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Monte Carlo studies in accelerator-driven systems for transmutation of high-level nuclear waste
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Energy Conversion and Management 49 (2008) 1966–1971 www.elsevier.com/locate/enconman
Monte Carlo studies in accelerator-driven systems for transmutation of high-level nuclear waste Basßar S ß arer a,*, M. Emin Korkmaz a, Mehtap Gu¨nay b, Abdullah Aydın c a
¨ niversitesi Fen-Edebiyat Faku¨ltesi Fizik Bo¨lu¨mu¨, Besßevler, Ankara, Turkey Gazi U b _ ¨ niversitesi Fen-Edebiyat Faku¨ltesi Fizik Bo¨lu¨mu¨, Malatya, Turkey Ino¨nu¨ U c ¨ niversitesi Fen-Edebiyat Faku¨ltesi Fizik Bo¨u¨mu¨, Kırıkkale, Turkey Kırıkkale U Available online 7 March 2008
Abstract A spallation neutron source was modeled using a high energy proton accelerator for transmutation of 239Pu, minor actinides 237Np, Am and long-lived fission products 99Tc, 129I, which are created from the operation of nuclear power reactors for the production of electricity. The acceleration driven system (ADS) is composed of a natural lead target, beam window, subcritical core, reflector, and structural material. The neutrons are produced by the spallation reaction of protons from a high intensity linear accelerator in the spallation target, and the fission reaction in the core. It is used a hexagonal lattice for the waste and fuel assemblies. The system is driven by a 1 GeV, 10 mA proton beam incident on a natural lead cylindrical target. The protons were uniformly distributed across the beam. The core is a cylindrical assembly. The main vessel is surrounded by a reflector made of graphite. The axes of the proton beam and the target are concentric with the main vessel axis. The structural walls and the beam window are made of the same material, stainless steel, HT9. We investigated the following neutronics parameters: spallation neutron and proton yields, spatial and energy distribution of the spallation neutrons, and protons, heat deposition, and the production rates of hydrogen and helium, transmutation rate of minor actinides and fission products. In the calculations, the Monte Carlo code MCNPX, which is a combination of LAHET and MCNP, was used. To transport a wide variety of particles, The Los Alamos High Energy Transport Code (LAHET) was used. Ó 2007 Elsevier Ltd. All rights reserved. 241
Keywords: Transmutation; Nuclear waste; ADS; Subcritical reactor
1. Introduction For the last 50 years, the nuclear systems have been producing increasing amounts of highly radioactive waste. The spent fuel consists of a wide variety of elements and isotopes. The majority of the long-lived isotopes come from only a few transuranic elements plutonium, neptunium, americium, curium and some fission products such as technetium and iodine. Fresh light water reactor (LWR) fuel consists of uranium oxide, UO2. About 95–97% of the uranium composition is 238U and 3–5% fissionable 235U. As the fuel in a *
Corresponding author. Tel./fax: +90 312 202 12 37. E-mail address: [email protected] (B. S ß arer).
0196-8904/$ - see front matter Ó 2007 Elsevier Ltd. All rights reserved. doi:10.1016/j.enconman.2007.09.029
reactor is burnt out, its composition changes. Every year a LWR, 1 GWh (e), discharges about 21 tons radioactive fuel with the following inventory: 20 tons of uranium containing 0.9% (180 kg) 235U, 200 kg of plutonium; 21 kg of minor actinides: 10 kg of neptunium, 10 kg americium, 1 kg curium: 760 kg fission products: 18 kg of 99Tc, 16 kg of 93Zr, 9 kg of 135Cs, 5 kg of 107Pd, and 3 kg 129I, which are long-lived elements [1]. The spent nuclear fuel is moved to a pool of water, where isotopes with short half-lives decay to safer levels and internal heat generation drops. The second step is to isolate the spent nuclear fuel from the biosphere by placing it in a geological repository. However, there are still the issues about the volume of waste and how to minimize release of radioactive material to the environment.
B. S ß arer et al. / Energy Conversion and Management 49 (2008) 1966–1971
Transmutation of plutonium and minor actinides and long-lived fission products is a promising concept to reduce the radioactive waste and its long-term radiotoxicity. Partitioning and transmutation are considered as ways of reducing the burden on a geological disposal. Since plutonium and the minor actinides are mainly responsible for the long-term radiotoxicity, when these nuclides are removed from the waste (partitioning) and fissioned (transmutation), the remaining waste loses most of its long-term radiotoxicity [2]. Transmutation is the process of bombarding a material with particles to form new atoms with higher masses and/ or to fission the material into atoms with smaller masses. It can reduce the mass, volume, activity, heatload, and/or raditoxicity of waste that must be sent to repository [3]. Many different technologies have been examined for transmutation, including a variety of reactors and ADSs. These systems are primarily distinguished by whether they have fast or thermal neutron spectrums. Transuranics transmutation can be planned in both thermal and fast reactors. However, a critical reactor with solid fuel and liquid coolant , thermal and fast, in which neutron production and neutron losses are in balance, may contain a quite limited amount of particular TRU mixture components like Np, Am or Cm isotopes, due to safety constraints. This is one of the mains reasons why the accelerator-driven systems are currently studied worldwide for nuclear waste burning. Accelerator-driven subcritical reactors have been proposed for many applications such as energy production, fertile-to-fissile transmutation and conversion of long-lived radioisotopes into stable or much shorter-lived isotopes [4– 7]. The transmutation technology to incinerate the longlived radioactive isotopes using an accelerator-driven subcritical reactor is one of the best solutions. ADSs can be designed to have a fast neutron energy spectrum, and they are subcritical so safety issues can be considered less significant than in a fast reactor. Another advantage of ADSs is that they can burn mixtures of material that would not maintain criticality in a reactor [8–14]. 2. The system Neutrons are produced by the spallation reaction of protons from a high intensity linear accelerator in the spallation target and the fission reaction in the core. It is used a hexagonal lattice for the waste and fuel assemblies. The system (Fig. 1) is driven by a 1 GeV, 10 mA proton beam incident on a natural lead cylindrical target, 20 cm radius, 70 cm height, and entering the target through a 5.3 cm radius hole. The protons were uniformly distributed across the beam of radius 3 cm. The core is cylindrical assembly, 2.3 m radius, 4.6 m high. The wall thickness of the main vessel is 2 cm. The main vessel is surrounded by a reflector made of graphite, 40 cm thick. The axes of proton beam and the target are concentric with the main vessel axis. The structural walls and beam window are made of the same material, stainless steel, HT9.
1967
Fig. 1. Horizontal and vertical sectional views of the present ADS design.
The vertical proton beam pipe, 5.3 cm outer radius and 5.0 cm inner radius, is inserted from the top into the main vessel down to the spallation target region. The hemispherical bottom end of the tube forms the beam window. The major reasons to choose sodium as coolant are its excellent thermal properties and good compatibility with stainless steel. Most breeder reactors use a hexagonal lattice for the fuel structure, to be consistent with breeder reactor designs, we used a hexagonal lattice for the fuel assemblies. 3. Neutronic analysis In the calculations, Monte Carlo code MCNPX [15], which is a combination of LAHET [16] and MCNP, was used. To transport a wide variety of particles, The Los Alamos High Energy Transport Code, LAHET, was used. These include charged particles such as protons and charged pions as well as neutrons with high energies.
1968
B. S ß arer et al. / Energy Conversion and Management 49 (2008) 1966–1971
MCNPX offers options based on three physics packages; the BERTINI [17–20] and ISABEL [21] models taken from the LAHET Code System, and the CEM package [22]. The MCNP transport code is used for radiation transport at low energies (<20 MeV). The transport of high energy particles (>150 MeV) was performed using nuclear models, while for the transport of low energy particles (<150 MeV) cross section libraries have been used. In an ADS, the energy is produced as a result of the multiplying nuclear cascades initiated by accelerated protons interacting with the spallation target. Spallation reactions in a target material proceed through three stages: In the first stage, an incident particle interacts with the nucleus through successive nucleon–nucleon hard collisions in a potential which describes the density of the nucleus as a function of radius. Intranuclear cascade (INC) model is used this stage, in which high energy particles are emitted. After the INC stage of the reaction the nucleus is left in an excited state. This is the starting point for the second or pre-equilibrium stage of the reaction. In the third stage, the evaporation phase of the reaction, the remaining nucleus is de-excited either by evaporation or by fission. We used Bertini model for nucleons and pions, and Isabel model for other particle types in the intranuclear cascade stage. Pre-equilibrium model has been used for the second stage of the reaction. The default physics module options were used with pre-equilibrium model after intranuclear cascade. The study was performed in three steps. In the first step, only the spallation target, proton beam pipe and beam window isolated from the core were considered. In the second step, all calculations have been made in absence of fissile materials and wastes. In the third step, the full reactor geometry was taken into account. In the calculations, we did not take time evaluation of the neutron spectrum and other quantities into account. The target was designed to maximize the neutron yield and to soften the axial power distribution. The target optimization means finding the maximum of neutron production inside the target and the maximum of neutron leakage to the core at given proton energy and target shape and material. Lead is a material highly transparent to neutrons; it has a very low capture cross section, the main isotope in natural lead 208Pb is double magic, and thus exceptionally stable, from the MeV region down to very low energies. But it has a moderate elastic neutron cross section. One of the fundamental parameters of the spallation target is the number of neutrons produced per beam particle incident on target. Calculation of number of neutrons produced in the target is done by summing up the absorption and leakage. Because of small neutron absorption of lead, the leakage is considered to be equal to the neutron production. Spallation neutron and proton yields per one incident proton are 27.4 and 1.2E 02, respectively. Other particle yields such as, pion and muon per one incident proton have been neglected due to their low particle fluxes.
The neutron leakage on the whole surface of the target and the main vessel for all three steps mentioned above is shown in Table 1. Some of the spallation neutrons produced by protons in the target can escape through the beam pipe, therefore it is desired to minimize the neutron escape through the beam pipe. The neutron leakage through some regions of the target top face (Fig. 2) for full system is shown in Table 2. The ratio of the neutron leakage into the proton beam vacuum to the total neutron leakage from the target top face is 1.54%. 3.1. Energy and spatial distributions The transmutation performance of a nuclear system mainly depends on flux level and neutron spectrum. To investigate the radial dependence for three calculation steps, neutron fluxes and heating were calculated in a set of coaxial cylinders, subdividing the target into parts and also many sections for z-dependence. The target was divided into 20 coaxial cylindrical sections and into 32 Table 1 The neutron leakage on the whole surface of the target and main vessel wall Neutron leakage (n/s)
Beam pipe + target
Beam pipe + target + coolant
Full system
Target Top face Side face Bottom face
3.25E+17 1.34E+18 4.02E+16
5.68E+17 2.78E+18 1.63E+17
1.44E+18 9.43E+18 1.03E+18
Main vessel Top face Side face Bottom Face
1.12E+14 – –
8.60E+15 4.12E+16 6.92E+15
8.73E+16 3.42E+17 8.15E+16
Fig. 2. Scheme of the target-beam pipe system.
Table 2 The neutron leakage on the target top face for full system Target top face
1. Region (n/s)
2. Region (n/s)
3. Region (n/s)
Total
1.43E+17
1.54E+16
1.28E+18
1.44E+18
B. S ß arer et al. / Energy Conversion and Management 49 (2008) 1966–1971
Neutron Flux (n/cm2.s)
10
15
10
13
10
11
Target
Beam pipe+target Beam pipe+target+coolant Full system
15
Neutron Flux (n/cm2.s)
8x10
15
7x10
15
6x10
15
5x10
15
4x10
15
3x10
15
2x10
15
1x10
14
1x10
-20
-15
-10
-5
0
5
10
15
20
r (cm)
Target
15
6x10
15
5x10
Beam pipe+target Beam pipe+target+coolant Full system
15
4x10
15
3x10
15
2x10
15
1x10
13
1x10
195
205
215
225
235
245
255
265
z (cm)
Fig. 4. The spatial variation of the neutron fluxes in the target: (a) radial, (b) axial.
The peak like a delta-function in the energy spectrum of protons in the beam window, Fig. 5, shows contribution of source protons. 3.2. Heat deposition
10
9
10
7
10
5
Beam pipe+target Beam pipe+target+coolant Full system
3
10 -10 10
10
-6
10
-2
10
2
Neutron Energy (MeV) 1012
Proton Flux (n/cm2.s)
Target
15
9x10
Neutron Flux (n/cm2.s)
horizontal sectors along the target axis. For the three systems, the neutron flux is low at energies lower than 1 MeV for target and beam window. The energy distributions of the spallation neutrons and protons in the target, the radial and the axial variations of the neutron flux in the target and energy distributions of the spallation neutrons and protons in the beam window are shown in Figs. 3–5, respectively. The energy distribution of the spallation neutrons are rather similar in the target and the spallation protons in the beam window for all three steps. In the target, the energy distribution for all three steps are the same both in shape and calculation values. The peaks of the neutron energy distributions appear at energy about 2 MeV for all three steps in target and the beam pipe. The both shapes and calculated values of the energy distribution of the spallation protons for all calculation steps in the whole energy range are the same in the target. But, in the beam window, the proton flux values for full system are some different than those of the other two systems in the energy range 100–1000 MeV. The neutron energy distributions of the systems beam pipe + target + coolant and full system for both target and beam window, are rather different than that of the beam pipe + target system. The neutron flux values of the beam pipe + target system begins from 1 eV and neutron flux values in the energy range 1 eV–2 MeV are substantially lower than those of the other two systems.
1969
Target
11
10
10
10
3.3. H and He production rates
9
10
Beam pipe+target Beam pipe+target+coolant Full system
108 7
10
100
The average heat deposition by neutrons in the some parts of the system has been calculated for the three calculation steps. The results are given in Table 3. The radial and axial variations of the heating in the target for three steps are shown in Fig. 6. It is observed that the amount of energy deposited in the target by neutrons is the mostly same for three steps. The average energy deposited by neutrons in the beam window, for each calculation steps, is greater about three times than in the case of target. There is a negligible energy deposition by neutrons in the other structural materials, beam pipe wall and especially main vessel wall.
101
102
103
Proton Energy (MeV)
Fig. 3. Energy distribution of the spallation neutrons and protons in the energy range 0–1000 MeV in the target.
LAHET writes each event into an output history file, HISTP. In this file each particle’s crossing of a geometric boundary as well as entry and exit coordinates (positions, angles and energies) are recorded in detail. Lahet also provides the postprocessing HTAPE code. The option IOPT 14, in HTAPE, provides an edit of hydrogen and helium gas production, by isotope, by element, and total. The
1970
B. S ß arer et al. / Energy Conversion and Management 49 (2008) 1966–1971
Neutron Flux (n/cm2.s)
1015
Table 4 Gas production in the target and beam window
Beam Window
1012
109
Beam pipe+target Beam pipe+target+coolant Full system 106 10-9
10-6
10-3
100
103
Target
Beam window
H H-2 H-3 Total H He-3 He-4 Total He
2798.21 304.1 181.67 3283.45 11.12 514.69 525.81
33,000 3770 810 37,200 565 3320 3880
Total H + He
3809.20
41,100
gas production rates in the target and beam window for full system are shown in Table 4. The results show that the gas production in the beam window is more intense than the target, as expected.
Neutron Energy (MeV) 15
10
Gas production (apmm/fpy)
Beam Window
14
2
Proton Flux (n/cm .s)
10
3.4. Transmutation
13
10
12
10
11
10
10
10
Beam pipe+target Beam pipe+target+coolant Full system
9
10
8
10 0 10
1
2
10
3
10
10
Proton Energy (MeV) Fig. 5. Energy distribution of the spallation neutrons and protons in energy range 0–1000 MeV in the beam window.
Table 3 Energy deposited by neutrons in the some parts of the system for three calculation steps Heating (W/cm3)
Beam pipe + target
Beam pipe + target + coolant
Full System
Target Beam window Beam pipe Main vessel wall
2.87 8.48 0.0 0.0
2.91 10.77 0.30 2.65E 04
3.19 9.61 0.49 2.14E 04
The core of the system mainly consists of the spallation target region and waste assembly region. The waste region consists of the pin-bundle type fuel assemblies and is cooled by liquid sodium. Fig. 1 shows the schematic view of the target and the fuel assembly. The fuel elements consist of a stainless steel cladding, 0.35 cm thick, characterized by an external radius of 1.35 cm and a total height of 120 cm, end cap included. The fuel is a cylinder, 119.30 cm height. As mentioned above, we used a hexagonal lattice for waste assemblies. The waste subassemblies, bundles, are arranged in an annular array of five rings, each waste subassembly contains seven fuel rods. There are 114 bundles in the system. Isotopes to be transmuted are minor actinides, such as 239Pu, 237Np, 241Am, and are long-lived fission products, such as 99Tc and 129I. The core configuration for the calculations is characterized by 10 bundles in ring 1 (239Pu); 20 bundles in ring 2 (237Np); 24 bundles in ring 3 (129I);
Target
3
Neutron Heating (W/cm )
Neutron Heating (W/cm3 )
Target 32.5 30.0 27.5 25.0 22.5 20.0 17.5 15.0 12.5 10.0 7.5 5.0 2.5 0.0 -20
Beam pipe+target Beam pipe+target+coolant Full system
-15
-10
-5
0
5
10
15
20
6.5 6.0 5.5 5.0 4.5 4.0 3.5 3.0 2.5 2.0 1.5 1.0 0.5 0.0
Beam pipe+target Beam pipe+target+coolant Full system
200
210
r (cm) Fig. 6. Radial and axial variation of the heating in the target.
220
230
z (cm)
240
250
260
B. S ß arer et al. / Energy Conversion and Management 49 (2008) 1966–1971 Table 5 Transmutation calculation results 99
129
237
239
241
910.434 1.455
182.509 0.806
1068.769 36.434
522.932 47.254
1082.229 4.212
Tc
Initial mass (kg) Total transmutation (kg/fpy)
I
Np
Pu
Am
30 bundles in ring 4 (99Tc); 30 bundles in ring 5 (241Am). All calculations are based on pure 99Tc,129I, 237Np,239Pu and 241Am isotopes. The neutron multiplication factor, keff,with its standard deviation was calculated by ‘‘kcode” card of the MCNP code and was found as, after running a large number of histories, 0.96646 ± 0.00119 for the core without the proton beam. Transuranic elements (TRU) and fission fragments (FF) are two main components of high-level nuclear waste, representing 1.1% and 4% of spent nuclear fuel, respectively. TRU, produced by neutron capture in the fuel can only be successfully incinerated by fission, while the radioactivitiy of the FF can be modified by neutron capture. Minor actinides should be transmuted mainly through fission reactions having the possibility increasing higher actinides, while the thermal capture is main transmutation reaction for long-lived fission productions. The main nuclides of minor actinides from power reactor, such as Np, Am and Cm, have threshold fission reaction, so, hard neutron spectrum is desirable. The capture reactions rates of LLFP increases in the thermal energy region of ADS is hard, so it is necessary to establish. The region with softer spectrum for transmutation of LLFP. However, the effect on the MA transmutation properties of ADS is significant. Results of transmutation calculation are given in Table 5. 4. Conclusions Accelerator-driven transmutation system was investigated. The system consists of 1 GeV, 10 mA proton accelerator and a subcritical core with an effective neutron multiplication factor of 0.96. For the neutronic analysis, we used the MCNPX code system. In the neutronic calculations, time evaluations have not been taken into account. The transmutation rates of minor actinides and long-lived
1971
fission products vary in the range 0.806–47.254 kg/fpy. The ADS proposed in this study produces energy from nuclear waste material, while a considerable amount of this spent fuel is incinerated. References [1] Garwin RL, Charpak G. Megawatts and megatons a turning point in the nuclear ages? New York: Alfred A. Knopf; 2001. [2] Salvatores M. Nuclear fuel cycle strategies including partitioning and transmutation. Nucl Eng Des 2005;235:805–16. [3] Westlen D. A cost benefit analysis of an accelerator driven transmutation system. Stockholm: Royal Institute of Technology; 2001. [4] Rubbia C, Rubio JA, Buono S, Carminati F, Fie´tier N, Galvez J, et al. CERN report, 1995, CERN/AT/95-44 (ET). [5] Van Atta CM, Lee JD, Heckrotte W. The electronuclear conversion of fertile to fissile material, Lawrance Livermore Laboratory Report, UCRL-52144, 1970. [6] Bowman CD, Arthur ED, Lisowski PW, Lawrence GP, Jensen RJ, Anderson JL, et al. Nucl Instrum Methods A 1992;320:336. [7] Salvatores M, Slessarev I, Tchistiakov A, Ritter G. Nucl Sci Eng 1997;126:333. [8] Rineiski A, Maschek W. Dynamics and reactivity control in accelerator driven systems for nuclear waste burning. Forschungszentrum (FZK), Institute for Nuclear and Energy Technologies, Karlsruhe, Nuclear Energy, 2006. [9] Vergnes J, Barbrault P, Lecarpentier D, et al. Limiting plutonium and minor actinides inventory: comparison between accelerator-driven system (ADS) and critical reactory. Proceedings of the international conference on future nuclear systems, Global’99, Jackson Hole, Wyoming. LaGrange Park, Illinois: American Nuclear Society; 1999. [10] Ait Abderrahim H et al. Nucl Instrum Methods A 2001;463:487–94. [11] Maschek W et al. J Nucl Mater 2003;320:147–55. [12] Haeck W et al. J Nucl Mater 2006;352:285–90. [13] Maschek W et al. Nucl Instrum Methods A 2006;562:863–6. [14] Haas D et al. Energy Convers Manage 2006;47:2724–31. [15] Waters LS, editor. MCNPX user’s manual version 2.3.0, April 2002 LA-UR-02-2607. [16] Prael RE, Lichtenstein H. User guide to LCS: the LAHET code system, Los Alamos National Laboratory Report LA-UR-89-3014, Revised September 15, 1989. [17] Bertini HW et al. Nucl Instrum Methods 1968;66. [18] Bertini HW. Intranuclear-cascade calculation of the secondary nucleon spectra from nucleon–nucleus interactions in the energy range 340–2900 MeV and comparisons with experiment. Phys Rev 1969;188:1711. [19] Boudard A, Cugnon J, Leray S, Volant C. Phys Rev C 2002;66:044615. [20] Junghans AR et al. Nucl Phys A 1998;629:635. [21] Yariv Y, Fraenkel Z. Phys Rev C 1981;24:48. [22] Mashnik SG, Toneev VD. MODEX – the program for calculation of the energy spectra of particles emitted in the reactions of preequilibrium and equilibrium statistical decays, Communication JINR P4-8417, Dubna, 1974, p. 25 [in Russian].
Energy Conversion and Management 49 (2008) 1966–1971 www.elsevier.com/locate/enconman
Monte Carlo studies in accelerator-driven systems for transmutation of high-level nuclear waste Basßar S ß arer a,*, M. Emin Korkmaz a, Mehtap Gu¨nay b, Abdullah Aydın c a
¨ niversitesi Fen-Edebiyat Faku¨ltesi Fizik Bo¨lu¨mu¨, Besßevler, Ankara, Turkey Gazi U b _ ¨ niversitesi Fen-Edebiyat Faku¨ltesi Fizik Bo¨lu¨mu¨, Malatya, Turkey Ino¨nu¨ U c ¨ niversitesi Fen-Edebiyat Faku¨ltesi Fizik Bo¨u¨mu¨, Kırıkkale, Turkey Kırıkkale U Available online 7 March 2008
Abstract A spallation neutron source was modeled using a high energy proton accelerator for transmutation of 239Pu, minor actinides 237Np, Am and long-lived fission products 99Tc, 129I, which are created from the operation of nuclear power reactors for the production of electricity. The acceleration driven system (ADS) is composed of a natural lead target, beam window, subcritical core, reflector, and structural material. The neutrons are produced by the spallation reaction of protons from a high intensity linear accelerator in the spallation target, and the fission reaction in the core. It is used a hexagonal lattice for the waste and fuel assemblies. The system is driven by a 1 GeV, 10 mA proton beam incident on a natural lead cylindrical target. The protons were uniformly distributed across the beam. The core is a cylindrical assembly. The main vessel is surrounded by a reflector made of graphite. The axes of the proton beam and the target are concentric with the main vessel axis. The structural walls and the beam window are made of the same material, stainless steel, HT9. We investigated the following neutronics parameters: spallation neutron and proton yields, spatial and energy distribution of the spallation neutrons, and protons, heat deposition, and the production rates of hydrogen and helium, transmutation rate of minor actinides and fission products. In the calculations, the Monte Carlo code MCNPX, which is a combination of LAHET and MCNP, was used. To transport a wide variety of particles, The Los Alamos High Energy Transport Code (LAHET) was used. Ó 2007 Elsevier Ltd. All rights reserved. 241
Keywords: Transmutation; Nuclear waste; ADS; Subcritical reactor
1. Introduction For the last 50 years, the nuclear systems have been producing increasing amounts of highly radioactive waste. The spent fuel consists of a wide variety of elements and isotopes. The majority of the long-lived isotopes come from only a few transuranic elements plutonium, neptunium, americium, curium and some fission products such as technetium and iodine. Fresh light water reactor (LWR) fuel consists of uranium oxide, UO2. About 95–97% of the uranium composition is 238U and 3–5% fissionable 235U. As the fuel in a *
Corresponding author. Tel./fax: +90 312 202 12 37. E-mail address: [email protected] (B. S ß arer).
0196-8904/$ - see front matter Ó 2007 Elsevier Ltd. All rights reserved. doi:10.1016/j.enconman.2007.09.029
reactor is burnt out, its composition changes. Every year a LWR, 1 GWh (e), discharges about 21 tons radioactive fuel with the following inventory: 20 tons of uranium containing 0.9% (180 kg) 235U, 200 kg of plutonium; 21 kg of minor actinides: 10 kg of neptunium, 10 kg americium, 1 kg curium: 760 kg fission products: 18 kg of 99Tc, 16 kg of 93Zr, 9 kg of 135Cs, 5 kg of 107Pd, and 3 kg 129I, which are long-lived elements [1]. The spent nuclear fuel is moved to a pool of water, where isotopes with short half-lives decay to safer levels and internal heat generation drops. The second step is to isolate the spent nuclear fuel from the biosphere by placing it in a geological repository. However, there are still the issues about the volume of waste and how to minimize release of radioactive material to the environment.
B. S ß arer et al. / Energy Conversion and Management 49 (2008) 1966–1971
Transmutation of plutonium and minor actinides and long-lived fission products is a promising concept to reduce the radioactive waste and its long-term radiotoxicity. Partitioning and transmutation are considered as ways of reducing the burden on a geological disposal. Since plutonium and the minor actinides are mainly responsible for the long-term radiotoxicity, when these nuclides are removed from the waste (partitioning) and fissioned (transmutation), the remaining waste loses most of its long-term radiotoxicity [2]. Transmutation is the process of bombarding a material with particles to form new atoms with higher masses and/ or to fission the material into atoms with smaller masses. It can reduce the mass, volume, activity, heatload, and/or raditoxicity of waste that must be sent to repository [3]. Many different technologies have been examined for transmutation, including a variety of reactors and ADSs. These systems are primarily distinguished by whether they have fast or thermal neutron spectrums. Transuranics transmutation can be planned in both thermal and fast reactors. However, a critical reactor with solid fuel and liquid coolant , thermal and fast, in which neutron production and neutron losses are in balance, may contain a quite limited amount of particular TRU mixture components like Np, Am or Cm isotopes, due to safety constraints. This is one of the mains reasons why the accelerator-driven systems are currently studied worldwide for nuclear waste burning. Accelerator-driven subcritical reactors have been proposed for many applications such as energy production, fertile-to-fissile transmutation and conversion of long-lived radioisotopes into stable or much shorter-lived isotopes [4– 7]. The transmutation technology to incinerate the longlived radioactive isotopes using an accelerator-driven subcritical reactor is one of the best solutions. ADSs can be designed to have a fast neutron energy spectrum, and they are subcritical so safety issues can be considered less significant than in a fast reactor. Another advantage of ADSs is that they can burn mixtures of material that would not maintain criticality in a reactor [8–14]. 2. The system Neutrons are produced by the spallation reaction of protons from a high intensity linear accelerator in the spallation target and the fission reaction in the core. It is used a hexagonal lattice for the waste and fuel assemblies. The system (Fig. 1) is driven by a 1 GeV, 10 mA proton beam incident on a natural lead cylindrical target, 20 cm radius, 70 cm height, and entering the target through a 5.3 cm radius hole. The protons were uniformly distributed across the beam of radius 3 cm. The core is cylindrical assembly, 2.3 m radius, 4.6 m high. The wall thickness of the main vessel is 2 cm. The main vessel is surrounded by a reflector made of graphite, 40 cm thick. The axes of proton beam and the target are concentric with the main vessel axis. The structural walls and beam window are made of the same material, stainless steel, HT9.
1967
Fig. 1. Horizontal and vertical sectional views of the present ADS design.
The vertical proton beam pipe, 5.3 cm outer radius and 5.0 cm inner radius, is inserted from the top into the main vessel down to the spallation target region. The hemispherical bottom end of the tube forms the beam window. The major reasons to choose sodium as coolant are its excellent thermal properties and good compatibility with stainless steel. Most breeder reactors use a hexagonal lattice for the fuel structure, to be consistent with breeder reactor designs, we used a hexagonal lattice for the fuel assemblies. 3. Neutronic analysis In the calculations, Monte Carlo code MCNPX [15], which is a combination of LAHET [16] and MCNP, was used. To transport a wide variety of particles, The Los Alamos High Energy Transport Code, LAHET, was used. These include charged particles such as protons and charged pions as well as neutrons with high energies.
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MCNPX offers options based on three physics packages; the BERTINI [17–20] and ISABEL [21] models taken from the LAHET Code System, and the CEM package [22]. The MCNP transport code is used for radiation transport at low energies (<20 MeV). The transport of high energy particles (>150 MeV) was performed using nuclear models, while for the transport of low energy particles (<150 MeV) cross section libraries have been used. In an ADS, the energy is produced as a result of the multiplying nuclear cascades initiated by accelerated protons interacting with the spallation target. Spallation reactions in a target material proceed through three stages: In the first stage, an incident particle interacts with the nucleus through successive nucleon–nucleon hard collisions in a potential which describes the density of the nucleus as a function of radius. Intranuclear cascade (INC) model is used this stage, in which high energy particles are emitted. After the INC stage of the reaction the nucleus is left in an excited state. This is the starting point for the second or pre-equilibrium stage of the reaction. In the third stage, the evaporation phase of the reaction, the remaining nucleus is de-excited either by evaporation or by fission. We used Bertini model for nucleons and pions, and Isabel model for other particle types in the intranuclear cascade stage. Pre-equilibrium model has been used for the second stage of the reaction. The default physics module options were used with pre-equilibrium model after intranuclear cascade. The study was performed in three steps. In the first step, only the spallation target, proton beam pipe and beam window isolated from the core were considered. In the second step, all calculations have been made in absence of fissile materials and wastes. In the third step, the full reactor geometry was taken into account. In the calculations, we did not take time evaluation of the neutron spectrum and other quantities into account. The target was designed to maximize the neutron yield and to soften the axial power distribution. The target optimization means finding the maximum of neutron production inside the target and the maximum of neutron leakage to the core at given proton energy and target shape and material. Lead is a material highly transparent to neutrons; it has a very low capture cross section, the main isotope in natural lead 208Pb is double magic, and thus exceptionally stable, from the MeV region down to very low energies. But it has a moderate elastic neutron cross section. One of the fundamental parameters of the spallation target is the number of neutrons produced per beam particle incident on target. Calculation of number of neutrons produced in the target is done by summing up the absorption and leakage. Because of small neutron absorption of lead, the leakage is considered to be equal to the neutron production. Spallation neutron and proton yields per one incident proton are 27.4 and 1.2E 02, respectively. Other particle yields such as, pion and muon per one incident proton have been neglected due to their low particle fluxes.
The neutron leakage on the whole surface of the target and the main vessel for all three steps mentioned above is shown in Table 1. Some of the spallation neutrons produced by protons in the target can escape through the beam pipe, therefore it is desired to minimize the neutron escape through the beam pipe. The neutron leakage through some regions of the target top face (Fig. 2) for full system is shown in Table 2. The ratio of the neutron leakage into the proton beam vacuum to the total neutron leakage from the target top face is 1.54%. 3.1. Energy and spatial distributions The transmutation performance of a nuclear system mainly depends on flux level and neutron spectrum. To investigate the radial dependence for three calculation steps, neutron fluxes and heating were calculated in a set of coaxial cylinders, subdividing the target into parts and also many sections for z-dependence. The target was divided into 20 coaxial cylindrical sections and into 32 Table 1 The neutron leakage on the whole surface of the target and main vessel wall Neutron leakage (n/s)
Beam pipe + target
Beam pipe + target + coolant
Full system
Target Top face Side face Bottom face
3.25E+17 1.34E+18 4.02E+16
5.68E+17 2.78E+18 1.63E+17
1.44E+18 9.43E+18 1.03E+18
Main vessel Top face Side face Bottom Face
1.12E+14 – –
8.60E+15 4.12E+16 6.92E+15
8.73E+16 3.42E+17 8.15E+16
Fig. 2. Scheme of the target-beam pipe system.
Table 2 The neutron leakage on the target top face for full system Target top face
1. Region (n/s)
2. Region (n/s)
3. Region (n/s)
Total
1.43E+17
1.54E+16
1.28E+18
1.44E+18
B. S ß arer et al. / Energy Conversion and Management 49 (2008) 1966–1971
Neutron Flux (n/cm2.s)
10
15
10
13
10
11
Target
Beam pipe+target Beam pipe+target+coolant Full system
15
Neutron Flux (n/cm2.s)
8x10
15
7x10
15
6x10
15
5x10
15
4x10
15
3x10
15
2x10
15
1x10
14
1x10
-20
-15
-10
-5
0
5
10
15
20
r (cm)
Target
15
6x10
15
5x10
Beam pipe+target Beam pipe+target+coolant Full system
15
4x10
15
3x10
15
2x10
15
1x10
13
1x10
195
205
215
225
235
245
255
265
z (cm)
Fig. 4. The spatial variation of the neutron fluxes in the target: (a) radial, (b) axial.
The peak like a delta-function in the energy spectrum of protons in the beam window, Fig. 5, shows contribution of source protons. 3.2. Heat deposition
10
9
10
7
10
5
Beam pipe+target Beam pipe+target+coolant Full system
3
10 -10 10
10
-6
10
-2
10
2
Neutron Energy (MeV) 1012
Proton Flux (n/cm2.s)
Target
15
9x10
Neutron Flux (n/cm2.s)
horizontal sectors along the target axis. For the three systems, the neutron flux is low at energies lower than 1 MeV for target and beam window. The energy distributions of the spallation neutrons and protons in the target, the radial and the axial variations of the neutron flux in the target and energy distributions of the spallation neutrons and protons in the beam window are shown in Figs. 3–5, respectively. The energy distribution of the spallation neutrons are rather similar in the target and the spallation protons in the beam window for all three steps. In the target, the energy distribution for all three steps are the same both in shape and calculation values. The peaks of the neutron energy distributions appear at energy about 2 MeV for all three steps in target and the beam pipe. The both shapes and calculated values of the energy distribution of the spallation protons for all calculation steps in the whole energy range are the same in the target. But, in the beam window, the proton flux values for full system are some different than those of the other two systems in the energy range 100–1000 MeV. The neutron energy distributions of the systems beam pipe + target + coolant and full system for both target and beam window, are rather different than that of the beam pipe + target system. The neutron flux values of the beam pipe + target system begins from 1 eV and neutron flux values in the energy range 1 eV–2 MeV are substantially lower than those of the other two systems.
1969
Target
11
10
10
10
3.3. H and He production rates
9
10
Beam pipe+target Beam pipe+target+coolant Full system
108 7
10
100
The average heat deposition by neutrons in the some parts of the system has been calculated for the three calculation steps. The results are given in Table 3. The radial and axial variations of the heating in the target for three steps are shown in Fig. 6. It is observed that the amount of energy deposited in the target by neutrons is the mostly same for three steps. The average energy deposited by neutrons in the beam window, for each calculation steps, is greater about three times than in the case of target. There is a negligible energy deposition by neutrons in the other structural materials, beam pipe wall and especially main vessel wall.
101
102
103
Proton Energy (MeV)
Fig. 3. Energy distribution of the spallation neutrons and protons in the energy range 0–1000 MeV in the target.
LAHET writes each event into an output history file, HISTP. In this file each particle’s crossing of a geometric boundary as well as entry and exit coordinates (positions, angles and energies) are recorded in detail. Lahet also provides the postprocessing HTAPE code. The option IOPT 14, in HTAPE, provides an edit of hydrogen and helium gas production, by isotope, by element, and total. The
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Neutron Flux (n/cm2.s)
1015
Table 4 Gas production in the target and beam window
Beam Window
1012
109
Beam pipe+target Beam pipe+target+coolant Full system 106 10-9
10-6
10-3
100
103
Target
Beam window
H H-2 H-3 Total H He-3 He-4 Total He
2798.21 304.1 181.67 3283.45 11.12 514.69 525.81
33,000 3770 810 37,200 565 3320 3880
Total H + He
3809.20
41,100
gas production rates in the target and beam window for full system are shown in Table 4. The results show that the gas production in the beam window is more intense than the target, as expected.
Neutron Energy (MeV) 15
10
Gas production (apmm/fpy)
Beam Window
14
2
Proton Flux (n/cm .s)
10
3.4. Transmutation
13
10
12
10
11
10
10
10
Beam pipe+target Beam pipe+target+coolant Full system
9
10
8
10 0 10
1
2
10
3
10
10
Proton Energy (MeV) Fig. 5. Energy distribution of the spallation neutrons and protons in energy range 0–1000 MeV in the beam window.
Table 3 Energy deposited by neutrons in the some parts of the system for three calculation steps Heating (W/cm3)
Beam pipe + target
Beam pipe + target + coolant
Full System
Target Beam window Beam pipe Main vessel wall
2.87 8.48 0.0 0.0
2.91 10.77 0.30 2.65E 04
3.19 9.61 0.49 2.14E 04
The core of the system mainly consists of the spallation target region and waste assembly region. The waste region consists of the pin-bundle type fuel assemblies and is cooled by liquid sodium. Fig. 1 shows the schematic view of the target and the fuel assembly. The fuel elements consist of a stainless steel cladding, 0.35 cm thick, characterized by an external radius of 1.35 cm and a total height of 120 cm, end cap included. The fuel is a cylinder, 119.30 cm height. As mentioned above, we used a hexagonal lattice for waste assemblies. The waste subassemblies, bundles, are arranged in an annular array of five rings, each waste subassembly contains seven fuel rods. There are 114 bundles in the system. Isotopes to be transmuted are minor actinides, such as 239Pu, 237Np, 241Am, and are long-lived fission products, such as 99Tc and 129I. The core configuration for the calculations is characterized by 10 bundles in ring 1 (239Pu); 20 bundles in ring 2 (237Np); 24 bundles in ring 3 (129I);
Target
3
Neutron Heating (W/cm )
Neutron Heating (W/cm3 )
Target 32.5 30.0 27.5 25.0 22.5 20.0 17.5 15.0 12.5 10.0 7.5 5.0 2.5 0.0 -20
Beam pipe+target Beam pipe+target+coolant Full system
-15
-10
-5
0
5
10
15
20
6.5 6.0 5.5 5.0 4.5 4.0 3.5 3.0 2.5 2.0 1.5 1.0 0.5 0.0
Beam pipe+target Beam pipe+target+coolant Full system
200
210
r (cm) Fig. 6. Radial and axial variation of the heating in the target.
220
230
z (cm)
240
250
260
B. S ß arer et al. / Energy Conversion and Management 49 (2008) 1966–1971 Table 5 Transmutation calculation results 99
129
237
239
241
910.434 1.455
182.509 0.806
1068.769 36.434
522.932 47.254
1082.229 4.212
Tc
Initial mass (kg) Total transmutation (kg/fpy)
I
Np
Pu
Am
30 bundles in ring 4 (99Tc); 30 bundles in ring 5 (241Am). All calculations are based on pure 99Tc,129I, 237Np,239Pu and 241Am isotopes. The neutron multiplication factor, keff,with its standard deviation was calculated by ‘‘kcode” card of the MCNP code and was found as, after running a large number of histories, 0.96646 ± 0.00119 for the core without the proton beam. Transuranic elements (TRU) and fission fragments (FF) are two main components of high-level nuclear waste, representing 1.1% and 4% of spent nuclear fuel, respectively. TRU, produced by neutron capture in the fuel can only be successfully incinerated by fission, while the radioactivitiy of the FF can be modified by neutron capture. Minor actinides should be transmuted mainly through fission reactions having the possibility increasing higher actinides, while the thermal capture is main transmutation reaction for long-lived fission productions. The main nuclides of minor actinides from power reactor, such as Np, Am and Cm, have threshold fission reaction, so, hard neutron spectrum is desirable. The capture reactions rates of LLFP increases in the thermal energy region of ADS is hard, so it is necessary to establish. The region with softer spectrum for transmutation of LLFP. However, the effect on the MA transmutation properties of ADS is significant. Results of transmutation calculation are given in Table 5. 4. Conclusions Accelerator-driven transmutation system was investigated. The system consists of 1 GeV, 10 mA proton accelerator and a subcritical core with an effective neutron multiplication factor of 0.96. For the neutronic analysis, we used the MCNPX code system. In the neutronic calculations, time evaluations have not been taken into account. The transmutation rates of minor actinides and long-lived
1971
fission products vary in the range 0.806–47.254 kg/fpy. The ADS proposed in this study produces energy from nuclear waste material, while a considerable amount of this spent fuel is incinerated. References [1] Garwin RL, Charpak G. Megawatts and megatons a turning point in the nuclear ages? New York: Alfred A. Knopf; 2001. [2] Salvatores M. Nuclear fuel cycle strategies including partitioning and transmutation. Nucl Eng Des 2005;235:805–16. [3] Westlen D. A cost benefit analysis of an accelerator driven transmutation system. Stockholm: Royal Institute of Technology; 2001. [4] Rubbia C, Rubio JA, Buono S, Carminati F, Fie´tier N, Galvez J, et al. CERN report, 1995, CERN/AT/95-44 (ET). [5] Van Atta CM, Lee JD, Heckrotte W. The electronuclear conversion of fertile to fissile material, Lawrance Livermore Laboratory Report, UCRL-52144, 1970. [6] Bowman CD, Arthur ED, Lisowski PW, Lawrence GP, Jensen RJ, Anderson JL, et al. Nucl Instrum Methods A 1992;320:336. [7] Salvatores M, Slessarev I, Tchistiakov A, Ritter G. Nucl Sci Eng 1997;126:333. [8] Rineiski A, Maschek W. Dynamics and reactivity control in accelerator driven systems for nuclear waste burning. Forschungszentrum (FZK), Institute for Nuclear and Energy Technologies, Karlsruhe, Nuclear Energy, 2006. [9] Vergnes J, Barbrault P, Lecarpentier D, et al. Limiting plutonium and minor actinides inventory: comparison between accelerator-driven system (ADS) and critical reactory. Proceedings of the international conference on future nuclear systems, Global’99, Jackson Hole, Wyoming. LaGrange Park, Illinois: American Nuclear Society; 1999. [10] Ait Abderrahim H et al. Nucl Instrum Methods A 2001;463:487–94. [11] Maschek W et al. J Nucl Mater 2003;320:147–55. [12] Haeck W et al. J Nucl Mater 2006;352:285–90. [13] Maschek W et al. Nucl Instrum Methods A 2006;562:863–6. [14] Haas D et al. Energy Convers Manage 2006;47:2724–31. [15] Waters LS, editor. MCNPX user’s manual version 2.3.0, April 2002 LA-UR-02-2607. [16] Prael RE, Lichtenstein H. User guide to LCS: the LAHET code system, Los Alamos National Laboratory Report LA-UR-89-3014, Revised September 15, 1989. [17] Bertini HW et al. Nucl Instrum Methods 1968;66. [18] Bertini HW. Intranuclear-cascade calculation of the secondary nucleon spectra from nucleon–nucleus interactions in the energy range 340–2900 MeV and comparisons with experiment. Phys Rev 1969;188:1711. [19] Boudard A, Cugnon J, Leray S, Volant C. Phys Rev C 2002;66:044615. [20] Junghans AR et al. Nucl Phys A 1998;629:635. [21] Yariv Y, Fraenkel Z. Phys Rev C 1981;24:48. [22] Mashnik SG, Toneev VD. MODEX – the program for calculation of the energy spectra of particles emitted in the reactions of preequilibrium and equilibrium statistical decays, Communication JINR P4-8417, Dubna, 1974, p. 25 [in Russian].