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This paper aims to determine the levels of exposure to neutron and photon radiation among the personnel engaged in cyclotron operation and the personnel engaged in the production of radiopharmaceuticals, with the use of the environmental radiation monitoring system (RMS) installed in the positron emission tomography laboratory. The annual exposures of employees operating the cyclotron measured with the use of the RMS system are: 1.39 ± 0.16 mSv in case of photon radiation and 2.61 ± 0.14 mSv in case of neutron radiation. In the case of employees in the radiopharmaceuticals’ production zone, the annual exposures measured by means of the RMS system are 0.15 ± 0.03 mSv in case of photon radiation and 0.11 ± 0.01 mSv in case of neutron radiation. The exposure levels among the personnel engaged in cyclotron operation and the personnel engaged in the production of radiopharmaceuticals are below the permissible radiation dose limits.

7 pags., 3 figs. -- 27th International Nuclear Physics Conference (INPC2019), The tetraneutron has attracted the attention of nuclear physicists during the past decades, but there is still no unambiguous confirmation of its existence or nonexistence. A new experiment based on 8 He(p,2p)7H{t+4n} reaction, with direct detection of the four neutrons, has been carried out at RIBF, which can hopefully help to draw a definite conclusion on the tetraneutron system., We acknowledge the support of the RIBF accelerator staff and the BigRIPS team for providing the high-quality beam. Z. H. Yang acknowledges the financial support from the Foreign Postdoctoral Researcher program of RIKEN. T. Aumann acknowledges the support by DFG via SFB 1245. P. Koseoglou acknowledges the support from BMBF (NUSTAR.DA grant No.05P 15RDFN1).

A new detector setup (NICE-detector) based on an organic plastic scintillator is presented. It will be used during experiments measuring neutron-induced reactions with a charged particle in the exit channel. The proposed design was tested at the Goethe University Frankfurt. One of the test cases was the capture cross-section of 209Bi at different astrophysically important energies, including thermal capture cross-section and resonance integral. This research presents the performance of the detector setup, as well as preliminary results for the calculated cross-sections. The preliminary results demonstrate that the newly developed NICE-detector can be used to determine capture cross-sections with sufficient accuracy, and might be adopted in the future as an alternative for charged particle measurements in different nuclear and astrophysical applications.

This study developed a computationally efficient and easy-to-implement analytical model to estimate the equivalent dose from secondary neutrons originating in the bodies (‘internal neutrons’) of children receiving intracranial proton radiotherapy. A two-term double-Gaussian mathematical model was fit to previously published internal neutron equivalent dose per therapeutic absorbed dose versus distance from the field edge calculated using Monte Carlo simulations. The model was trained using three intracranial proton fields of a 9-year-old girl. The resulting model was tested against two intracranial fields of a 10-year-old boy by comparing the mean doses in organs at risk of a radiogenic cancer estimated by the model versus those previously calculated by Monte Carlo. On average, the model reproduced the internal neutron organ doses in the 10-year-old boy within 13.5% of the Monte Carlo at 3–10 cm from the field edge and within a factor of 2 of the Monte Carlo at 10–20 cm from the field edge. Beyond 20 cm, the model poorly estimated H/DRx, however, the values were very small, at

We present fitting equations for estimating effective dose per unit fluence at any photon energy between 10 keV and 10 GeV and any neutron energy between 0.001 eV and 10 GeV. These new equations are based on the latest radiation protection quantities for external radiation exposure found in International Commission on Radiological Protection (ICRP) Publication 116 and incorporate the latest definition of effective dose as described in ICRP Publication 103. The ICRP 116 dose coefficients were fit to piecewise polynomial functions. A total of 8 irradiation geometries were considered: the six in ICRP 116 and two additional geometries presented elsewhere in the literature. The fitting functions generally reproduce the ICRP 116 data to within 3% or better. The functions were used to modify the Monte Carlo N-Particle radiation transport code version 6 (MCNP6) and were applied to a sample problem. The results are intended to be used as a basis for revising the American National Standards Institute/American Nuclear Society 6.1.1-1991 standard.

A compilation of fluence-to-absorbed dose conversion coefficients for the dose to the lens of the eye is presented. The compilation consists of both previously published data and newly calculated values: photon data (5 keV–50 MeV for both kerma approximation and full electron transport), electron data (10 keV–50 MeV), and positron data (1 keV–50 MeV) – neutron data will be published separately. Values are given for angles of incidence from 0° up to 90° in steps of 15° and for rotational irradiation. The data presented can be downloaded from this article's website and they are ready for use by Report Committee (RC) 26. This committee has been set up by the International Commission on Radiation Units and Measurements (ICRU) and is working on a ‘proposal for a redefinition of the operational quantities for external radiation exposure’.

Plants, algae, and their derivatives (paper, textiles, etc.) are complex systems that are chiefly composed of a web of cellulose fibers. The arrangement of solvents within the polymeric structure is of great importance since cellulose degradation is strongly influenced by water accessibility and external agents. In this paper we develop a model that is able to deconvolve the scattering contributions of both polymeric structures and solvent clusters trapped along the polymeric fibers. The surface morphology of cellulose fibers and the spatial distribution of water-filled pores and their dimensions have been recovered from small angle neutron scattering and atomic force microscopy data in function with paper degradation. In addition to providing a boost to the effort to preserve cellulose-supported material (included cultural heritage), the relevance of our model resides in the exploitation of a large number of biopolymer networks that are known to share structures similar to that of cellulose.

ÖZET Radyasyon etkilerinin canlılar üzerinde yarattığı büyük tehlikeden dolayı, nükleer enerji kullanılmasında radyasyon korunmasına büyük önem verilmiş ve çeşitli zırh- lama teknikleri ile zırh malzemeleri geliştirilmiştir. Bu amaçla geliştirilen zırh malzemesi, nötronların zırhlama işleminde endüstriyel atık malzemesi olan kâğıdın kullanılmasıyla gerçekleştirilmiştir. Kâğıdın endüstriyel alanda kolayca kullanılabilen ekonomik bir malzeme oluşu ve yüksek oranda selüloz (C6H12O5) hammaddesini içermesi onu seçici kılmaktadır. Kâğıdın zırh malzemesi olarak başlı başına kullanımı zırhlama tekniğinde yetersiz kalmaktadır. Bundan dolayı katkı malzemesi olarak bor elementi gibi nötronları fazla sıyla zayıflatan yeni bir zırh malzemesi geliştirilmesi amaçlanmıştır. Katkı malzemesi olarak seçilen bor elementinin nötron yutma tesir kesidinin yüksek, yakalama reaksiyonunda yayınladıkları gama enerjisinin düşük olması kağıt ile iyi bir zırh malzemesi oluşturmasını sağlamıştır. Deneysel çalışmanın ilk aşamasında sadece kağıt mal zemesinden sayımlar alınmış ve nötronların en fazla soğu rulduğu zırh kalınlığı seçilmiştir. Bu zırhlama etkisinin aynısını veya fazlasını daha az kalınlıkta zırhlama malzemesi kullanılarak gerçekleştirmek amaçlanmıştır. Farklı konsantrasyonlarla hazırlanmış zırh malzemesinden sayı mi ar alınmış ve ideal zırh kalınlığı belirlenmiştir. Kağıt malzemeye katılan borik asidin nötronlara karşı zırhlama yeteneğinden yararlanarak daha az kalınlıkta bir zırh malzemesi geliştirilmiştir. Düşük konsantrasyonlarda borik asit kullanılarak daha ince kağıt ile oldukça etkin bir zırh malzemesini üretmenin mümkün olabileceği görülmüştür. İV INDUSTRIAL RESIDUAL PAPER USES THE BOR ELEMENT IN NEUTRON SHIELDING TO DEVELOPE AN EFFECTIVE SHIELD ABSTRACT Nowadays, the use of radiation and radioactive materials increases rapidly and the field of applications spreads widely. Because of the increasing of the usage of radiation, several radiation risk factors come along in it. Radiation effect on living things is so important and dangerous. Therefore to protect the human being from the radiation, the most important factor is shielding. New shielding techniques were developed for the protection of radiation. The use of shielding makes the possibility of usage of radiation all over the world. Of course for the development of shielding, the type and the activity of the radiation should be known. It was the main aim of this work to develop a new shielding materials for neutrons in order to protect the living tissues from radiation. In this work, neutrons from Howitzer neutron source (NH-3) produced by (a, n) reactions used. Experimental set up in this work, made from neutron Howitzer, BF3 detector, high voltage supply, amplifier and counting system. The body of neutron Howitzer source contains an irradiation canal in which BF3 detector and shielding material are placed. The BF3 detector is placed behind the shielding material for counting the neutrons, penetrated from the matter. The BF3 detector which is placed inside of the neutron Howitzer source is connected to high voltage supply and amplifier. The detector that takes the voltage from the high voltage supply transforms to voltage signals in itself. Because of the weakness of these signals amplification became necessary and the amplifier used for this purpose.In this work, it was aimed to use the paper in order to use the industrial residual material for the process of neutron shielding. Different quantities of the paper are chosen because of their inexpensive and widely uses in many industries. The most important reason for choosing the paper material in this work is its components. The paper material is formed by high percentage cellulose (C6H12O5). As it shown in the cellulose chemical formula, carbon and especially hydrogen elements are in large amount which makes the cellulose has a big advantageous to slow down the neutrons. For the shielding of the neutrons, the industrial residual paper are used with bor element to produce a new cheap material which is not being tested before. Bor element is not found by naturally and also it is not an element that is found in the crust of earth. However Turkey is one of the top countries that is so rich with underground bor element sources. The main criteria of choosing of additive materials should have high neutron absorption cross section and the low energy gammas emitted in the capture reactions. It was observed that in this conditions bor element with the paper material can be the best shielding for neutron protection. The aim of this experiment is to form an ideal shielding material by using bor element and paper materials together. Therefore the produced shielding is going to reduce the intensity of radiation by absorption. Generally, instead of stopping radiations completely it is preferred reducing their intensities into maximum permissible level. After getting positive results from developed shielding material, it is planned to produce a new shielding material. In the first stage of experimental work, it is aimed to know how effective the paper material is as a shielding material against neutrons without using any other additive elements. Since paper material is found to be not effective vienough for neutrons by itself in the shielding techniques it is planned to used bor element as an additive material. In the stage of preparing shielding material, boric acid solution has been made in different concentrations (0,25, 0,5, 0,75, 1, 3%), and different amounts of bor elements were used. Paper material in different solutions has been absorbed by different chemical reactions and thus the shielding material has been formed. Different levels of boric acid are worked in order to find the ideal amount of boric acid concentration to mix with paper material. Separate shielding material is prepared for each boric acid concentration levels. The counts taken from the paper material are examined, and it was satisfactorily observed that absorption of neutrons were increased as the paper thickness increases. But one of the aim of this work is to form thinner and more effective shielding material against radiation. First of fall the counts are taken from the shielding material having 0,25 % boric acid concentration and these counts are seen more effective in neutrons shielding comparing with the paper material. Increasing the thickness of shielding material having 0,25 % boric acid concentration, the increasing is observed in absorption of neutrons. It is observed that effective results to absorb neutron are achieved the paper material with 0,25 % boric acid concentration paper when compared with some thickness, and without the boric acid. But the differences between them are not in great values. In order to form more effective shielding material comparing with the paper material the more boric acid concentration is needed. Therefore, shielding material having 0,5 % boric acid concentration are prepared when the results are compared with different shielding materials having 0,25 % boric acid concentration observed the values more effective. When boric concentration is increased to 0,75 %, similar Vllimprovement are observed comparing with 0,25 % and 0,5 % boric acid concentration. These different boric acid concentrations (0,25 %, 0,5 % and 0,75 %) couldn't absorb neutrons at the desired level. When compared with the paper material it is found that the great differences are not shown on different boric acid concentrations. When boric acid concentration level is increased to 1 %, very different values are observed compared with 0,25 %, 0,5 % and 0,75 % boric acid concentration used in shielding materials. Increasing the thickness of shielding material, the high values of absorption are observed. These high values for absorption of neutrons show that how effective the boric acid is in the paper material. The shielding material having 1 % boric acid concentration proved that absorption neutrons are more effective than 0,25 %, 0,5 % and 0,75 % boric acid concentration. This results leaded to measure the results from 3 % and 5 % boric acid concentration. The results are taken separately for 3 % and 5 % boric acid concentration. When these counts are compared with 1 % boric acid concentration different absorptions were seen. Differences of concentrations in shielding material gives very similar absorption results and therefore there is no need to use the high concentrated sampled. Finally, it was thought that 1 % boric acid concentration is preferable. The usage of 1 % boric acid concentration, shielding material is more economical than the other higher level concentration shielding materials. The main aim of this work is to use the shielding material developed in the industrial field. 1 % boric acid concentration is chosen because all three types of concentration (1 %, 3 % and 5 %) give similar results. After choosing the ideal boric acid concentration level the last stage of experimental work is to decide the ideal thickness of neutron shielding material. This is done viiiby the help of the results of thickness of papers and the boric acid concentration. First of fall, it is found that 90,4 % neutron absorption observed by using 5 cm thickness of paper material without any additive materials. The final aim of this work is to achieve the same or higher level of shielding with thinner shielding materials. Then it is started to use different boric acid concentrations with different thickness of paper material and different level of boric acid concentration the shielding material which is formed in this thickness and concentration showed the absorption of neutron in 95,5 %. Since success of shielding for neutrons it is found to be sufficient to use 1 % boric acid concentration instead of using 5 cm thickness of paper material by itself. As a result thinner shielding material is achieved by using effectiveness of boric acid concentration as a shielding to absorb neutron with paper material. Therefore it is possible to produce effective shielding material by using lower concentrations of boric acid with a thinner paper. IX 66