Romanian Journal of Physics 65, 302 (2020)

RADIATION PROTECTION ASSESSMENT FOR COMMISSIONING A FAST NEUTRON SOURCE BASED ON THE p(7Li,n)7Be INVERSE REACTION AT THE 9 MV TANDEM ACCELERATOR OF IFIN–HH

IRINA DINESCU1,2, GHEORGHE CĂTA-DANIL2 1 Horia Hulubei National Institute for Physics and Nuclear Engineering, 30 Reactorului,

P.O. Box MG – 6, RO – 077125 Bucharest – Măgurele, Romania E-mail: irina.stiru@nipne.ro

2 University “Politehnica” of Bucharest, Physics Department, 313 Splaiul Independentei, RO – 060042, Bucharest, Romania

E-mail: gheorghe.cata-danil@physics.pub.ro

Received January 13, 2020

Abstract. This paper reports the results for a radiation protection assessment done at the 9 MV Tandem Accelerator of Horia Hulubei National Institute for Physics and Nuclear Engineering for commissioning of a fast neutron source based on the 7Li(p,n)7Be inverse reaction. The assessment is based on the results obtained via numerical simulation with the use of FLUKA Monte Carlo simulation code. We have calculated the secondary radiation fields generated in different areas of the Tandem building during the acceleration process and the residual activity induced after a certain operation time.

Key words: fast neutron source, FLUKA, radiation protection.

1. INTRODUCTION

A fast neutron beam with well-defined parameters is highly required by many basic and applied physics studies [1]. Recently, such a beam has been obtained at the Institute for Nuclear Physics (IPN) in Orsay using the p(7Li,n)7Be inverse reaction [2]. In Bucharest, at Horia Hulubei National Institute for Physics and Nuclear Engineering (IFIN–HH) a 9 MV Tandem Accelerator is running since 1976 and Li beams are frequently used over the last decade. Therefore, the technical requirements to build a fast neutron source at Bucharest, in the first steps similar to the one built at IPN Orsay, are fulfilled. Besides the technical feasibility, an evaluation in terms of radiation protection has to be made so the main objective of this study is to investigate the radiological protection situation and to evaluate the existing shielding for the commissioning of a fast directional neutron source based on the p(7Li,n)7Be inverse reaction at the 9 MV Tandem Accelerator of IFIN–HH. For the purpose of this study we used the Monte Carlo numerical simulation code FLUKA [3, 4]. Due to the natural monoenergetic collimation of the neutron source, provided by the reaction itself, an entire class of experiments could be possible, such as high-resolution spectroscopy of neutron-induced reaction, nuclear cross section measurements, nuclear medicine or other fast neutron dosimetric applications.

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In this paper we report the results obtained via numerical simulation in terms of dose rates and residual activity induced by the fast neutron source obtained via the p(7Li,n)7Be inverse reaction. The accelerator structure, beam lines configuration and experimental areas specific to the 9 MV Tandem building have been considered and implemented in the numerical simulation. By taking into account relevant beam parameters such as the emittance, type, energy and intensity of the accelerated particle, the isotopic composition of the irradiated targets and structural materials we have calculated the secondary radiation fields generated in different areas of the 9 MV Tandem Accelerator building during the acceleration process and the residual activity induced after a certain accelerator operation time. Finally, discussions regarding the radiological impact during commissioning of a fast neutron source are made.

2. THE LICORNE NEUTRON SOURCE AT ORSAY

In 2013, at the IPN at Orsay was developed a unique, directional neutron source called LICORNE (Lithium Inverse Cinematiques Orsay NEutron source) which is based on fast neutron production by using the inverse kinematics p(7Li,n)7Be reaction. Conventional quasi-mono-energetic neutron sources produce neutrons via direct kinematics reaction on light nuclei, such as 7Li(p,n)7Be, but the problem that arises is that the resulting neutron flux does not have a well-defined directionality and less than 1 percent of it can be used for further experiments. The main advantage of the inverse kinematics is the natural forward collimation of the reaction ejectiles which results in increased neutron fluxes. The resulting fluxes depend on numerous factors but the most important ones are the 7Li beam emittance, energy and intensity and the thickness of the polypropylene (C3H6)n target. Assuming 7Li beam currents of 100 pnA, fluxes up to maximum 107 neutrons/s/sr are to be expected [2].

3. RADIOLOGICAL ASPECTS OF THE 9 MV TANDEM ACCELERATOR AT IFIN–HH

The 9 MV Tandem Accelerator is a complex installation that has as its main component the Van de Graaff acceleration system that allows a maximum voltage of 9 MV on the terminal. It can accelerate a wide range of ions with beam intensities between 100 pA and 2 μA [5] and the emittance of the Source of Negative Ions by Cesium Sputtering (SNICS II) is reported to be from 5 to less than 3 πmm mR (MeV)1/2 for 80% of the beam, depending upon the beam mass [6]. The radiological risk at the accelerator can occur during its operating time due to prompt radiation (primary beam, secondary beam that can be composed of X or gamma radiation and neutrons or charged particles, scattered radiation, bremsstrahlung radiation) and due to residual radioactivity, as a result of activation of the components

Radiation protection assessment at IFIN–HH Article no. 302 3

of the accelerator. The walls of the rooms where the accelerator tank and the beam transport systems are located, have been designed in such a way in order to fully attenuate the prompt radiation [7]. The source of the ionizing secondary radiations can be a result of the interactions of primary accelerated particles with the beam line components, as further described. The accelerating structure at the 9 MV Tandem Accelerator, as seen in Fig. 1, can be divided into 4 major parts in terms of secondary sources of ionizing radiation: i. The low energy transport line consisting of three main components, the ion sources, the inflection magnet and the Faraday cup, is treated as a low source of secondary ionizing radiation, mostly bremsstrahlung X-rays and characteristic X-rays. The Faraday cup has the biggest contribution in terms of generated doses for this area. ii. The accelerator tank, is considered as a medium source of secondary ionizing radiation, high energy bremsstrahlung X-rays, due to the fact that the walls of the tank attenuate significantly part of the generated ionizing radiation. iii. The high energy transport line consisting of another Faraday cup, the object slits, the analyzing magnet and the image slits, is treated as a high source of secondary ionizing radiation, mostly composed of gamma rays and neutrons, due to the fact that on this part of the accelerating structure we deal with the highest energy of the accelerated species. iv. The experimental area is composed of two halls with seven separate beam lines which lead to different reaction chambers. The ion beam coming from the image slits will be commuted by a switching magnet to one of the seven extensions. This last part of the accelerating structure is treated also as a high intensity source of ionizing radiations but is highly dependent on the type, on the intensity and on the energy of the accelerated particle. For the purpose of this study we considered as experimental line the beam line number seven located in the Experimental Hall II (see Fig. 1). Based on the nature of the activity of the personnel involved in experiments and operation of the Tandem Accelerator, there is a strict delimitation of the working areas in the 9 MV Tandem Accelerator building, as seen in Fig. 1: i. Controlled area which is defined as an area subject to special rules for protection against ionizing radiation or of preventing the radioactive contamination and to which access is controlled and restricted over time [8]. ii. Supervised area which is defined as an area subject to the adequate surveillance in the purpose of protection against ionizing radiation. In order to have a comprehensive evaluation of the risks involved during experiments, area monitoring is mandatory for both controlled and supervised areas, by means of ambient dose equivalent measurements, but also for individuals, by means of personal dose equivalent, in order to fulfill the acceptance criteria for the dose limits imposed by norms [8]. The active area system monitoring consisting of 12 gamma type BDKG–04 (G1–G12) and 4 neutron type BDKN–04 (N1–N4) detectors, which are connected

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to a computer dose display system, provides the ambient dose equivalent. All detectors are attached either on walls either or on a special built support and are positioned in key position, where are expected the highest doses due to primary or secondary sources of ionizing radiation, as seen in Fig. 1.

Fig. 1 – Fixed dose monitoring system at the 9 MV Tandem Accelerator of IFIN–HH.

The BDKG-04s are wide range X-ray and gamma radiation scintillation smart probes which are used for measuring the ambient dose equivalent and dose equivalent rate for continuous X-ray and gamma radiation in energy range 15 keV – 3 MeV up to 10 Sv/h [9]. The BDKN-04s are 3He proportional counters incorporated in a polyethylene moderator which are used for measuring the ambient dose equivalent and dose equivalent rate for neutrons in the energy range 0.025 eV – 14 MeV up to 10 mSv [10]. Both gamma and neutron detectors present a high immunity to external magnetic and electric fields. This feature is an important one and is necessary due to the fact that some of the detectors are placed in close vicinity of the analysing magnet of the 9 MV Tandem Accelerator where the highest dose rates are expected. According to the radiation protection regulation [8] the effective dose rate should not exceed 1 mSv/y for all areas and accelerator installations, which are within the premises and freely accessible for workers and visitors. Taking into account that in a working year there are 2000 working hours, the dose rate should not exceed 0.5 μSv/h for a public area and 3 μSv/h for a supervised area.

Radiation protection assessment at IFIN–HH Article no. 302 5

4. METHODOLOGY DESCRIPTION

As mentioned before the goal of the present study is twofold: one consists of scoring the neutron, gamma and total dose rates resulting from performing the p(7Li,n)7Be inverse reaction at the 9 MV Tandem Accelerator and second consists of estimating the residual activity of the accelerator components due to the beam and products of the same reaction. The targets characteristics, energy and beam intensity considered for this study are similar to the ones used at Orsay [2]. Two different types of targets were studied for the purpose of the present study. The first one consists of a simple layer of polypropylene of 4.4 μm thickness with a 50 μm gold backing. The second one consists of a 2.25 μm layer of TiH2 placed between two layers of gold of 200 nm and 50 μm thickness. We performed the simulations for 7Li beams with energies of 13.7 MeV and 16 MeV and considered as irradiation profile 2 weeks of active beam time with an intensity of 100 pnA. For the numerical simulations we have employed the FLUKA code which is a Monte Carlo code, used for calculation of particle transport and interaction with matter, with a broad spectrum of applications such as shielding, activation, dosimetry, radiotherapy, etc. [3, 4]. We also used the FLuka Advanced InteRface (FLAIR) that is a handy graphical user interface to run FLUKA [11]. The code’s input consists of different cards, each card belonging to particular categories. For simpler cases, the primary particle properties can be easily defined with only two default input cards: BEAM and BEAMPOS used for specifying the type of particle, its energy, starting position and direction. For the purpose of this simulation we also included the following cards: PHYSICS – used for enabling the COALESCENCE mechanisms and the new FLUKA EVAPORATION model [3, 4], EMFFLUO, used for activating photoelectric interaction [3, 4] and EMFRAY with Compton (bind+prof) used for activating Rayleigh scattering and Compton binding and profile function corrections in selected regions [3, 4]. For the calculation of the dose rates we used the USRBIN card [3, 4]. The doses are calculated by folding fluence with EWT74 conversion coefficients. This option uses effective dose sets from ICRP74 and Pelliccioni data calculated with ICRP radiation weighting factors wr for worst possible geometry of the irradiation [3, 4]. In the simulation input was also included the RADDECAY card which requests simulation of radioactive decays [3, 4] and IRRPROFI card which helped us define in terms of time and intensity the irradiation profile used for radioactive decay calculations [3, 4] and the residual nuclei produced were scored with RESNUCLEI card [3, 4]. The layout of Tandem building was implemented in the numerical simulation, as can be seen in Fig. 2. For the purpose of this study the following components were introduced in the numerical simulation: the accelerator tank, the high energy object slits, the analyzing magnet, the high energy image slits, the beam transport line (from tank exit to the reaction chamber) and the targets. These components

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were included because they are the ones interacting with the 7Li beam and they could contribute to the secondary field of generated ionizing radiation.

Fig. 2 – The geometry considered in the numerical simulation as seen in Flair Geoviewer Mode. It includes the accelerator and building components relevant for the radiation protection analysis.

5. RESULTS

5.1. DOSE RATES

For the first case we used a 7Li beam with an intensity of 100 pnA at 13.7 MeV. The dose rate results for the 1st considered target, the polypropylene with gold backing, can be seen in Fig. 3 for gamma particles, in Fig. 5 for neutrons and in Fig. 7 for all scored particles. The results for the 2nd target, the TiH2 sandwiched between two layers of gold, can be seen in Fig. 4 for gamma particles, in Fig. 6 for neutrons and in Fig. 8 for all scored particles. For the second case we used again a 7Li beam with an intensity of 100 pnA but this time at an energy of 16 MeV. The dose rate results for the 1st considered target, the polypropylene with gold backing, can be seen in Fig. 9 for gamma particles, in Fig. 11 for neutrons and in Fig. 13 for all scored particles. The results for the 2nd target, the TiH2 sandwiched between two layers of gold, can be seen in Fig. 10 for gamma particles, in Fig. 12 for neutrons and in Fig. 14 for all scored particles.

Radiation protection assessment at IFIN–HH Article no. 302 7

The simulation results for both analyzed cases have proven the efficiency of the existing shielding walls for the public area, which include the offices areas and the hallway, where the dose rates should not exceed 0.5 μSv/h. As it can be seen in Fig. 7, for the 1st considered target and in Fig. 8, for the 2nd target considered, for both targets using a 7Li beam at 13.7 MeV the simulated dose rate is around 0.1 μSv/h. For the 7Li beam at 16 MeV, the simulated dose rate is around 0.01 μSv/h as it can be seen in Fig. 13, for the 1st considered target and in Fig. 14, for the 2nd target considered.

1. 7Li beam @ 13.7 MeV with an intensity of 100 pnA

1st target – Polypropylene (PP) with gold backing

2nd target – TiH2 sandwiched between 2 layers of gold

Fig. 3 – Gamma dose for PP target (μSv/h).

Fig. 4 – Gamma dose for TiH2 target (μSv/h).

Fig. 5 – Neutron dose from PP target (μSv/h).

Fig. 6 – Neutron dose for TiH2 target (μSv/h).

Fig. 7 – ALL-PART dose for PP target (μSv/h).

Fig. 8 – ALL-PART dose for TiH2 target (μSv/h).

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2. 7Li beam @ 16 MeV with and intensity of 100 pnA

1st target – Polypropylene with gold backing

2nd target – TiH2 sandwiched between 2 layers of gold

Fig. 9 – Gamma dose for PP target (μSv/h).

Fig. 10 – Gamma dose for TiH2 target (μSv/h).

Fig. 11 – Neutron dose for PP target (μSv/h).

Fig. 12 – Neutron dose for TiH2 target (μSv/h).

Fig. 13 – ALL-PART dose for PP target (μSv/h).

Fig. 14 – ALL-PART dose for TiH2 target (μSv/h).

Some differences in dose rates may be observed in Figs. 3–14 for the two types of targets analyzed depending on the type of particle scored but given the fact that we are mainly interested in the dose rates for the public area, where no differences can be seen, it can be concluded that the existing shield is efficient for both analyzed cases.

Radiation protection assessment at IFIN–HH Article no. 302 9

As expected, the highest doses, around 102–103 μSv/h are in those parts of the accelerating structure where the beam interacts with the components: the image slits, the analyzing magnet and the object slits.

5.2. RESIDUAL ACTIVITY ESTIMATION

The operational licensing procedure requires also an evaluation of the residual radioactivity after certain periods of beam time. As mentioned before, we created two numerical simulations for both analyzed cases: one with 7Li beam with an intensity of 100 pnA at 13.7 MeV and one 7Li beam with an intensity of 100 pnA at 16 MeV. For both energies, we considered as irradiation time 2 weeks of active beam time. The activation of the materials located in the accelerator hall does not depend on the type of target used for the neutron production but only on the energy of the initial accelerated particle given the fact that the intensity of the 7Li beam is 100 pnA for all analyzed cases, thus in the presented results only a discussion from the energy point of view is made. The results for the residual activity for the accelerating structure, made of stainless steel, simulated for cooling time periods of immediately after beam stop, 1 hour, 1 day, 1 week, 1 month and after 1 year of accelerator operation in the conditions already described before, are compared with the limits given by safety norms [8] as seen in Table 1, for 7Li beam at 13.7 MeV and in Table 2, for 7Li beam at 16 MeV. From Table 1 we can observe that radionuclides with half-lives longer than 100 days are produced (65Zn, 57Co, 54Mn and 3H) due to the interaction of the 7Li beam at 13.7 MeV with the accelerating structure. From Table 2 we can observe that the same radionuclides with half – lives longer that 100 days are produced (65Zn, 57Co, 54Mn, 3H plus 60Co) but with higher contributions due to the fact that the energy is slightly increased for the second case analyzed. The object and image slits are made of tantalum and for both energies considered, the reaction is below the Coulomb barrier so those components cannot be activated during the accelerating process. The air activation in Experimental Hall II depends on the type of target used so for this part we used again the two targets described before. A 30 cm radius spherical detector made of air was placed inside the Experimental Hall II in the numerical simulation in order to check whether the air in the room is activated or not. After analyzing the results no air activation is produced in the scored detector thus neither in the room for either of the two types of targets used. Although some of the radionuclides have the activity higher than the exclusion level, safety norms state that if the total activity of the sample is lower than the sum of the exclusion levels of the radionuclides form that sample, the

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material can be considered industrial waste. For some radionuclides the literature does not include any exclusion levels but the contribution of those is taken into account when calculating the total activity of the analyzed sample.

Table 1

Residual activity for 7Li @ 13.7 MeV – 2 weeks irradiation

Isotope Half-life Activity at STOP

Activityafter

1 hour cooling

Activityafter 1 day

cooling

Activityafter

1 week cooling

Activityafter

1 monthcooling

Activityafter

1 year cooling

Relative error

Exclusion levels from

norms (T1/2) (Bq/cmc) (Bq/cmc) (Bq/cmc) (Bq/cmc) (Bq/cmc) (Bq/cmc) (%) (Bq/cmc)

Zn-65 244.26 d 8.88E+01 8.88E+01 8.85E+01 8.70E+01 8.15E+01 3.15E+01 11.35 3.00E+03

Zn-63 38.47 m 1.46E+03 4.97E+02 7.90E-09 0.00E+00 0.00E+00 0.00E+00 7.14

Cu-62 9.47 m 1.33E+04 2.49E+02 1.13E+01 2.16E-04 8.28E-23 0.00E+00 7.65

Co-61 1.65 h 3.47E+03 2.28E+03 1.45E-01 7.76E-28 0.00E+00 0.00E+00 9.96 1.00E+05

Cu-61 3.33 h 3.22E+05 2.62E+05 2.19E+03 2.16E-10 0.00E+00 0.00E+00 1.24

Co-57 271.79 d 1.01E+04 1.01E+04 1.01E+04 9.96E+03 9.38E+03 4.00E+03 0.70 5.00E+04

Mn-56 2.57 h 2.96E+03 2.26E+03 4.67E+00 7.21E-17 0.00E+00 0.00E+00 10.24 1.00E+04

Co-56 7.27 d 2.53E+03 2.53E+03 2.51E+03 2.38E+03 1.93E+03 9.59E+01 6.04 4.00E+03

Fe-55 2.73 y 1.20E+02 1.20E+02 1.21E+02 1.21E+02 1.19E+02 9.41E+01 5.62 3.00E+04

Co-55 17.53 h 1.53E+03 1.47E+03 5.93E+02 2.00E+00 4.46E-10 0.00E+00 14.12 9.00E+03

Mn-54 312.2 d 3.25E+03 3.25E+03 3.24E+03 3.20E+03 3.04E+03 1.45E+03 1.77 1.00E+04 Mn-52 // Mn-52 m

5.591 d // 21.1 m 1.99E+03 1.98E+03 1.76E+03 8.35E+02 4.58E+01 4.43E-17 9.32 6.00E+03 //

1.00E+04 Ar-37 35.04 d 6.26E+02 6.25E+02 6.13E+02 5.45E+02 3.43E+02 4.58E-01 11.00 –

S-35 87.32 d 1.71E+02 1.71E+02 1.70E+02 1.62E+02 1.35E+02 9.52E+00 12.12 4.00E+04 Cl-34 // Cl-34 m

1.52 s // 32 m 4.93E+03 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 7.70

P-33 25.34 d 4.45E+03 4.44E+03 4.33E+03 3.67E+03 1.94E+03 2.05E-01 5.26 4.00E+04

Cl-33 2.511 s 2.62E+03 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 11.64

P-32 14.262 d 3.44E+03 3.44E+03 3.28E+03 2.45E+03 7.85E+02 5.31E-03 4.72 4.00E+03

P-30 2.498 m 7.56E+04 4.44E-03 0.00E+00 0.00E+00 0.00E+00 0.00E+00 2.20

Al-28 2.2414 m 5.78E+02 5.06E-06 0.00E+00 0.00E+00 0.00E+00 0.00E+00 15.57

F-18 109.77 m 4.25E+03 2.91E+03 4.78E-01 9.67E-25 0.00E+00 0.00E+00 10.54 1.00E+04

F-17 64.49 s 1.33E+03 2.08E-14 0.00E+00 0.00E+00 0.00E+00 0.00E+00 15.45

H-3 12.33 y 1.22E+01 1.22E+01 1.22E+01 1.22E+01 1.22E+01 1.16E+01 7.13 2.00E+05

TOTAL 1.86E+04 1.24E+042.41E+032.12E+03 1.66E+032.49E+02 4.00E+04

Radiation protection assessment at IFIN–HH Article no. 302 11

Table 2 Residual activity for 7Li @ 16 MeV – 2 weeks irradiation

Isotope Half-life Activity at STOP

Activityafter

1 hour cooling

Activityafter 1 day

cooling

Activityafter

1 week cooling

Activityafter

1 month cooling

Activityafter

1 year cooling

Relative error

Exclusion levels from

norms (T1/2) (Bq/cmc) (Bq/cmc) (Bq/cmc) (Bq/cmc) (Bq/cmc) (Bq/cmc) (%) (Bq/cmc)

Ga-68 67.65 m 3.47E+03 1.88E+03 1.35E-03 4.70E-42 0.00E+000.00E+00 13.41 Ga-67 3.26 d 1.73E+04 1.72E+04 1.40E+04 3.92E+03 2.70E+01 3.53E-30 4.93 Ga-66 9.49 h 3.02E+03 2.81E+03 5.24E+02 1.42E-02 2.11E-20 0.00E+00 9.58 Zn-65 244.26 d 4.37E+03 4.37E+03 4.36E+03 4.28E+03 4.01E+031.55E+03 1.60 3.00E+03 Ga-65 15.2 m 4.04E+04 2.62E+03 1.22E-24 0.00E+00 0.00E+000.00E+00 2.60 Co-61 1.65 h 2.39E+04 1.57E+04 1.00E+00 5.35E-27 0.00E+000.00E+00 3.14 1.00E+05 Cu-61 3.33 h 1.47E+06 1.20E+06 1.00E+04 9.87E-10 0.00E+000.00E+00 0.62

Co-60 // Co-60 m

5.27 y // 10.4 m 1.14E+02 1.14E+02 1.14E+02 1.14E+02 1.13E+021.00E+02 4.00 1.00E+03 //

1.00E+06 Cu-60 23.7 m 4.80E+04 8.31E+03 2.46E-14 0.00E+00 0.00E+000.00E+00 1.97 Fe-59 44.5 d 3.23E+02 3.23E+02 3.18E+02 2.90E+02 2.01E+021.10E+00 13.51 6.00E+03 Cu-59 81.5 s 1.15E+04 5.81E-10 0.00E+00 0.00E+00 0.00E+000.00E+00 3.93

Co-58 // Co-58 m

70.86 d // 9.04 h 1.96E+05 1.96E+05 1.94E+05 1.83E+05 1.46E+055.51E+03 0.35 1.00E+04 //

3.00E+05 Mn-57 85.4 s 1.28E+04 2.61E-09 0.00E+00 0.00E+00 0.00E+000.00E+00 5.86 Co-57 271.79 d 3.33E+04 3.33E+04 3.32E+04 3.27E+04 3.08E+041.31E+04 0.71 5.00E+04 Mn-56 2.57 h 1.17E+04 8.95E+03 1.85E+01 2.85E-16 0.00E+000.00E+00 6.00 1.00E+04 Co-56 7.27 d 2.09E+04 2.09E+04 2.07E+04 1.97E+04 1.59E+047.92E+02 1.43 4.00E+03 Fe-55 2.73 y 4.43E+02 4.43E+02 4.44E+02 4.44E+02 4.37E+023.46E+02 2.66 3.00E+04 Co-55 17.53 h 4.23E+03 4.06E+03 1.64E+03 5.51E+00 1.23E-09 0.00E+00 8.83 9.00E+03 Mn-54 312.2 d 1.17E+04 1.17E+04 1.16E+04 1.15E+04 1.09E+045.18E+03 0.57 1.00E+04

Mn-52 // Mn-52 m

5.591 d // 21.1 m 8.89E+03 8.84E+03 7.85E+03 3.73E+03 2.05E+02 1.98E-16 5.95 6.00E+03 //

1.00E+04 Cr-51 27.7 d 6.29E+02 6.29E+02 6.14E+02 5.28E+02 2.94E+02 6.80E-02 9.93 3.00E+05 Ar-37 35.04 d 9.39E+02 9.38E+02 9.21E+02 8.18E+02 5.14E+02 6.87E-01 10.69 – S-35 87.32 d 3.71E+02 3.71E+02 3.68E+02 3.51E+02 2.92E+022.06E+01 11.66 4.00E+04

Cl-34 // Cl-34m

1.52 s // 32 m 7.97E+03 0.00E+00 0.00E+00 0.00E+00 0.00E+000.00E+00 3.23

P-33 25.34 d 8.67E+03 8.66E+03 8.43E+03 7.16E+03 3.77E+03 4.00E-01 3.57 4.00E+04 Cl-33 2.511 s 4.36E+03 0.00E+00 0.00E+00 0.00E+00 0.00E+000.00E+00 4.79 P-32 14.262 d 9.72E+03 9.70E+03 9.25E+03 6.91E+03 2.22E+03 1.08E-02 2.29 4.00E+03 Si-31 157 m 1.31E+03 1.00E+03 2.29E+00 6.69E-17 0.00E+000.00E+00 17.61 6.00E+04 P-30 2.498 m 1.27E+05 7.48E-03 0.00E+00 0.00E+00 0.00E+000.00E+00 2.04 F-18 109.77 m 5.67E+03 3.88E+03 6.37E-01 1.29E-24 0.00E+000.00E+00 7.88 1.00E+04 F-17 64.49 s 2.20E+03 3.45E-14 0.00E+00 0.00E+00 0.00E+000.00E+00 11.89 H-3 12.33 y 4.03E+01 4.03E+01 4.03E+01 4.03E+01 4.01E+013.81E+01 3.81 2.00E+05

TOTAL 6.69E+04 4.50E+04 8.78E+037.45E+035.82E+037.21E+02 5.48E+04

Article no. 302 Irina Dinescu, Gheorghe Căta-Danil 12

6. CONCLUSIONS

A future development of a fast neutron source at the 9 MV Tandem Accelerator at IFIN–HH in Bucharest was assessed from a radiological point of view. The reaction type, target and beam properties considered for this radiation protection study are similar to the ones used at the Orsay LICORNE facility. FLUKA Monte Carlo simulation code with the Flair interface has been used in order to estimate the radiation doses in different areas of the Tandem building and residual activity of the accelerator components as a result of commissioning a fast neutron source based on the p(7Li,n)7Be inverse reaction. All relevant components of the 9 MV Accelerator and of the Tandem building were implemented in the numerical simulation input. The analysis of the simulation results indicate that the shielding walls of the Tandem building are efficient and that the resulting dose rates are below the limits stated in legislation, therefore radiation protection requirements are satisfied for constructing such a facility. The results for the residual activity for the accelerating structure, made of stainless steel, simulated for cooling time periods of immediately after beam stop, 1 hour, 1 day, 1 week, 1 month and after 1 year of accelerator operation in the conditions already described before, are compared with the limits given by safety norms which state that if the total activity of the sample is lower than the sum of the exclusion levels of the radionuclides form that sample, the material can be considered industrial waste. For some radionuclides the literature does not include any exclusion levels but the contribution of those is taken into account when calculating the total activity of the analyzed sample. A further more detailed radioprotection evaluation will be done once a more precise characterization of the neutron source, as applied to the conditions of the 9 MV Tandem Accelerator of IFIN–HH, will be established. Acknowledgements. We would like to thank dr. Nicolae Marius Mărginean for all the support.

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Radiation protection assessment at IFIN–HH Article no. 302 13

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