LOF transient, slow reactivity insertion transient and fastreactivity insertion transient (Woodruff et al., 1996). Overall goodagreements between PARET/ANL and RELAP5 code have beenconfirmed.

The safety analysis is intended to ensure that abrupt leakage orbeam tube break leading to fast positive reactivity insertion main-tains the fuel integrity in good condition. The safety parametersshould not exceed the their limits, such as 200 and 145 �C for fueland clad temperatures, respectively. In addition, the flow instabil-ity parameter should be higher than 1.48 for transient conditions.

2. MCNP benchmark

In order to give confidence on the accuracy of MCNP code, it isrecommended to validate the code with experimental data. Theexperimental data of the first criticality of the MPR GAS is usedto validate the code with nuclear data ENDF/B-VI. The first critical-ity was composed of 9 fresh fuel elements (FEs) and 6 control ele-ment (CEs). On the other hand, the first core was composed of 12fresh FEs and 6 CEs. The meat material of FE and CE were madeof uranium oxide in aluminum dispersion (U3O8-Al). The uraniumoxide fuel has uranium density in the meat of 2.96 g/cm3 which isenriched by 19.75%. One standard fuel element contains 250 g U235

and 1264 g U.The first criticality was attained where 5 control elements were

fully up while the remaining one control element located in C-8was at 475 mm (Liem, 1999). Fig. 1 shows the core configurationof first criticality. MCNP5 calculations were also compared withdata of first core. The geometrical domain covers the core and allwater within the reactor pool.

The fuel element consists of 21 fuel plates cladded with AlMg2.The fuel plate has 60 cm of active length. The total plate length is62.5 cm. The geometrical model of fuel element also includesupper plenum, handling, lower plenum and lower end fitting.The control element has 15 fuel plates having the same dimensionand material as those of fuel element. Both side rows of control

Fig. 1. Core configuration model of the first criticality.

Please cite this article in press as: Khakim, A. Safety analysis of MTR type researAnn. Nucl. Energy (2014), http://dx.doi.org/10.1016/j.anucene.2014.08.018

element are inserted with fork-type absorber made of AgInCdcladded with SS-321. The absorber blades can freely move upand down to control the neutron population (Liem et al., 1998).

Table 1 presents validation results of MCNP–ENDF/B-VI withthe experimental data of the first criticality. Compared to the firstcriticality, the MCNP5 code agrees very well with difference of0.23%. Regarding excess reactivity of the first core, MCNP5 codepredicts accurately with relative difference of 0.43%. The resultsof MCNP5 code a bit overestimate compared to experimental data.

3. Parametric study

To calculate the reactor criticality with MCNP, KCODE card hasto be defined. It contains information on the number of particles tobe simulated, initial guess of keff, number of cycles to be skippedbefore calculating the accumulative keff (this is important espe-cially when the initial guess of keff is not good), and total numberof cycles to be calculated. The number of particles to be simulatedis adjusted to the complexity of core system. There at least is oneparticle to be simulated within the fissile material. The more par-ticles to be simulated the smaller the standard deviation. TheKCODE card forms:

KCODE nsrck rkk ikz kct

where nsrck: number of particles for each cycle; rkk: initial guess forkeff; ikz: number of cycles to be skipped before calculating keff; kct:total number of cycles for calculation.

Taking the appropriate input data for KCODE card is critical tothe accuracy of keff calculations. Therefore, parametric studies wereperformed to determine the appropriate values of nsrck and kct forthe system of interest. Low values of nsrck and kct will result ininaccurate calculated keff. In contrast, too high on nsrck and kct leadto very time-consuming calculations. The optimum numbers are tobe indentified to obtain accurate yet reasonably quick calculations.

Fig. 2 shows the effective multiplication factor keff for variousnumbers of simulated particles (nsrck). The keff values fluctuatefor nsrck lower than 200,000, and eventually they reach stable val-ues for higher than 200,000. Therefore, for this core system300,000 is taken for the next calculations.

As MCNP5 applies Monte Carlo method, the output of the calcu-lations is written along with their standard deviations. Standarddeviation indicates the result’s statistical relative error. Fig. 3addresses the standard deviation against the number of simulatedparticles (nsrck). Standard deviation can also be additionally con-sidered when deciding the number of simulated particles. As canbe seen, the standard deviation at nsrck = 300,000 is 8.0 � 10�5

which is very low relative to the keff. The increase in the nsrck from300,000 to 400,000 does not significantly reduce standard devia-tion, in contrary it even increases the CPU time significantly from37.89 to 77.70 h for kct = 300.

Fig. 4 describes the keff values over various numbers of totalsimulated cycles (kct) with MCNP5 for the core system. Fluctuation

Table 1Validation results of MCNP5 with first core.

Core configuration Experimentdata

MCNP & ENDF/B-VI

First criticality (9 FEs; 6 CEs) Keff 1.0 1.00238 ± 0.002C/E

1.00238

Full core (12 FEs; 6 CEs all up) Keff 1.09242 1.09714 ± 0.0002C/E

1.001

Full core (CEs all down) Keff – 0.91875 ± 0.0013

C/E: comparison of computation over experiment.

ch reactor during postulated beam tube break inducing positive reactivity.


Fig. 2. keff values at various number of simulated particles (nsrck).

Fig. 3. Standard deviation versus the number of simulated particles (nsrck).

Fig. 4. The keff value versus total number of cycles to be simulated (kct).

Fig. 5. Standard deviation versus total number of cycles (kct).

A. Khakim / Annals of Nuclear Energy xxx (2014) xxx–xxx 3

on keff value takes place for kct less than 300, and stable values areachieved for kct higher or equal to 300. Therefore, hereafter 300 istaken for total number of cycles to be simulated (kct).

Standard deviation decreases as kct increases, however it leadsto significant increase in CPU time for the calculations. Forkct = 300, the standard deviation is considered to be low enough(see Fig. 5). Thus, it is reasonable to take kct = 300 for thecalculations.

Hereafter, the KCODE card of the MCNP5 input will form:

KCODE 300;000 1:0 75 300

4. Calculation models

The neutronic calculation with MCNP5 is to identify the maxi-mum reactivity induced from beam tube break. The equilibriumcore of MPR GAS is modeled in 3-dimensional geometry with thereactor tank as outer most boundary.

Fig. 6 shows MCNP5 geometrical model of typical working corewith tangential (A) and radial (B) beam tubes, and fuel element (C)and control element (D). Tangential beam tubes are located 10 cm


above core mid plane, namely S2 and S6 forming 90� and 270�counter clock wise, respectively. The radial beam tubes S1, S3, S4and S5 of Fig. 6(B) in clock wise direction are located 10 cm belowthe midplane. All tubes have 16 cm of inner diameter, except S5 of19.5 cm. On the concrete side, the tubes are equipped with a plugwhich functions to prevent the water loss due to beam tube leak-age. In addition, they have valves which function to fill water whenidle and drain it when used for irradiation. All 6 tubes penetratetwo outer layers of beryllium block reflectors, but not the mostinner layer. The induction of positive reactivity is assumed to occurdue to fast beam tube leakage or break when they are used for irra-diation in such the water then occupies the tubes. The reactivityinsertion is derived from the difference of reactivity change fromvoided (filled with air) to flooded (filled with water) conditionsof particular tubes.

The core consists of 10 � 10 grids surrounded by berylliumreflectors. The equilibrium core is made up of 40 standard fuel ele-ments and 8 control elements. The fuel meat materials are made ofU3Si2Al. Two core sides are surrounded by beryllium blocks com-prised of three layers. The other two sides are surrounded by beryl-lium elements which are inserted into the core lattice (Liem andSembiring, 2010). Axially other than fuel plates, parts of the fuelelement modeled with MCNP5 are upper plenum together withfuel handling, lower plenum, and lower end fitting. The absorbermaterial of AgInCd is cladded with stainless steel. In the calcula-tions of the reactivity due to the tube break, all control elementsare in the position of fully withdrawn.

The reactor dynamics analyses due to reactivity insertion wereperformed with PARET/ANL code. The reactor power behavior inthe code is governed by point reactor kinetic equations:

dUðtÞdt

¼ qðtÞ � b½ �K

UðtÞ þXI

i¼1

kiCiðtÞ þ SðtÞ ð1Þ

dCiðtÞdt

¼ b f i

KUðtÞ � kiCiðtÞ ; i ¼ 1;2; . . . ; I ð2Þ

where t: time; U: reactor power; q: reactivity of the system; b:effective delayed neutron fraction; K: prompt neutron generationtime; ki: decay constant for group i; Ci: concentration of delayedneutron precursors of group i; fi: fraction of delayed neutrons ofgroup i, bi/b.

Reactivity feedback is calculated as the sum of that feedbackthrough the mechanism of fuel rod expansion, moderator densityeffect and fuel temperature effects (Doppler effects). The totalcompensated reactivity qc is:

qc ¼ qRod þ qMD þ qDop ð3Þ

The reactivity term used in reactor kinetic equation is:

q ¼ qin � qc ð4Þ



Fig. 6. Geometrical model of core with beam tubes, fuel element and control element.

Table 2Input parameters for reactor dynamics calculation.

Parameter Value

Core mass flux, kg/m2s 3761Coolant inlet temperature, �C 44.5Radial power peaking factor 2.6Axial power peaking factor 1.6Engineering hot spot factor 1.2Engineering hot channel factor 1.215Delayed neutron fraction (beff) 0.00719Over power trip signal, %, MW 114% (34.2 MW)Delay time, s 0.5Meat material U3Si2Al


where qin is external reactivity (e.g. control rods). In this calcula-tion, the reactivity from beam tube break is inserted throughexternal reactivity (qin) by the user in the input file.

The partial differential equation for diffusion of heat withineach fuel element is

@

@tgðu; rÞuðr; tÞ½ � ¼ r:kðu; rÞruðr; tÞ þ Sðr; tÞ ð5Þ

where the symbol u(r, t) represents temperature as a function ofradial position r and time t. Volumetric heat capacity and thermalconductivity are denoted by symbols g(u,r) and k(u,r), respectively,and are treated as functions of both position and temperature. Theterm S(r, t) is the heat source per unit volume.

The hydrodynamics calculations are governed by three equa-tions representing the laws of conservation of mass, momentumand energy.

@q@t¼ � @G

@zð6Þ

@G@tþ @

@zG2

q0

!¼ � @p

@z

� �� fv jGjG

2De� qg ð7Þ

q00@H@t

� �þ G

@H@z

� �¼ q

rhð8Þ

where: z = axial spatial variable; q = volume weighted two-phasedensity of coolant; G = mass flow rate of coolant; q0 = effectivedensity of coolant for momentum considerations; p, f = pressure,friction factor; v = specific volume of coolant; De = equivalent diam-eter of the coolant channel; g = gravitational constant; q0 0 = effective


slip flow density of coolant; H = enthalpy of coolant; rh = hydraulicradius of coolant channel; q = thermal energy gained by coolant.

The reactor was assumed to have been operating with initialpower of 1 and 30 MW, when the beam tubes abruptly break.The control elements were assumed unable to compensate a fastreactivity change; consequently they could not automaticallyprevent the power change. By design, the maximum controllablereactivity insertion rate is 2.2 � 10�4%dk/k.s. The calculations wereperformed for duration of insertion time in such that the reactivityinsertion rate is higher than 2.2 � 10�4%dk/k.s.

The reactor core was divided into two cooling channels repre-senting average and hot channels. Hot channel is a conservativelyassumed channel among all channels in the core that leads toextreme condition. The channel is the product of average channelwith multiplication factor stemming from axial and radial powerpeaking factor, engineering hot spot factor and engineering hot



Table 3Reactivity contribution by tube leak.

Beam tube Reactivity insertion (%dk/k)

S1 (radial) +0.034S2 (tangential) +0.010S3 (radial) +0.032S4 (radial) +0.025S5 (radial) +0.062S6 (tangential) +0.009All tubes +0.157

A. Khakim / Annals of Nuclear Energy xxx (2014) xxx–xxx 5

channel factor. In the radial direction, fuel plate is divided into7 nodes, while axially the cooling channel and fuel plate aredivided into 21 nodes. Power distribution in the axial direction isassumed to be cosine, where the ratio between the peak valueand the average is equal to the axial power peaking factor. Theother input parameters are listed in Table 2.

Fig. 7. Time history of power and fuel temperature for initial power of 1 MW.

Fig. 8. Time history of power and fuel temperature for initial power of 30 MW.

5. Results and discussions

When used for irradiation, a number of neutrons go out of thecore through the tubes and collide with the irradiation samples.When the tubes suddenly filled with water due to the tube rupture,the neutrons are reflected back to the core, leading to higher neu-tron population in the core. As the water also acts as moderator,more neutrons in the beam tubes will be shifted to the lowerregion. Both the increase in the neutron population and thedecrease in neutron leakage add positive reactivity into the core.

In general, radial beam tubes give higher reactivity feedbackthan that of tangential ones. The neutron emission from the centerof the core is isotropic. The radial oriented beam tubes providesentrance surface which is perpendicular to the neutron emission.Thus, the neutron current flying through the tubes is higher. Con-sequently, the radial beam tubes provide higher neutron currentfor irradiation purposes than that of tangential ones. This leadsto higher reactivity feedback induced by radial tubes than thetangential ones.

Among the radial tubes, S5 gives the highest feedback, i.e.,0.062%dk/k, as shown in Table 3, due to having bigger diameterand perpendicular orientation toward the core. Overall beam tubesprovide reactivity feedback by 0.157%dk/k, which means the leak-age occurs simultaneously in all tubes, for instance due to an earthquake. The tube giving the highest reactivity feedback (i.e., S5) isassumed to undergo fast leakage leading to reactivity insertion of0.08623 $ (0.00062:0.00719). Three cases are simulated for inser-tion duration of 1, 2 and 3 s. A combination of five beam tube breaksimultaneously at the same time is ignored due to its low probabil-ity of occurence.

In each case, the transients start at 5 s. For the initial power of1 MW (Fig. 7), the increase in power does not bring reactor toscram, as neither period nor floating limit value trip signal isachieved. The reactor reaches the maximum power of 1.104 MWat insertion rate of 0.08623 $/s. The reactivity insertion ends whenthe whole volume of the tube is flooded with water, and eventuallythe power decreases. The maximum fuel, clad and coolant temper-atures are 51.16, 49.89 and 46.17 �C, respectively, at the peakpower. However, these temperatures are far below the maximumacceptable limit of 200 and 145 �C for fuel and clad, respectively.In addition, the flow stability parameter (S) of 45.0 is much higherthan the minimum acceptable limit of 1.48.

As for the initial power of 30 MW (see Fig. 8), the reactor doesnot scram either, as neither the floating limit value nor over powertrip signal is reached. The reactor reaches the maximum power of32.74 MW, well below the over power trip signal of 34.2 which is114% of nominal power. The power eventually decreases as reac-tivity insertion terminates. The maximum fuel, clad and coolant


temperatures stand at 175.28, 136.75 and 94.17 �C, respectively.Even with the maximum power of 32.74 MW the maximum tem-perature limits are not exceeded. On the other hand, the minimumflow stability parameter (S) of 2.79 is well above the minimumacceptable limit 1.48.

6. Conclusions

The leakage in radial beam tubes gives higher positive reactivityfeedback than tangential tubes. Among the radial tubes, S5 givesthe highest reactivity feedback of 0.062%dk/k. Therefore, S5 is used




for postulated tube break to be evaluated. Both with the initialpower of 1 and 30 MW, no trip signal is initiated. As for initialpower of 30 MW, the maximum fuel and clad temperature reach175.28, and 136.75 �C, respectively. The minimum flow stabilityparameter reaches 2.79. All safety parameters are under acceptablelimits. Thus, the reactor can be maintained secured duringreactivity initiated accident due to fast break of beam tube S5.

Acknowledgement

The author would like to express gratitude to Nuclear EnergyRegulatory Agency of Indonesia (BAPETEN) for supporting andfunding the research during fiscal year of 2012.

References

Hedayat, A., Davilu, H., Jafari, J., 2007. Loss of coolant accident analyses on Tehranresearch reactor by RELAP5/Mod3.2 code. Prog. Nucl. Energy 49, 511–528.


IAEA, 1992. Research Reactor Core Conversion Guidebook Volume 3: AnalyticalVerification. IAEA-TECDOC-643.

Khakim Azizul, 2012. Perhitungan Kritikalitas Gudang Uranium Fasilitas InstalasiPabrikasi Elemen Bakar Reaktor Riset (IPEBRR). In: Proceedings SeminarNasional ke-18 Teknologi dan Keselamatan PLTN serta Fasilitas Nuklir,UPI-Bandung.

Liem, P.H., 1999. Validation of Batan’s standard 3-D diffusion code, Batan-3Diff, onthe first core of RSG GAS. Atom Indonesia 25, 2.

Liem, P.H., Sembiring, T.M., 2010. Design of transition cores of RSG GAS (MPR-30)with higher loading silicide fuel. Nucl. Eng. Des. 240, 1433–1442.

Liem, P.H., Bakrie, A., Sembiring, T.M., Prayoto, R., Nabbi, 1998. Fuel managementstrategy for the new equilibrium silicide core design of RSG GAS (MPR-30).Nucl. Eng. Des. 180, 207–219.

Woodruff, W.L., 1989. Additional capabilities and benchmarking with the SPERTtransient for heavy water applications of the PARET code. In: Proceedings of1989 International Meeting on Reduced Enrichment for Research and TestReactors, Berlin, Germany, pp. 357–365.

Woodruff, W.L., Hanan, N.A., Smith R.S., Matos, J.E., 1996. Comparison of the PARET/ANL and RELAP5/MOD3 code for the analysis of IAEA benchmark transients. In:International Meeting on Reduced Enrichment for Research and Test Reactors,Seoul.

X-5 Monte Carlo Team, 2003. MCNP – A General Monte Carlo N-Particle TransportCode. Version 5: User’s Guide.


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