Fission Target Design and Integration of Neutron Converter for Eurisol-Ds Project J

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FISSION TARGET DESIGN AND INTEGRATION OF NEUTRON CONVERTER FOR EURISOL-DS PROJECT J. Bermudez, O. Alyakrinskiy, M. Barbui, L.B. Tecchio Laboratori Nazionali di Legnaro I.N.F.N. Viale dell'Università 2, 35020 Legnaro (PADOVA) ITALY. F. Negoita, L. Serbina, E. Udup “Horia Hulubei” - National Institute for Physics and Nuclear Engineering (IFIN-HH) Str. Atomistilor 407, P.O. Box MG-6, 077125 Bucharest-Magurele, Romania

Abstract A study of a new fission target for EURISOL-DS is presented with a detailed description of the target. Calculations of several configurations were done using Monte Carlo code FLUKA aimed to obtaining 1015 fissions/s on single target. In Eurisol, neutrons inducing the fission reactions are produced by a proton beam 1GeV- 4mA interacting with a mercury converter. The target configuration was customized to gain fission yield from the large amount of low energy neutrons produced by the Hg converter. To this purpose, the fissile material is composed by discs of 238- Uranium carbide enriched with 15 g of 235-U. Studies of several geometries were done in order to define the shape and composition of uranium target, taking into account the mechanical and space constraints. Furthermore different configurations of reflector and moderators materials were considered to increase the thermal neutrons confined around the target and so enhance the performance of the system. The final configuration consists in six modular target containers inclined respect to the vertical, containing the fission targets, ion sources with RIB extraction and steering elements, cryopanels and devices for handling. The cross sections of the modules are rectangular with 10 mm coating of water for cooling. The UCx fission targets are cylindrical dwelling in a graphite holder inside a tantalum container. The analysis includes also the shielding, reflectors, isolators, and connections. A detailed study of reflector materials, moderator materials, fission target geometry, fissile materials and containers were done. Neutronic calculations, fission rates, energy deposited on main parts of the system and neutron distribution results are reported.

1 I. INTRODUCTON

It has been asserted that the increase of primary beam intensity does not necessarily lead to an increase to intensity of the secondary Radioactive Ion Beams (RIB) [1]. In order to profit of the production potential of a 1 GeV proton beam of MW power the concept of charged particle to neutron converter technique has been chosen. Here, the power of the primary proton beam is dissipated and disposed in a primary cooled target (i.e. mercury converter) and the resultant neutron flux induces fission products in a thick ISOL target (fission target) without destruction of the latter by overheating.

Designing this converter and the surrounding fission targets is one of the main scope of this project. Converter and fission targets design is conceived in a modular way so that its individual parts can be rapidly replaced and serviced by means of remote handling. Referring to the concept previously adopted by the PIAFE [2] and MAFF [3] planned facilities, profiting of the fact that several technological aspects were developed within such projects a possibly solution based on the same concept has been proposed for the EURISOL fission target assembly [4].

Conceptually, a target filled with 235U is inserted, through a channel created in the shielding, close to the neutron converter in a position where the neutron flux shown has the maximum intensity. The neutron flux is moderated in energy to optimize the number of fissions induced in the target. Six fission targets are foreseen to be installed around the converter: 2 above the neutron converter, 2 on the right side and 2 to the left side. The spatial mapping of the neutron flux gives the placement of the fission target elements. The vertical placement of these assemblies (see Fig. 1) has several advantages, one of them is that the alignment of ion beam optics elements is better preserved in operation, which is essential for good extraction and transport of secondary beam.

The proton beam parameters correspond to 4 MW power delivered to the mercury converter, with a maximum current of 4 mA. Combined iron+concrete shielding with a total thickness of about 6 meters has to be considered.

Fig. 1. Schematically arrangement of the six fission targets around the neutron converter. Each target is placed inside a vacuum tube. In the right figure, only one of the rectangular vacuum tube in which the targets are placed is shown.

2 Loading and unloading the fission targets is accomplished by a mobile tube mounted on the top of a fix tube; the fission target and all other elements in the fixed tube are pull into the mobile one and moved into a hot cell where mounting/dismounting of the fission targets (as well as of other elements) can be performed under visual control.

All components are placed inside a vacuum tube embedded in heavy concrete shielding. A double wall tube is proposed to provide a coupling water flow in order to evacuate the heat produced by fission in target and keep at least the upper part of tube at normal temperatures. Large surface cryogenic-panels are distributed inside the vacuum tube to maintain a good quality vacuum and trapping the radioactivity in a confined region.

Two geometries of the vacuum tube, cylindrical and rectangular, were considered in calculations. The rectangular shape was adopted, since the fission targets can be placed closer to the converter in higher neutron flux. It is considered more suitable than cylindrical shape also from the safety point of view allowing taking advantage of the available space to separate the services (water cooled high voltage and high current bars) from the target and the RIB line. The modification of the tube cross section provides the space for the inclusion of a moderator material in the nearest area around the target and could represent a profit for increasing the fission rate. Modifications were done keeping the UC targets closer to the converter in the most intense neutron flux zone.

Different fission target materials were chosen to get a in-target production rate of 1015 fission per second : MKLN (special graphite), POCO foam (graphite foam) and high density UC pellets. In the first two cases, the fissile material is highly enriched 235U uranium dispersed in a graphite matrix with an apparent density is about 2 g/cm3. For the high density UC pellets the fissile material consists mainly of 238U enriched with about 2% of 235U with a density of 12 g/cm3. The target assembly is shown in Figure 2.

Main insulator Cryopanel

Water cooled, high voltage, RIB line high current bars Extraction electrodes

Flexbile Ion source connectors Transfer tube

UCx material Vacuum tube External container Fig. 2. Fission target assembly (left); detail of the finned target and thermal calculation (right).

3 The fuel is housed in a graphite primary container, 200 mm long with 35 mm of diameter surrounded by a tantalum container acting as protection and as heat radiator. The fuel consists of 86 discs, 1 mm thick and in between 1,3 mm thick grafoil discs with the aim to keep a high thermal conductivity for a better heat dissipation. The target presents a central hole of 8 mm diameter. Different kind of ion sources (laser ion source, plasma ion source, ECR,) are planned to be installed close to the target to ionize (charge state 1+) the selected fission products. The fission target has been designed to operate at 100 kV respect to the ground potential. To dissipate the heat power (30 kW) released by the fission reaction in the fuel the target container has a finned construction which increase the emissivity and gives a good power dissipation. This allows the target to remain at the required high temperature of around 2000 °C and to have large temperature drop across the fins.

Transport of secondary beam through the shielding inside the about 6 meters long tube is assured by several electrostatic quadrupole lenses, one doublet each 2 meters, with apertures of 60 mm equal to secondary beam line diameter. At upper end of this vertical beam line a 90° electrostatic deviation turns the secondary beam in a horizontal plane toward the beam purification and acceleration area. In order to extract all the 6 beams in the same direction, the vertical tubes have to be inclined. Increasing the distance between the tubes at upper end is also necessary to allow installation of gate valves, connectors, etc. Resulting geometry can be seen in Fig 3.

Fig. 3: Geometry of fission target modules. Incident proton beam line is red. The 6 radioactive beams are delivered on 6 parallel lines (green) at about 7 meters above. Around neutron converter (not visible) a neutron reflector (dark gray) can be used to increase neutron flux at fission targets positions.

In the present report the fission target design optimized for the such geometry is discussed mainly from the point of view of fission yields. Results on neutron fluxes as weel as power deposition are also presented, while other important issues associated to fission targets such as release properties or handling are treated is other reports.

4 II. FISSION TARGET AND MODULE DESING. ONE TARGET SIMULATIONS

The design study of the target and the optimization of its geometry, using different configurations were done with the Monte Carlo code FLUKA [5, 6, 7] in order to achieve improvements on the performances. The performance of targets in the configuration shown in Fig. 3 was analyzed simulating one target. Cross section views of the geometrical model used for Monte Carlo simulations, including container module with the UC fission target, the beam line and the services starting above the top of the UCx target, is presented on Fig. 4.

Fig. 4: Cross section views of rectangular container module. a) UC Target, b) water, c) BeO insulator, d) beam line extraction, e) Hg converter.

A. THE EXTERNAL REFLECTOR

A reflector is proposed in order to confine the neutron flux near the fission target, to optimize the neutrons-target interactions, to enhance the fission yield and assure the dose control. The reflector has a cylindrical shape with an external diameter of 1.0 m and internal diameter of 200 mm and is surrounding the neutron converter. The reflector length is of 1.5 m. The fission targets are inserted in the reflector and located close to the converter. Different materials (steel, graphite, beryllium oxide and concrete) were considered for the reflector.

Preliminary calculations were done using steel as reflector material. In fact, due to the high neutron absorbtion cross section of iron, steel guarantees a good neutron confinement. However, as expected, the results show a number of fissions below the planned rate of 1015 fissions/s per target. This behavior is attributable to the characteristic low neutron reflectivity of iron, which on this configuration acts as an absorber for the thermal neutrons required to induce fissions processes. This result moves up to the reconsideration of the reflector materials and to a detailed study of their properties in order to optimize the configuration and dimensions requested for the reflector.

Looking through the characteristic neutron cross sections, beryllium oxide and graphite were chosen for this study. Beryllium and carbon have low neutron absorption compared to iron and their capability of thermalize neutrons is well known (see fig. 5).

Fig. 5: n,total Cross sections for Iron, carbon and Beryllium.

Calculations were performed using one fission target and its corresponding container placed at the top-over the converter. The total power deposited on target and on reflector, the fission rate and the neutron fluence were studied simultaneously. The configuration was prepared using water as moderator around the target. The total power deposited on the reflector was also determined to understand the heating process around of the target and evaluate the cooling requirements. The results obtained are shown in Table 1.

Table 1: Total power deposited in the reflector and in the target, fission rate and neutron fluence for 4 reflector materials. Total P Total P Neutron Reflector Fission rate reflector target Fluence Material (fissions/mA) (kW/mA) (kW/mA) (n/cm2/mA)

Fe 173 16.7 5.89E+14 6.30E+13

BeO 162 18.3 6.50E+14 6.04E+13

Graphite 128 17.6 6.25E+14 5.92E+13

Concrete 174 8.2 2.89E+14 4.03E+13

When using Fe and concrete as reflector materials, the total power deposited inside the Reflector is high, due to the high neutron absorption. On the other hand, when using BeO and graphite the power deposited in the reflector is lower whereas the power released in the target increases proportionally to the increased number of fissions.

Looking at the fission rates results, BeO seems to be the most suitable to maximize the fissions on target. In fact, a fission rate of 6.5E+14 fissions/mA implies

6 a fission rate beyond the goal when 4 mA will apply. Using graphite similar rates are reached (6.25E+14 fissions/mA). This is considerably significant due to commercial availability and low cost of graphite.

In all cases, a considerable quantity of power is released on the reflector mainly by charged particles escaping from the MMW converter. Using a graphite reflector, the lowest value of total power deposited on reflector (128 kW/mA) is obtained together with a high fission rate (6.25E+14 fissions/mA), that is enough to satisfy the requested rate.

B. OPTIMIZING THE INTERNAL MODERATOR

A local moderator is placed around the fission target in order to slow down the neutrons and to take advantage of the favorable fission cross-section of low energy neutrons impinging on the fission target. The study evaluates the improvements and real contribution of increasing the amount of thermal neutrons. The results can move forward to evaluate substantial changes on the mechanical structure to favor reaching of higher rates and to establish the final configuration of the six targets of the facility.

For such a purpose three different materials were investigate as possible moderator: graphite, water and BeO. The moderator is associated to a reflector creating the most suitable and performing configuration reflector-moderator. In principle, water is a good neutron moderator, but the use of water near the target could represent a real problem for safety reasons.

Three different thicknesses of the local moderator placed around each target were studied (50, 100, 120 mm). The power deposited on targets, the number of fissions and the neutron fluence were determined considering the two targets located at the top of the mercury converter. The results obtained are presented on Table 2.

Thickness of moderator

The energy deposited in two targets as a function of the moderator thickness is shown in Fig. 6. The evaluation of the performance without reflector is presented as “concrete reflector”, that means fullfilling the reflector region with the usual shielding material. For thise case was studied the total power only for 50 mm of moderator

30 30 Graphitewater FeWATER 28 28 Graphgraph FeGraphite 26 26 GraphAir FeAir GraphBeO 24 FeBeO 24 22 Iron 22 20 20 18 18 16 16 14 14 12 12 10 10 8 8 Graphite 6 6

Total Power (kW/mA) Power Total 4 4 2 2

40 50 60 70 80 90 100 110 120 130 40 50 60 70 80 90 100 110 120 130 30 30 28 28 26 26 24 24 22 22 20 20 18 18 16 16 14 14 12 BeO 12 10 10 BeOwater Concretewater 8 8 BeOgraphi Concrete Concretegraph 6 BeOAir 6 ConcreteAir Total Power (kW/mA) Power Total 4 BeOBeO 4 ConcreteBeO 2 2

40 50 60 70 80 90 100 110 120 130 40 50 60 70 80 90 100 110 120 130 Thickness (mm) Thickness (mm) Fig. 6. Total power deposited in two targets for different thickness of moderator, using a) Fe reflector, b) Graphite reflector, c) BeO reflector and d) without reflector (only for 50 mm moderator).

According to the results on Figure 6, the thickness of moderator does not affect significantly the fission rates. Further comprehension of the processes taking place on target can be found out from the fission rate calculated on targets. Figure 7 shows the results obtained for different thickness of moderator combined to different reflectors.

Table 2. Performances for different reflector and moderator materials. Neutron Thickness Power Fission rate Reflector Moderator Fluence (mm) (kW/mA) (fissions/mA) (neu/cm2/mA) 1 50 16.7 5.89E+14 6.30E+13 2 Water 100 14.6 5.13E+14 5.68E+13 3 120 14.6 5.13E+14 5.66E+13 4 50 13.0 4.54E+14 8.40E+13 5 Grafite 100 15.2 5.31E+14 8.51E+13 6 120 15.8 5.54E+14 8.54E+13 Fe 7 50 8.19 2.79E+14 7.90E+13 Air 8 100 (no moderator) 7.37 2.51E+14 7.47E+13 9 120 7.21 2.46E+14 7.22E+13 10 14.9 5.21E+14 8.43E+13 50 11 BeO 100 18.0 6.31E+14 8.56E+13 12 18.8 6.63E +14 8.66E+13 13 50 18.3 6.50E+14 6.04E+13 14 Water 100 15.5 5.49E+14 5.50E+13 15 120 15.3 5.41E+14 5.47E+13 16 50 23.2 8.25E+14 8.58E+13 17 Grafite 100 23.6 8.38E+14 8.59E+13 18 BeO 120 23.3 8.31E+14 8.61E+13 19 50 21.6 7.63E+14 8.18E+13 Air 20 100 (no moderator) 20.4 7.25E+14 7.73E+13 21 120 19.3 6.81E+14 7.48E+13 22 50 24.2 8.63E+14 8.56E+13 BeO 23 100 25.4 9.06E +14 8.85E+ 13 24 50 17.6 6.25E+14 5.92E+13 Water 25 100 14.9 5.26E+14 5.36E+13 26 50 18.9 6.69E+14 7.95E+13 Grafite 27 100 19.3 6.88E+14 8.13E+13 Grafite 28 Air 50 16.4 5.8E+14 7.14E+13 29 (no moderator) 100 14.6 5.15E+14 5.90E+13 30 50 20.3 7.19E+14 8.18E+13 BeO 31 100 21.8 7.75E+14 8.40E+13 32 Water 50 8.21 2.89E+14 4.03E+13 33 Grafite 50 5.66 1.95E+14 7.90E+13 Concrete Air 34 (no reflector) 50 (no moderator) 3.45 1.17E+14 3.81E+13 35 BeO 50 6.92 2.40E+14 4.98E+13

1,00E+015 1,00E+015 GraphiteWater FeWater GraphiteGraphite 9,00E+014 Fegraphite 9,00E+014 GraphiteAir FeAir 8,00E+014 Iron 8,00E+014 GraphiteBeO FeBeO 7,00E+014 7,00E+014

6,00E+014 6,00E+014

5,00E+014 5,00E+014

4,00E+014 4,00E+014

3,00E+014 3,00E+014

2,00E+014 2,00E+014 Graphite

1,00E+014 1,00E+014 Fission rate (fissions/mA) 40 50 60 70 80 90 100 110 120 130 40 50 60 70 80 90 100 110 120 130 1,00E+015 1,00E+015

9,00E+014 9,00E+014

8,00E+014 8,00E+014

7,00E+014 7,00E+014

6,00E+014 6,00E+014 ConcreteWater ConcreteGraphite 5,00E+014 5,00E+014 ConcreteAir 4,00E+014 4,00E+014 ConcreteBeO

3,00E+014 3,00E+014 BeOWater BeO 2,00E+014 BeOGraphite 2,00E+014 Concrete BeOAir 1,00E+014 BeOBeO 1,00E+014 Fission rate (fissions/mA) 40 50 60 70 80 90 100 110 120 130 40 50 60 70 80 90 100 110 120 130 Thickness (mm) Thickness (mm) Fig. 7. Fission rate for different thickness of moderator, using a) Fe reflector, b) Grafite Reflector, c) BeO reflector and d) without reflector (only for 50 mm moderator).

The results show the fission rates in two targets for different thickness of moderator. A decreasing of the fission yield due to increasing of thickness was confirmed when using water moderator or no moderator.

On the other hand, whatever moderator material and thickness, graphite and BeO reflectors give rise to more than 5.00E+14 fissions/mA (in two targets). The only one similar behavior is observed on the values reached in the configuration of Fe reflector, combined to 50 mm thick of water.

The poorest performance is clearly obtained using concrete, revealing a maximum of 3.0E+14 fissions/mA for two targets.

As a general tendency we note that the reflector material play the most important role in the fission yield determination, whereas small changes are produced by the chose of the local moderator. This is clear looking at the energy spectra of neutrons inside the target.

Considering the neutron energy distribution on the fission target for water, graphite and air combined to graphite reflector which present high reflectivity as demonstrated above, water gives the lowest amount of thermal neutron (see Figure 8).

10 − Graphite − Air

− H2O − BeO

Fig. 8. Neutron energy distribution on 2 -targets, using graphite as external reflector and graphite, water and beryllium oxide as moderators.

For comparison, the results of power deposition on target obtained for the multiple combinations of reflectors and moderator materials (for 50 mm thick) are presented in Figure 9.

Power deposited over target for different internal reflector materials

BeO graphite Air water 30

graphite 25 23,2 water graphite 24,2 18,9 20 water Air 20,3 21,6 BeO BeO graphite 17,6 18,3 BeO 13 15 Air 16,7 16,4 water (kW/mA) 14,9 10 graphite 5,66 water Air 5 BeO 8,21 8,19 6,92 Air 3,45 0 Grafite BeO Concrete Fe

External reflector

Fig. 9. Total power deposited on two targets for all reflectors and moderators.

Analogous analysis can be done for power deposited on reflector (see Figure 10).

Power deposited over reflector for different moderator materials

BeO graphite Air water 250

graphite

BeO 200 218 Air graphite 220 graphite Air BeO Air 179 water water BeO 175 150 165 174 173 graphite 161 water BeO Air 162 (kW/mA) 100 126 130 water 128

0 Grafite BeO Concrete Fe External reflector

Fig. 10. Total power deposited on reflector calculated using two targets.

The power deposited on graphite and BeO reflectors are the lowest of the group while results of iron and concrete reflectors reveals more than 160kW/mA deposited on reflector evidencing the advantage of using BeO and graphite. The graph show minor differences between moderators materials with the same reflector and major differences between reflector materials performance.

For further analysis, also the fission rates determined for different configurations are summarized on Figure 11. The concrete results correspond to calculations with no reflector. Similarly, calculations done using air represent no moderator material on the configuration.

The fission results using graphite reflector combined to air are comparable with the results obtained using BeO combined with water. The results obtained with iron reflector present a maximum combined to water moderator, but comparable with fission rates obtained using graphite reflector with no moderator. In any case the highest rates are obtained using BeO reflector, but is considerably interesting the results obtained using graphite reflector combined to graphite moderator.

Observing the graphs is possible to summarize:

1.- Regarding to the external reflector, beryllium oxide gives the highest fission rates for all moderator materials studied. The use of graphite as external reflector shows a performance similar to using beryllium oxide. All cases showed more than 6.00E+14 fissions/mA for two targets. Probably, BeO is the best reflector material but graphite was suggested because is a more practical solution and much less expensive;

12 2.- Iron and concrete show the lowest values for fission rate. The configuration presenting the poorest performance is in the absence of reflector and moderator. (referred as concrete reflector and air moderator);

3.- Concerning the moderator, an interesting result is observed using air moderator joint to graphite and/or beryllium oxide as external reflectors. The performance of the system is similar for air and water, but using air can be avoided risks associated to using water around a high voltage system;

4.- Variations obtained changing the moderator material on each external reflector is not significant. This performance leads to discard the moderator and to establish that a proper choice of the external reflector material plays a relevant role on the total fission rates.

Fission rate on 2 Targets for different reflectors and moderators

BeO 1.E+15 Graphite 9.E+14 Air 8.E+14 Water

) 7.E+14

6.E+14

5.E+14

4.E+14

Fissions (Fiss/mA 3.E+14

2.E+14

1.E+14

0.E+00 BeO Graphite Iron Concrete External reflector

Fig. 11: Fission rate obtained using different external reflectors and moderators.

III. THE SIX TARGETS GEOMETRY

A. Optimizing the Geometry and target materials

In the design study of the fission target, after the study of moderator and reflectors and using the characteristics of the MAFF concept target, a configuration of six targets was prepared for simulations. The conceptual design of the full facility is presented on Figure 12 with details on target handling area. The general configuration used for FLUKA calculations is presented on Figure 13, referring the position of converter and positions of targets.

Fig. 12: General view of the conceptual design of the fission targets system.

Fig. 13: Geometry of six targets inclined: Cross section view of the system (left) and detail of the target positions (right), as for the FLUKA input.

Based on the results discussed above, a full structure of the fission target was designed for two different configurations: with graphite reflector and without any reflector (concrete), and without any moderator. The configuration considers six targets prepared with the correspondent rectangular module containers. Three targets are located at z=0 and the other three at z=15 along the proton beam direction. Both groups were tilted respect to the vertical. The study includes calculations of the fission rates, the neutron energy distribution on target and the power deposited on target, on reflector, and on some components. A cross section view of the targets can be seen in Figure 14.

14 1 rectangular tube,2 cryo-panels, 3 connectors B

A C

Fig. 14: Configuration of the six targets geometry. A) Y-Z cross section, B) X-Z planes and major components, c) X-Y planes.

The fission rate and neutron fluence obtained for six targets are presented on the Table 3. The results for the three targets placed at z=0 position and for the three targets at z=15 are presented separately.

Table 3. Fissions rates and neutron fluence obtained for six targets. 3-Targets 3-Targets Fission Total at Z=0 at Z=15 Rate Fissions/mA Fissions/mA Fissions/mA Graphite 8.94E+14 9.25E+14 1.82E+15

Concrete 3.40E+14 3.89E+14 7.29E+14

3-Targets 3-Targets Neutron Total Fluence at Z=0 at Z=15 Fluence n/cm2/mA n/cm2/mA n/cm2/mA Graphite 1.08E+14 1.14E+14 2.22E+14

Concrete 7.22E+13 7.78E+14 1.50E+14

15 Using graphite as external reflector, more than twice fissions on target can be obtained respect to the configuration with no reflector. The highest rate is 1.82E+15 fissions/mA. A graphical representation of fissions density using graphite as reflector can be seen on Figure 15.

Fig. 15: Graphic of fission density for the 2 groups of targets.

The total power deposited on targets is presented on Table 4, as well as the power deposited on the main parts of the system is presented on Table 5. The higher energy deposition on target (51.5 kW/mA) was obtained with the graphite reflector. The total power density on targets is presented on Figure 16.

kW/cm3/mA

Fig. 16: Total power density on six targets inclined.

16 Table 4. Total power deposited over six fission targets. power deposited power deposited Reflector on reflector on target Material (kW/mA) (kW/mA) Graphite 112 51.5

Concrete 156 21.1

The study was extended to determine the total power deposited over the single parts of the system, in order to understand the heating processes and the heat dissipation required on the services. Figure 17 shows the rectangular tube with double wall for cooling, the cryo-panels, two groups of connectors and the beam line for RIB extraction.

Connect1 Wall-ext Connect2

CryoP3

CryoP2

Wall-int CryoP1

Fig. 17. Scheme of the regions for FLUKA simulations (left) considering the main parts of the mechanical design (right)

The total power deposited on components and materials of service are represented graphically on Figure 18. The results are integrated over the full volume of the target module. The nominal values obtained for each region, are presented on Table 5.

Fig. 18. Total power deposited over main parts of the target module.

17 Table 5. Total power deposited over main parts of the system. Reflector Graphite Concrete 3-Targets 3-Targets 3-Targets 3-Targets Region at Z=0 at Z=15 at Z=0 at Z=15 (kW/mA) (kW/mA) (kW/mA) (kW/mA) Wall-ext 0.98 1.33 0.68 1.02

H2O 4.72 5.45 3.78 4.50 Wall-int 0.92 1.22 0.62 0.88 Connect1 0.45 0.54 0.16 0.22 Connect2 1.01 1.32 0.31 0.53 CrypP1 9.90E-03 1.39E-02 6.28E-03 1.13E-02 CryoP2 6.63E-03 1.13E-02 3.87E-03 7.86E-03 CryoP3 1.02E-02 1.49E-02 6.33E-03 1.06E-02

Simulations also show that a power of few kW is released inside to the bar connectors which require to be cooled.

The gradient of power released on the reflector was established measuring inside reflector at 4 different horizontal distances from the center of the converter in 1000 cm3 of volume. A schematic view is shown on Figure 19. The volumes are identified as A,B,C and D. For each one are described the Cartesian position starting from the target. Results are presented on Table 6.

A B C Cartesian coordinates in cm from the target A: B: C: D: X: from 25 to 30 X: from 10 to 15 X: from 0 to 5 X: from -30 to -25 Y: from -20 to -15 Y: from -20 to -15 Y: from -20 to -15 Y: from -20 to -15 Z: from -20 to 20 Z: from -20 to 20 Z: from -20 to 20 Z: from -20 to 20

Fig. 19: General scheme representing the 4 positions studied on the reflector. Cartesian coordinates of each are represented. A) section front view along X-Y, b) section lateral view along Z-Y, c) isometric view.

18 Table 6. Power deposited on 4 points of the reflector. Power deposited/cm3 Position Position Position Position Reflector A B C D Material Power Power Power Power (W/mA) (W/mA) (W/mA) (W/mA) Graphite 0.152 0.345 0.507 0.142

Concrete 0.20 0.60 0.96 0.25

The position “C” nearest to the converter reveals the highest values, showing that the highest power deposited on reflector is located closer to the Hg converter. While 25 cm far from the center, the power deposited is 3.5 times lower. This behavior was observed using graphite reflector and also concrete (no reflector). Results obtained using concrete were almost double than using graphite as reflector.

The UCx material for fission target

The configuration with six inclined targets was studied using two different UCx material compositions. This calculation was prepared with the aim to analyze the advantage of using a light target.

Two material compositions defined as Material-1 and Material-3 are considered. Material-1 is uranium carbide composed by high density 238-Uranium enriched with 15 g of 235-Uranium. Material-3 is uranium carbide prepared using carbon and only 15 g of 235-Uranium. The different materials were defined varying the densities, the isotopic composition and the ratio of carbon versus uranium. The characteristics of the fissile materials studied are presented on Table 7. Results are exposed below showing the performances shown by both materials.

Table 7. Target material composition Material Composition Properties 15 g 235U Apparent density: 4.4 g/cm3 Material-1 37.2 g C Volume: 180 cm3 740 g 238U 15 g 235U Density: 1.89 g/cm3 Material-3 324 g C Volume: 180 cm3

In this way it is possible to observe the contribution of 238-U isotope to the fission products yield. Analyses were carried out using graphite reflector and concrete (no reflector). The total power deposited on target and on reflector, number of fissions and neutron fluence were calculated. The results are presented on Table 8.

19 Table 8. Calculation over six targets using material-1and material-3 Power on Power on Target Fissions Neutron F Reflector Target 2 reflector Material (Fission/mA) (neutron/cm /mA) (kW/mA) (kW/mA)

Mat-1 Graphite 51.5 1.82E+15 2.22E+14 112

Mat-1 Concrete 21.1 7.29E+14 1.50E+14 156

Mat-3 Graphite 50.5 1.83E+15 2.33E+14 113

Mat-3 Concrete 20.1 7.16E+14 1.56E+14 156

Figure 20 presents the histograms of the energy deposited on target using both target materials and both reflector materials.

60 MAT-1 51,5 50

MAT-3 40 50,5

MAT-1 30 21,11 Energy (kW/mA) 20 MAT-3 20,08 10

Concrete

Graphite

Concrete

E- Tar get Graphite

Fig. 20. Power deposited on six targets for two materials targets and concrete or graphite reflector.

The influence of reflector material on energy deposited can be seen observing the power deposited on target. Both target compositions show similar behavior using graphite reflector or concrete. Again we note that major differences are produced by the reflector, while the two target compositions do not produce relevant differences.

The same behavior was found on the fission rate, as shown in Figure 21.

20 MAT-1 2,0E+15 MAT 3 1,82E+15 1,83E+15 1,6E+15

1,2E+15 A MAT-1 7,29E+14 MAT-3 8,0E+14 7,16E+14 Fission /m

4,0E+14

0,0E+00

Concrete Graphite Fiss: Concrete Graphite Fission Rate

Fig. 21. Fission rate on six targets for two target materials and concrete or graphite as reflector.

The neutron fluence results (Figures 22 and 23) were correlated with the fission rate results for all cases. The general behavior was confirmed: the effects of the material used as reflector is more important than effects observed changing the composition of the target material. Z=0 Z=15 n/cm2/mA

Fig. 22. Neutron fluence on targets.

It is necessary to consider practical constrain related to the production of uranium carbide using only the 235-U isotope. However it is clear that the contribution of 238-U over the fission yield is less significant than that observed using only 235-U. This result allows to understand that the using 15g of 235-U the fission processes is directly depending of the reflector material.

21 2,5E+14 MAT-3 2MAT-1 2,33E+14 ,22E+14

MAT-1 2,0E+14 MAT-3 1,50E+14 1,56E+14 1,5E+14

1,0E+14 Neutrons/cm2/mA 5,0E+13

0,0E+00

Concrete Graphite Concrete Neutron Fluence Graphite

Fig. 23. Neutron Fluence on six targets for two target materials and graphite reflector or concrete

The energy spectra of the neutrons measured inside the target are presented in Figure 24 for both configurations discussed above. Looking at the spectra it is clear that the number of low energy neutrons on target obtained using the concrete reflector is significantly lower than what obtained with the graphite reflector. Moreover in the thermal region, even if it is less evident, the graphite reflector keeps the tendency of showing more neutrons. Taking into account the errors concerning to the use of Monte Carlo models, no significant differences are observed between the performances of the two target materials.

Mat 3- Graphite: 2.33E+14 Neutrons/cm2/mA Mat 1- Graphite: 2.22E+14 Neutrons/cm2/mA Mat 3- Concrete: 1.56E+14 Neutrons/cm2/mA Mat 1- Concrete: 1.50E+14 Neutrons/cm2/mA

Fig. 24. neutron energy distribution for four configurations of target-reflector couple.

22 CONCLUSIONS

On the framework EURISOL-DS project neutronic calculations were done using FLUKA Monte Carlo code. A six fission targets system connected to an Hg converter was prepared and evaluated variables aimed to optimization of the general configuration. The summary of results can be present for different subjects:

Geometry:

• Using a rectangular shape target module was observed the best performance due to the higher volume available on it. This geometry allows the distribution of the target and all the services in two regions separately. • The preparation of the rectangular target module obliged to incline the targets in order to achieve all the beams extraction on same direction. • The 86 discs distributed in 200 mm height and 35 mm diameter using graphite between discs, dwelling in tantalum and graphite container was the target configuration with the best performance. • Diameter of reflector was fixed at 500 mm around Hg converter.

Reflector Material:

• The study of several combinations of reflector and moderator materials made clear that, on the proposed geometry, the moderation effect was irrelevant even using several dimensions of it. Instead of was demonstrated the determinant role of the appropriate selection of the reflector material. • Concrete and graphite were evaluated as possible reflector materials using six target inclined on rectangular full target-containers. Energy deposited on reflector as well as in main parts of the system were calculated. • Graphite reveals the best performance as reflector material leading to higher fission rates. However was observed a high deposition of total power on reflector (~128 kW/mA) suggesting the application of a heating dissipation system. • The results shown a power density on the reflector of 0.5 W/cm3/mA on the center, close to the Hg converter, while is decreasing with the distance from it. At 25 cm from the center the deposited power density is 3.5 times lower.

Target Material

• The exclusion of 238U from the UC target composition but keeping the 15 g 235U was evaluated to observe its influence on the fission rate. Not relevant differences were verified on results obtained for both materials. The results disclose the lower contribution on 238U to the fission rate. However it is necessary consider the material properties involved on the production of UC materials.

Acknowledgement

We acknowledge the financial support of the European Community under the FP6 "Research Infrastructure Action-Structuring the European Research Area" EURISOL DS Project contract no 515768 RIDS. The EC is not liable for the use that can be made of the information contained herein.

REFERENCES

1- D. Ridikas, W. Mittig and N. Alamanos; “RNB production with a powerful proton accelerator”, report DAPNIA-SPHN-2000-59, CEA Saclay (2000).

2- PIAFE; Project Report, Ed. By Ulli Koester and Jean-Alain Piston, March 1998.

3- H.J. Maier et al., Nuclear Instruments and Methods in Nuclear Research A 438 (1999)185.

4- O. Alyakrinskiy, M. Barbui, F. Negoita, L. Serbina, L.B. Tecchio, E. Udup; “EURISOL Fission Target Design adapting MAFF concept”, http://www.eurisol.org.

5- G. Battistoni, S. Muraro, P.R. Sala, F. Cerutti, A. Ferrari, S. Roesler, A. Fasso`, J. Ranft; FLUKA: a multi-particle transport code", Proceedings of the Hadronic Shower Simulation Workshop 2006,Fermilab 6--8 September 2006, M. Albrow, R. Raja eds.,AIP Conference Proceeding 896, 31-49, (2007).

6- A. Fasso`, A. Ferrari, J. Ranft, and P.R. Sala; "The FLUKA code: Description and benchmarking", CERN-2005-10 (2005), INFN/TC_05/11, SLAC-R-773.

7- V.Vlachoudis "FLAIR: A Powerful But User Friendly Graphical Interface For FLUKA" Proc. Int. Conf. on Mathematics, Computational Methods & Reactor Physics (M&C 2009), Saratoga Springs, New York, 2009.