Dose Rate Calculations Research Articles

ANALYSIS OF DOSE RATES AROUND THE SLOVENIAN SILO-TYPE LILW REPOSITORY USING ADVANTG

The ADVANTG code was used to analyze dose rates from the proposed Slovenian silo-type low and intermediate level waste (LILW) repository. Detailed calculations of dose rates are challenging as gamma-sources are located in thick concrete containers and the effect of backscattering in the air has to be considered. Compared to analog MCNP simulations maximum relative speed-up of up to 9500 was achieved by using ADVANTG. We have analyzed the influence of different repository configurations on the final annual γ-dose rates. For different measurement positions, which were located from 10 m to 150 m from the outer edge of the repository, γ-dose rates were calculated. The results show that for all configurations of the open repository where the waste from different sources is deposited, except using the waste from Krško NPP, the calculated annual γ-dose rates for all measurement positions were lower than the prescribed limit values for professionals of 20 mSv/year. To ensure that the γ-dose rates are lower than the prescribed limit values even for the case with the waste from Krško NPP an additional analysis was made where the concrete casks, in which the waste is stored, were modelled in more detail and the quantity of additional grout on top of the waste was varied.

APPLICATION OF THE JSIR2S CODE PACKAGE FOR SHUTDOWN DOSE RATE CALCULATIONS ON JET

In this paper we present a computational exercise for shut-down dose rate calculations for the JET tokamak using the in-house developed JSIR2S code package as part of its validation. The computation is performed in two parts: neutron transport and transport of secondary gamma radiation. In order to calculate neutron activation reaction rates with sufficiently low variance, hybrid variance reduction techniques using the ADVANTG code have been utilized. Probability based sampling of secondary source particles was performed. Calculated gamma dose rates after shut down are compared with dose rate measurements performed on site using ionization chambers. The C/E agreement for 1st octant is between 0.8 to 1 while statistically meaningfull results for the 2nd octant are yet to be obtained.

Dose rate calculations at beam tube no. 5 of the JSI TRIGA mark II research reactor using Monte Carlo method

Monte Carlo N-Particle (MCNP) transport code accelerated by AutomateD VAriaNce reducTion Generator (ADVANTG) code was used to simulate neutron and prompt gamma particles emitted from TRIGA research reactor during operation. Firstly, the method was validated by measuring dose rates around open beam port number 5 was unplugged. Neutron and gamma dose rates inside the reactor hall in the vicinity of the beam port were calculated and compared to the measurements. Due to the satisfactory agreement, the method was later used to design external shielding for the same beam port when it was upgraded – special mechanism was installed that allows irradiation of larger samples. Computational analysis of the proposed shielding configuration provided acceptable dose rate levels inside the reactor hall. When the shield was constructed, calculated dose rates were confirmed by the actual measurements. No modifications were needed.

Estimation of Radioactivity Release Activity Using Non-Linear Kalman Filter-Based Estimation Techniques

The estimation of radioactivity release following an accident in a nuclear power plant is crucial due to its short and long-term impacts on the surrounding population and the environment. In the case of any accidental release, the activity needs to be estimated quickly and reliably to effectively plan a rapid emergency response and design an appropriate evacuation strategy. The accurate prediction of incurred dose rate during normal or accident scenario is another important aspect. In this article, three different non-linear estimation techniques, extended Kalman filter, unscented Kalman filter, and cubature Kalman filter are proposed in order to estimate release activity and to improve the prediction of dose rates. Radionuclide release rate, average wind speed, and height of release are estimated using the dose rate monitors data collected in proximity of the release point. Further, the estimates are employed to improve the prediction of dose rates. The atmospheric dispersion phenomenon of radioactivity release is modelled using the Gaussian plume model. The Gaussian plume model is then employed for the calculation of dose rates. A variety of atmospheric and accident related scenarios for single source and multiple sources are studied in order to assess the efficacy of the proposed filters. Statistical measures have been used in order to validate the performance of the proposed approaches.

Dose-rate considerations for the INTRABEAM electronic brachytherapy system: Report from the American association of physicists in medicine task group no. 292.

The purpose of this report is to provide detailed guidance on the dosimetry of the INTRABEAM® (Carl Zeiss Medical AG, Jena, Germany) electronic brachytherapy (eBT) system as it stands at the present time. This report has been developed by the members of American Association of Physicists in Medicine (AAPM) Task Group 292 and endorsed by the AAPM. Members of AAPM Task Group 292 on Electronic-Brachytherapy Dosimetry have reviewed pertinent publications and user manuals regarding the INTRABEAM system dosimetry and manufacturer-supplied dose calculation protocols. Formal written correspondence with Zeiss has also provided further clarification. Dose-rate calculations for the INTRABEAM system are highly dependent on choice of dosimetry protocol. Even with careful protocol selection, large uncertainties remain due to the incomplete characterization of the ionization chambers used for verification with respect to their energy dependence as well as manufacturing variations. There are two distinct sets of dose-rate data provided by Zeiss for the INTRABEAM system. One dataset (Calibration V4.0) is representative of the physical dose surrounding the source and the other dataset (TARGIT) has been adjusted to be consistent with a clinical trial named TARGIT (TARGeted Intraoperative RadioTherapy). The adjusted TARGIT doses are quite dissimilar to the physical doses, with differences ranging from 14% to 30% at the surface of a spherical applicator, depending on its diameter, and up to a factor of two at closer distances with the smaller needle applicators. In addition, ion chamber selection and associated manufacturing tolerances contribute to significant additional uncertainties. With these substantial differences in dose rates and their associated uncertainties, it is important for users to be aware of how each value is calculated and whether it is appropriate to be used for the intended treatment. If users intend to deliver doses that are the same as they were in 1998 at the onset of the TARGIT trial, then the TARGIT dose-rate tables should be used. The Calibration V4.0 dose rates may be more appropriate to use for applications other than TARGIT trial treatments, since they more closely represent the physical doses being delivered. Users should also be aware of the substantial uncertainties associated with the provided dose rates, which are due to beam hardening, chamber geometry, and selection of the point-of-measurement for a given ionization chamber. This report serves to describe the details and implications of the manufacturer-recommended dosimetry formalism for users of the INTRABEAM system.

Shutdown dose rate benchmarking using modern particle transport codes

Shutdown dose rate calculations provide an essential input to the design and research of fusion power plant technology. They allow the estimation of dose to personnel and equipment during planned and unplanned maintenance. The mesh coupled rigorous 2 step (MCR2S) methodology used at Culham Centre for Fusion Energy (CCFE) was originally developed to link the MCNP particle transport code and the FISPACT-II inventory code. As new particle transport codes are developed there is a strong motivation to move towards a code agnostic approach. This paper details the integration of MCR2S with the FISPACT-II API and two other transport codes, Serpent 2 and OpenMC. Two benchmarks, the FNG shutdown dose rate experimental benchmark and the ITER computational benchmark, have been performed and compared to results produced with MCNP. In general, the results show that MCNP, Serpent 2 and OpenMC give shutdown dose rate results similar to the FNG experiment for both experimental campaigns. However, all codes appeared to slightly overestimate the dose rates for Campaign 1 (all results had a C/E between 1 and 1.5) and underestimate the dose rates for Campaign 2 (all results had a C/E between 0.6 and 1). Differences were seen between OpenMC and MCNP for the ITER port plug benchmark, where the lack of variance reduction in OpenMC meant that the neutron flux estimates at the rear of the model were not converged. This led to differences of up to 13% in the shutdown dose rates. It was shown that Serpent 2 and MCNP, where variance reduction was used, gave shutdown dose rates within 3% of each other. Although some areas for development of the Serpent 2 and OpenMC transport codes have been highlighted, overall the comparisons give confidence that the implementation of these two transport codes into the MCR2S work-flow has been carried out successfully.

Engineering Calculations and Preparation for Manufacturing of ITER Equatorial Port #11

The current integration state of ITER Equatorial Port #11 (EP#11) is described. This port is the first plasma port and thus the integration phase is close to the final stage. The new modular design of neutron shielding is implemented in the Diagnostic shielding modules’ (DSM) construction to achieve the maximum shielding efficiency within the weight restrictions and keeping fully operable the tenant diagnostics systems. The same approach was applied to Interspace structure frame design to allow the safe hand-on maintenance near the highly activated components. The current iteration of the engineering calculations, including the neutron transport and shutdown dose rate (SDDR), thermohydraulics, electromagnetics, seismic, and structure integrity calculations, showed that the current design of EP#11 satisfy all ITER requirements. The special metalworking, welding, and various other technologies are tested during the process of preparation for the manufacturing of ITER EP#11. Production technologies of the full-size DSM mock-up and various elements of its infrastructure were tested. The multizoned clean area assembly hall for the final assembly is close to getting completed and the testing of vacuum items has to be sent to ITER.

Radiological risk assessment in the terrestrial ecosystem: comparative study of two software tools used for dose rate calculations.

In this study, two software tools, namely the ERICA Assessment Tool and the RESRAD-BIOTA code, are used for the calculation of the radiological exposure of non-human organisms. For the purposes of the analysis, data retrieved from field studies are used. The site-specific measurements were performed on organisms (mammals-sheep and goats of Bovidae spp.) collected from free-ranged grazing regions in Greece. Plants (grass) of Poaceae spp. and soil samples were also collected from these regions. Natural radionuclides (226Ra, 228Ra, and 228Th) of lithogenic origin and 137Cs, resulted from human activities (Chernobyl and Fukushima nuclear power plant accidents and global fallout), were detected in all samples. The measured activity concentrations were used as input to the two software tools, the ERICA Assessment Tool and the RESRAD-BIOTA code. The results of the simulations provided the external, internal, and total dose rates received by the organisms due to the exposure to the radionuclides. The assessments indicated that out of all detected radionuclides, 228Th is the main contributor to the external dose and 226Ra and 228Ra are the main contributors to the internal dose. The comparative analysis of the two tools revealed significant differences in the calculated doses. In fact, external and internal doses calculated by RESRAD-BIOTA are higher than the values calculated by the ERICA Tool, due to the dose conversion coefficients (DCCs) used for the dose calculation. RESRAD-BIOTA provides more conservative values, but ERICA Tool provides lower uncertainty due to the higher flexibility in the design of the phantom organism. On a risk assessment basis, there is no significant impact, due to organisms' exposure to radioactivity. However, further consideration of the exposure levels is required due to the potential effects of protracted low-level ionizing radiation on the various levels of life's organization.

Progress in development of advanced D1S dynamic code for three-dimensional shutdown dose rate calculations

Abstract The Advanced Direct 1-Step (AdD1S) is one of the most validated tools for the evaluation of Shutdown Dose Rate in complex fusion tokamak machines. The present work is built on the experience in analyzing tokamaks using prior versions of the code, where the activation analysis was limited to the first-step reactions. In the light of the neutron irradiation scenarios foreseen for fusion power reactors and the increasing importance of safety requirements, the possibility to treat the multi-step reactions introduced in the present version represents a novel and relevant development in the field. Moreover, a new Python library, PyD1S, has been developed to optimize the MCNP-FISPACT interface and data analysis. In this paper, the recent developments and the first applications are presented.

Shut-down dose rate analysis for the activated target assembly in IFMIF-DONES during maintenance

• The activation and shutdown dose analyses of the IFMIF-DONES. • Transport and activation calculations using McDeLicious code. • The dose rate maps in the access cell as well as the shielding assessments is presented. • The analysis of the shielding for activated target of DONES. This paper presents the activation and shutdown dose analyses of the IFMIF-DONES (International Fusion Materials Irradiation Facility – DEMO-Oriented NEutron Source) Target assembly. To this end, a series of coupled transport and activation calculations are performed using McDeLicious code, which is based on the MCNP Monte Carlo code for simulating the neutron production from d-Li reactions, and the FISPACT inventory code. The latest DONES test cell model has been employed and FENDL neutron cross-section data library has been used. Shut-down dose rate calculations were performed using the R2Smesh code system developed at KIT. The obtained dose rate maps in the access cell as well as the shielding assessments will be presented in the paper.

Nuclide inventory of FIR 1 TRIGA research reactor fuel

FiR1, a TRIGA Mark II type research reactor, was operated in Finland for years 1962–2015 and will be decommissioned during the following few years. Data on the fuel nuclide inventories is needed to estimate fuel rod activities and dose rates for safe fuel removal, transportation and final disposal. This paper presents comparison calculations for FiR1 reactor fuel nuclide activity inventory performed with VTT’s Monte Carlo based continuous energy reactor physics code Serpent2 with internal burnup and depletion calculation routine, and with the point-depletion code ORIGEN-S using MCNP5 pre-calculated three-group neutron fluxes as spectrum weight factors. Results are presented at the time of final reactor shutdown and after a cooling period of nine months, which corresponds to the systematic dose rate measurements for all the fuel rods performed in 2016.Results of the Serpent2 calculations include radionuclide inventory data for each individual fuel rod. The reactor operating history, except the last four and half years, was modelled in yearly burnup steps so that the yearly energy production became correct. The final four and half years of operation from the beginning of 2011 were modelled in more detail. All changes in the fuel loading were taken into account according to reactor bookkeeping.ORIGEN-S point-depletion calculations were not conducted for each individual fuel rod. The ORIGEN-S calculations were based on MCNP5 calculated averaged neutron spectrum and fission thermal power for the three different fuel rod types used in the reactor. However, with ORIGEN-S, the operating history was modelled in high detail, as ORIGEN-S calculations are fast and computing time was not a limiting factor.The difference between ORIGEN-S and Serpent2 in the sums of all the studied nuclides for the whole core 12 h after shutdown was 16%. Differences between activities for the nuclides contributing 99% to the activity of all examined nuclides were roughly around 30%. For Gd-153, Am-244 and Sn-126 the differences were several hundred percents. The main cause for differences in the activity inventory may be due to differences in neutron flux which arises from the use of one fuel loading in the calculation of the flux for the ORIGEN-S model. Serpent2 applied all 46 historical loadings during which e.g. some of the fuel rods were removed from the reactor and some (not necessarily the same number) were put back. Significant differences arise most likely also from nuclear data and the different way the branching ratios to isomeric states are determined in the two codes. For nuclides with half lives less than 250 days, the somewhat crude model for the operating history may also cause considerable uncertainty.Rod-wise burnup data was used to perform Serpent dose rate calculations for all individual rods nine months after final reactor shutdown. The difference between the calculated and measured dose rates was on average 21 percent. These were also compared to rod-type wise Serpent dose rates that applied the fuel inventories calculated with ORIGEN-S. The relative differences between measured dose rates and results from both calculations varied from −20 to 57 percent.

Shutdown dose calculations for the IFMIF-DONES lithium loop cell using variance reduction techniques

Abstract Shutdown dose rate (SDR) calculations for the lithium loop cell (LLC) of the IFMIF-DONES (International Fusion Material Irradiation Facility- DEMO Oriented NEutron Source) have been performed in this work. A new version of Rigorous 2-Step code R2Smesh-3.0 was utilized, which shows improvements on efficiency, robustness, and reliability in SDR simulations. The quality of the dose map in the LLC has been improved using a weight-window mesh, combining with an intensity-based source sampling technique. The dose rate in most of the area of the LLC is below 0.5 μSv/h at one day after shutdown. Strong gamma streaming through the empty lithium pipe is a possible concern for the lithium loop maintenance.

Calculation of radium-223 and actinium-225 α-emitter radiopharmaceuticals dose rates in treatment of metastatic castration-resistant prostate cancer.

There is limited information regarding the α-emitter radiopharmaceuticals dose calculation used in the setting of men with prostate cancer (PCa). The present study investigates the α-emitter radiopharmaceuticals absorbed dose distribution in the body organs. The α-emitter radiopharmaceuticals dose coefficient and absorbed doses biokinetics distribution, which are used for the treatment of PCa in all over the world, were performed using the "Internal Dose Assessed by Computer" (IDAC-Dose 2.1) program. The results of absorbed dose distribution in any organ of the body, were compared in studied α-emitter radiopharmaceuticals. The absorbed dose value of 223Ra radiopharmaceutical in the prostate organ was evaluated 9.47E-9 Gy/Bq. The maximum and minimum absorbed doses due to biokinetics distribution of 223Ra were found in the thymus (9.53E-8 Gy/Bq) and eye lenses (1.30E-10 Gy/Bq) organs, respectively. Furthermore, the 225Ac absorbed dose in the prostate organ was obtained 1.91E-9 Gy/Bq, where this value is 1% of total body dose. While the absorbed dose distribution of 225Ac in body organs shows the highest concentration in the spleen (1.47E-8 Gy/Bq) and lowest in the eye lenses (7.93E-12 Gy/Bq). The absorbed dose in the body organs due to 223Ra and 225Ac α-emitter radiopharmaceuticals which are used in metastasized castration-resistant prostate cancer (mCRPC), calculated in this study. The results of this study will assist in evaluating and analyzing human body organ doses from application of 223Ra and 225Ac that used in mCRPC patients.

Comparison of calculated and measured basic dosimetric parameters for total body irradiation with 6- and 18-MV photon beams

AbstractPurpose:Total body irradiation (TBI) is an external beam radiation therapy in which large field size and extended source skin distances (SSDs) are applied to deliver a therapeutic dose to the whole body. As measurements in such situations are not common and have more uncertainties in comparison to standard dosimetry situations, it is more precise if calculated beam data can be used instead of measurements taken under TBI situations. The purpose of this study is to compare calculated beam data [percentage depth dose (PDD) and dose rate] with those obtained from simulated treatment measurements.Materials and methods:PDD and dose rates were measured for the 6- and 18-MV photon beams under TBI and standard conditions using 9,000 cm3 water phantom and ion chambers (Markus and Farmer). The results were then compared with the calculated PDD and dose rate data. The beam flatness was also measured under TBI and standard conditions for both 6- and 18-MV photon beams, and the results were then compared.Results:A comparison of the measurement and calculated beam data shows that the difference between calculated and measured PDD values is −6·97 and −4·14% for the 6- and 18-MV photon beams, respectively. The ratio of calculated to measured dose rate was 1·09 and 1·02 for the 6- and 18-MV photon beams, respectively. The beam flatness under TBI conditions was 4·59% for 6-MV and 5·37% for 18-MV photon beam, whereas under standard conditions, these values were 1·50 and 1·98% for 6- and 18-MV radiation beams, respectively.Findings:According to the results, due to a high error level in dose rate and PDD calculations, these parameters must be directly measured under TBI conditions; however, regarding the obtained results, direct measurement is not necessary for the 18-MV photon beam.

Calculation of the effective external dose rate to a person staying in the resettlement zone of the Vetka district of the Gomel region of Belarus based on in situ and ex situ assessments in 2016–2018

The aim of this study was to perform a preliminary assessment of the expected effective dose rate from external exposure to an adult individual staying at that part of the radioactively contaminated territory of the Vetka district of the Gomel region of the Republic of Belarus, from where residents had been resettled after the Chernobyl accident. For this assessment, in summer 2016 and 2018 soil samples were taken from 19 sites located in forests (7 plots), virgin meadows (4 plots), cultivated meadows (6 plots) and vegetable gardens (2 plots), with the subsequent estimation of the inventory and vertical distribution of 137Cs in the soil. The values of 137Cs inventory in the soil ranged from 452 to 1620 kBq m−2 (mean = 904 kBq m−2, median = 964 kBq m−2). The results of the measurement of soil samples were used to calculate values of the air kerma rate, normalized to the inventory of radioactive caesium in the soil. On average, the normalized indicator of the air kerma rate from the man-made source was higher in forests (1.13 nGy h−1 per kBq m−2) compared to virgin meadows (0.95 nGy h−1 per kBq m−2). Normalized air kerma rate in cultivated meadows and vegetable gardens was approximately two times lower than the corresponding indicator for virgin meadows. Using a field gamma spectrometer-dosemeter, ambient dose equivalent rate of gamma radiation in the air was measured at the surveyed sites and the contributions of the technogenic and natural components to the dose rate were estimated. Additionally, such measurements were performed on asphalted surfaces (5 sites) and inside two wooden houses. The measured values of the total ambient dose equivalent rate at a height of 1 m above the ground, asphalted surface or house floor varied from 160 to 2260 nSv h−1. The lowest levels were recorded over asphalted surfaces and inside houses, and the highest ones at forest and virgin meadow sites. The contribution of the technogenic component to the total dose rate varied from 61.9% to 98.8% (mean = 88.9%; n = 26). The effective dose of anthropogenic radiation calculated from the results of in situ measurements in a forest, virgin meadow, cultivated meadow, kitchen garden, asphalted area and house was 0.59, 0.80, 0.34, 0.29, 0.06 and 0.06 μSv h−1, respectively. Similar values for land plots were calculated based on ex situ analysis of soil samples. It can be expected that, starting from 2020, the average effective external dose of a person staying in the resettlement zone of the Vetka district will not exceed 1 mSv year−1.

Extension of Monte Carlo code MCS to spent fuel cask shielding analysis

The shielding analysis of the spent fuel dry storage cask TN-32 is carried out using the continuous-energy Monte Carlo neutron- and photon-transport code MCS developed by the Computational Reactor Physics and Experiment laboratory of the Ulsan National Institute of Science and Technology. The shielding analysis involves the transport simulation of the decay neutrons and gammas emitted by the spent fuel to obtain the neutron and gamma dose rate values outside of the cask. This deep-penetration problem is solved by means of weight window population control as variance reduction. The TN-32 cask calculations are repeated with the MCNP6 Monte Carlo code for verification purpose and a good agreement within three standard deviations is observed between MCS and MCNP6. The neutron dose rate on the cask surface and 2 m away from the cask can be calculated with MCS, whereas the photon dose rate calculation for positions outside of the cask will be further improved by employing more sophisticated variance reduction techniques.

Alpha dose rate calculations for UO2 based materials using stopping power models

Abstract Accurate dose rate models for UO2 based materials in contact with water are important in the modeling of the radiolytically promoted dissolution of spent fuel. Dose rates of α-doped UO2 and un-irradiated MOX fuel were modelled using the ASTAR and SRIM stopping power databases. Dose rates were calculated as a function of distance from the active surface. Comparisons with common dose rate calculation models and the combined Bethe-Bloch and Lindhard–Scharff (LS) equation were performed. It was shown that the ASTAR and SRIM databases could more accurately simulate an α-spectrum compared to the Bethe-Bloch-LS equation. A comparison between the continuous slowing down approximation (CSDA) and the radial projection algorithm in the SRIM program was performed, and it was shown that CSDA overestimates the range of the α-particles by a few percent. This leads to an overestimation of the α-dose rate at distances close to the maximum range of the α-particle in water. A relationship between the average dose rate to specific α-activity ratio as a function of α-energy was obtained from the calculations, which can easily be implemented in alpha dose rate calculations of a UO2 based materials.

The remote-controlled spectrometric system for searching and characterization of high-level radioactive waste

In 2003, the work on environmental rehabilitation of the former coastal technical base of the Russian Navy in Gremikha town, with technical and financial support of the State Atomic Energy Corporation “Rosatom” and the French Alternative Energies and Atomic Energy Commission had begun. In 2017, the task of searching, characterization and packaging in transport containers for export to the long-term storage of high-level waste containing 152Eu was set as part of this program. These wastes were presented as fragments of reactor control and safety system rods of first generation nuclear submarine reactors and were placed in temporary storage containers mixed with other radioactive waste and construction garbage on the territory of the Gremikha branch of the North-West Center “SevRAO”. To accomplish the task, the spectrometric system “Gamma-Pioneer” was developed and made at National Research Center “Kurchatov Institute”. The system consist of a collimated spectrometric gamma-ray CdZnTe detector 500 mm3 volume, a dosimeter for continuous measurement of ambient equivalent dose rate at the instrument location and a color video camera with zoom lens. The device is placed in a protective frame and installed as an attachment implements on robot Brokk-90. Communication between the measuring system and the control computer is carried out using USB extender and 100 meters cable. The video camera is connected to the control computer via Ethernet. To search for the most active sources of gamma radiation, an equivalent dose rate calculation program using the results of measured instrumental spectra was developed. The characterization of the extracted fragments was carried out by activity evaluation of the main radionuclides polluting the samples under investigation. The paper describes design and operating principles of the system, as well as methods for the equivalent dose rate calculation and the activity evaluation of the fragments of reactor control and safety system rods based on spectral measurements obtained using the developed radiometric system

Assessment of shutdown dose rates at the ESS target cooling system using SCALE6.2

The production of high-energy neutrons at the European Spallation Source through the spallation process may cause an erosion of the tungsten target. The eroded particles could be released into the target helium cooling system which contains four kind of filters. Among them, the auxiliary filters called “getters” are designed to capture volatile elements and remaining dust. In this work, the ORNL’s SCALE6.2 modelling and simulation suite for nuclear safety analysis is applied to assess shutdown dose rates and determine if added shielding and/or robotic arms are needed for their maintenance. SCALE6.2 is well adapted to treat this problem as it allows for isotope selection regarding source term calculation. Dose rates are determined by an ORIGEN2 source term and a MAVRIC shielding sequence calculation. As SCALE6.2 is non-standard software for ESS, the results are verified against MCNP, which is the baseline tool for neutronics analysis at ESS. Dose rate calculations show that additional shielding and/or robot arm are not needed to remove the getters after 3 months of cooling time, following 5400 h of operation at 5 MW beam power. At a distance of 1 mm from the getter, the dose rate is 0.2 mSv/h in the most conservative estimation.

Determination of backscattered neutrons/gammas from open beam port of a research reactor

Neutron and gamma dose rate calculations were carried out around horizontal beam tube no. 5 at the Jožef Stefan Institute (JSI) TRIGA Mark II research reactor. Results were compared to the experimental measurements in order to verify the computation model. In addition, another set of calculations and measurements was carried out, where an additional shield made out of concrete and paraffin was installed. With that configuration, we were able to study neutron and gamma scattering.