Journal of Nuclear Materials 429 (2012) 263–269

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Journal of Nuclear Materials

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Use of Th and U in CANDU-6 and ACR-700 on the once-through cycle: Burnupanalyses, natural U requirement/saving and nuclear resource utilization

Mehmet Türkmen ⇑, Okan H. ZabunogluDepartment of Nuclear Engineering, Hacettepe University, 06800 Beytepe, Ankara, Turkey

a r t i c l e i n f o a b s t r a c t

Article history:Received 6 April 2012Accepted 6 June 2012Available online 16 June 2012

0022-3115/$ - see front matter � 2012 Elsevier B.V. Ahttp://dx.doi.org/10.1016/j.jnucmat.2012.06.008

⇑ Corresponding author. Tel.: +90 312 297 73 00; faE-mail address: tm@hacettepe.edu.tr (M. Türkmen

Use of U and U–Th fuels in CANDU type of reactors (CANDU-6 and ACR-700) on the once-through nuclearfuel cycle is investigated. Based on the unit-cell approximation with the homogeneous-bundle/coremodel, utilizing the MONTEBURNS code, burnup computations are performed; discharge burnups aredetermined and expressed as functions of 235U and Th fractions, when applicable. Natural UraniumRequirement (and Saving) and Nuclear Resource Utilization are calculated for varying fuel compositions.Results are analyzed to observe the effects of 235U and Th fractions, thus to reach conclusions about use ofTh in CANDU-6 and ACR-700 on the once-through cycle.

� 2012 Elsevier B.V. All rights reserved.

1. Introduction

The only natural fissile isotope is 235U, which occurs as about0.7 weight percent (w/o) of natural U. 99.3 w/o of natural U is238U, which is not fissile but fertile. By neutron capture and twosubsequent beta decays, 238U in a nuclear reactor can be convertedto 239Pu. 239Pu is a fissile isotope not occurring in nature. Almost allnuclear fuels contain a combination of 235U and 238U, and thus in anuclear reactor, while 235U sustains fission chain reaction, 238U,being converted to 239Pu, contributes to energy production andextends the fuel residence time significantly.

Th in nature is about 100 w/o 232Th, which is not a fissile butfertile isotope. By neutron capture and two succeeding beta decays,232Th can be converted to 233U, which is a fissile isotope not exist-ing in nature. Since 232Th is not fissile, a nuclear reactor cannot bemade critical by using Th alone, just as it cannot be made criticalwith 238U alone. 232Th can replace 238U in current nuclear reactors,and subsequently the fissile 233U produced from 232Th partly re-places the fissile 239Pu produced from 238U. Such a replacementmay be expected to result in a long-term expansion of nuclear fuelresources, by noting that Th is more abundant than U (about 3 to 4times more in earth’s crust).

Although Th has a thermal absorption cross section 2.75 timesthat of 238U, it exhibits better thermal–hydraulic and chemicalproperties than U in nuclear fuel cycles. The thermal conductivityof ThO2 is about 50% higher than that of UO2 over a large temper-ature range, and its melting temperature is 340 �C higher than thatof UO2. As a consequence, all thermally activated processes, such ascreep and fission gas diffusion will be reduced. Fission-gas release

ll rights reserved.

x: +90 312 299 21 22.).

from ThO2 fabricated with proper control of microstructure will belower than that from UO2 operating under similar ratings. ThO2 ischemically very stable and does not oxidize easily, which offersadvantages for normal operation, postulated accidents and inwaste management [1]. Reactor-grade ThO2 can be blended withUO2 in required fractions to produce (U–Th)O2 fuels, which canbe irradiated in current/future nuclear reactors.

In addition to long-term extension of nuclear fuel resources, allthe advantages which may arise from better characteristics of Thhave also triggered efforts for use of Th in nuclear reactors so far,and will probably continue to do so more in future.

Pressurized Heavy Water Reactors (PHWR, also known as CAN-DU: CANadian Deuterium Uranium) have originally been designedto burn natural U; however, they can also use slightly enriched U.In any case, the fraction of 238U is the highest in PHWRs. That iswhy they attract the most attention when use of Th (to replace238U) is considered.

In this study, use of Th in CANDU type of reactors (CANDU-6and ACR-700) is to be investigated for the once-through fuel cycle.For different Th fractions in U fuel, with varying U enrichment val-ues, burnup computations are performed and discharge burnupsare determined for both CANDU-6 and ACR-700. Results are to becompared to observe the effect of Th fraction. Natural UraniumRequirement (NUR), Natural Uranium Saving (NUS) and NuclearResource Utilization (NRU) are also calculated and compared fordifferent fuel compositions in both reactors.

2. Reference cases (CANDU-6 and ACR-700) and the model

Short descriptions of reference CANDU-6 and ACR-700, withtheir fuels and burnup, are presented below; more comprehensivedata are given in Table A.1 in Appendix A.

264 M. Türkmen, O.H. Zabunoglu / Journal of Nuclear Materials 429 (2012) 263–269

2.1. CANDU-6

The CANDU-6 [2] is a 600-MWe nuclear power plant with a40-year design life at a plant capacity factor of 85%. It is a heavy-water-cooled, heavy-water-moderated pressure-tube reactor,designed to burn natural U (NU). The CANDU-6 fuel bundle has37 fuel elements. Each element carries NU in the form of cylindri-cal pellets of UO2 enclosed with a zircaloy-4 sheath. The CANDUfuel bundle is relatively small (0.5 m in length, 10 cm in diameter)and easy to handle (a2bout 20 kg).

For the reference CANDU-6, a 37-element bundle containing NUis burned to 7154 MWd/tonU on the once-through cycle [2].

2.2. ACR-700

The ACR-700 [2] is a 700-MWe nuclear power plant with a 60-year design life at a plant capacity factor of 90%. It is a light-water-cooled, heavy-water-moderated pressure-tube reactor, designed toburn slightly-enriched U (SEU). The ACR-700 reactor consists of284 fuel channels and has a total thermal power output of1982 MW. The fuel channels are in a compact array with a squarelattice pitch of 24 cm. Each fuel channel hosts 12 fuel bundles.CANFLEX-ACR (CANdu FLEXible) fuel bundles contain 43-fuelelements, each 49.53 cm in length, with uniform enrichment. Thebundle contains fuel elements of two different sizes. The centerand inner ring consist of 8 elements with a diameter of 13.5 mm,while the outer two rings contain 35 elements with a smallerdiameter 11.5 mm. The outer three rings of fuel elements containU pellets with 2.1 w/o 235U, while the central fuel element contains7.5 w/o Dysprosium (Dy, a burnable poison) homogeneouslyblended with NU, (U-Dy)O2.

For the reference ACR-700, a 43-element bundle containing 2.1w/o enriched U is irradiated to an average burnup of 20500 MWd/tonU on the once-through cycle [2].

2.3. The model

Three general approaches can be noted for burning (U–Th) fuelin PHWRs: ‘‘mixed-core’’, ‘‘mixed-bundle’’ and ‘‘homogeneous-bundle/core’’.

In the ‘‘mixed-core’’ approach, a large number of channelsloaded with ‘‘driver’’ fuel containing SEU only would provide theexternal source of neutrons for a fewer number of channels fueledwith ThO2. On-power refueling enables the ThO2 fuel to remain inthe core much longer than the driver fuel. The large disparitybetween the properties of the driver fuel and the ThO2 channelswould make the fuel management particularly challenging; itsapplicability requires more practice and experience [3].

The ‘‘mixed-bundle’’ approach, devised by AECL, provides apractical means to utilize Th in operating CANDU reactors. The‘‘mixed-fuel bundle’’ contains ThO2 in the inner 8 elements of aCANFLEX bundle, and SEU in the outer two rings (35 elements).Since ThO2 and the driver fuel are parts of the same bundle,separate residence times cannot be achieved.

In a ‘‘homogeneous-bundle/core’’ approach, a homogeneousmixture of (U–Th)O2 is used in all fuel elements of a bundle andbundles are uniformly distributed throughout the core. This isthe simplest model, therefore the most appropriate for computa-tional analyses.

Although mixed approaches are likely to produce better resultsin favor of Th, they pose several difficulties in computational stud-ies. The mixed-core case usually requires near-full core modeling,which is cumbersome anyway, and even more so for on-line refu-eling with different dwell times. The mixed-bundle model can behandled with more ease than the mixed-core; yet, results areexpected to be worse than those by the mixed-core from the

viewpoint of resource utilization [3]. The mixed-bundle approach,with varying fuel-element arrangements and compositions, islikely to be considered in a future study.

In this study, for burnup analyses of U–Th fuels, ‘‘the homoge-neous-bundle/core’’ model is used. Also note that there exist sev-eral results obtained for CANDU-6 using the homogeneousapproach in earlier studies, which provide a means for comparison[1,4–6].

3. Computational tools, procedure and related expressions

3.1. Codes used

The MCNP5 code, coupled to the ORIGEN by the MONTEBURNScode (shortly described in Appendix B), is used to model the fuelbundle. Computations are carried out for a single fuel bundle andthen results are adjusted to full core accordingly.

3.2. Calculation of reference leakage reactivity values

According to the Non-Linear Reactivity Model described in Dris-coll et al. [6], the reactivity as a function of burnup can beexpressed by:

qðBÞ ¼ q0 þ A1Bþ A2B2 þ A3B3 þ . . . . . . :þ AkBk ð1Þ

where q(B) is the reactivity at specific burnup B and q0, A1, . . ., Ak

are coefficients of the kth degree polynomial.In equilibrium condition for on-line refueling:

qs � qL ¼ qc ¼ 0) qL ¼1Bd

Z Bd

0qðBÞ dB ð2Þ

where qL is the leakage reactivity, qc is reactivity of the core, qs isreactivity of the system, and Bd is discharge burnup.

Using the MONTEBURNS code, reactivity data in terms of themultiplication factor and corresponding burnup values are ob-tained. By fitting the points on the reactivity-burnup curve, coeffi-cients of Eq. (1) are determined. Then, with the reference dischargeburnup values at hand for CANDU-6 and ACR-700 (Section 2),Eq. (2) yields the leakage reactivity.

The leakage reactivity for CANDU-6 at the reference burnup of7.154 MWd/kgU is determined to be 0.0479 (4.79%). Note thatleakage reactivity values given in Driscoll et al. [6] and Bodansky[7] are in the range of 4.4 to 4.9% for CANDU reactors.

The leakage reactivity for ACR-700 at the reference burnup of20.5 MWd/kgU is obtained to be 0.094467 (9.4467%).

qL is basically sensitive to the fuel loading and managementpatterns. It is assumed that qL values do not change for the pur-poses of this study. In other words, the qL values determined abovefor each reactor are taken as the reference values throughout thisstudy.

3.3. Calculation of discharge burnup

Once the leakage reactivity is determined, in order to calculatedischarge burnup for a given fuel composition, the following proce-dure is applied. Utilizing MONTEBURNS, with an input of isotopicfractions in fuel mixture under consideration, reactivity-versus-burnup datum points are obtained; and by fitting to Eq. (1), coeffi-cients q0, A1, A2. . . are calculated. Then, with the reference leakagereactivity values at hand for CANDU-6 and ACR-700 (Section 3.2),Eq. (2) is solved for Bd. A computational flow chart is presentedin Fig. 1.

Fig. 1. Computational flow diagram.

Fig. 2. Discharge burnup as a function of enrichment for U fuels.

M. Türkmen, O.H. Zabunoglu / Journal of Nuclear Materials 429 (2012) 263–269 265

3.4. Relations for Natural Uranium Requirement (NUR) and Saving(NUS), and Nuclear Resource Utilization (NRU)

Natural Uranium Requirement (NUR) is the amount of naturaluranium (NU) needed to produce a unit amount of electricalenergy, in tonNU/GWe-y. Natural Uranium Saving (NUS) is definedas, for a fuel composition of interest, percent reduction in NUR of areference case.

Nuclear Resource Utilization (NRU) is defined as the ratio ofamount of material fissioned to resource input [8]. Taking intoaccount the fact that fissioning of 1 gram of fissile material yieldsapproximately 1 MWth-day, NRU can be related to burnup by

NRU ¼ Bd

RES=LOADð3Þ

where RES/LOAD is the amount of resources in HM required to pro-duce a unit mass of fuel load in HM and Bd is in MWth-d/kgHM. Incase of (U–Th) fuels, fuel load is comprised of SEU and Th, and re-sources needed to produce LOAD consist of NU and Th.

RESLOAD

¼ h100þ 1� h

100

� �xp � xT

xF � xT

� �ð4Þ

where h is the weight percent (w/o) of Th in fuel load as HM; and xp,xT and xF are, respectively, 235U weight percents of product, tail andfeed in U enrichment process. All process losses are ignored.

4. Results

In both CANDU-6 and ACR-700; for SEU fuels with differentenrichments and for U–Th fuels with different Th loads and varyingU enrichments, discharge burnups are computed. NUR (and NUS)

and NRU are calculated and comparisons are made when meaning-ful. The reference core geometry parameters and leakage reactivityare kept unchanged.

4.1. Burnup versus enrichment for SEU fuels (with no Th)

For U fuels with no Th, results of the burnup calculations arepresented in Fig. 2 as burnup versus enrichment plots. Note that,in case of ACR-700, discharge burnup values are averaged overmass for a bundle since the center fuel element is different fromthe others.

Table 1Coefficients for uranium fuel (in Eq. (5)).

Coefficient D0 D1

CANDU-6 �9.7212 25.536ACR-700 �11.847 15.482

Fig. 3. Discharge burnup as a function of 235U fraction for varying Th contents in U–Th fuels.

Table 2Coefficients for uranium–thorium fuel [in Eq. (6)].

Coefficient CANDU-6 ACR-700

A0 �3.890 �1.285A1 �3.083 � 10�2 �3.766 � 10�2

A2 �6.084 � 10�4 �0.118 � 10�4

B0 37.78 24.66B1 0.2624 0.3483B2 4.4230 � 10�3 0.5199 � 10�3

C0 �19.39 �25.90C1 �0.4239 �0.9736C2 �9.327 � 10�3 �1.649 � 10�3

Fig. 4. Natural U Requirement (NUR) as a function of discharge burnup.

2 For NU and SEU (1.2 w/o) fuels in CANDU-6, IAEA-Tech. Report-407 [1] notes NUR

266 M. Türkmen, O.H. Zabunoglu / Journal of Nuclear Materials 429 (2012) 263–269

A linear expression, in the following form, adequately describesthe relation between enrichment (w/o 235U) and dischargeburnup.1

Bd ¼ D1 � eþ D0 ð5Þ

where D0 and D1 are constants in MWd/kg and e is 235U enrichmentin weight percent. Do and D1 are presented in Table 1 for both CAN-DU-6 and ACR-700. Eq. (5) is valid for 0.711 6 e 6 2.0 w/o withmaximum error of ± 0.87 MWd/kg in CANDU-6 and for1.7 6 e 6 5.2 w/o with maximum error of ± 1.17 MWd/kg in ACR-700.

2.1-w/o SEU in ACR-700 can extract three times more burnupthan NU fuel in CANDU-6 while the core volume decreases nearlya third, together with about one-third reduction in the fuel-loadmass per unit energy produced. 3.1-w/o SEU fuel in ACR-700reaches a burnup of 37900 MWd/ton. This value is 18% higher thanthat of a standard PWR-1000, yielding 32150 MWd/ton with 3.1 w/o SEU. [9].

4.2. Burnup versus Th content and 235U fraction for U–Th fuels

A homogeneous mixture of Th and SEU is used as fuel in all ele-ments of a bundle. For 10, 30, 50, 70 and 90 w/o Th contents, withvarying 235U fractions, discharge burnups are calculated. The re-sults are illustrated in Fig. 3.

As Th content increases, a higher 235U fraction is required toaccumulate the same discharge burnup that can be obtained fromSEU fuels (with no Th). For a specific Th percent, discharge burnupincreases as 235U fraction increases.

The discharge burnup can be related to Th percent and 235Ufraction by a second order polynomial in the following form.

Bdðef ; hÞ ¼ ðA0 þ A1hþ A2h2Þe2

f þ ðB0 þ B1hþ B2h2Þef þ ðC0

þ C1hþ C2h2Þ ð6Þ

1 For U fuels in CANDU-6, discharge burnup values that can be obtained usingEq. (5) agree very well with the plotted Combustion Engineering (CE) data in the ploton page 166 in Driscoll et al. [6].

where ef is 235U fraction in weight percent (235U in total); h is Thweight percent in total fuel; A0, A1, A2, B0, B1, B2, C0, C1, and C2 areconstants [in MWd/kg], given in Table 2. Eq. (6) is valid for1.0 6 e 6 3.1 w/o with maximum error of ± 0.540 MWd/kg in CAN-DU-6 and for 2.1 6 e 6 4.6 w/o with maximum error of ± 0.420MWd/kg in ACR-700.

4.3. Natural Uranium Requirement (NUR) and Saving (NUS)

NURs are shown in Fig. 4. NURs of CANDU-6 for NU and SEU (1.2w/o) fuels are found to be 159.4 and 106.8 ton NU per GWe-y,respectively; SEU (2.1 w/o) fuel of ACR requires 211.9 tonNU/GWe-y.

For all fuel types in both CANDU-6 and ACR-700, as burnup goesup, NUR comes down sharply first and then levels off. In CANDU-6,SEU with enrichment between 1.3 and 1.6 w/o yields a minimumvalue of NUR, about 106 tonNU/GWe-y. At the same discharge bur-nup, ACR-700 requires considerably more NU than does CANDU-6.

In (U–Th) fuels, as Th fraction goes up at any given burnup, NURincreases; however, as burnup increases, the differences diminish,and they all seem to approach more or less the same value.

Results obtained here for CANDU-6 are consistent with the re-sults from earlier studies.2 No result for ACR-700 is available inthe literature to render comparisons. Change of NUS with dischargeburnup is shown in Fig. 5. Since NUS is inversely proportional to

values of 157 and 116 tonNU/GWe-y, respectively; while Gupta et al. [4] mention 168and 104 tonNU/GWe-y. For a fuel composition of 70-w/o Th and 3-w/o 235U inCANDU-6, Gupta et al. [4] present a NUR value of 120 tonNU/GWe-y by thehomogeneous-bundle/core model; for the same composition, NUR is here calculatedto be 126 tonNU/GWe.

Fig. 5. Natural U Saving (NUS) as a function of discharge burnup.

Fig. 6. Nuclear Resource Utilization (NRU) as a function of discharge burnup.

Table 3Summary of results for selected cases.

Reactor Burnup CR NUR NUS NRU

Fuel composition MWth-d/kgHM

TonNU/GWe-y

% MWth-d/kgHM

CANDU-6NU (Ref.) 7.15 0.771 159.4 0 7.15SEU (w/o 235U)

1.20 22.0 0.696 106.8 33.0 10.71.32 25.0 0.693 105.9 33.5 10.81.60 31.6 0.670 105.7 33.7 10.82.00 40.2 0.640 107.7 32.4 10.6

U–Th1.58 w/o 235U

+ 30 w/o Th25 0.745 139.6 12.5 7.5

1.79 w/o 235U+ 50 w/o Th

25 0.764 164.5 No 6.1

2.00 w/o 235U+ 70 w/o Th

25 0.778 204.3 No 5.1

ACR-700SEU(w/o 235U)

2.10 (Ref.) 20.5 0.515 211.9 0 5.12.33 25.0 0.510 198.2 5.2 5.52.50 27.9 0.505 189.3 10.6 5.73.10 37.9 0.488 176.5 16.7 6.14.00 50.0 0.472 175.6 17.0 6.25.20 67.4 0.445 172.4 18.6 6.3

U–Th2.83 w/o 235U

+ 30 w/o Th25.0 0.558 252.0 No 4.1

3.10 w/o 235U+ 50 w/o Th

25.0 0.570 285.0 No 3.5

3.33 w/o 235U+ 70 w/o Th

25.0 0.575 301.0 No 3.2

4.19 w/o 235U+ 30 w/o Th

50.0 0.523 179.1 10.9 5.5

4.41 w/o 235U+ 50 w/o Th

50.0 0.536 203.0 5.8 5.2

4.53 w/o 235U+ 70 w/o Th

50.0 0.545 210.3 1.2 4.9

‘‘Ref.’’ within brackets in the italic rows represents the reference cases.CR: Conversion Ratio, NUR: Natural Uranium Requirement, NUS: Natural UraniumSaving, and NRU: Nuclear Resource Utilization.

M. Türkmen, O.H. Zabunoglu / Journal of Nuclear Materials 429 (2012) 263–269 267

NUR, the opposite of what is observed for NUR can be tracked forNUS values.

4.4. Nuclear Resource Utilization (NRU)

NRU factors are shown in Fig. 6.3 For all fuel types in both reac-tors, NRU increases with discharge burnup first at lower burnups,and then levels off. Though, for CANDU-6, SEU with enrichmentabout 1.3 w/o exhibits a fuzzy maximum.4

For the same discharge burnup, NRU for CANDU-6 is signifi-cantly greater than that for ACR-700 in all cases. In both reactors,NRU decreases as Th fraction increases; the higher the burnup,the smaller the differences.

A summary of results for selected fuel compositions and/orcombinations is tabulated in Table 3.

5. Conclusions

For the once-through cycle in CANDU-6 and ACR-700, from theresults outlined above, the following conclusions can be arrived at.

3 For U fuels in CANDU-6, the NRU values plotted in Fig. 6 are in very goodconformity with the results that can be obtained using Eq. (5.1) on page 137 inDriscoll et al. [6].

4 According to CANDU-6 Program Team [5], the maximum NRU in CANDU-6 occursat an enrichment level of 1.2 w/o.

- About effect of burnup

In all cases, as burnup is increased from a relatively low value(e.g., from around the reference value), NUR comes down andNRU goes up significantly. At high burnups, this trend tends tosmooth out.

- About CANDU-6 and ACR-700

For the same discharge burnup, CANDU-6 is superior to ACR-700 from respects of NUR (and NUS) and NRU. Yet; in general,NUR decreases and NRU increases as discharge burnup goes up,and ACR-700 has a greater ability to reach higher burnups. Then,to sum up, CANDU-6 is more suitable for lower burnup applica-tions while ACR-700 is attractive for higher burnup. Also note thatACR-700 is a GEN III + reactor and the advancements in its designmakes it more readily adaptable, in practice, to higher burnupsand usage of Th.

- About effect of Th fraction

As Th fraction in (U–Th) fuels of both reactors increases, a high-er 235U fraction is required to accumulate the same discharge bur-nup, because the thermal absorption cross-section of 232Th is 2.75times that of 238U.

Table A.1Design parameters for CANDU-6 and ACR-700 [2].

Reactor and fuel channel CANDU-6 ACR-700

Thermal power (MWth) 2061 1982Gross nominal power (MWe) 713 731Reactor pressure (MPa) 11.1 12.6Reactor core length (mm) 5944 5940Number of fuel channels 380 284Reactor core radius (mm) 3800 2600Fueling rate (channels per day) 1.97 2.8Average fuel burnup (MWd/kgU) 7.154 20.5Lattice pitch (mm) 286 220Number of fuel bundles per channel 12 12Pressure tube inner radius (mm) 51.689 51.689Pressure tube outer radius (mm) 52.1233 58.169Calandria tube inner radius (mm) 64.478 75.50Calandria tube outer radius (mm) 65.875 78.00

MaterialPellet data

Material NaturalUO2

2.1 w/o UO2*

Nominal density (g/cm3) 10.60 10.65Sheath material Zircaloy-4 Zircaloy-4

Density (g/cm3) 7.48 7.48Calandria tube material Zircaloy-2 Zircaloy-4

Density (g/cm3) 6.55 6.44Pressure tube material Zr-2.5%Nb Zr-2.5%NbDensity (g/cm3) 6.57 6.57Coolant D2O H2OCoolant atom purity (%) 99.10 99.75Moderator D2O D2OModerator atom purity (%) 99.85 99.85

Fuel bundle and elementNumber of fuel elements per bundle 37 43 (total)Bundle U mass (kg) 19.10 17.98Bundle Zr mass (kg) 2.206 2.3Bundle weight (kg) 21.67 22.7Bundle length (mm) 482 495.3Fuel pin 13.10 mm

OD11.5 mm OD 13.5 mm

ODPellet data

Pellet OD (mm) 12.20 10.65 12.58Pellet length (mm) 16.60 10.60 16.00

Fuel element dataNumber of elements 37 14 + 21 = 35

(two outerrings)

1*+7 = 8(centralandinnermostring)

Number of pellets per stack 29 45 30Stack length (mm) 482.0 481.1 481.1

⁄The central fuel element consists of 7.5 w/o Dy in natural U.

268 M. Türkmen, O.H. Zabunoglu / Journal of Nuclear Materials 429 (2012) 263–269

A higher Th fraction in fuel-load directly results in a loweramount of SEU required; however, despite this lower SEU require-ment, NUR increases with Th fraction. The higher NUR that iscaused by the increased enrichment of SEU outweighs the effectof the reduction in SEU need, and the net result becomes an in-creased NUR (refer to Fig. 4 and Table 3). At high burnups, NURsfor all fuel compositions tend to approach the same level.

U–Th fuel yields lower NUR than NU fuel in CANDU-6. AndU–Th fuel yields lower NUR than SEU fuel (2.1 w/o enriched) inACR-700. However, it should be emphasized that, if the SEU por-tion of the U–Th fuel in each case were used alone (that is, withoutTh), NUR values would be much lower.

Concisely, once-through use of Th in a straightforward, homo-geneous way in CANDU-6 or ACR-700 increases NUR (Natural Ura-nium Requirement) and decreases NRU (Nuclear ResourceUtilization). Then, Th is not a good replacement for 238U in thePHWRs on the once-through cycle. To rephrase it from an eco-nomic perspective; even if Th were free of charge, use of it insteadof 238U in PHWRs on the once-through would pose an economicburden, not to mention an advantage.

6. Discussion

All the results/conclusions are based on the homogeneous-bun-dle/core model and the once-through cycle, which were initiallyselected for all analyses in this study.

Other models, mainly mixed-bundle and/or mixed-core, involv-ing use of U–Th fuel rods at certain locations in a bundle and/or useof bundles containing U–Th rods at certain locations in a reactor,where U–Th fuels have higher and/or longer exposures, are likelyto improve the case for Th. It may be worthwhile to put forwarda comparison of all these fuel management schemes. Yet, such astudy would require several challenges in modeling and comput-ing. As stated in Section 2.3, the mixed-bundle approach is simplerthan the mixed-core both in practice and in computations; and itmay be investigated in a future study.

Still, even by the best approach to use Th instead of 238U, itappears that there is not much to be gained by insisting on theonce-through cycle. Closed cycles, which involve reprocessingand recycling of 233U produced from 232Th, perhaps with novelscenarios, may result in favor of Th usage, since conversion ratiosin case of U–Th fuels are higher than those in U fuels (as seen inTable 3). In U–Th fuels, as Th fraction goes up, conversion ratio alsogoes up. In other words, when more Th is added, more fertile atomsare converted to fissile ones.

Several recycling scenarios which involve Th/Pu and Th/233Ufuels in CANDUs and PWRs have been looked into by Gupta et al.[4] and, Nuttin et al. [10]. In addition, use of SEU, Pu and Th inthe Indian Advanced Heavy Water Reactor (AHWR) [11], which isa Th-based boiling light-water reactor moderated by heavy water,has been studied by Prasad et al. [12,13].

On one hand; because Th is not a good replacement for 238U, in anuclear world where even reprocessing of spent LWR fuels is notadopted to an effective extent, it seems at least in the short runthat no sufficient incentive (especially for the industry) is thereto go after closed fuel cycles particularly devised to make betteruse of Th.

On the other hand; because Th is the ONLY replacement for238U, pessimistic views about Th, being somewhat fed from unsat-isfactory attempts in utilization of a remarkable potential, willkeep attracting minds of researchers and academics.

In pursuit of practical (and maybe novel) recycling scenarios,with a touch on issues in the reprocessing, we are investigatinguse of U (and Pu, optionally) in spent U–Th fuels from CANDUand ACR. It is probable that such a study is going to yield resultsworthwhile putting forward in a follow-up article.

Appendix A. Data for CANDU-6 and ACR-700

Comprehensive data for CANDU-6 and ACR-700 are tabulated inTable A.1.

Appendix B. Description of the codes used

MCNP5/Monte Carlo N–Particle code [14] can be used forneutron, photon, electron, or coupled neutron/photon/electrontransport. The code treats an arbitrary three-dimensional configu-ration of materials in geometric cells. Point-wise cross-section dataare used. For neutrons, all reactions given in a particular cross-sec-tion evaluation (such as ENDF/B-VI) are accounted for. Thermalneutrons are described by both the free gas and S (a, b) models.For photons, the code accounts for incoherent and coherent scat-tering, and also other possible basic reactions. MCNP5 includesthe tallies: surface current and flux, volume flux (track length),

M. Türkmen, O.H. Zabunoglu / Journal of Nuclear Materials 429 (2012) 263–269 269

point or ring detectors, particle heating, fission heating, pulseheight tally for energy or charge deposition and mesh tallies.

MONTEBURNS2 [15] is a fully automated tool that links theMonte Carlo transport code MCNP with the radioactive decayand burnup code ORIGEN2. MCNP5 provides neutron cross-sectionsets to MONTEBURNS2. OrigenS module calculates fuel composi-tions with regard to exposure time of fuel. MONTEBURNS2 pro-duces a large number of criticality and burnup results based onvarious material feed/removal specifications, power(s), and timeintervals. Various results from MCNP, ORIGEN2, and other calcula-tions are then output successively as the code runs.

Appendix C. Supplementary data

Supplementary data associated with this article can be found, inthe online version, at http://dx.doi.org/10.1016/j.jnucmat.2012.06.008.

References

[1] IAEA, Heavy Water Reactors – Status and Projected Development, IAEATechnical Reports Series No:407, 2002.

[2] AECL, ACR-700 Technical Description, 10810-01371-TED-001Revision 0, 2003.<http://pbadupws.nrc.gov/docs/ML0320/ML032030391.pdf>.

[3] IAEA, Thorium fuel cycle – Potential benefits and challenges, IAEA TECDOC-1450, 2005.

[4] H.P. Gupta, S.V.G. Menon, S. Banerjee, J. Nucl. Mater. 383 (2008) 54–62.[5] CANDU 6 Program Team: Reactor Development Business Unit, CANDU-6

Technical Summary, 2005. <http://www.rwe.com/web/cms/contentblob/2914/data/8756/Technical-Summary-CANDU.pdf>.

[6] M.J. Driscoll, T.J. Downar, E.E. Pilat, Linear Reactivity Model for Nuclear FuelManagement, ANS, Illinois, 1990.

[7] D. Bodansky, Nuclear Energy: Principles, Practice and Prospects, Springer, NewYork, 2004.

[8] J.R. Lamarsh, A.J. Baratta, Introduction to Nuclear Engineering, third ed.,Prentice Hall, New Jersey, 2001.

[9] L. Özdemir, B.B. Acar, O.H. Zabunoglu, Ann. Nucl. Energy 38 (2011) 540–546.

[10] A. Nuttin, P. Guillemin, A. Bidaud, N. Capellan, R. Chambon, S. David, O.Méplan, J.N. Wilson, Ann. Nucl. Energy 40 (2012) 171–189.

[11] R.K. Sinha, A. Kakodkar, Nucl. Eng. Des. 236 (2006) 683–700.[12] N. Prasad, A. Kumar, U. Kannan, A. Kumar, P.D. Krishnani, R.K. Sinha, Study for

use of LEU along with Thorium in advanced heavy water reactor (AHWR) toenhance proliferation resistance characteristics of fuel, in: Symposium onInternational Safeguards Preparing for Future Verification Challenges, IAEA-CN-184(2010)/207, Vienna, Austria, November 2010.

[13] N. Prasad, A. Kumar, P.D. Krishnani, R.K. Sinha, Core Design Optimisation inAdvanced Heavy Water Reactor for Achieving Self-Sustenance in 233U,PHYSOR 2010: Advances in Reactor to Power the Nuclear Renaissance,Pittsburg, Pennsylvania, USA, May, 2010.

[14] X-5 MONTE CARLO TEAM, MCNP – A General Monte Carlo N-Particle TransportCode, Version 5, LA-CP-03-0245, LANL, 2003.

[15] D.L. Poston, H.R. Trellue, User’s Manual, Version 2.0 for MONTEBURNS Version1.0, LA-UR-99-4999, 1999.