Thermal Neutron Fluence Research Articles

Neutron-induced modifications on Hostaphan and Makrofol wettability and etching behaviors

Understanding the nature of polymers used as nuclear detectors is crucial to enhance their behaviors. In this work, the induced modifications in wettability and etching properties of Hostaphan and Makrofol polymers irradiated by different fluences of thermal neutrons are investigated. The wetting properties are studied by contact angle technique which showed the spread out of various liquids over the irradiated polymers surfaces (wettability enhanced). This wetting behavior is attributed to the induced changes in surface free energy (SFE), morphology, roughness, structure, hardness, and chemistry. SFE values are calculated by three different models and found to increase after neutrons irradiation associated with differences depending on the used model. These differences result from the intermolecular interactions in the liquid/polymer system. Surface morphology and roughness of both polymers showed drastic changes after irradiation. Additionally, surface structure and hardness of pristine and irradiated polymers were discussed and correlated to the surface wettability improvements. The changes in surface chemistry are examined by Fourier transform infrared spectroscopy (FTIR), which indicate an increase in surface polarity due to the formation of polar groups. The irradiated polymers etching characteristics and activation energies are discussed as well. Lastly, it is evident that thermal neutrons show efficiency in improving surface wettability and etching properties of Hostaphan and Makrofol in a controlled way.

Triple ionization chamber method for clinical dose monitoring with a Be-covered Li BNCT field.

Fast neutron, gamma-ray, and boron doses have different relative biological effectiveness (RBE). In boron neutron capture therapy (BNCT), the clinical dose is the total of these dose components multiplied by their RBE. Clinical dose monitoring is necessary for quality assurance of the irradiation profile; therefore, the fast neutron, gamma-ray, and boron doses should be separately monitored. To estimate these doses separately, and to monitor the boron dose without monitoring the thermal neutron fluence, the authors propose a triple ionization chamber method using graphite-walled carbon dioxide gas (C-CO2), tissue-equivalent plastic-walled tissue-equivalent gas (TE-TE), and boron-loaded tissue-equivalent plastic-walled tissue-equivalent gas [TE(B)-TE] chambers. To use this method for dose monitoring for a neutron and gamma-ray field moderated by D2O from a Be-covered Li target (Be-covered Li BNCT field), the relative sensitivities of these ionization chambers are required. The relative sensitivities of the TE-TE, C-CO2, and TE(B)-TE chambers to fast neutron, gamma-ray, and boron doses are calculated with the particle and heavy-ion transport code system (PHITS). The relative sensitivity of the TE(B)-TE chamber is calculated with the same method as for the TE-TE and C-CO2 chambers in the paired chamber method. In the Be-covered Li BNCT field, the relative sensitivities of the ionization chambers to fast neutron, gamma-ray, and boron doses are calculated from the kerma ratios, mass attenuation coefficient tissue-to-wall ratios, and W-values. The Be-covered Li BNCT field consists of neutrons and gamma-rays which are emitted from a Be-covered Li target, and this resultant field is simulated by using PHITS with the cross section library of ENDF-VII. The kerma ratios and mass attenuation coefficient tissue-to-wall ratios are determined from the energy spectra of neutrons and gamma-rays in the Be-covered Li BNCT field. The W-value is calculated from recoil charged particle spectra by the collision of neutrons and gamma-rays with the wall and gas materials of the ionization chambers in the gas cavities of TE-TE, C-CO2, and TE(B)-TE chambers (10B concentrations of 10, 50, and 100 ppm in the TE-wall). The calculated relative sensitivity of the C-CO2 chamber to the fast neutron dose in the Be-covered Li BNCT field is 0.029, and those of the TE-TE and TE(B)-TE chambers are both equal to 0.965. The relative sensitivities of the C-CO2, TE-TE, and TE(B)-TE chambers to the gamma-ray dose in the Be-covered Li BNCT field are all 1 within the 1% calculation uncertainty. The relative sensitivities of TE(B)-TE to boron dose with concentrations of 10, 50, and 100 ppm 10B are calculated to be 0.865 times the ratio of the in-tumor to in-chamber wall boron concentration. The fast neutron, gamma-ray, and boron doses of a tumor in-air can be separately monitored by the triple ionization chamber method in the Be-covered Li BNCT field. The results show that these doses can be easily converted to the clinical dose with the depth correction factor in the body and the RBE.

Peripheral neutron dose model verification for real IMRT cases

Introduction Peripheral doses are directly related to second cancer risk after radiotherapy. Our group developed a methodology to estimate neutron contribution to peripheral organ doses by terms of two general models, namely abdomen and head and neck [PMB-2012;57:6167–6191]. Purpose This work aims to verify the validity of these models in real treatments, in order to evaluate the need of further improvements for specific locations beside the two generic ones. Materials and methods Neutron doses were calculated in 12 representative organs from measured thermal neutron fluences with TNRD detectors at 16 points inside the phantom [MedPhys-2014;41:112105], for two high energy (15 MV) treatments (lung and prostate). Following the methodology described in [RO107:234–241], these neutron doses were estimated by terms of number of delivered MU and facility characterization. Abdomen model was used for the prostate case while both (abdomen and headn42:276–281]. Results Values generally agreed within the 30% uncertainty range established for the models and the 15% for the measurement. Abdomen model has shown to fit better for the lung. Further studies should be needed to improve generic models in some specific locations such as skin or organs close to the field-edge. Conclusion The generic model has shown to be good enough to cover frequent high-energy specific treatments as those studied here. It seems to be no need of more specific models, while some improvements have to bee done for particular points.

Fuel temperature prediction using a variable bypass gap size in the prismatic VHTR

The temperature gradient and hot spot temperatures were calculated in the prismatic very high temperature reactor as a function of the variable bypass gap size. Many previous studies have predicted the temperature of the reactor core based on a fixed bypass gap size. The graphite matrix of the assemblies in the reactor core undergoes a dimensional change during the operation due to thermal expansion and neutron fluence. The expansion and shrinkage of the bypass gaps change the coolant flow fractions into the coolant channels, the control rod holes, and the bypass gaps. Therefore, the temperature of the assemblies may differ compared to those for the fixed bypass gap case. The temperature gradient and the hot spot temperatures are important for the design of reactor structures to ensure their safety and efficiency. In the present study, the temperature variation of the PMR200 is studied at the beginning (BOC), middle (MOC), and end (EOC) of cycles 07 and 14. CORONA code which has been developed in KAERI is applied to solve the thermal-hydraulics of the reactor core of the PMR200. CORONA solves a fluid region using a one-dimensional formulation and a solid region using a three-dimensional formulation to enhance the computational speed and still obtain a reasonable accuracy. The maximum temperatures in the fuel assemblies using the variable bypass gaps did not differ much from the corresponding temperatures using the fixed bypass gaps. However, the maximum temperatures in the reflector assemblies using the variable bypass gaps differ significantly from the corresponding temperatures using the fixed bypass gaps.

Measurements of neutron radiation and induced radioactivity for the new medical linear accelerator, the Varian TrueBeam

Contemporary linear accelerators applied in radiotherapy generate X-ray and electron beams with energies up to 20 MeV. Such high-energy therapeutic beams induce undesirable photonuclear (γ,n) and electronuclear (e,e'n) reactions in which neutrons and radioisotopes are produced. The originated neutron can also induce reactions such as simple capture, (n,γ), reactions that produce radioisotopes. In this work measurements of the non-therapeutic neutrons and the induced gamma radiation were carried out in the vicinity of a new medical accelerator, namely the Varian TrueBeam. The TrueBeam is a new generation Varian medical linac making it possible to generate the X-ray beams with a dose rate higher than in the case of the previous models by Varian. This work was performed for the X-ray beams with nominal potentials of 10 MV (flattening filter free), 15 MV and 20 MV, and for a 22 MeV electron beam. The neutron measurements were performed by means of a helium chamber and the induced activity method. The identification of radioisotopes produced during emission of the therapeutic beams was based on measurements of the energy spectra of gammas emitted in decays of the produced nuclei. The gamma energy spectra were measured with the use of the high-purity germanium detector. The correlation between the neutron field and the mode and nominal potential was observed. The strongest neutron fluence of 3.1 × 106 cm−2 Gy−1 and 2.0 × 106 cm−2 Gy−1 for the thermal and resonance energies, respectively, was measured during emission of the 20 MV X-ray beam. The thermal and resonance neutron fluence measured for the 15 MV X-rays was somewhat less, at 1.1 × 106 cm−2 Gy−1 for thermal neutrons and 6.7 × 105 cm−2 Gy−1 for resonance neutrons. The thermal and resonance neutron fluences were smallest for the 10 MV FFF beam and the 22 MeV electron beam and were around two orders of magnitude smaller than those of the 20 MV X-ray beam. This work has shown that the neutron reactions are dominant because of relatively high cross sections for many elements used in the accelerator construction. The detailed analysis of the measured spectra made it possible to identify 11 radioisotopes induced during TrueBeam delivery. In this work the following radioisotopes were identified: 56Mn, 122Sb, 124Sb, 131Ba, 82Br, 57Ni, 57Co, 51Cr, 187W, 24Na and 38Cl.

Measurements of Thermal and Resonance Neutron Fluence and Induced Radioactivity Inside Bunkers of Medical Linear Accelerators in the Center of Oncology in Opole, Poland

Measurements of Thermal and Resonance Neutron Fluence and Induced Radioactivity Inside Bunkers of Medical Linear Accelerators in the Center of Oncology in Opole, Poland

LiCaAlF6 scintillators in neutron and gamma radiation fields

Intentionally doped LiCaAlF6 (LiCAF) single crystals are prospective scintillators, especially for thermal neutron detection through the 6Li(n,t)4He nuclear reaction. Four different LiCAF scintillator samples were tested in various neutron and gamma fields. Two of the tested samples were LiCAF:Eu and LiCAF:Eu,Na single crystals, and another two samples were made of LiCAF:Eu micro crystals dispersed in transparent rubber, with different rubber dimensions. All LiCAF samples contain lithium enriched to[Formula: see text]Li. A plutonium–beryllium source was used as a neutron source. The neutron spectrum was modified by moderator and filter to get different ratios between thermal, epithermal and fast neutron fluence rates. The MCNP code was used for calculations of the fluence rates for different configurations. Radionuclides [Formula: see text]Cs and [Formula: see text]Co were applied as gamma radiation sources. The light signal from the scintillator was evaluated with a photomultiplier and a multichannel analyzer. The purpose of this work was to study the characteristics of LiCAF scintillators, especially the ability to discriminate signals from neutron and gamma radiation, which is the basic scintillator condition for neutron detection in mixed neutron-gamma radiation fields. Generally, the discrimination can be done by the pulse height and/or the pulse shape of the evaluated signals. Both methods can be used for a LiCAF scintillator. However, only the pulse height discrimination method is discussed in this paper. The possibility of fast neutron detection with LiCAF scintillators was also tested.

CaSO4:Dy microphosphor for thermal neutron dosimetry

Dysprosium-doped calcium sulphate (CaSO4:Dy) microphosphor was synthesized by acid re-crystallization method and its thermoluminescence (TL) properties irradiated with thermal neutrons was studied. Structural and morphological characteristics have been studied using X-ray diffraction and SEM which mainly exhibits a orthorhombic structure with particle size of 200 to 250µm. Moreover, thermal neutron dosimetric characteristics of the microphosphor such as thermoluminescence glow curve, TL dose–response have been studied. This microphosphor powder represents a TL glow peak (Tmax) centered at around 240°C. The TL response of CaSO4:Dy microphosphor as a function of thermal neutron fluence is observed to be very linear upto the fluence of 52×1011n/cm2 and further saturates. In addition, TL glow curves were deconvoluted by computerized glow curve deconvolution (CGCD) method and corresponding trapping parameters have been determined. It has been found that for every deconvoluted peak there is change in the order of kinetics. Overall, the experimental results show that the CaSO4:Dy microphosphor can have potential to be an effective thermal neutron dosimetry.

Radiotherapy dose enhancement using BNCT in conventional LINACs high-energy treatment: Simulation and experiment

AimTo employ the thermal neutron background that affects the patient during a traditional high-energy radiotherapy treatment for BNCT (Boron Neutron Capture Therapy) in order to enhance radiotherapy effectiveness. BackgroundConventional high-energy (15–25MV) linear accelerators (LINACs) for radiotherapy produce fast secondary neutrons in the gantry with a mean energy of about 1MeV due to (γ, n) reaction. This neutron flux, isotropically distributed, is considered as an unavoidable undesired dose during the treatment. Considering the moderating effect of human body, a thermal neutron fluence is localized in the tumour area: this neutron background could be employed for BNCT by previously administering 10B-Phenyl-Alanine (10BPA) to the patient. Materials and methodsMonte Carlo simulations (MCNP4B-GN code) were performed to estimate the total amount of neutrons outside and inside human body during a traditional X-ray radiotherapy treatment.Moreover, a simplified tissue equivalent anthropomorphic phantom was used together with bubble detectors for thermal and fast neutron to evaluate the moderation effect of human body. ResultsSimulation and experimental results confirm the thermal neutron background during radiotherapy of 1.55E07cm−2Gy−1.The BNCT equivalent dose delivered at 4cm depth in phantom is 1.5mGy-eq/Gy, that is about 3Gy-eq (4% of X-rays dose) for a 70Gy IMRT treatment. ConclusionsThe thermal neutron component during a traditional high-energy radiotherapy treatment could produce a localized BNCT effect, with a localized therapeutic dose enhancement, corresponding to 4% or more of photon dose, following tumour characteristics. This BNCT additional dose could thus improve radiotherapy, acting as a localized radio-sensitizer.

Study of suitability of Fricke-gel-layer dosimeters for in-air measurements to characterise epithermal/thermal neutron beams for NCT

The reliability of Fricke gel dosimeters in form of layers for measurements aimed at the characterization of epithermal neutron beams has been studied. By means of dosimeters of different isotopic composition (standard, containing 10B or prepared with heavy water) placed against the collimator exit, the spatial distribution of gamma and fast neutron doses and of thermal neutron fluence are attained. In order to investigate the accuracy of the results obtained with in-air measurements, suitable MC simulations have been developed and experimental measurements have been performed utilizing Fricke gel dosimeters, thermoluminescence detectors and activation foils. The studies were related to the epithermal beam designed for BNCT irradiations at the research reactor LVR-15 (Řež). The results of calculation and measurements have revealed good consistency of gamma dose and fast neutron 2D distributions obtained with gel dosimeters in form of layers. In contrast, noticeable modification of thermal neutron fluence is caused by the neutron moderation produced by the dosimeter material. Fricke gel dosimeters in thin cylinders, with diameter not greater than 3mm, have proved to give good results for thermal neutron profiling. For greater accuracy of all results, a better knowledge of the dependence of gel dosimeter sensitivity on radiation LET is needed.

피션트랙 검출기법에 의한 고체 및 액체상태 지질물질의 우라늄 정량과 그 적용성

피션트랙 검출에 의해 고체 및 액체상태 지질물질 내 우라늄을 정밀정량하는 최적기법을 제안하고 그 적용성을 검증하였다. 우라늄 정량은 중성자 조사에 의해 유도시킨 <TEX>$^{235}U$</TEX>의 핵분열 흔적을 트랙 디텍터에 기록하여 고배율 현미경 하에서 계수함으로써 이루어진다. 우라늄-친화력이 좋은 암석(예: 화강암, 석탄)을 분말 펠릿시료로 만들어 건식 검출하면, 그다지 높지 않은 우라늄함량(<5 ppm, 즉 <TEX>${\mu}g\;g^{-1}$</TEX>)에서도 빈번하게 함유된 함-우라늄광물에 기인한 트랙 군집현상으로 인해 시료의 전체평균 함량 결정이 쉽지 않았으며, 트랙 균질부와 군집부를 별도로 검토해야 한다. 시료의 균질성이 유지된다면 백운모-Lexan 디텍터에 의한 중복 측정과 여러 차례 중성자 조사에 의한 반복 측정에서 높은 재현성이 확인되었다. 건고 액체시료 경우에 <TEX>$10^1ppm$</TEX> 수준 및 그 이상의 우라늄함량에서 흔히 나타나는 불균질 현상을 극복하기 위해, 진공 석영튜브를 이용한 습식 검출법을 제안한다. 습식 검출에서 우라늄 균질성은 <TEX>$10^0ppm$</TEX> 수준 이하에서 회복되며, 측정하한은 <TEX>$10^2ppb$</TEX> (i.e. <TEX>$ng\;g^{-1}$</TEX>) 수준까지 무난한 것으로 입증되었다. The fission track registration techniques for accurate determinations of uranium in solid- and liquid-state geological materials were recommended and their applicability were examined. The determination of uranium can be achieved by optical counting of neutron-induced fission tracks of <TEX>$^{235}U$</TEX> registered on solid-state track detectors under high magnification. In a dry registration method using powdered pellets of rocks (e.g., granite and coal) showing good uranium-affinity, it was not easy to decide an overall mean concentration over the total sample owing to track-clusters caused by frequent presence of uranium-bearing minerals. Separate scanning for homogeneous and track-clustered parts may be an alternative choice. Assuring the homogeneity over the whole sample, high reproducibilities were confirmed both from duplicate detections using mica and Lexan polycarbonate detectors and from multiple measurements at different thermal neutron fluences. The wet registration method using sealed quartz tubes is recommended to overcome the common heterogeneity in uranium concentrations of <TEX>$10^1ppm$</TEX> and more. Adopting the wet registration, the uranium homogeneity was recovered below the <TEX>$10^0ppm$</TEX> level and the lower detection limit was proved to reach without difficulty the <TEX>$10^2ppb$</TEX> (i.e. <TEX>$ng\;g^{-1}$</TEX>) level.

SU‐E‐T‐365: Estimation of Neutron Ambient Dose Equivalents for Radioprotection Exposed Workers in Radiotherapy Facilities Based On Characterization Patient Risk Estimation

Purpose:Previous measurements with Bonner spheres1 showed that normalized neutron spectra are equal for the majority of the existing linacs2. This information, in addition to thermal neutron fluences obtained in the characterization procedure33, would allow to estimate neutron doses accidentally received by exposed workers, without the need of an extra experimental measurement.Methods:Monte Carlo (MC) simulations demonstrated that the thermal neutron fluence distribution inside the bunker is quite uniform, as a consequence of multiple scatter in the walls⁴. Although inverse square law is approximately valid for the fast component, a more precise calculation could be obtained with a generic fast fluence distribution map around the linac, from MC simulations⁴. Thus, measurements of thermal neutron fluences performed during the characterization procedure3, together with a generic unitary spectra2, would allow to estimate the total neutron fluences and H*(10) at any point⁵. As an example, we compared estimations with Bonner sphere measurements1, for two points in five facilities: 3 Siemens (15–23 MV), Elekta (15 MV) and Varian (15 MV).Results:Thermal neutron fluences obtained from characterization, are within (0.2–1.6×10⁶) cm−2•Gy−1 for the five studied facilities. This implies ambient equivalent doses ranging from (0.27–2.01) mSv/Gy 50 cm far from the isocenter and (0.03–0.26) mSv/Gy at detector location with an average deviation of ±12.1% respect to Bonner measurements.Conclusion:The good results obtained demonstrate that neutron fluence and H*(10) can be estimated based on: (a) characterization procedure established for patient risk estimation in each facility, (b) generic unitary neutron spectrum and (c) generic MC map distribution of the fast component.[1] Radiat. Meas (2010) 45: 1391 – 1397; [2] Phys. Med. Biol (2012) 5 7:6167–6191; [3] Med. Phys (2015) 42:276 ‐ 281. [4] IFMBE (2012) 39: 1245–1248. [5] ICRU Report 57 (1998)

SU‐E‐T‐195: Commissioning the Neutron Production of a Varian TrueBeam Linac

Purpose:The purpose of this work is the characterization of a new Varian TrueBeam™ facility in terms of neutron production, in order to estimate neutron equivalent dose in organs during radiotherapy treatments.Methods:The existing methodology [1] was used with the reference SRAMnd detector, calibrated in terms of thermal neutron fluence at the reference field operated by PTB (Physikalisch‐Technische‐Bundesanstalt) at the GeNF (Geesthacht‐Neutron‐Facility) with the GKSS reactor FRG‐1 [2]. Thermal neutron fluence for the 5 available possibilities was evaluated: 15 MV and 10&amp;6 MV with and without Flattening Filter (FF and FFF, respectively). Irradiation conditions are as described in [3]. In addition, three different collimator‐MLC configurations were studied for 15 MV: (a) collimator of 10×10 cm2 and MLC fully retracted (reference), (b) field sizes of 20×20 cm2 and 10×10 cm2 for collimator and MLC respectively, and (c) collimator and MLC aperture of 10×10 cm2.Results:Thermal fluence rate at the “reference point” [3], as a consequence of the neutron production, obtained for (a) conformation in 15 MV is (1.45±0.11) x10⁴ n•cm2/MU. Configurations (b) and (c) gave fluences of 96.6% and 97.8% of the reference (a). Neutron production decreases up to 8.6% and 5.7% for the 10 MV FF and FFF beams, respectively. Finally, it decreases up to 2.8% and 0.1% for the 6 MV FF and FFF modes, respectively.Conclusion:This work evaluates thermal neutron production of Varian TrueBeam™ system for organ equivalent dose estimation. The small difference in collimator‐MLC configuration shows the universality of the methodology [3]. A decrease in this production is shown when decreasing energy from 15 to 10 MV and an almost negligible production was found for 6 MV. Moreover, a lower neutron contribution is observed for the FFF modes.[1]Phys Med Biol,2012;57:6167–6191.[2]Radiat Meas,2010;45:1513–1517.[3]Med Phys,2015;42:276–281.

Fricke-gel dosimetry in epithermal or thermal neutron beams of a research reactor

Fricke–xylenol–orange gel has shown noticeable potentiality for in-phantom dosimetry in epithermal or thermal neutron fields with very high fluence rate, as those characteristic of nuclear research reactors. Fricke gels in form of layers give the possibility of achieving spatial distribution of gamma dose, fast neutron dose and dose due to charged particles generated by thermal neutron reactions. The thermal neutron fluence has been deduced from the dose coming from the charge particles emitted by neutron reactions with the isotope 10B. Some measurements have been performed for improving the information on the relative sensitivity of Fricke gel dosimeters to the particles produced by 10B reactions, because at present the precision of dose evaluations is limited by the scanty knowledge about the dependence of the dosimeter sensitivity on the radiation LET. For in-air measurements, the dosimeter material can produce an enhancement of thermal neutron fluence. Measurements and Monte Carlo calculations have been developed to investigate the importance of this effect.

Experimental characterization of semiconductor-based thermal neutron detectors

In the framework of NESCOFI@BTF and NEURAPID projects, active thermal neutron detectors were manufactured by depositing appropriate thickness of 6LiF on commercially available windowless p–i–n diodes. Detectors with different radiator thickness, ranging from 5 to 62μm, were manufactured by evaporation-based deposition technique and exposed to known values of thermal neutron fluence in two thermal neutron facilities exhibiting different irradiation geometries. The following properties of the detector response were investigated and presented in this work: thickness dependence, impact of parasitic effects (photons and epithermal neutrons), linearity, isotropy, and radiation damage following exposure to large fluence (in the order of 1012cm−2).

International key comparison of thermal neutron fluence measurements—CCRI(III)-K8

After more than thirty years a new key comparison of thermal neutron fluence measurements was organized by section III of CCRI. The comparison was carried out by rotating four transfer instruments among the four participants (CIAE, PTB, NMIJ and NPL).The stability of the detectors was repeatedly verified by the pilot laboratory between the measurements. Each of the four transfer devices had a different dependence of the fluence response on the neutron energy. Hence the comparison was also sensitive to the knowledge of the spectral distributions of the facilities used by the participants for their measurements. The results of the comparison showed signs of inconsistencies which could not be resolved during the analysis. Therefore the arithmetic mean of the results was used to calculate the key comparison reference value.Main text.To reach the main text of this paper, click on Final Report. Note that this text is that which appears in Appendix B of the BIPM key comparison database kcdb.bipm.org/.The final report has been peer-reviewed and approved for publication by the CCRI, according to the provisions of the CIPM Mutual Recognition Arrangement (CIPM MRA).

Photo neutron dose equivalent rate in 15 MV X-ray beam from a Siemens Primus Linac.

Fast and thermal neutron fluence rates from a 15 MV X-ray beams of a Siemens Primus Linac were measured using bare and moderated BF3 proportional counter inside the treatment room at different locations. Fluence rate values were converted to dose equivalent rate (DER) utilizing conversion factors of American Association of Physicist in Medicine's (AAPM) report number 19. For thermal neutrons, maximum and minimum DERs were 3.46 × 10-6 (3 m from isocenter in +Y direction, 0 × 0 field size) and 8.36 × 10-8 Sv/min (in maze, 40 × 40 field size), respectively. For fast neutrons, maximum DERs using 9” and 3” moderators were 1.6 × 10-5 and 1.74 × 10-5 Sv/min (2 m from isocenter in +Y direction, 0 × 0 field size), respectively. By changing the field size, the variation in thermal neutron DER was more than the fast neutron DER and the changes in fast neutron DER were not significant in the bunker except inside the radiation field. This study showed that at all points and distances, by decreasing field size of the beam, thermal and fast neutron DER increases and the number of thermal neutrons is more than fast neutrons.

Applicability of self-activation of an NaI scintillator for measurement of photo-neutrons around a high-energy X-ray radiotherapy machine.

The applicability of the activation of an NaI scintillator for neutron monitoring at a clinical linac was investigated experimentally. Thermal neutron fluence rates are derived by measurement of the I-128 activity generated in an NaI scintillator irradiated by neutrons; β-rays from I-128 are detected efficiently by the NaI scintillator. In order to verify the validity of this method for neutron measurement, we irradiated an NaI scintillator at a research reactor, and the neutron fluence rate was estimated. The method was then applied to neutron measurement at a 10-MV linac (Varian Clinac 21EX), and the neutron fluence rate was estimated at the isocenter and at 30cm from the isocenter. When the scintillator was irradiated directly by high-energy X-rays, the production of I-126 was observed due to photo-nuclear reactions, in addition to the generation of I-128 and Na-24. From the results obtained by these measurements, it was found that the neutron measurement by activation of an NaI scintillator has a great advantage in estimates of a low neutron fluence rate by use of a quick measurement following a short-time irradiation. Also, the future application of this method to quasi real-time monitoring of neutrons during patient treatments at a radiotherapy facility is discussed, as well as the method of evaluation of the neutron dose.

A new online detector for estimation of peripheral neutron equivalent dose in organ.

Peripheral dose in radiotherapy treatments represents a potential source of secondary neoplasic processes. As in the last few years, there has been a fast-growing concern on neutron collateral effects, this work focuses on this component. A previous established methodology to estimate peripheral neutron equivalent doses relied on passive (TLD, CR39) neutron detectors exposed in-phantom, in parallel to an active [static random access memory (SRAMnd)] thermal neutron detector exposed ex-phantom. A newly miniaturized, quick, and reliable active thermal neutron detector (TNRD, Thermal Neutron Rate Detector) was validated for both procedures. This first miniaturized active system eliminates the long postprocessing, required for passive detectors, giving thermal neutron fluences in real time. To validate TNRD for the established methodology, intrinsic characteristics, characterization of 4 facilities [to correlate monitor value (MU) with risk], and a cohort of 200 real patients (for second cancer risk estimates) were evaluated and compared with the well-established SRAMnd device. Finally, TNRD was compared to TLD pairs for 3 generic radiotherapy treatments through 16 strategic points inside an anthropomorphic phantom. The performed tests indicate similar linear dependence with dose for both detectors, TNRD and SRAMnd, while a slightly better reproducibility has been obtained for TNRD (1.7% vs 2.2%). Risk estimates when delivering 1000 MU are in good agreement between both detectors (mean deviation of TNRD measurements with respect to the ones of SRAMnd is 0.07 cases per 1000, with differences always smaller than 0.08 cases per 1000). As far as the in-phantom measurements are concerned, a mean deviation smaller than 1.7% was obtained. The results obtained indicate that direct evaluation of equivalent dose estimation in organs, both in phantom and patients, is perfectly feasible with this new detector. This will open the door to an easy implementation of specific peripheral neutron dose models for any type of treatment and facility.

Thermal conductivity changes upon neutron transmutation of 10B doped diamond

10B doped p-type diamond samples were subjected to neutron transmutation reaction using thermal neutron flux of 0.9 × 1013 cm−2 s−1 and fast neutron flux of 0.09 × 1013 cm−2 s−1. Another sample of epilayer grown on type IIa (110) single crystal diamond substrate was subjected to equal thermal and fast neutron flux of 1014 cm−2 s−1. The defects in the diamond samples were previously characterized by different methods. In the present work, thermal conductivity of these diamond samples was determined at room temperature by transient thermoreflectance method. The thermal conductivity change in the samples as a function of neutron fluence is explained by the phonon scattering from the point defects and disordered regions. The thermal conductivity of the diamond samples decreased more rapidly initially and less rapidly for larger neutron fluence. In addition, the thermal conductivity in type IIb diamond decreased less rapidly with thermal neutron fluence compared to the decrease in type IIa diamond subjected to fast neutron fluence. It is concluded that the rate of production of defects during transmutation reaction is slower when thermal neutrons are used. The thermal conductivity of epilayer of diamond subjected to high thermal and fast neutron fluence is associated with the covalent carbon network in the composite structure consisting of disordered carbon and sp2 bonded nanocrystalline regions.