Effect of void-fraction on characteristics of several thorium fuel cycles in BWR Abdul Waris ⇑ , Mohamad Ali Shafii, Syeilendra Pramuditya, Rizal Kurniadi, Novitrian, Zaki Su’ud Nuclear Physics and Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jl. Ganesa 10, Bandung 40132, Indonesia article info Article history: Available online 3 April 2012 Keywords: Thorium fuel Time-independent burnup BWR Void fraction SRAC 2002 abstract Study on effect of void-fraction on characteristics of several thorium fuel cycles in boiling water reactor (BWR) has been carried out. Four fuel cycle scenarios are evaluated and the comparison with once- through-cycle BWR was also conducted. The 1367 nuclides have been employed in the time independent burnup calculation scheme where 129 of them are heavy metals (HMs). The burnup code was facilitated with the calculation method for determining the required 233 U concentration for criticality of the system inherently. The code was coupled with SRAC 2002 cell calculation code to compose a time independent- cell burnup code. In the cell calculation, 26 HMs, 66 fission products (FPs) and one pseudo FP have been utilized. The JENDL 3.2 library has been employed in this study. The results show that the profile of the required 233 U concentration for criticality depends on the neutron spectra and the microscopic cross-sec- tion. However, the profile of the conversion ratio likely depends on the number density of heavy nuclides. Ó 2012 Elsevier Ltd. All rights reserved. 1. Introduction Thorium, which has approximately three times as abundant as uranium [1], is a fertile material, since 232 Th is able to capture thermal neutrons to generate 233 U. From the neutronics viewpoint, 233 U is the best fissile isotope, as a thermal reactor fuel. The g value (the average number of neutron produced per neutron absorbed in fuel) of 233 U, in all energy range, is superior compared to those of 235 U and 239 Pu. Therefore, thorium based fuel cycle can be used in all proven reactor types [2]. Thorium oxide (ThO 2 ) has better stability and able to be utilized with high temperature, and can be expected to gain high burnup since it has longer durability due to its melting point of 3050 °C compared to that of UO 2 (2700–2800 °C) [2]. In the past, the thorium fuel development into industrial scale has been slowdown due to economic and technical problems such as: some difficulties in thorium extraction from the ores and high gamma radiation associated with the short lived daughter products of 232 U, which is always associated with 233 U, as well as necessi- tates remote reprocessing which contributes to high fuel fabrica- tion price [2]. However, recently, the thorium-based nuclear fuels have been getting refreshed attentiveness as a manner to deal with prolifera- tion and wastes concerns associated with commercial nuclear power [3,4]. Accumulation of massive stockpiles of plutonium has become serious public and political concern. One alternative solution for the management of plutonium is to recycle it in reactors. When the plutonium is loaded in reactors in the form of uranium/plutonium mixed oxide (MOX), second-generation pluto- nium is also generated. Recycling of plutonium in combination with thorium in reactor is a possible solution to this problem [4]. As a matter of perspective, originally, only plutonium was recov- ered from spent light water reactors (LWRs) fuels to initiate the ‘‘U-238 – Pu breeding cycle’’ in fast breeder reactor (FBR) or to be recycled in other reactor systems. However, from the prolifera- tion aspect, for future fuel cycle schemes it is likely, that Pu together with the minor actinides (MA) should be partitioned from spent fuel and also should be recycled together [5]. As a part of revisiting the thorium-based nuclear fuel for the present and future nuclear energy systems, from the neutronics point of view, the present study evaluates the effect of void fraction on characteristics of several scenario of trans-uranium (TRU) con- fining not only plutonium in boiling water reactor (BWR). The light water reactors especially BWR was selected because LWRs will still rule the nuclear energy systems up to 2050 [6]. Even though the use of 232 Th and 233 U as initial loaded fuel nuc- lides is rather traditional scenario for thorium fuel cycle, since 233 U does not occur naturally, the present study is a part of our whole study on thorium utilization in Generation III and VI reactors [7,8]. The proxy study on 233 U free thorium fuel cycle in BWR has been published in Ref. [9]. 2. Calculation method 2.1. Equilibrium burnup scheme In the present study, a time independent burnup scheme which sometimes called as ‘‘the equilibrium burnup’’ has been employed. 0196-8904/$ - see front matter Ó 2012 Elsevier Ltd. All rights reserved. http://dx.doi.org/10.1016/j.enconman.2012.01.026 ⇑ Corresponding author. E-mail address: awaris@fi.itb.ac.id (A. Waris). Energy Conversion and Management 63 (2012) 11–16 Contents lists available at SciVerse ScienceDirect Energy Conversion and Management journal homepage: www.elsevier.com/locate/enconman
Transcript
Energy Conversion and Management 63 (2012) 11–16
Contents lists available at SciVerse ScienceDirect
Effect of void-fraction on characteristics of several thorium fuel cycles in BWR
Abdul Waris ⇑, Mohamad Ali Shafii, Syeilendra Pramuditya, Rizal Kurniadi, Novitrian, Zaki Su’udNuclear Physics and Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jl. Ganesa 10, Bandung40132, Indonesia
Study on effect of void-fraction on characteristics of several thorium fuel cycles in boiling water reactor(BWR) has been carried out. Four fuel cycle scenarios are evaluated and the comparison with once-through-cycle BWR was also conducted. The 1367 nuclides have been employed in the time independentburnup calculation scheme where 129 of them are heavy metals (HMs). The burnup code was facilitatedwith the calculation method for determining the required 233U concentration for criticality of the systeminherently. The code was coupled with SRAC 2002 cell calculation code to compose a time independent-cell burnup code. In the cell calculation, 26 HMs, 66 fission products (FPs) and one pseudo FP have beenutilized. The JENDL 3.2 library has been employed in this study. The results show that the profile of therequired 233U concentration for criticality depends on the neutron spectra and the microscopic cross-sec-tion. However, the profile of the conversion ratio likely depends on the number density of heavy nuclides.
� 2012 Elsevier Ltd. All rights reserved.
1. Introduction reactors. When the plutonium is loaded in reactors in the form of
Thorium, which has approximately three times as abundant asuranium [1], is a fertile material, since 232Th is able to capturethermal neutrons to generate 233U. From the neutronics viewpoint,233U is the best fissile isotope, as a thermal reactor fuel. The g value(the average number of neutron produced per neutron absorbed infuel) of 233U, in all energy range, is superior compared to those of235U and 239Pu. Therefore, thorium based fuel cycle can be usedin all proven reactor types [2].
Thorium oxide (ThO2) has better stability and able to be utilizedwith high temperature, and can be expected to gain high burnupsince it has longer durability due to its melting point of 3050 �Ccompared to that of UO2 (2700–2800 �C) [2].
In the past, the thorium fuel development into industrial scalehas been slowdown due to economic and technical problems suchas: some difficulties in thorium extraction from the ores and highgamma radiation associated with the short lived daughter productsof 232U, which is always associated with 233U, as well as necessi-tates remote reprocessing which contributes to high fuel fabrica-tion price [2].
However, recently, the thorium-based nuclear fuels have beengetting refreshed attentiveness as a manner to deal with prolifera-tion and wastes concerns associated with commercial nuclearpower [3,4]. Accumulation of massive stockpiles of plutoniumhas become serious public and political concern. One alternativesolution for the management of plutonium is to recycle it in
ll rights reserved.
uranium/plutonium mixed oxide (MOX), second-generation pluto-nium is also generated. Recycling of plutonium in combinationwith thorium in reactor is a possible solution to this problem [4].As a matter of perspective, originally, only plutonium was recov-ered from spent light water reactors (LWRs) fuels to initiate the‘‘U-238 – Pu breeding cycle’’ in fast breeder reactor (FBR) or tobe recycled in other reactor systems. However, from the prolifera-tion aspect, for future fuel cycle schemes it is likely, that Putogether with the minor actinides (MA) should be partitioned fromspent fuel and also should be recycled together [5].
As a part of revisiting the thorium-based nuclear fuel for thepresent and future nuclear energy systems, from the neutronicspoint of view, the present study evaluates the effect of void fractionon characteristics of several scenario of trans-uranium (TRU) con-fining not only plutonium in boiling water reactor (BWR). The lightwater reactors especially BWR was selected because LWRs will stillrule the nuclear energy systems up to 2050 [6].
Even though the use of 232Th and 233U as initial loaded fuel nuc-lides is rather traditional scenario for thorium fuel cycle, since 233Udoes not occur naturally, the present study is a part of our wholestudy on thorium utilization in Generation III and VI reactors[7,8]. The proxy study on 233U free thorium fuel cycle in BWRhas been published in Ref. [9].
2. Calculation method
2.1. Equilibrium burnup scheme
In the present study, a time independent burnup scheme whichsometimes called as ‘‘the equilibrium burnup’’ has been employed.
Power output (thermal) 3000 MWAverage fuel cell power density 50 W cm�3
Radius of fuel pellet 0.529 cmRadius of fuel rod 0.615 cmPin pitch 1.444 cmVoid fraction 20–70%Fuel type 232ThO2 + 233UO2
Cladding Zircaloy-2Coolant H2O
12 A. Waris et al. / Energy Conversion and Management 63 (2012) 11–16
This burnup scheme has been introduced and intensively used inour previous studies concerning LWRs [7–14]. The equilibriumburnup model is as a powerful method to evaluate the characteris-tics of the nuclear energy system since it can cope with all poten-tial produced nuclides in any nuclear system. In the equilibriumburnup model, we assumed that refueling is a continuous processand the number density of each nuclide in the reactor is constant[13].
The detail explanation of the equilibrium burnup calculationscheme for BWR can be found in Refs. [7,9] and can be summarizedconcisely as below.
The equilibrium burnup equation can be modestly expressed asthe following matrix equation.
Mn ¼ s; ð1Þ
where the elements of matrix M comprise all the transmutationparameters of all nuclides such as natural decay constants andmicroscopic transmutation cross-sections. The n and s stand forthe vectors of the number density of nuclides in the reactor coreand the supply rate of fuel nuclides (232Th and 233U), respectively.
To evaluate the performance of the investigated fuel cycles, wehave introduced the nuclide importance values [11,13]. The nu-clide importance vectors f and a can be calculated from the follow-ing adjoint equations [13].
Mtf ¼ /mrf ;
Mta ¼ /ra;ð2Þ
where Mt is the adjoint matrix of M, / is the neutron flux, rf and ra
are the vectors of microscopic fission cross-sections and micro-scopic absorption cross-sections, correspondingly. The m representsthe number of neutrons produced in each fission reaction. We havecalled f and a as fission neutron importance and absorbed neutronimportance, respectively [13]. The fission neutron importance rep-resents the number of neutrons produced from fission of one nu-cleus of the studied nuclide and its family members (reactionproducts) during its existence in the reactor. The absorbed neutronimportance represents the number of neutrons absorbed by one nu-cleus of the studied nuclide and its family members during its pres-ence in the reactor [13].
By employing the nuclide importance vectors, the infinite mul-tiplication factor, k, can be expressed as the following equation.
k ¼ ðmrf ;nÞaðra;nÞ
¼ ðf; sÞaða; sÞ ; ð3Þ
where a is a correction parameter for estimating neutron absorp-tion of non-fuel nuclides such as coolant and structural materials.
To determine the criticality of the system, the neutron leakagefrom the system should be evaluated. For current BWR, the neu-tron leakage is estimated about 2.5% of produced neutrons. This va-lue is based on our previous study on the equilibriumcharacteristics of BWR with the uranium fuel cycle [7,9]. Thenthe following condition is employed for the criticality conditionin the present study.
k ¼ 1:025 ¼ kc ð4Þ
The whole calculations were conducted by using our equilib-rium burnup code. This equilibrium burnup code is coupled witha SRAC [15] cell-calculation code to become a coupled cell-burnupcalculation code system. We have employed 1367 nuclides in theequilibrium burnup calculation with 129 of them are heavy metals(HM) and the rest are fission products (FP). For both calculationschemes, the JENDL 3.2 library has been used [16].
Table 1 represents the design parameters of the GeneralElectric’s BWR/6 reactor, which has been adopted in the presentstudy [17].
2.2. Conversion ratio
One of neutronics parameters that may be change as well as thevoid fraction changes is the conversion ratio. The conversion ratiofor thorium fuel reactor is defined by the following equation [7].
CR ¼ Capture rate of ð232Th� 233Paþ 234UÞAbsorption rate of ð233U þ 235UÞ ð5Þ
This equation is based on production rate and consumption rateof the fertile and fissile nuclides, respectively, with the contribu-tion from intermediate nuclides such as 234U and 233Pa. The valueof the conversion ratio is calculated by using the equilibriumnuclide number densities and microscopic cross-sections.
3. Results and discussion
In the present study, four scenarios of the TRU confining in BWRare evaluated. They are Plutonium confining, Plutonium and MinorActinides (MA) confining, All HM confining, and All HM except Ura-nium confining. Here, we used ‘‘confining’’ instead of ‘‘recycling’’ or‘‘incineration’’. One of the advantages of the equilibrium burnupmodel is it can be used for non-conventional fuel loading scheme.For example, the last scenario means that all HMs except uraniumare maintained in the core without any reprocessing process anduranium only is discharged with the annual rate of 33%. Exceptfor All HM confining, uranium is discharged in the other scenarios.In addition, the calculation results on once through cycle (OTC)BWR are presented for comparison.
The neutron flux for OTC-BWR, Pu-confining, Pu&MA-confining,and All-HM-except-U-confining cases, in other words for all caseswhere uranium is discharged (U-discharging cases) are exactlysame and increases with the raising of void fraction from1.72 � 1014 (#/cm2/s) to 3.43 � 1014 (#/cm2/s) for 20% to 70% ofvoid fraction, respectively. Similar trend also happened for All-HM-confining case, the scenario where uranium is confined (U-confining case), but the values are smaller from 1.65 � 1014 (#/cm2/s) to 3.24 � 1014 (#/cm2/s) for 20% to 70% of void fraction, cor-respondingly. In summary, the neutron flux raises with the declin-ing of moderator volume.
Fig. 1 shows the influence of void fraction on the number den-sity (ND) of selected HM for Pu-confining case. For Pu-confining,the equilibrium number density of Pu, Am and Cm have the highestvalue at 42% of void fraction. Figs. 2 and 3 show the number den-sity of selected HM for all evaluated cases with the void fraction of42% and 70%, correspondingly. In general, the equilibrium numberdensity of Pu and MA in the core is much smaller than that of U andTh for OTC-BWR, Pu-confining, and Pu&MA-confining cases. ForAll-HM-except-U-confining case, the number density of U, Pu,and MA increases with the increasing of void fraction. Impact ofthe void fraction enlarging on the number density of U, Pu, andMA for All-HM-confining case is similar to All-HM-except-U-con-fining case, except for 242Pu, 243Am, and 244Cm. For all void frac-tions, All-HM-except-U-confining case gives the largest numberdensity of all selected HM.
A. Waris et al. / Energy Conversion and Management 63 (2012) 11–16 13
Figs. 4 and 5 demonstrate the neutron spectra for all evaluatedcases with the void fraction of 20% and 70%, respectively. As can beseen from these figures, the neutron spectra become harder withthe escalating of void fraction for all five cases. The neutron spectraare similar for four cases (OTC-BWR, Pu-confining, Pu&MA-confin-ing, and All-HM-except-U-confining) for the same void fraction. Inother words, the neutron spectra are identical for the U-discharg-ing cases. For All-HM-confining case (U-confining case), the neu-tron spectrum is harder compared to those of others for thesame void fraction.
The microscopic absorption cross-section has similar tendencycompared to the neutron spectra. Figs. 6 and 7 present the micro-scopic absorption cross-section for all evaluated cases with thevoid fraction of 20% and 70%, correspondingly. The later mentionedparameter becomes smaller with the mounting of void fraction forall evaluated cases. These cross-sections are identical for the fourcases (OTC-BWR, Pu-confining, Pu&MA-confining, and All-HM-ex-cept-U-confining) for the same void fraction. In other words, these
values are equal for the U-discharging cases. For All-HM confiningcase (U-confining case), the microscopic absorption cross-section islesser compared to those of others for the same void fraction. Thelatter fact becomes clearer for the void fraction that superior than20%.
Fig. 8 illustrates the required 233U concentration for criticality ofall evaluated scenarios as a function of void fraction. The required233U concentration for criticality of the Pu-confining and thePu&MA-confining cases are equal as that of the OTC-BWR case,from 1.94 wt.% for 20% void fraction to 3.03 wt.% for 70% void frac-tion. Moreover, for All-HM-except-U-confining case, the required233U concentration enlarges with the growing of void fraction, from51.96 wt.% to 61.31 wt.% for 20% and 70% void fraction, respec-tively. In general, the required 233U concentration for criticality in-creases with the decreasing of moderator volume for the abovefour U-discharging cases. However, for All-HM-confining case,the required 233U concentration diminishes with the escalating ofvoid fraction, from 11.83 wt.% to 6.78 wt.% for 20% and 70% of voidfraction, correspondingly. In other words, the required 233U
Fig. 8. Required 233U concentration for all cases.
14 A. Waris et al. / Energy Conversion and Management 63 (2012) 11–16
concentration for criticality reduces with the rising of moderatorvolume for the U-confining case.
The conversion ratio (CR) of all evaluated cases as a function ofvoid fraction is presented in Fig. 9. The conversion ratio of the OTC-BWR, the Pu-confining and the Pu&MA-confining scenarios areidentical, from 0.93 for 20% void fraction to 0.98 for 70% void frac-tion. For All-HM-confining scenario, CR improves with the enhanc-ing of void fraction, from 0.89 to 0.95 for 20% to 70% of voidfraction, correspondingly. In other words, for these four scenariosCR raises with the lessening of moderator volume. However, forAll-HM-except-U-confining case, CR is small enough and it de-clines with the boosting of void fraction, from 0.53 to 0.45 for20% and 70% of void fraction, respectively.
As mention in Section 2.2, the CR was evaluated by employingthe equilibrium number density and the microscopic cross-sec-tions. As can be seen from Figs. 4–7, the influence of the neutronspectra as well as the microscopic cross-sections on CR seems less
sensitive compared to the number density. Figs. 10–12 demon-strate the macroscopic capture cross-section (Rc) of 232Th, 233Paand 234U, and macroscopic absorption cross-section (Ra) of 233Uand 234U for Pu&MA-confining, All-HM-confining, and All-HM-ex-cept-U-confining, respectively. For OTC-BWR, Pu-confining, andPu&MA-confining scenarios, CR raises with the increment of voidfraction may due to Ra decrement of 235U. Similar trend happensto All-HM-confining case may due to capture-rate reduction of233Pa. For All-HM-except-U-confining scenario, CR declines withthe augmentation of void fraction due to larger absorption-rateof 233U.
Interestingly, for All-HM-except-U-confining case, the required233U concentration for criticality raises together with the diminish-ing of CR with the lessening of moderation. And for All-HM-confin-ing case, the required 233U concentration for criticality reduces andCR increases with the enlarging of void fraction. However, for threecases (OTC-BWR, Pu-confining, and Pu&MA-confining), the 233Uconcentration for criticality and CR both increase with the
A. Waris et al. / Energy Conversion and Management 63 (2012) 11–16 15
increasing of void fraction. The required 233U concentration forcriticality of U-discharging cases increase with the boosting of voidfraction may due to similar pattern changing of the neutron spectraand the microscopic cross-section.
4. Conclusion
Study on effect of void-fraction on equilibrium characteristics ofseveral thorium fuel cycles in BWR has been conducted. Four fuelcycle scenarios are evaluated and the comparison with OTC BWRwas also performed. The required 233U concentration for criticality,the neutron flux, microscopic absorption cross-section, and theneutron spectra of Pu-confining and Pu&MA-confining cases areequal compared to that of the OTC BWR case.
It may be concluded that the required 233U concentration forcriticality pattern depends on the neutron spectra and the micro-scopic cross-section. On the other hand, CR pattern very likelydepends on the number density of heavy nuclides in BWR core.
Acknowledgment
This study is fully funded by Institut Teknologi Bandung’sResearch Grant No. 174/K01.07/PL/2007 and Institut TeknologiBandung’s IMHERE Program 2011.
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