Annals of Nuclear Energy 46 (2012) 29–36

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Annals of Nuclear Energy

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Effect of pebble packing on neutron spectrum and the isotopic compositionof HTGR fuel

Mehmet Türkmen, Üner Çolak ⇑Nuclear Engineering Department, Hacettepe University, Beytepe, Ankara 06800, Turkey

a r t i c l e i n f o

Article history:Received 29 November 2011Received in revised form 8 March 2012Accepted 14 March 2012Available online 10 April 2012

Keywords:HTGRSpherical HTGR fuelNeutron spectrumFission products

0306-4549/$ - see front matter � 2012 Elsevier Ltd. Ahttp://dx.doi.org/10.1016/j.anucene.2012.03.016

⇑ Corresponding author. Tel.: +90 312 2977300; faxE-mail address: ucolak@hacettepe.edu.tr (Ü. Çolak

a b s t r a c t

Fission products play an important role in the safety and fuel integrity of high-temperature gas-cooledreactor (HTGR) and they depend on temperature, burnup, neutron energy distribution, and fast fluence.Energy distribution of neutrons in a fuel region determines the isotopic distribution of the fission prod-ucts to be produced. The local concentrations of these isotopes are considered to be functions of temper-ature and burnup as well as the amount transported from the kernel to the coating layers where theyinteract and may degrade layers. Thus, the integrity of the fuel particle may be lost and fission productscan be released into the reactor coolant inventory. In this study, it is the main purpose to perform neutronenergy spectrum shift in spherical HTGR fuels and to investigate its effect on fission products. Moreover,it is also intended to analyze the effect of unit cell geometries on criticality of the system.

The calculations for group fluxes based on ENDF4 library with 27 neutron energy groups are accom-plished by the MCNP5 neutron transport code. Burnup and criticality analyses are performed by usingthe MONTEBURNS2 code (MCNP5 coupled with ORIGENS). To simplify the neutron transport problem,instead of full core modeling, two fundamental unit cell arrangements, body-centered cubic (BCC) andhexagonal close-packed (HCP) lattices, are considered to be as reference geometry models. Unit cellsare defined with proper boundary conditions and random packing for TRISO particles provided by thestochastic geometry card specified in MCNP5 for HTGR pebbles is used.

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1. Introduction

Nuclear reaction rates in a reactor are affected by the energydistribution of neutrons. Normally, the coexistence of fuel andmoderator materials and their volumetric fractions play an impor-tant role for the determination of fuel’s isotopic composition. Uponfission reactions, fissile isotopes are consumed and a wide range offission products are generated. Furthermore, a number of actinideisotopes are produced as a result of sequential capture and decayevents.

Isotopic composition of fuel provides an important insightespecially for reprocessing activities. Furthermore, this informa-tion may be useful for the fuel performance considerations. Forinstance, isotopes of Cs and I as fission products have criticalimportance for stress corrosion cracking in metallic cladding appli-cations in LWRs.

TRISO coated particle fuel provides a safe and reliable alterna-tive for high-temperature gas-cooled reactors (HTGRs). This willbe a viable option especially for high temperature process heatapplications and hydrogen generation. Primary pressure boundary,

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SiC layer, may have corrosive interaction with metallic fissionproducts such as Pd which results in undesired thinning. Similarly,silver diffused through coating layers will be transported in thecoolant circuit and deposited at cool sections. This is an importantsafety concern for such reactors. The composition of uranium andactinide isotopes could be valuable information for transmutationand reprocessing purposes.

Neutron energy spectrum is affected by various parametersincluding fuel-to-moderator (F/M) ratio. This study deals withthe analysis of the effect of F/M on neutron energy spectrum andchange in the isotopic composition of fuel at the discharge burnup.Spherical fuel and moderator elements with different fractions andgeometric configurations were taken into account for the analysis.It is intended in this study to demonstrate any possible impact ofarrangement of spheres as well as F/M ratio on the neutron spec-trum shift, and in turn, influence on the isotopic composition offuel. Criticality and depletion calculations were carried out bythe computer code MONTEBURNS2.

MONTEBURNS2 code (Poston and Trellue, 1999) which linksMCNP5, Monte Carlo transport code, with ORIGEN2, depletion,burnup, and radioactive decay code, is used for the calculation ofcore inventory and of burnup as a function of criticality. In a spec-ified geometry, power, and time frame, this code gives a great dealof information about burnup, effective multiplication factor,

Table 1Core characteristics.

Parameter Value

Fuel kernel diameter 500 lmParticle material type UO2

UO2 density 10.4 g/cm3

Coating material PyC/PyC/SiC/PyCDensity of materials in the layer 1.05/1.9/3.2/1.9Average pebble-bed packing fraction BCC: 0.61 HCP: 0.74Fuel enrichment 9.6% UO2 equilibrium coreFuel pebble outer radius 3.0 cmThickness of fuel free zone 0.5 cmGraphite matrix and fuel free zone density 1.74 g/cm3

Total HM loading per fuel pebble 9 g (equilibrium)Calculated side length of cubical lattice 0.1635 cmCalculated lattice cell pitch BCC: 7.18 cm HCP: 7.4641 cmCoated particle radius 0.46 mmPacking fraction of coated particles 0.093 (Kim et al., 2005)Number of passes 6 (or 10)Fuel residence time 933 daysBurnup 95.0 MWD/kgUF/M 1:1Average helium temperature 771 �CAverage moderator temperature 817 �CAverage fuel temperature 830 �C

30 M. Türkmen, Ü. Çolak / Annals of Nuclear Energy 46 (2012) 29–36

one-group cross-section sets, material composition, activity, heatload, and toxicity for the relevant isotopes.

The MONTEBURNS2 code has been used in many studies withsufficient accuracy for the modeling analysis of HTGR cores. A com-parative analysis for actinide inventory in a spent nuclear fuel for apebble-bed HTR using three different MCNP-based depletion codes(Monteburns2, MCNPX2.6.0, and BGCore) was reported by Bom-boni et al. (2010).

For a specific design of VHTR (Very High Temperature Reactor),Tsvetkov et al. (2008) worked on spectrum shifting to improve per-formance of the advanced actinide fuels and focused on TRU(Transuranic)-impact on a single-batch mode.

First core loading criticality calculations for a pebble bed gascooled reactor, HTR-10, was reported by S�eker and Çolak (2003)using MCNP. The study showed that criticality analysis resultswere in good agreement with experimental results and success-fully performed for this type of reactors using Monte Carlo simula-tions. Jeong and Chang (2008) performed an analysis to calculatefission product inventory for HTR-10 reactor. They also usedMONTEBURNS in their analysis. However, they only consideredfixed geometry represented by a body centered cubic (BCC) latticeand an F/M ratio of 57/43.

2. Geometry modeling and characteristics

The coated fuel particles with diameter of about 0.5 mm aremade of a spherical uranium dioxide fuel kernel surrounded bythe layers of pyrolytic carbon and silicon carbide. A pebble withabout 15,000 coated fuel particles (CFP) contains approximately9 g enriched UO2. A CFP includes, in turn, a UO2 kernel of0.025 cm radius, a buffer layer of 0.009 cm thick with PyC (pyro-lytic carbon), an inner layer of 0.004 cm thick with PyC, a SiC (sil-icon carbide) layer of 0.0035 cm thick and an outer layer of0.004 cm thick with PyC.

In this study, the stochastic geometry modeling is used in peb-ble bed type of high-temperature gas-cooled reactors. Spatial dis-tribution of spherical fuels in HTGR changes with the axial andradial position. Although the placement of spheres is in a randommanner with a packing fraction of 0.61 (Terry and Ougouag,2003; Auwerda et al., 2010), it turns from a BCC structure to HCPstructure towards the bottom of the core (Terry, 2001). Randomlylocated fuel kernels in pebbles (Brown et al., 2005) are consideredwhen the eigenvalue problems are solved in BCC and HCP unitcells. The models used are adjusted to preserve the quantity of fuelmaterial and packing fraction. Furthermore, the number of fuel andmoderator pebbles which are randomly packed throughout thecore is assigned such that the fuel-to-moderator ratio is to be 1/1in the specified unit cells to define the reference case. In a referenceBCC unit cell, the length of lattice pitch is calculated to be 7.18 cm.In this lattice, cubical lattice pitch which covers the TRISO particleis found to be 0.1635 mm, as mentioned in previous study by Kimet al. (2007). In the reference HCP unit cell, the length of latticepitch is computed as 7.464 cm with an axial ratio c/a ratio of1.633/1. Atomic packing factors for BCC and HCP are, in turn,0.61 and 0.74. Some important core characteristics of the HTGRare listed in Table 1. A visual representation of geometric modelfor indicated unit cells is shown in Fig. 1.

3. MCNP modeling

Neutron energy spectrum is divided into 27 group neutron ener-gies based on ENDF/B-IV library. Group energies are listed in Table2. Energy dependent flux in fuel kernel is obtained by using MCNP5(X-5 Monte Carlo Team, 2003) F4 tally flux averaged over a cell forneutrons. The neutron data library is primarily based on data from

ENDF/B-VI.8, although some data come from other sources (ENDF/B-VI.0, LLNL, etc.). An infinite square array of these fuel pins is thenmodeled by surrounding these annuli with helium and using a peri-odic boundary condition in parallel planes and a white boundarycondition in z direction. S (a,b) treatment data for graphite at1200 K is used to model thermal scattering at low energies. In themain study, the .12c cross section set is used in the evaluation ofreaction rates of fuel at the temperature of 1200 K. The .12c and.09c cross section sets are used in evaluation of reaction rates ofgraphite and helium at about 1200 and 900 K, respectively.

Monte Carlo calculations are performed by using 5000 neutronsper stage for 50 passive cycles and 250 active cycle run about1250,000 sampled neutrons per calculation. The continuous energythree-dimensional Monte Carlo code MCNP5 is used for calculations.

All calculations are for equilibrium condition and enricheduranium fuel contains 9.6% U-235 in U. The weight percents ofisotopes for fuel, clad and graphite matrix are given in Table 3.

ORIGEN2 used for the fuel depletion calculation lacks the neu-tron cross-section sets for HTGRs, and also, neutron cross-sectionsets are mostly burnup dependent. Thus, in this work, they are gen-erated for two unit cells by means of the Monte-Carlo method.

The reactivity values using linear (LRM) and non-linear reactiv-ity model (NLRM) (Driscoll et al., 1990) are obtained as a functionof burnup. The selection of the model depends on the characteris-tics of burnup variation as a function of time during the operation.In general, the reactivity of fuel is evaluated in this analysis using asecond-order polynomial fit with sufficient accuracy.

Here, fuel enriched to 9.6% U-235 gains an average dischargeburnup of 95.0 MWd/kgU (suggested by Boer et al., 2009) for twoselected reference unit cells. The number of pass of pebblesthroughout core life is six in a multi-pass fuel cycle scheme andcore leakage reactivity in all unit cells is found as 15.9%. Mulderand Teuchert (2008) indicate this value as 15.6%. In this respect,results are well-consistent with other studies. As a result, it isassumed that the value of leakage remains unchanged and thisvalue is used in later calculation steps.

The full-core model based on BCC and HCP unit cell approaches isachieved for current HTGR core design. Fuels at various ratios pro-vide criticality for the overall core reactivity of 0.1590 (keff = 1.1891)and zero-system reactivity under the equilibrium condition.

The average power produced from one pebble is set to about9.16 kW. Number of pass for a fuel pebble is taken as six with equal

Fig. 1. MCNP model for BCC and HCP unit cells.

Table 2Neutron energies used in spectrum calculations.

Group # Energy (MeV) Group # Energy (MeV)

1 1E�11 15 1E�52 1E�8 16 3E�53 3E�8 17 1E�44 5E�8 18 5.5E�45 1E�7 19 3E�36 2.25E�7 20 1.7E�27 3.25E�7 21 0.18 4E�7 22 0.49 8E�7 23 0.910 1E�6 24 1.411 1.13E�6 25 1.8512 1.30E�6 26 313 1.77E�6 27 6.43414 3.05E�6 28 20Resonance 1 eV–0.01 MeVThermal <0.025 eVContinuum 0.01 MeV–25 MeVFast <1 eV

Table 3Isotopic weight percent.

Isotope Weight percent (%)

Fuel region (9.6% enriched)U-234 0.06705U-235 8.46125U-238 79.6098O-16 11.8619

PyC and graphite matrixC 100SiCSi 70.0C 30.0

M. Türkmen, Ü. Çolak / Annals of Nuclear Energy 46 (2012) 29–36 31

pass history of 155.5 days. It is also assumed that the fuel burns ata constant power during irradiation. For the fuel material, thenumber of automatic tally isotopes to be obtained for the spentfuel content is set to 81. In addition, the discharged fuel at theEnd of Life (EOL) are then cooled 150 days.

4. Results

4.1. Neutron spectrum shift

Due to the energy dependency of cross-sections in neutron–nucleus reactions, it is necessary to find out the energy distributionof neutrons in order to determine the rate of reactions of neutronswith matter. The investigation is started with the assessment ofneutron energy distribution in BCC and HCP unit cells in fuelkernels.

According to Fig. 2, there is no significant change in cell-averageflux at very high energies due to the fuel arrangement. This is dueto direct fission neutrons being the principal source at this energylevel.

The neutron spectrum shift effect is more influential at reso-nance and low energy values of continuum region as the moremoderator pebbles are added to the system. Furthermore, thermalneutron flux distribution up to 0.025 eV decreases gradually as themoderator pebbles decrease.

This leads to scale up the level in fast neutron flux distributionand compensated by the less neutrons which slow down to thethermal energy region. Furthermore, the probability of neutroncapture in resonance decreases due to enhancement in lethargyof neutrons with moderator insertion by slowing down moreneutrons to below the resonance zone. That is, the neutronsscattered from energies above the resonance to energies belowthe resonance increase. Such an event causes an increment ineffective multiplication factor to be described in following section.This is, also, related with neutron reaction rates since the fuelcomposition made of actinides, FPs, and light elements stronglydepends on neutron flux spectrum.

Fig. 2. Weighted neutron flux spectrum for various F/M ratios in fuel kernel (a) BCC unit cell and (b) HCP unit cell.

Table 5Discharge burnup as a function of F/M ratio for various unit cells.

HCP unit cell BCC unit cell

F/M ratio Bd (MWd/kgU) F/M ratio Bd (MWd/kgU)

1/5 77.280 1/7 68.2301/3 85.220 1/3 86.3502/4 91.840 3/5 93.2203/3 95.000 1/1 95.0004/2 88.130 5/3 90.7703/1 81.730 3/1 79.8605/1 70.670 7/1 64.470

Fig. 3. For BCC, variation of system reactivity with burnup during irradiation.

Table 4Parameters for various F/M ratios in BCC unit cell.

F/Mratio

Bd (MWd/kgU)

Reactivity(BOL)

Reactivity(EOL)

EFPD(days)

1/7 68.230 0.205882 �0.23233 168.001/3 86.350 0.233275 �0.29530 424.403/5 93.220 0.229521 �0.25595 691.801/1 95.000 0.215699 �0.19955 933.005/3 90.770 0.197147 �0.14396 1119.603/1 79.860 0.178077 �0.10492 1176.607/1 64.470 0.157702 �0.07901 1107.00

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4.1.1. Effect on core design parametersThe net effect of spectrum change is to increase in effective

multiplication factor (keff) as the number of moderator pebblesincrease in the unit cell. The change of multiplication factors withburnup is represented in terms of reactivities in Fig. 3. Effectivemultiplication factor (illustrated as reactivity in the third andfourth column of Table 4) hits a peak value of 1.31975 (equals toa positive reactivity of 0.24228) in the neighborhood of F/M ratioof 1/3 at the Beginning of Life (BOL) and ends with 0.77742 (equalsto a negative reactivity of �0.28650) at EOL. Over-moderated con-figurations offer higher excess reactivity than the reference case;however, irradiation period shortens by a half value due to rapidconsumption of fissile material and fertile-to-fissile conversion fac-tor becomes smaller. In case of hard spectrum (under-moderation),fuel offers longer effective full power days (EFPDs) (as seen fromthe fifth column of Table 4), lower excess reactivity and higher con-version factor. Average burnup values at discharge for various con-figurations are listed in the second column of Table 5. For the peakvalue of keff, discharge burnup is about 86.3 MWd/kgU.

4.1.2. Effect on production/consumption of actinidesIn parallel with changing the spectrum described in a detailed

way in the previous section, isotope concentrations changesignificantly. This change originated from neutron reaction ratesis illustrated in Fig. 4. Peak F/M ratios of interested actinides arelisted in Table 6. From the table, it can be deduced that uraniumpenalty (U-236) in uranium reaches on its highest value at F/M

ratio of 3/5; but, uranium utilization be maximum. Also, most ofthe actinides reach their maximum values in between the F/M ratioof 5/3 and 2/1. In case of Pu-242, responsible for predominantly Pupenalty, this value is 1/1. At the same time, maximum fissile Puproduction and consumption occurs at 5/3 ratio. As a result, mostof the actinides reach peak values beyond the F/M ratio of 5/3; but,the selected spectrum from the aspect of uranium utilizationshould be in the F/M ratio of 3/5.

4.1.3. Effect on fission product productionConcentration of considered FPs as the spectrum changes is

given in Fig. 5. The fission products displayed in the figure aselemental (not isotopic) distribution are Xe, Cs, Nd, Tc, Ru, Mo,Sm, Pd, Rh, Pm, Eu, Ag, I, Kr. It is clear that they typically decreasewhen a softer or harder spectrum is used than the referencespectrum. The last column of Table 6 gives the peak F/M ratios

Fig. 4. For BCC, isotopic constituents (gram) of actinides vs. F/M ratio [fis: fissioned, cap: captured].

Table 6F/M ratio peak values for various actinides and FPs.

Isotope F/M:5/3 Isotope F/M:2/1 Isotope F/M:1/1 Isotope Peak values

Am-243 Peak Cm-242 Peak Pu-242 Peak Kr-85 3/5Cm-245 Peak Pu-238 Peak I-131 3/5 (max. slope)Cm-244 Peak Pu-240 Peak Sr-90 3/5Np-237 Peak Pd-107 5/3Pu-239 Max. slope Isotope F/M:3/5 Cs-137 1/1Pu-241 Max. slope U-236 Peak Ag-110m 2/1U-238 Min. U-235 Min.

Fig. 5. For BCC, elemental constituents (gram) of FPs vs. F/M ratio.

M. Türkmen, Ü. Çolak / Annals of Nuclear Energy 46 (2012) 29–36 33

for the specific isotopes of interested FP produced by various fis-sionable isotopes. The results demonstrate that not all FPs reachtheirs maximum at the same F/M ratio, they stretch out in a widerange spectrum. However, there are some proper spectra for someselected isotopes of FPs so that less interaction rate in the claddingmaterial and less diffusion from the fuel to the clad region wouldoccur due to FPs.

Release process mechanisms and fission yields of interested FPisotopes are described in following paragraphs and listed in Table 7.

Related with Fig. 4 and Table 7, following outcomes can bededucted:

1. Cs-137 with a half-life of 30 y is produced mainly from thedirect fast fissioning of U-238 and fissioning (with 0.0253 eVand 1.0 MeV neutrons) of Pu-239; thus, the production ofCs-137 isotope might be reduced by using soft spectra. Such aconsequence shown in Fig. 6 leads to a decrease in the produc-tion of fissile Pu-239 isotopes, and hence, generation of Cs-137.All fission process with any kind of actinides yields the samefission product yield with 6.2% for this isotope.

2. Similar to Cs-137, the greatest part of produced Ag-109 (stable)(or Ag-110m (250 days)) isotopes with a yield of 1.9% comesfrom fast/thermal fissioning of Pu-239, but; U-235 fission con-tribution is responsible for the smallest part with less than0.1%. In this respect, using soft spectrum is enough to reducethe production of Ag-109.

3. Sr-90 (28.8 y) and Kr-85 (10.76 y) isotopes are generated by fastand thermal fissioning of fissile U-235 and Pu-239. Also, asomewhat contribution to the production of these isotopes isdue to fast fissioning of fertile U-238. For thermal neutronfission of Sr-90, the fission product yield from U-235 is 5.9%,but from Pu-239 only 2.1%. For fast neutron fission, the fissionproduct yield is 5.43% from U-235 and 3.24% from U-238, but,from Pu-239 only 2.0%. For thermal neutron fission of Kr-85,

Table 7Fission yield (% per fission) for various isotopes.

Fissile with Neutron Energy Cs-137 (30.07 y) Ag-109 (stable) Pd-107 (6.5E6 y) Kr-85 (10.78 y) Sr-90 (28.78 y) I-131 (8.02 days)

U-235 (thermal) 6.268 0.034 0.140 0.274 5.904 2.884U-235 (fast) 6.203 0.115 0.327 0.269 5.434 3.180U-238 (fast) 6.091 0.268 1.296 0.140 3.243 3.239Pu-239 (thermal) 6.727 1.879 3.361 0.126 2.101 3.847Pu-239 (fast) 6.505 1.919 3.053 0.128 2.035 3.870

Fig. 6. For BCC and HCP, constituents (gram) of considered FP isotopes vs. F/M ratio.

Table 8Critic enrichment level at 95.0 MWd/kgU in BCC unit cell.

F/Mratio

Critic enrichment(%)

Reactivity(BOL)

Reactivity(EOL)

EFPD(days)

1/7 12.08 0.23184 �0.30978 2341/3 10.41 0.23935 �0.29620 4673/5 9.76 0.22968 �0.24842 7001/1 9.60 0.21570 �0.19955 9335/3 9.94 0.19980 �0.14893 11653/1 11.15 0.18464 �0.10937 14007/1 13.15 0.17199 �0.09290 1634

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the fission product yield from U-235 is 0.27%, but from Pu-239only 0.13%. For fast neutron fission, the fission product yield is0.27% from U-235 and 0.14% from U-238, but, from Pu-239 only0.13%. According to given fission yields and from the relatedresults shown in Fig. 6, generation rate of these isotopedecreases under the hard spectra.

4. Pd-107 isotope with a half-life of 6.5 million year yields higherin fast fission or in fission heavier nuclei. The main actinide per-taining to generation is Pu-239 with a yield of 3.36% for thermalneutron fission and 3.1% for fast neutron fission; however, lessthan 0.3% for other actinides. Thus, generation rate of fissile plu-tonium clearly describes the accumulation of this isotope.

5. As a crucial isotope, I-131 (8 days) is equally produced by fis-sion reaction of U-235, U-238, and Pu-239. The yield contribu-tion of these isotopes is respectively 2.88%, 0.0% and 3.85% inthermal fission, and, 3.12%, 3.24% and 3.87% in fast fission.Due to large decay constant and equal contributions from fis-sionable actinides, there is no way to take under control ofthe quantity of this isotope. However, the accumulation of thisisotope at high F/M values increases is mainly attributed toextending irradiation duration. Decay process becomes less sig-nificant in long irradiation period.

As a result, it is likely to reduce the quantity and the effect ofselected FP isotopes, mentioned above in the enumerated items,by using a proper spectrum. For instance, production rate of allof actinides, Ag-110m and Pd-107 isotopes goes down under thespectrum that corresponds to F/M ratio of 3/5.

4.2. Unit cell differences

Although unit cells have different geometrical arrangements,the obtained results show that there is no distinguishable differ-ence between the unit cells.

4.3. Criticality search

4.3.1. Enrichment adjustmentIn previous analyses, core configurations are searched for var-

ious fuel-to-moderator pebble ratios at a constant U-235 enrich-ment. These configurations reach different discharge burnupvalues according to their moderator pebble content and also, pro-vide different excess reactivity to the core in the course of theircore life-time. Thus, in this part of the work, critic uraniumenrichment level at the discharge burnup of 95 MWd/kgU atEOL is obtained so that configurations would provide same ex-cess reactivity in overall. The level of required critical enrichmentfor each core configuration, only in BCC unit cell, is given inTable 8. The results show that all of configurations accumulatethe same discharge burnup at EOL. However, the number ofburn-up days (presented in the last column) is, still, different.In other words, configurations undergo a shorter or longer num-ber of pass in a constant pebble movement speed. In the sametable, BOL and EOL reactivities are given to illustrate the reactiv-ity differences.

Fig. 7. For BCC, isotopic constituents (gram) of FPs and actinides vs. F/M ratio [fis: fissioned, cap: captured, con: consumed].

M. Türkmen, Ü. Çolak / Annals of Nuclear Energy 46 (2012) 29–36 35

4.3.2. Spent fuel content analysisIn this part, isotopic composition of the spent fuel that contains

considered actinides and FPs is analyzed for various core configu-rations. Discharge values are illustrated in Fig. 7.

From the figures, following remarks can be drawn:

1. The quantity of all Pu isotopes except Pu-242 (due to its neu-tron absorption cross-section at higher neutron energies) in apebble goes up steadily as F/M ratio increase.

2. The actinides are generated at low quantities when the highmoderator content-core designs are in use; but, low moderatorcontent-core designs account for higher content of actinides.The production rate is limited by the use of moderator contentso that their values remain constant, even the moderator con-tent is reduced.

3. Uranium utilization is at utmost at F/M ratio of 3/5 and the pen-alty originated from U-236 is minimized at this ratio.

4. An increase in moderator leads to an increase in the populationof Ag-110m and Pd-107, and leads to a decrease in the popula-tion of I-131 and Kr-85, however, no effect is observed inCs-137.

Pu contribution to the core power increases with increasing F/Mratio. At low ratios, main contribution to the core power comesfrom U-235 fission.

5. Conclusions

The main purpose of this study is to estimate how the accumu-lation of fission products changes under the certain neutron spec-trum at the end of irradiation. In this sense, this article is aimed tominimize the effect of fission products on the protective layers ofTRISO particles by employing various neutron energy spectrum.Additionally, criticality and actinide management analyses areinvestigated with changing fuel-to-moderator ratio to indicatewhether the configurations are proliferation resistant.

Concisely, the results indicate that it is not likely to select anoptimal neutron spectrum that would simultaneously minimizethe effect of all of fission products. Due to different peak valuesof Ag-110m, Cs-137, and Pd-107 isotopes, it is difficult to suggestan optimum neutron spectrum from the point of minimization ofhighly diffusive fission products. However, it is proposed that the

optimal fuel loading should be put into practice in 5/1 or upperratio from the safety aspect of the fuel. In that case, most of thehighly diffusive fission products accumulates in low quantitiescompared to their peak values so that the accumulation rate ofthese isotopes reduces up to 30% in the fuel region of micro-parti-cles. This also means that reaction rate with coating layer reducessignificantly. Such hard spectra will help to avoid earlier corrosionof coating layers, reduce certain isotopes to accumulate at cold sec-tions of reactor coolant circuit and prevent some other undesiredinteractions. On the other hand, fissile Pu production in fuel ker-nels becomes an important matter for the proliferation concern.However, current studies show that enhanced proliferation resis-tance of HTGR avoids the full separations of Pu by reprocessingtechnique.

To protect the SiC layer corrosion from Pd and Ag diffusion asmuch as possible, it is recommended that the optimum neutronspectra should be lower than that for the F/M ratio of 1/3. The corewould be loaded with fuel-to-moderator ratio of 3/5 from the ura-nium utilization point of view. In this case, Kr-85, Sr-90, and I-131isotopes accumulates in the fuel kernel and highly diffuse to thecoating layer. F/M ratio should be less than 1/1 for better actinidefuel management. Most of the actinides including Pu burns effi-ciently under soft spectrum. Such configurations would assist theadvanced fuel cycle studies on actinide management.

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