Double differential neutron yield (DDNY) for projectiles at116 MeV bombarded on thick target is reported. Thick target neutron yield data in energy range MeV/nucleon is very sparsely available in literature. Time of flight technique was used to measure the kinetic energy distribution of the emerging neutrons. Angular distribution of emission was also studied using five small organic liquid scintillator detectors EJ 301, simultaneously, at five angles at a distance of 150 cm from the target and apart covering forward () to backward () angles. The experimental measurements were compared with statistical reaction code PACE2 and FLUKA outputs. The PACE2 and FLUKA calculations reproduce the measured thick target neutron spectra fairly well implying that compound nuclear emissions dominate the reaction process at these energies. The detailed energy distribution of emitted neutrons can be used for source term estimations in shielding calculation for the facility and it will also be useful in dose estimation due to secondary neutrons in the heavy or light ion therapy.

An experiment was conducted in Agricultural and Horticultural Research Station (AHRS), Bavikere, Tarikere taluk during Kharif 2016 to evaluate the K uptake Pattern and availability of nutrients as affected by level and time of K application under drill sown and transplanted condition. The factors comprising of two methods of establishment (drill and transplanting), three levels of potassium application (25, 37.5 and 50 kg ha-1) and two different time of application of potassium (basal and split). The results revealed that basal application of potassium (50 kg ha-1) under transplanted condition registered higher K uptake at different growth stages and also lower available nutrients after the harvest of crop due to higher grain and straw yields.

Double differential neutron yield (DDNY) for 12 C 6 + projectiles at116 MeV bombarded on thick 12 C target is reported. Thick target neutron yield data in energy range ∼ 10 MeV/nucleon is very sparsely available in lit-erature. Time of flight technique was used to measure the kinetic energy distribution of the emerging neutrons. Angular distribution of emission was also studied using five small organic liquid scintillator detectors EJ 301, simultaneously, at five angles at a distance of 150 cm from the target and 30 ◦ apart covering forward (0 ◦ ) to backward (120 ◦ ) angles. The experimental measurements were compared with statistical reaction code PACE2 and FLUKA outputs. The PACE2 andFLUKAcalculationsreproducethemeasuredthicktarget neutron spectra fairly well implying that compound nuclear emissions dominate the reaction process at these energies. The detailed energy distribution of emitted neutrons can be used for source term estimations in shielding calculation for the facility and it will also be

The radiation shielding analysis of the external concrete enclosure of a feature long beamline of the Advanced Photon Source Upgrade project is described in this work. The beamline has 2 undulators that can operate in two different configurations and the source terms are estimated for the worst configuration using the SRW code in the OASYS package as well as the STAC8 code. The shielding analysis is carried out using the STAC8 code, and for the case of pink beam, with the FLUKA code also. Effective dose rates on contact with the outside surface of the concrete walls for pink, reflected full spectrum and monochromatic beams are calculated. For the monochromatic beams, 2 sets of discrete bandwidths (BW) from the Double Crystal Monochromator (DCM) and the full reflectance of the Double Multilayer Monochromator (DMM) are used. Further, due to the size of the direct beam and the energy domain of interest in the direct and scattered fields, the importance of the equivalent dose to the lens of the eye is discussed with respect to the shielding calculations of such facilities. The dose rates outside the concrete hutch in the lateral directions calculated using STAC8 is about 50% higher than the FLUKA results for the case of pink beam but may be considered as a reasonable agreement due to the differences in the calculational methodologies. Those dose rates in the lateral and forward directions are higher for the pink beam to allow full occupancy from the radiation protection point of view. For the full spectrum after 3 reflections, the effective dose rates are less than 0.5 μSv/h in the lateral direction but is higher in the forward region below about 22° with respect to the beam direction. For the DCM beams, except for the scenario with one mirror and 0.1%BW, all dose rates are less than 0.5 μSv/h outside the concrete hutch in the lateral directions. Here, the higher harmonics are what contributes to the dose outside the shield and the bandwidths assumed to derive the photon flux thus becomes important. The dose rates with the DMM are generally lower compared to the 0.1% BW DCM beam but are higher than the results obtained with XOP calculated bandwidths.

The CSSI beamline of the APSU will have a 20.5 m long, 2.75 m diameter SS end station designed to accept pink beam. The radiation shielding analysis of this end station is analyzed here using STAC8 and FLUKA codes. Effect of mirror properties such as reflectivity, coating, inclination and roughness as well as variation in the steel composition are also studied. 8 mm thick SS is found to be adequate if the source is defined by a pinhole and the white beam is reflected with two mirrors inclined at 3.0 mradian or more. The dose rates from monochromatic beams are found to be below the desired levels when the XOP calculated bandwidth (BW) are used while the use of a flat 0.1% BW requires the mirror inclinations to be above 2.5 mradian or more. STAC8 results are consistently higher than the FLUKA results by about 1.5-2.0 times.

Green fodder is the natural diet for livestock for a sustainable dairy farming and dairy cattle requires green fodder for high milk yield. However, it cannot available throughout the year especially in dry land situation, it is difficult to have access for green fodder due to many reasons; non-availability of irrigated lands for fodder production, higher labour cost and small land holdings. Thus, hydroponic maize fodder (HMF) is an effective solution for fodder scarcity and is very promising for sustainable livestock production in different regions of the world and nowadays it plays significant role in augmenting fodder shortage of small holding dairy farmers in scarce rainfall areas. Hydroponics fodder production involves growing of plants without soil for a short duration (7 day). In hydroponic fodder production system, it can be possible to grow 13-14 kg of green fodder from one kg of maize seeds. In this study results revealed that feeding of HMF to lactating cows with partially reduced concentrate mixture increased the digestibility of nutrients of crude protein (13.62 vs 10.24 %), ether extract (3.24 VS 2.13 %), nitrogen free extract (64.23 vs 50.23 %) and lower crude fiber (13.86 vs 23.25 %), total ash (3.12 vs 9.02 %) and acid insoluble ash (0.23 vs 1.12 %) and significantly higher mean body weight (431.77), before test milk yield (11.76 kg/day), during test milk yield (14.56 kg/day), 0 % Fat corrected milk (9.16 kg/day), fat (4.46 %) solids not fat (9.11 %) was recorded with cows fed with HMF 14 kg/day + 10 kg dry sorghum straw /groundnut haulm + concentrate feed (4.0 kg) as compared to conventional green fodder (CGF) respectively and further, there was increase in milk yield per day to the tune of 16.5 % more in fed with HMF (Rs. 435) than CGF (Rs. 363) leading to increase in net profit (Rs. 241/day/animal).

The neutron ambient dose equivalent measured at 0 degrees, 30 degrees, 90 degrees and 120 degrees with respect to the incident beam for 116 MeV C-12 projectiles incident on a thick carbon target are reported here. The measurements were carried out with a conventional moderated BF3 proportional counter based rem-meter and a spectrometry based neutron probe (N-probe) dose meter that measures the fluence and folds it with the appropriate fluence to dose conversion coefficients to give the ambient dose equivalent. The ambient dose equivalents are also estimated using the neutron spectra calculated by the PACE2 statistical nuclear reaction model code and folding it with the fluence to ambient dose equivalent conversion coefficients. The dose results are also obtained using an empirical technique reported in the literature. The doses measured by the rem-meter and N-probe, and that estimated from the PACE2 results agreed within 10% of each other while the results from the empirical relation under-predicted the dose rates in the forward direction and over predicted in the backward direction. The angular distribution of the dose from the empirical relation also did not agree with the results from the other techniques. The measured and the PACE2 calculated neutron spectra are not similar in the fluence, shape, peak energy or the slope. The integral doses are still comparable because of the compensation by the higher fluence in the PACE2 results and a corresponding higher fluence to ambient dose equivalent conversion coefficients at those energies.

The thick target neutron yield from a O-16(6+) beam on a thick Al-27 target at 120- and 142-MeV (7.5- and 8.8-MeV/nucleon, respectively) incident energies is presented and compared with theoretical calculations. The theoretical estimates for the yield of a high-energy neutron was obtained using a preequilibrium (PEQ) heavy-ion reaction model (HION), whereas that for the low-energy part was obtained using the standard evaporation models PACE4 and EMPIRE 3.2. In the present data a significant PEQ contribution, similar to 2% to 3% of the evaporation contribution, is observed even at projectile energies below 10 MeV/nucleon. The measured energy spectrum of neutrons for E > 20 MeV (the predominant PEQ region) is in good agreement with the PEQ model code HION.

The neutron ambient dose equivalent outside the radiation shield of a proton therapy cyclotron vault is estimated using the unshielded dose equivalent rates and the attenuation lengths obtained from the literature and by simulations carried out with the FLUKA Monte Carlo radiation transport code. The source terms derived from the literature and that obtained from the FLUKA calculations differ by a factor of 2–3, while the attenuation lengths obtained from the literature differ by 20–40%. The instantaneous dose equivalent rates outside the shield differ by a few orders of magnitude, not only in comparison with the Monte Carlo simulation results, but also with the results obtained by line of sight attenuation calculations with the different parameters obtained from the literature. The attenuation of neutrons caused by the presence of bulk iron, such as magnet yokes is expected to reduce the dose equivalent by as much as a couple of orders of magnitude outside the shield walls.

Energy distributions of neutrons emitted from the interaction of 20 MeV protons incident on a thick Be target were estimated at two angles (0 degrees and 90 degrees) with respect to the incident beam using activation foils as threshold detectors. A recently developed unfolding code GAMCD based on Genetic Algorithm and Monte Carlo methods was used to estimate the neutron spectra by unfolding the counts obtained from the activation foils. These results were compared with the unfolding codes, MAXED and GRAVEL. In the GAMCD code, a guess spectrum as a-priori information need not be provided as input unlike in MAXED and GRAVEL. Results obtained from all these codes matched each other reasonably well. Out of 30 reactions studied experimentally only the selected 15 were found sufficient to generate acceptable spectra in the case of neutrons emitted from the p+Be reaction at 20 MeV. A peak in the energy distributions around 3 MeV at both the measured angles is contributed by the three body breakup process while a broad hump between 6 MeV and 10 MeV only in the forward direction suggests contribution from the pre-equilibrium emissions from the Be-9(p, n)B-9 channel, which also contributes to the small peak observed around 15 MeV. The code GAMCD was found to perform satisfactorily for the present data set.

Neutronyieldfrom115MeV 12C projectilesbombardingathick 27Al targethasbeenmeasuredusingthe time of flight technique.NuclearreactionmodelcodePACEandtheFLUKAMonteCarlocodeareusedto calculate theyieldandtheresultsarecomparedwiththeexperimentaldata.Theenergyformaximum neutron emissioninexperimentalmeasurementandreactioncodeoutputhasaslightdisagreementin the extremeforwardemissionanglebutinallotheranglesithasaclosematch.Theslopeofthe distribution ingeneralshowsgoodmatchbetweentheexperimentalandthereactioncoderesultsas well asFLUKAcalculations.Themaximumenergyoftheemittedneutronsisobservedtodecreasewith the increasingemissionangles.

Neutron yield from 115 MeV 12C projectiles bombarding a thick 27Al target has been measured using the time of flight technique. Nuclear reaction model code PACE and the FLUKA Monte Carlo code are used to calculate the yield and the results are compared with the experimental data. The energy for maximum neutron emission in experimental measurement and reaction code output has a slight disagreement in the extreme forward emission angle but in all other angles it has a close match. The slope of the distribution in general shows good match between the experimental and the reaction code results as well as FLUKA calculations. The maximum energy of the emitted neutrons is observed to decrease with the increasing emission angles.

The scarcity and the high cost of 3He has spurred the use of various detectors for neutron monitoring. A new lithium yttrium borate scintillator developed in BARC has been studied for its use in a neutron rem counter. The scintillator is made of natural lithium and boron, and the yield of reaction products that will generate a signal in a real time detector has been studied by FLUKA Monte Carlo radiation transport code. A 2 cm lead introduced to enhance the gamma rejection shows no appreciable change in the shape of the fluence response or in the yield of reaction products. The fluence response when normalized at the average energy of an Am–Be neutron source shows promise of being used as rem counter.

A prototype two element eye lens dosimeter badge based on highly sensitive α-Al 2 O 3 :C optically stimulated luminescence dosimeters (OSLDs) was designed and developed. The badge consists of a plastic card in which two thin α-Al 2 O 3 :C optically stimulated luminescence (OSL) discs are placed. The OSLDs in the plastic card (when inserted into plastic cassette) are covered with energy compensation filters made up of 0.3 mm thick Cu and 1.35 mm thick Teflon discs positioned symmetrically on both sides of the dosimeter. The OSLD badge is useful for monitoring doses from photons and beta particles. In this paper, theoretical studies using Monte Carlo method as well as using the analytical technique have been performed to study the energy response of the bare α-Al 2 O 3 :C based disc dosimeters. These dosimeter discs have been found to exhibit over-response by a factor of ~ 3.4 at ~ 33 keV photon energy, whereas, beyond 80 keV photon energy, the response is nearly energy independent. Studies have also been performed to find the energy response of the α-Al 2 O 3 :C disc dosimeters under different metal filters, viz., Al, Cu, Sn etc., and under various thicknesses of Teflon. From theoretical simulations, it has been found that 0.3 mm thick Cu is sufficient to correct the over-response in lower energy region within acceptable limits. Further, Teflon disc (DuPont, USA) having thickness of 1.35 mm is found to be the optimized choice as filter for the second dosimeter disc. It is worth mentioning that the ratio of the response of the OSL disc under Teflon to that under Cu filter indicates average energy of X-ray photons and same is used to correct the over-response as well as to estimate the quantity H p (3). Also for higher photon energy region, the readout of the dosimeter disc under Teflon filter directly measures the quantity H p (3). Same holds true for beta particles having maximum beta energy, Emax beyond 0.7 MeV.

In mixed oxide (MOX) fuel fabrication facilities, the fabrication of MOX fuel involves various metallurgical operations such as mixing and milling of weighed quantities of uranium and plutonium oxides, precompaction, granulation of precompacts, final compaction of granules, sintering of green pellets, followed by fuel pin fabrication. All these operations are carried out in glove boxes, which have complex geometries due to housing of various equipments. Plutonium, a source of neutrons, is handled in large quantities in various forms such as powder, granules, pellets and pellets encapsulated in pins during these operations. It is obvious that extensive knowledge on the neutron spectral distributions in workplace is required from radiation protection and shielding point of view. In this paper, a brief introduction to the source of neutrons in MOX fuel handling facilities, studies that include experimental measurements of neutron spectra of various forms of MOX, contribution of neutron fluence in various energy groups and its dose equivalent, establishment of a simulation procedure for glove boxes handling Pu using FLUKA Monte Carlo codes, comparison of simulation results with the actual experimental measurements are presented. The results indicate that bare PuO 2 and MOX sources require 3 mm of lead shield to eliminate gamma interference in MICROSPEC with N-probe neutron spectrometer. Furthermore, the studies reveal that dose equivalent contribution from MOX pellets is very significant in the energy group of 1.0-1.5 MeV unlike PuO 2 powder and fuel pins, which exhibit significant contribution in the energy groups of both 1.0-1.5 MeV and 2.0-3.0 MeV. Also, the fuel pins show high neutron fluence in the energy group of 0.0-0.01 MeV, but they do not contribute significantly to dose equivalent. A good agreement between experimentally measured data and FLUKA simulated results has been observed.

A systematic study on the measurement of neutron spectra emitted from the interaction of protons of various energies with a thick beryllium target has been carried out. The measurements were carried out in the forward direction (at 0° with respect to the direction of protons) using CR-39 detectors. The doses were estimated using the in-house image analyzing program autoTRAK_n, which works on the principle of luminosity variation in and around the track boundaries. A total of six different proton energies starting from 4 MeV to 24 MeV with an energy gap of 4 MeV were chosen for the study of the neutron yields and the estimation of doses. Nearly, 92% of the recoil tracks developed after chemical etching were circular in nature, but the size distributions of the recoil tracks were not found to be linearly dependent on the projectile energy. The neutron yield and dose values were found to be increasing linearly with increasing projectile energies. The response of CR-39 detector was also investigated at different beam currents at two different proton energies. A linear increase of neutron yield with beam current was observed.

Monte Carlo simulations have been carried out using the FLUKA code to improve the neutron ambient dose equivalent [H*(10)] response of the ZReC (zirconium-lined portable neutron counter responding satisfactorily to neutrons up to 1 GeV) by introducing various neutron absorbers in the system such as cadmium, gadolinium, natural boron, enriched B and borated polythene. It was found that ZReC can be effectively used as a portable neutron monitor by introducing any one of the following perforated layers: 5 mm thick natural boron, 0.5 mm thick enriched B or 1 cm high-density polythene mixed with 50 % boron by weight. The integral response of the instrument was also calculated for some typical reference neutron fields. The relative ambient dose equivalent response of the said system is also found comparable with that of the existing LINUS neutron monitor.

During extravehicular activities (EVA) astronauts are exposed to intense radiation from solar flare events (SFE).Hence, the radiation shielding efficacy of space suits, in particular the helmets and extra vehicular visor assembly (EVVA) are imperative to safety of astronauts and ultimate success of the space mission.At West German Proton Therapy Centre Essen (WPE) we have replicated the historical solar flare spectra by manipulating the treatment planning system (TPS) of the 230 MeV proton-therapy cyclotron and developed novel proton irradiation and dosimetry methods for the testing of space borne devices as mentioned above.We have used a single foil of radiochromic film (type EBT 2) to estimate the proton depth dose distribution in the human eye.The optimal thickness of EVVA (acryl glass) for the protection from solar flares was estimated to be 2.35 cm.A maximum SFE proton dose up to 20 Gy found to be well assessed using the EBT 2 films.

Monte Carlo simulations have been carried out using the FLUKA code to improve the neutron ambient dose equivalent [H*(10)] response of the ZReC (zirconium-lined portable neutron counter responding satisfactorily to neutrons up to 1 GeV) by introducing various neutron absorbers in the system such as cadmium, gadolinium, natural boron, enriched (10)B and borated polythene. It was found that ZReC can be effectively used as a portable neutron monitor by introducing any one of the following perforated layers: 5 mm thick natural boron, 0.5 mm thick enriched (10)B or 1 cm high-density polythene mixed with 50 % boron by weight. The integral response of the instrument was also calculated for some typical reference neutron fields. The relative ambient dose equivalent response of the said system is also found comparable with that of the existing LINUS neutron monitor.