Radiation Protection Dosimetry (2014), Vol. 162, No. 4, pp. 459 –462 Advance Access publication 18 February 2014
doi:10.1093/rpd/ncu016
RADIATION DOSE ASPECTS IN THE HANDLING OF EMERGING NUCLEAR FUELS G. Nicolaou* Laboratory of Nuclear Technology, Department of Electrical and Computer Engineering, School of Engineering, Demokritus University of Thrace, Xanthi 67100, Greece *Corresponding author: [email protected]
The occupational annual dose levels, encountered at fabrication of emerging nuclear fuels, have been studied. Emerging fuels for the single and multiple recycling of Pu and MA have resulted in considerably higher gamma and neutron doses in comparison with commercial fuels. The occupational dose limit is exceeded at fabrication by a single fuel rod in all fuel cases with 241Am and Cm isotopes present in their composition. In the absence of these isotopes, 2 – 4 adjacent fuel rods are sufficient to exceed the limit. Self-shielding within the fuel reduces significantly only the gamma dose that would have been delivered otherwise. Hence, only the first row of fuel rods in an assembly contributes to the dose, whereas in the case of neutrons, all fuel rods contribute.
INTRODUCTION
MATERIALS AND METHODS
The management of nuclear waste, whether as spent nuclear fuel discharged from reactors or high-level waste from the reprocessing of this fuel, is confronted with the radiotoxicity potential owing to transuranium (TRU) isotopes in the waste. Hence, the possibility of partitioning TRU isotopes out of the waste and transmuting them into less hazardous isotopes, relying on existing nuclear power stations, has been proposed and studied(1 – 5). In this context, ‘oncethrough cycle’ or ‘self-generated’ recycling strategies have emerged as potential partitioning and transmutation (P&T) fuel cycles of minor actinides (MA), namely neptunium (Np), americium (Am) and curium (Cm), and plutonium (Pu). The inclusion of Pu and MA in a uranium oxide matrix fuel, for their subsequent recycling, is expected to increase the radiation dose levels associated with such fuels at fabrication. Knowledge of these dose levels would allow measures to be conceived for the handling of such fuels during fabrication, packaging, delivery and loading to reactors. In this work, occupational radiation dose levels, due to the gamma rays and neutrons emitted by single rods of such fuels, have been calculated for different single and multiple recycling schemes. They are subsequently compared with the dose levels encountered at fabrication of fuels in existing commercial fuel cycles. The effect on the dose, owing to self-shielding effects within the fuel itself, attenuation in the cladding and screening between the fuels, is considered in the calculations and assessed. The necessary calculations have been performed using the Monte Carlo N-Particle code MCNPX(6).
Nuclear fuels considered Emerging nuclear fuels in the nuclear reactors pressurised water reactor (PWR) and liquid metal fast breeder reactor (LMFBR) have been considered, within the nuclear waste management option of the P&T of TRU. Existing commercial fuels, in the same reactor types, have been included in the study for comparison purposes. The following are the compositions of the fuels, ‘at fabrication’, considered (Table 1): (1) Transmutation schemes: mixed oxide (MOX) fuels containing homogeneously mixed low concentrations of either 237Np or 241Am (cases T1 and T2) for their homogeneous transmutation in LMFBR, and uranium (U) matrix fuels with large concentrations of 237Np and 241Am for their heterogeneous transmutation in LMFBR (T3 and T4); (2) Self-generated recycling schemes: self-generated Pu and TRU recycle in PWR (S1 and S2), and self-generated TRU recycle in LMFBR (S3); (3) Commercial fuels: 3.5 % of 235U-enriched U fuel and MOX UPu in PWR (R1 and R2), and MOX UPu in LMFBR (R3). The plutonium, used in the PWR- and LMFBRMOX fuel cases R2, R3, T1 and T2, was a 1st-generation Pu from the PWR-U case R1 (55 GWd tU21). The isotopic composition (wt %) of Pu retrieved from the reprocessing of a spent fuel with 1.5 y of cooling time is 238/239/240/241/242 of 3.2/48/24/15.2/ 9.6. In the self-generated recycling schemes (cases S1 –S3), the Pu and TRU compositions in the fuels, at the fabrication stage, are the ones after the 5th
# The Author 2014. Published by Oxford University Press. All rights reserved. For Permissions, please email: [email protected]
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Received 2 September 2013; revised 11 December 2013; accepted 29 January 2014
G. NICOLAOU Table 1. Commercial and emerging fuel compositions at fabrication. Reactor
Fuel type
Fuel charge compositiona
R1 R2 R3 T1 T2 T3 T4 S1 S2 S3
PWR PWR LMFBR LMFBR LMFBR LMFBR LMFBR PWR PWR LMFBR
UO2 MOX MOX MOX with 1 % Np MOX with 1 % Am U with Np and Am U with Np self-generated Pu recycle self-generated TU recycle self-generated TU recycle
UO2 U0.96Pu 0.04 U0.80Pu0.20 U0.795Pu0.195Np0.01 U0.795Pu0.195Am0.01 U0.60Np0.20Am0.20 U0.55Np0.45 U0.96Pu0.04 U0.95Pu0.04Am0.003Cm0.002Np0.005 U0.8Pu0.186Am0.0067Cm0.0058Np0.0015
Case R1—235U 3.5 % enriched; all other fuel cases with depleted U (0.2 % generation Pu from PWR.
a
recycling step (cases S1 and S2) or the 16th recycling step (case S3). These recycling steps correspond to the periods required in order to achieve ‘equilibrium’ in the composition of the recycled nuclides of interest in the fuel(7). At this stage, the amount of nuclides being formed equals the one recycled during irradiation. The Pu and MA, used in the recycling and transmutation schemes, were retrieved from reprocessing carried out on spent fuel having a cooling time of 1.5 y. Fuel fabrication took place 1 y after the retrieval of Pu and MA, avoiding the build-up of 208Tl in large amounts. Computational methodology The required 1st-generation Pu and the ‘equilibrium’ Pu and TRU isotopic compositions, after the 5th and 16th recycle steps, were obtained through depletion calculations. The calculations for the PWR fuel cycles were carried out using the module ORIGEN-ARP, which is a zero-dimensional isotope generation and depletion code within SCALE 6.1(8), employing nuclear data from the ENDF/B-VI library(9). Since an LMFBR model is not included in ORIGEN-ARP, the ORIGEN-2.2(10) depletion code coupled with nuclear data libraries from ENDF/B-V(11) was used for the calculations based on the LMFBR. The depletion calculations were carried out with target burnup values of 55 and 110 GWd tU21 for the PWR and LMFBR cases, respectively. Photon spectra with energies up to 2.75 MeV and the number of neutrons emitted by the different fuels at fabrication are obtained from these depletion calculations. The photon spectra and neutron emission rates are then used as input to the MCNPX code in order to describe the fuel as the radiation source term in the required dosimetry calculations. Dosimetry calculations have been performed using the general purpose 3-D MCNPX code coupled with the ENDF/B-VI nuclear data. The physical description
235
U); cases R2, R3, T1 and T2 with 1st-
Table 2. Physical description of the fuel rods and assemblies considered. PWR Assembly type Fuel rod Height (m) Active height (m) Active mass (g) Pellet diameter (mm) Pellet height (cm) Pellet mass (g) Clad material Clad thickness (mm) Fuel rod pitch (mm)
LMFBR
Square 17` 17 rods
hexagonal 271 rods
4 3.81 2961 9.5
2.7 1.6 995 8.5
1 7.8 Zr 0.57 12.6
1 6.2 Stainless steel 0.7 9.7
of the fuel rods and assemblies for the purpose of this study are given in Table 2(12, 13). The calculations were performed on the basis of the composition and certain physical properties (e.g. density and size) of the fuel material and the cladding of the fuel rod. Hence, self-shielding effects within the fuel rod and screening between the fuels are considered in the calculations. The neutron spectrum emitted by the fuel is modelled with the Maxwellian fission spectrum option of the MCNPX code(6). The doses (mSv h21), due to the neutrons and photons, were calculated using the F2 and Fm2 tallies coupled with the DE and DF cards. The F2 tally describes the neutron flux over a surface, whereas the D cards convert the absorbed dose to equivalent dose. Calculations were performed with 108 histories yielding accuracy on the dose calculations of ,1 %.
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Fuel case
NUCLEAR FUEL DOSIMETRY
RESULTS AND DISCUSSION
Table 3. Calculated annual doses from an unshielded fuel rod and a pellet at fabrication, at 1 m and 50 cm, respectively. Fuel case
R1 R2 R3 T1 T2 T3 T4 S1 S2 S3
Gamma dose (mSv)
Neutron dose (mSv)
1 fuel rod
1 pellet
1 fuel rod
1 pellet
0.034 5.1 10.9 10.7 53.9 338.6 0.02 5.1 36 49.7
0.00014 0.13 0.38 0.42 1.69 10.9 0.00048 0.2 0.64 1.53
0.00026 03 0.8 0.8 0.9 1.4 0.00164 0.6 103 135
0.0000024 0.1 0.3 0.3 0.3 0.4 0.00052 0.1 6.45 19.8
Figure 1. Shielding effect on the gamma spectrum emitted by the fuel rod in case S3.
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The annual occupational doses from the fuels for the transmutation and self-recycling schemes are compared with the commercial ‘once-through’ cycles (Table 3). The comparisons are performed at fabrication, on the basis of a pellet and a fuel rod of the different fuel materials considered in this study (Table 1). The doses are calculated at a distance of 50 cm and 1 m from a pellet and a fuel rod, respectively, at a point symmetrically placed in front of them. An annual occupational limit of 20 mSv is considered(14), in 2000 working hours per year(15). The occupational doses, due to either a pellet or a fuel rod, are increased considerably, in the recycling and transmutation schemes, owing to the inclusion of Pu and MA in the fuel. Hence, the gamma dose in the fuel cases (S2) and (T2, T3 and S3) is higher than R2 and R3, respectively. Furthermore, the inclusion of the Cm isotopes has increased the neutron dose in cases S2 and S3. The presence of 241Am (t12 ¼432.2 y) in particular and self-generated multiple-recycled Pu (fuel cases T3 T2, S3 and S2) render the highest doses (Table 3). The dominance of 241Am is demonstrated in fuel case T2, with the dose being four times higher than that of R3. This is due to the 1 % 241Am present in the former, despite the fact that both these LMFBRMOX fuels have nearly the same Pu content. Furthermore, the 20 % 241Am content in T3 renders the highest doses of all recycling cases with Pu and MA, by factors ranging from 6 to 50. The highest gamma dose occurs from material T3 with the highest amount in 241Am, followed by S3 (high amount of self-generated multiple-recycled Pu and 241Am), T2 (high amounts of Pu and 241Am) and S2 (low amount of self-generated multiple-recycled Pu and 241Am). Cases T1 and R3, without any 241Am, are comparable on the basis of their Pu content. Fuel cases R1 and T4, without Pu and 241Am, render doses lower by at least two orders of magnitude.
Besides 241Am, the self-generated multiple recycling schemes show a higher dose than their corresponding ‘once-through’ cases (S2 over R2 and S3 over R3) owing to their different Pu isotopic composition. Multiple recycling of self-generated Pu in a PWR results in a composition of Pu where the even isotopes have been significantly increased. The increase in 238Pu (t12 ¼87.74 y) by 3 times contributes significantly to both gamma and neutron doses. The soft gamma rays from the Pu isotopes and 241Am, which are the main contributors to the dose, can be easily shielded by 1-mm lead (Pb). The handling of a single PWR or LMFBR fuel rod at fabrication encounters doses which exceed, in fuel cases T2, T3, S2 and S3, the occupational dose limit by factors up to 30 (Table 3). Cases R3 and T1 would exceed the dose limit if two rods are adjacent, whereas four adjacent rods would be required in fuel cases R2 and S1. However, the soft gamma spectrum is again easily shielded by thin Pb (2–4 mm). The occupational neutron dose limit is exceeded in the case of single fuel rods from fuel cases S2 and S3 owing to their Cm content. This would demand a shielding of polyethylene up to 15 cm. In all other cases with Pu, 13–30 rods are needed in order to exceed the occupational limit. The effect of self-shielding by the fuel material and the shielding by the cladding as well as of the shielding by adjacent rods is shown in Figure 1, for the fuel case S3. The gamma spectrum emitted by the fuel rod is considered, under the assumption that there is no absorption within the fuel material, the cladding (void/ void) and adjacent rods. The self-absorption of the lower gamma energies is clearly indicated when only cladding and U/cladding are included in the calculations (void/cladding and U/cladding). The shielding by the presence of 1 or 2 rods in front of S3 further reduces the low gamma energies (þ1 rod and þ2 rods). The outcome is of particular importance considering
G. NICOLAOU
CONCLUSIONS The annual occupational radiation dose levels associated with the handling, at fabrication, of emerging nuclear fuels has been studied. The inclusion of Pu and MA for their single and multiple recycling in PWR and LMFBR has resulted in elevated gamma and neutron doses in comparison with commercial fuels. On the basis of a single fuel rod, the occupational dose limit exceeded in all fuel cases with 241Am and Cm isotopes in their charge composition. In the absence of 241Am, 2 –4 fuel rods are required in the recycling and transmutation schemes to exceed the limit. The use of freshly purified plutonium will give rise to a spectrum dominated by soft gamma rays, which is easily shielded by 1- 2-mm Pb, whereas the growth of 208TI is avoided. Self-shielding within a fuel rod reduces significantly only the gamma dose that would have been delivered otherwise. Hence, only the first row of the fuel rods within an assembly contributes to the dose. On the contrary, all fuel rods in the assembly contribute to the dose owing to neutrons. The handling of a single emerging nuclear fuel rod would require a shielding of Pb up to 4 mm and of polyethylene up to 15 cm depending on the amount of Pu, Am and Cm. In a production line, with several fuel pellets and rods been handled to fabricate rods and assemblies, fissile material mass balance and the geometrical setup of their handling(16) would influence the
occupational dose and hence the necessity of any shielding. Hence, on safety grounds, besides criticality prevention measures, radiation protection could be envisaged during powder preparation, manufacture and non-destructive testing of pellets, fuel rods and assemblies, and MOX scrap recovery(17). REFERENCES 1. Salvatores, M. Transmutation: issues, innovative options and perspectives. Prog. Nucl. Ener. 40, 375 (2002). 2. Nishihara, K., Nakayama, S., Morita, Y., Oigawa, H. and Iwasaki, T. Impact of partitioning and transmutation on LWR high-level waste disposal. J. Nucl. Sci. Techn. 45, 84 (2008). 3. Gonza´lez-Romero, E. M. Impact of partitioning and transmutation on the high level waste management. Nucl. Engin. Des. 241, 3436 (2011). 4. Salvatores, M. and Palmiotti, G. Radioactive waste partitioning and transmutation within advanced fuel cycles: achievements and challenges. Prog. Part. Nucl. Phys. 66, 144 (2011). 5. Pillon, S. Actinide-bearing fuels and transmutation targets. Comp. Nucl. Mater. 3, 109 (2012). 6. Pelowitz, D. B. MCNPXTM user’s manual Version 2.5.0, Oak Ridge National Laboratory (2005). 7. Koch, L. and Nicolaou, G. Comparison of possible partitioning and transmutation schemes when added to the existing nuclear fuel cycles. In: Proceedings Meeting of the Technical Committee to IAEA on Safety and Environmental Aspects of Partitioning and Transmutation of Actinides and Fission Products, IAEA. 29 November– 2 December 1993, p. 195– 201, IAEA-TECDOC-783 (1993). 8. SCALE: A comprehensive modeling and simulation suite for nuclear safety analysis and design, ORNL/TM-2005/ 39, version 6.1. Oak Ridge National Laboratory (2011). 9. Rose, P. F. ENDF-201, ENDF/B-VI Summary Documentation, fourth edn. Brookhaven National Lab.BNLNCS-17541 (1991). 10. Croff, A. G. ORIGEN2: a versatile computer code for calculating the nuclide compositions and characteristics of nuclear materials. Nucl. Technol. 62, 335–352 (1983). 11. ENDF/B-V Library. National Nuclear Data Centre, Brookhaven National Laboratory, Upton NY (1979). 12. Knief, R. E. Nuclear Engineering: Theory and Technology of Commercial Nuclear Power, second edn. Taylor & Francis (1992). 13. Todreas, N. E. and Kazimi, M. S. Nuclear Systems I: Thermal Hydraulic Fundamentals, second edn. Taylor & Francis (1993). 14. Wrixon, A. D. New ICRP recommendations. J. Radiol. Prot. 28, 161 (2008). 15. NUREG. Standard review plan for the review of an application for a Mixed Oxide (MOX) fuel fabrication facility. NUREG-1718, US Nuclear Regulatory Commission (2000). 16. IAEA. Safety of uranium fuel fabrication facilities specific safety guide. IAEA Safety Standards Series No. SSG-6, International Atomic Energy Agency Vienna (2010). 17. IAEA. Experiences and trends of manufacturing technology of advanced nuclear fuels. IAEA-TECDOC 1686, International Atomic Energy Agency Vienna (2012).
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that the number of low-energy gammas is 3 orders of magnitude higher than the other energies. The neutrons arriving at the point of interest are reduced by 3 % owing to self-shielding with the fuel rod. Fuel rods, prior to their loading into a reactor, are assembled together to form a nuclear fuel assembly. Modelling with MCNPX has indicated that the first row of fuel rods contributes to 99 % of the gamma dose at 1 m from the assembly. The self-absorption within the assembly results in a gamma dose of 15 times lower than the one with each row of fuel rods contributing equally. Regarding the neutron emission by the assembly and its contribution to the neutron dose, self-absorption within the assembly is less significant now. Hence, there is a contribution by all fuel rod rows, whereas each one shields the one behind by 3 %. The above-mentioned analyses have been performed at 1 y after reprocessing. Hence, the build-up of the thallium isotope 208Tl (t12 ¼3 min, 583, 860 and 2.6 MeV photons) in large amounts, from the decay of 236Pu (t12 ¼2.6 y), was avoided. For longer times, 208 TI builds up significantly and may eventually require the use of further Pb as shielding. Over a period of 10 y after reprocessing, the 208TI build-up would increase the gamma dose at 1 m from a single fuel rod by a factor of 7.
doi:10.1093/rpd/ncu016
RADIATION DOSE ASPECTS IN THE HANDLING OF EMERGING NUCLEAR FUELS G. Nicolaou* Laboratory of Nuclear Technology, Department of Electrical and Computer Engineering, School of Engineering, Demokritus University of Thrace, Xanthi 67100, Greece *Corresponding author: [email protected]
The occupational annual dose levels, encountered at fabrication of emerging nuclear fuels, have been studied. Emerging fuels for the single and multiple recycling of Pu and MA have resulted in considerably higher gamma and neutron doses in comparison with commercial fuels. The occupational dose limit is exceeded at fabrication by a single fuel rod in all fuel cases with 241Am and Cm isotopes present in their composition. In the absence of these isotopes, 2 – 4 adjacent fuel rods are sufficient to exceed the limit. Self-shielding within the fuel reduces significantly only the gamma dose that would have been delivered otherwise. Hence, only the first row of fuel rods in an assembly contributes to the dose, whereas in the case of neutrons, all fuel rods contribute.
INTRODUCTION
MATERIALS AND METHODS
The management of nuclear waste, whether as spent nuclear fuel discharged from reactors or high-level waste from the reprocessing of this fuel, is confronted with the radiotoxicity potential owing to transuranium (TRU) isotopes in the waste. Hence, the possibility of partitioning TRU isotopes out of the waste and transmuting them into less hazardous isotopes, relying on existing nuclear power stations, has been proposed and studied(1 – 5). In this context, ‘oncethrough cycle’ or ‘self-generated’ recycling strategies have emerged as potential partitioning and transmutation (P&T) fuel cycles of minor actinides (MA), namely neptunium (Np), americium (Am) and curium (Cm), and plutonium (Pu). The inclusion of Pu and MA in a uranium oxide matrix fuel, for their subsequent recycling, is expected to increase the radiation dose levels associated with such fuels at fabrication. Knowledge of these dose levels would allow measures to be conceived for the handling of such fuels during fabrication, packaging, delivery and loading to reactors. In this work, occupational radiation dose levels, due to the gamma rays and neutrons emitted by single rods of such fuels, have been calculated for different single and multiple recycling schemes. They are subsequently compared with the dose levels encountered at fabrication of fuels in existing commercial fuel cycles. The effect on the dose, owing to self-shielding effects within the fuel itself, attenuation in the cladding and screening between the fuels, is considered in the calculations and assessed. The necessary calculations have been performed using the Monte Carlo N-Particle code MCNPX(6).
Nuclear fuels considered Emerging nuclear fuels in the nuclear reactors pressurised water reactor (PWR) and liquid metal fast breeder reactor (LMFBR) have been considered, within the nuclear waste management option of the P&T of TRU. Existing commercial fuels, in the same reactor types, have been included in the study for comparison purposes. The following are the compositions of the fuels, ‘at fabrication’, considered (Table 1): (1) Transmutation schemes: mixed oxide (MOX) fuels containing homogeneously mixed low concentrations of either 237Np or 241Am (cases T1 and T2) for their homogeneous transmutation in LMFBR, and uranium (U) matrix fuels with large concentrations of 237Np and 241Am for their heterogeneous transmutation in LMFBR (T3 and T4); (2) Self-generated recycling schemes: self-generated Pu and TRU recycle in PWR (S1 and S2), and self-generated TRU recycle in LMFBR (S3); (3) Commercial fuels: 3.5 % of 235U-enriched U fuel and MOX UPu in PWR (R1 and R2), and MOX UPu in LMFBR (R3). The plutonium, used in the PWR- and LMFBRMOX fuel cases R2, R3, T1 and T2, was a 1st-generation Pu from the PWR-U case R1 (55 GWd tU21). The isotopic composition (wt %) of Pu retrieved from the reprocessing of a spent fuel with 1.5 y of cooling time is 238/239/240/241/242 of 3.2/48/24/15.2/ 9.6. In the self-generated recycling schemes (cases S1 –S3), the Pu and TRU compositions in the fuels, at the fabrication stage, are the ones after the 5th
# The Author 2014. Published by Oxford University Press. All rights reserved. For Permissions, please email: [email protected]
Downloaded from http://rpd.oxfordjournals.org/ at Main Library of Gazi University on May 6, 2015
Received 2 September 2013; revised 11 December 2013; accepted 29 January 2014
G. NICOLAOU Table 1. Commercial and emerging fuel compositions at fabrication. Reactor
Fuel type
Fuel charge compositiona
R1 R2 R3 T1 T2 T3 T4 S1 S2 S3
PWR PWR LMFBR LMFBR LMFBR LMFBR LMFBR PWR PWR LMFBR
UO2 MOX MOX MOX with 1 % Np MOX with 1 % Am U with Np and Am U with Np self-generated Pu recycle self-generated TU recycle self-generated TU recycle
UO2 U0.96Pu 0.04 U0.80Pu0.20 U0.795Pu0.195Np0.01 U0.795Pu0.195Am0.01 U0.60Np0.20Am0.20 U0.55Np0.45 U0.96Pu0.04 U0.95Pu0.04Am0.003Cm0.002Np0.005 U0.8Pu0.186Am0.0067Cm0.0058Np0.0015
Case R1—235U 3.5 % enriched; all other fuel cases with depleted U (0.2 % generation Pu from PWR.
a
recycling step (cases S1 and S2) or the 16th recycling step (case S3). These recycling steps correspond to the periods required in order to achieve ‘equilibrium’ in the composition of the recycled nuclides of interest in the fuel(7). At this stage, the amount of nuclides being formed equals the one recycled during irradiation. The Pu and MA, used in the recycling and transmutation schemes, were retrieved from reprocessing carried out on spent fuel having a cooling time of 1.5 y. Fuel fabrication took place 1 y after the retrieval of Pu and MA, avoiding the build-up of 208Tl in large amounts. Computational methodology The required 1st-generation Pu and the ‘equilibrium’ Pu and TRU isotopic compositions, after the 5th and 16th recycle steps, were obtained through depletion calculations. The calculations for the PWR fuel cycles were carried out using the module ORIGEN-ARP, which is a zero-dimensional isotope generation and depletion code within SCALE 6.1(8), employing nuclear data from the ENDF/B-VI library(9). Since an LMFBR model is not included in ORIGEN-ARP, the ORIGEN-2.2(10) depletion code coupled with nuclear data libraries from ENDF/B-V(11) was used for the calculations based on the LMFBR. The depletion calculations were carried out with target burnup values of 55 and 110 GWd tU21 for the PWR and LMFBR cases, respectively. Photon spectra with energies up to 2.75 MeV and the number of neutrons emitted by the different fuels at fabrication are obtained from these depletion calculations. The photon spectra and neutron emission rates are then used as input to the MCNPX code in order to describe the fuel as the radiation source term in the required dosimetry calculations. Dosimetry calculations have been performed using the general purpose 3-D MCNPX code coupled with the ENDF/B-VI nuclear data. The physical description
235
U); cases R2, R3, T1 and T2 with 1st-
Table 2. Physical description of the fuel rods and assemblies considered. PWR Assembly type Fuel rod Height (m) Active height (m) Active mass (g) Pellet diameter (mm) Pellet height (cm) Pellet mass (g) Clad material Clad thickness (mm) Fuel rod pitch (mm)
LMFBR
Square 17` 17 rods
hexagonal 271 rods
4 3.81 2961 9.5
2.7 1.6 995 8.5
1 7.8 Zr 0.57 12.6
1 6.2 Stainless steel 0.7 9.7
of the fuel rods and assemblies for the purpose of this study are given in Table 2(12, 13). The calculations were performed on the basis of the composition and certain physical properties (e.g. density and size) of the fuel material and the cladding of the fuel rod. Hence, self-shielding effects within the fuel rod and screening between the fuels are considered in the calculations. The neutron spectrum emitted by the fuel is modelled with the Maxwellian fission spectrum option of the MCNPX code(6). The doses (mSv h21), due to the neutrons and photons, were calculated using the F2 and Fm2 tallies coupled with the DE and DF cards. The F2 tally describes the neutron flux over a surface, whereas the D cards convert the absorbed dose to equivalent dose. Calculations were performed with 108 histories yielding accuracy on the dose calculations of ,1 %.
460
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Fuel case
NUCLEAR FUEL DOSIMETRY
RESULTS AND DISCUSSION
Table 3. Calculated annual doses from an unshielded fuel rod and a pellet at fabrication, at 1 m and 50 cm, respectively. Fuel case
R1 R2 R3 T1 T2 T3 T4 S1 S2 S3
Gamma dose (mSv)
Neutron dose (mSv)
1 fuel rod
1 pellet
1 fuel rod
1 pellet
0.034 5.1 10.9 10.7 53.9 338.6 0.02 5.1 36 49.7
0.00014 0.13 0.38 0.42 1.69 10.9 0.00048 0.2 0.64 1.53
0.00026 03 0.8 0.8 0.9 1.4 0.00164 0.6 103 135
0.0000024 0.1 0.3 0.3 0.3 0.4 0.00052 0.1 6.45 19.8
Figure 1. Shielding effect on the gamma spectrum emitted by the fuel rod in case S3.
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The annual occupational doses from the fuels for the transmutation and self-recycling schemes are compared with the commercial ‘once-through’ cycles (Table 3). The comparisons are performed at fabrication, on the basis of a pellet and a fuel rod of the different fuel materials considered in this study (Table 1). The doses are calculated at a distance of 50 cm and 1 m from a pellet and a fuel rod, respectively, at a point symmetrically placed in front of them. An annual occupational limit of 20 mSv is considered(14), in 2000 working hours per year(15). The occupational doses, due to either a pellet or a fuel rod, are increased considerably, in the recycling and transmutation schemes, owing to the inclusion of Pu and MA in the fuel. Hence, the gamma dose in the fuel cases (S2) and (T2, T3 and S3) is higher than R2 and R3, respectively. Furthermore, the inclusion of the Cm isotopes has increased the neutron dose in cases S2 and S3. The presence of 241Am (t12 ¼432.2 y) in particular and self-generated multiple-recycled Pu (fuel cases T3 T2, S3 and S2) render the highest doses (Table 3). The dominance of 241Am is demonstrated in fuel case T2, with the dose being four times higher than that of R3. This is due to the 1 % 241Am present in the former, despite the fact that both these LMFBRMOX fuels have nearly the same Pu content. Furthermore, the 20 % 241Am content in T3 renders the highest doses of all recycling cases with Pu and MA, by factors ranging from 6 to 50. The highest gamma dose occurs from material T3 with the highest amount in 241Am, followed by S3 (high amount of self-generated multiple-recycled Pu and 241Am), T2 (high amounts of Pu and 241Am) and S2 (low amount of self-generated multiple-recycled Pu and 241Am). Cases T1 and R3, without any 241Am, are comparable on the basis of their Pu content. Fuel cases R1 and T4, without Pu and 241Am, render doses lower by at least two orders of magnitude.
Besides 241Am, the self-generated multiple recycling schemes show a higher dose than their corresponding ‘once-through’ cases (S2 over R2 and S3 over R3) owing to their different Pu isotopic composition. Multiple recycling of self-generated Pu in a PWR results in a composition of Pu where the even isotopes have been significantly increased. The increase in 238Pu (t12 ¼87.74 y) by 3 times contributes significantly to both gamma and neutron doses. The soft gamma rays from the Pu isotopes and 241Am, which are the main contributors to the dose, can be easily shielded by 1-mm lead (Pb). The handling of a single PWR or LMFBR fuel rod at fabrication encounters doses which exceed, in fuel cases T2, T3, S2 and S3, the occupational dose limit by factors up to 30 (Table 3). Cases R3 and T1 would exceed the dose limit if two rods are adjacent, whereas four adjacent rods would be required in fuel cases R2 and S1. However, the soft gamma spectrum is again easily shielded by thin Pb (2–4 mm). The occupational neutron dose limit is exceeded in the case of single fuel rods from fuel cases S2 and S3 owing to their Cm content. This would demand a shielding of polyethylene up to 15 cm. In all other cases with Pu, 13–30 rods are needed in order to exceed the occupational limit. The effect of self-shielding by the fuel material and the shielding by the cladding as well as of the shielding by adjacent rods is shown in Figure 1, for the fuel case S3. The gamma spectrum emitted by the fuel rod is considered, under the assumption that there is no absorption within the fuel material, the cladding (void/ void) and adjacent rods. The self-absorption of the lower gamma energies is clearly indicated when only cladding and U/cladding are included in the calculations (void/cladding and U/cladding). The shielding by the presence of 1 or 2 rods in front of S3 further reduces the low gamma energies (þ1 rod and þ2 rods). The outcome is of particular importance considering
G. NICOLAOU
CONCLUSIONS The annual occupational radiation dose levels associated with the handling, at fabrication, of emerging nuclear fuels has been studied. The inclusion of Pu and MA for their single and multiple recycling in PWR and LMFBR has resulted in elevated gamma and neutron doses in comparison with commercial fuels. On the basis of a single fuel rod, the occupational dose limit exceeded in all fuel cases with 241Am and Cm isotopes in their charge composition. In the absence of 241Am, 2 –4 fuel rods are required in the recycling and transmutation schemes to exceed the limit. The use of freshly purified plutonium will give rise to a spectrum dominated by soft gamma rays, which is easily shielded by 1- 2-mm Pb, whereas the growth of 208TI is avoided. Self-shielding within a fuel rod reduces significantly only the gamma dose that would have been delivered otherwise. Hence, only the first row of the fuel rods within an assembly contributes to the dose. On the contrary, all fuel rods in the assembly contribute to the dose owing to neutrons. The handling of a single emerging nuclear fuel rod would require a shielding of Pb up to 4 mm and of polyethylene up to 15 cm depending on the amount of Pu, Am and Cm. In a production line, with several fuel pellets and rods been handled to fabricate rods and assemblies, fissile material mass balance and the geometrical setup of their handling(16) would influence the
occupational dose and hence the necessity of any shielding. Hence, on safety grounds, besides criticality prevention measures, radiation protection could be envisaged during powder preparation, manufacture and non-destructive testing of pellets, fuel rods and assemblies, and MOX scrap recovery(17). REFERENCES 1. Salvatores, M. Transmutation: issues, innovative options and perspectives. Prog. Nucl. Ener. 40, 375 (2002). 2. Nishihara, K., Nakayama, S., Morita, Y., Oigawa, H. and Iwasaki, T. Impact of partitioning and transmutation on LWR high-level waste disposal. J. Nucl. Sci. Techn. 45, 84 (2008). 3. Gonza´lez-Romero, E. M. Impact of partitioning and transmutation on the high level waste management. Nucl. Engin. Des. 241, 3436 (2011). 4. Salvatores, M. and Palmiotti, G. Radioactive waste partitioning and transmutation within advanced fuel cycles: achievements and challenges. Prog. Part. Nucl. Phys. 66, 144 (2011). 5. Pillon, S. Actinide-bearing fuels and transmutation targets. Comp. Nucl. Mater. 3, 109 (2012). 6. Pelowitz, D. B. MCNPXTM user’s manual Version 2.5.0, Oak Ridge National Laboratory (2005). 7. Koch, L. and Nicolaou, G. Comparison of possible partitioning and transmutation schemes when added to the existing nuclear fuel cycles. In: Proceedings Meeting of the Technical Committee to IAEA on Safety and Environmental Aspects of Partitioning and Transmutation of Actinides and Fission Products, IAEA. 29 November– 2 December 1993, p. 195– 201, IAEA-TECDOC-783 (1993). 8. SCALE: A comprehensive modeling and simulation suite for nuclear safety analysis and design, ORNL/TM-2005/ 39, version 6.1. Oak Ridge National Laboratory (2011). 9. Rose, P. F. ENDF-201, ENDF/B-VI Summary Documentation, fourth edn. Brookhaven National Lab.BNLNCS-17541 (1991). 10. Croff, A. G. ORIGEN2: a versatile computer code for calculating the nuclide compositions and characteristics of nuclear materials. Nucl. Technol. 62, 335–352 (1983). 11. ENDF/B-V Library. National Nuclear Data Centre, Brookhaven National Laboratory, Upton NY (1979). 12. Knief, R. E. Nuclear Engineering: Theory and Technology of Commercial Nuclear Power, second edn. Taylor & Francis (1992). 13. Todreas, N. E. and Kazimi, M. S. Nuclear Systems I: Thermal Hydraulic Fundamentals, second edn. Taylor & Francis (1993). 14. Wrixon, A. D. New ICRP recommendations. J. Radiol. Prot. 28, 161 (2008). 15. NUREG. Standard review plan for the review of an application for a Mixed Oxide (MOX) fuel fabrication facility. NUREG-1718, US Nuclear Regulatory Commission (2000). 16. IAEA. Safety of uranium fuel fabrication facilities specific safety guide. IAEA Safety Standards Series No. SSG-6, International Atomic Energy Agency Vienna (2010). 17. IAEA. Experiences and trends of manufacturing technology of advanced nuclear fuels. IAEA-TECDOC 1686, International Atomic Energy Agency Vienna (2012).
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that the number of low-energy gammas is 3 orders of magnitude higher than the other energies. The neutrons arriving at the point of interest are reduced by 3 % owing to self-shielding with the fuel rod. Fuel rods, prior to their loading into a reactor, are assembled together to form a nuclear fuel assembly. Modelling with MCNPX has indicated that the first row of fuel rods contributes to 99 % of the gamma dose at 1 m from the assembly. The self-absorption within the assembly results in a gamma dose of 15 times lower than the one with each row of fuel rods contributing equally. Regarding the neutron emission by the assembly and its contribution to the neutron dose, self-absorption within the assembly is less significant now. Hence, there is a contribution by all fuel rod rows, whereas each one shields the one behind by 3 %. The above-mentioned analyses have been performed at 1 y after reprocessing. Hence, the build-up of the thallium isotope 208Tl (t12 ¼3 min, 583, 860 and 2.6 MeV photons) in large amounts, from the decay of 236Pu (t12 ¼2.6 y), was avoided. For longer times, 208 TI builds up significantly and may eventually require the use of further Pb as shielding. Over a period of 10 y after reprocessing, the 208TI build-up would increase the gamma dose at 1 m from a single fuel rod by a factor of 7.
