Monte Carlo Code MCNP4B Research Articles

Optimization of Beam Shaping Assembly design for Boron Neutron Capture Therapy based on a transportable proton accelerator

The optimization of a Beam Shaping Assembly design for Boron Neutron Capture Therapy (BNCT), based on a commercial available proton accelerator, has been simulated. The primary goal is to improve the flux of the epithermal neutrons, which is the major drawback in all designed BNCT facilities with portable neutron source, satisfying simultaneously all the other relevant parameters.A BNCT facility has been simulated using the MCNP4B Monte Carlo code. A wide set of material were examined as moderators, while three different materials are used as fast neutron, gamma ray and thermal neutrons filters.Fast neutrons from 9Be(p,n)9B reaction are moderated through the 7LiF and TiF3 moderators and further reduced using 60Ni as fast neutron filter. The undesired gamma rays and thermal neutrons are removed using Bi and Cd filters respectively. The therapeutic efficacy of the proposed facility was calculated through the dosimetric evaluation in a head phantom.According to the results of the calculations the proposed unit satisfies the recommended by IAEA parameters. Only the flux of the epithermal neutrons remains below the recommended threshold, but is significantly improved.

Comparison of different geometric configurations and materials for neutron radiography purposes based on a 241Am/Be neutron source

The present work examines two different geometric configurations and three different lining materials that are suitable for thermal neutron radiography purposes based on a 241Am/Be neutron source. The same source was also used for fast neutron radiography. Appropriate collimators were simulated for each of the radiography modes, comparing the effectiveness of Cadmium, Gadolinium, and Boral as lining materials for thermal neutron radiography and evaluating the efficiency of Iron and Tungsten as interior wall materials of the collimator in the case of fast neutron radiography. The presented facilities have been simulated for a wide range of parameter values to characterize neutron radiography using the MCNP4B Monte Carlo code.

Optimization study for BNCT facility based on a DT neutron generator

Background : A Boron Neutron Capture Therapy (BNCT) facility, based on a DT neutron generator, with the final goal to find out a potenal, alternave, soluon to exisng BNCT treatment facilies which a re based on nuclear reactors is examined. Materials and Method s: With the aim of the MCNP4B Monte Carlo code different beam-shaping assembly (BSA) configuraons were considered. Lead was selected as reflector material while CF 2, D 2O, Fluental, PbF 4, PbF 2, BiF 3, BiF 5, MgF 2, Al 2O3, AlF 3, TiF 3, BeD 2, CaF 2 and 7 LiF were examined as spectrum shi3ers. In order to improve t he quality of the beam �tanium, nickel-60, iron andtanium alloy (Ti 6Al 14 V) were simulated as fast neutrons filters while lead and bismuth were considered as gamma filters. Results: An extensive set of calculaons performed with MCNP4B Monte Carlo code have shown that the combinaon of 7 LiF which accommodates a conic part made of D 2O, then followed by a TiF 3 layer is the opmum moderator design. The use of three different materials for further reducon of fast neutrons, thermal neutrons and gamma rays is necessary. 60

Radiological shielding design for the TOF polarisation analysis spectrometer Pelican at the OPAL reactor

The shielding for the time-of-flight Polarisation Analysis Spectrometer (Pelican) located at the OPAL reactor (ANSTO) was designed using the Monte Carlo code MCNP 4B. The proposed shielding design has produced compact shielding assemblies, such as the Pelican monochromator drum, lead snout interfaced to guide shielding, and shielding around the Optical System, Sample Chamber and Detector Chamber Systems. These arrangements are consistent with safety requirements, floor load limits, and cost constraints. The neutron dose rates around the Pelican instrument were reduced to < 0.5 μSv/h and the gamma dose rates were lowered to a safe working level of ≤ 3 μSv/h. The Optical System, with its components and the Detector Chamber were housed in borated polyethylene to shield from background signal noise.

Monte Carlo estimation of patient effective dose in diagnostics procedures using 131I

Therapeutic or diagnostic radiopharmaceutical capsule containing Na131I stays in stomach for 15 minutes before the absorption starts, long enough to make possible risky exposure. During the oral application it is reasonable to measure effective dose in stomach. Direct measurements of organ doses are not possible so there is a strong recommendation to estimate them by calculation. The main goal is the 131I risk assessment. Monte Carlo code MCNP4b was used to model the transport of gamma and beta particles emitted by radionuclide 131I considered as a point source at the bottom of the stomach. Absorbed energy per unit transformation in stomach and surrounding organs has been calculated. The dose equivalents in these organs have been calculated in aim to determine the effective doses using appropriate tissue weighting factor values. Obtained results had not significant importance for radiation protection but they were important for establishment of new calibration procedures as a part of QA and QC programs in radiopharmaceuticals production and control.

Spectrally average conversion coefficients for air kerma to ambient dose equivalent for clinical linear accelerator

This work aims to calculate the conversion coefficients from air kerma to ambient dose equivalent, H*(10)/ K air for photon beams produced by linear accelerators, such as the Clinac-4, Clinac-6, Clinac-18 and Clinac-2500, after transmission through primary barriers of radiotherapy treatment rooms. Concrete walls of thickness 1.0, 1.5 and 2.0 m were irradiated with 30 cm×30 cm primary beam spectra. The transmitted spectra were calculated to obtain the conversion coefficients for beams found in radiotherapy services. The calculations were done using the MCNP-4B Monte Carlo code. The results indicate the need to use a factor of about 1.20 to obtain the ambient dose equivalent for radiation surveys near primary barriers using instruments calibrated in air kerma.

A New Approach to Make Collapsed Cross Section for Burnup Calculation of Subcritical System

A general-purpose transport and burnup code system for precise analysis of subcritical reactors like a fusion-fission (FF) hybrid reactor was developed and used for analyzing their performance. The FF hybrid reactor is a subcritical system, which has a concept of fusion reactor with a blanket region containing nuclear fuel and has been under discussion by author’s group for years because the present burnup calculation system mainly consists of a general-purpose Monte Carlo code MCNP-4B, a point burnup code ORIGEN2. JENDL-3.3 pointwise cross section library and JENDL Activation Cross Section File 96 were used as base cross section libraries to make group constant for burnup calculation.A new method has been proposed to make group constant for the burnup calculation as accurate as possible directly using output data of the neutron transport calculation by MCNP and evaluated nuclear data libraries. This method is strict and a general procedure to make one group cross sections in Monte Carlo calculations, while it takes very long computation time. Some speed-up techniques were discussed for the present group constant making process so as to decrease calculation time. Adoption of postprocessing to make group constant improved the calculation accuracy because of increasing number of cross sections to be updated in each burnup cycle.The present calculation system is capable of performing neutronics analysis of subcritical reactors more precise than our previous one. However, at the moment, it still takes long computation time to make group constants. Further speed-up techniques are now under investigation so as to apply the present system to neutronics design analysis for various subcritical systems.

Rapid Transmutation of High-Level Nuclear Wastes in a Catalyzed Fusion-Driven System

The aim of this study is to investigate the high-level waste (HLW) transmutation potential of fusion-driven transmuter (FDT) based on catalyzed D–D fusion plasma for various fuel fractions. The Minor actinide (MA) (237Np, 241Am, 243Am and 244Cm) and long-lived fission product (LLFP) (99Tc, 129I and 135Cs) nuclides discharged from high burn-up pressured water reactor-mixed oxide spent fuel are considered as the HLW. The volume fractions of the MA and LLFP are raised from 10 to 20% stepped by 2% and 10 to 80% stepped by 5%, respectively. The transmutation analyses have been performed for an operation period (OP) of up to 6 years by 75% plant factor (η) under a first-wall neutron load (P) of 5 MW/m2 by using two different computer codes, the XSDRNPM/SCALE4.4a neutron transport code and the MCNP4B Monte Carlo code. The numerical results bring out that the considered FDT has a high neutronic performance for an effective and rapid transmutation of MA and LLFP as well as the energy generation along the OP.

Modification of PROMETHEUS Reactor as a Fusion Breeder and Fission Product Transmuter

This study presents the analyses of the fissile breeding and long-lived fission product (LLFP) transmutation potentials of PROMETHEUS reactor. For this purpose, a fissile breeding zone (FBZ) fueled with the ceramic uranium mono-carbide (UC) and a LLFP transmutation zone (TZ) containing the 99TC and 129I and 135Cs isotopes are separately placed into the breeder zone of PROMETHEUS-H design. The neutronic calculations are performed by using two different computer codes, the XSDRNPM/SCALE4.4a neutron transport code and the MCNP4B Monte Carlo code. A range of analyses are examined to determine the effects of the FF, the fraction of 6Li in lithium (Li) and the theoretical density (TD) of Li2O in the tritium breeder zone (TBZ) on the neutronic parameters. It is observed that the numerical results obtained from both codes are consistent with each other. It is carried out that the profiles of fission power density (FPD) are flattened individually for each FF (from 3 to 10%). Only, in the cases of FF ≥ 8%, the system is self sufficient from the point of view of tritium generation. The results bring out that the modified PROMETHEUS fusion reactor has capabilities of effective fissile breeding and LLFP transmutation, as well as the energy generation.

New burnup calculation system for fusion–fission hybrid reactor

Fusion–fission hybrid reactor is a new energy system based on a fusion reactor. Recently, it is thought to be able to apply to nuclear waste incineration. For the neutronics analysis, the author's group developed their own neutronics and burnup calculation system, which consists of a general-purpose Monte Carlo code MCNP-4B and a point burnup code ORIGEN2. Since the fusion–fission hybrid reactor is operated in a sub-critical state, the cross-section library attached in ORIGEN2 cannot be utilized as well known. One-group cross-sections for burnup calculation had to, thus, be evaluated by a tally function of MCNP in our calculation system. However, all the necessary nuclides could not be dealt with appropriately by the above procedure because of several limitations of MCNP. In the present study, a new procedure has thus been developed to collapse cross-sections by directly using tracking information of neutron accumulated in MCNP calculation. From the test calculation, it was found that the new collapsing procedure successfully removed approximation in the previous evaluation method of one-group cross-sections, which was employed in the calculation system the authors developed previously. Also, ongoing feasibility studies for speedup of the system were briefly discussed, because this kind of cross-section evaluation procedure requires a long computation time at present. The present procedure would be a general method to evaluate collapsed cross-section for burnup with a transport calculation result of continuous energy Monte Carlo code.

Evaluation and comparison of absorbed dose for electron beams by LiF and diamond dosimeters

The absorbed dose response of LiF and diamond thermoluminescent dosimeters (TLDs), calibrated in 60Co γ-rays, has been determined using the MCNP4B Monte Carlo code system in mono-energetic megavoltage electron beams from 5 to 20 MeV. Evaluation of the dose responses was done against the dose responses of published works by other investigators. Dose responses of both dosimeters were compared to establish if any relation exists between them. The dosimeters were irradiated in a water phantom with the centre of their top surfaces (0.32×0.32 cm 2), placed at d max perpendicular to the radiation beam on the central axis. For LiF TLD, dose responses ranged from 0.945±0.017 to 0.997±0.011. For the diamond TLD, the dose response ranged from 0.940±0.017 to 1.018±0.011. To correct for dose responses by both dosimeters, energy correction factors were generated from dose response results of both TLDs. For LiF TLD, these correction factors ranged from 1.003 up to 1.058 and for diamond TLD the factors ranged from 0.982 up to 1.064. The results show that diamond TLDs can be used in the place of the well-established LiF TLDs and that Monte Carlo code systems can be used in dose determinations for radiotherapy treatment planning.

Improvement and calibration of a SSNT personal dosemeter and study of importance of albedo factor for dose calculation

The Neutriran albedo neutron dosemeter has been improved and calibrated for neutron personal dosimetry. The Monte Carlo code MCNP4b was used to calculate the thermal neutrons backscattered from the body (albedo factor). Backscattering from the wall, ceiling and floor in calibration room was considered also via simulation by MCNP4C. A semi automated counting system applying a high-resolution scanner was used for counting of tracks. An 241Am source was used to produce similar alpha particles from 10B (n,alpha)7Li reaction for the optimisation of scanner parameters to distinguish and separate the tracks in SSNTD, which lead to a better distinction between etched alpha tracks and, consequently, a higher linear region of dose characteristic.

Radiation shielding for neutron guides

Models of the neutron guide shielding for the out of bunker guides on the thermal and cold neutron beam lines of the OPAL Reactor (ANSTO) were constructed using the Monte Carlo code MCNP 4B. The neutrons that were not reflected inside the guides but were absorbed by the supermirror (SM) layers were noted to be a significant source of gammas. Gammas also arise from neutrons absorbed by the B, Si, Na and K contained in the glass. The proposed shielding design has produced compact shielding assemblies. These arrangements are consistent with safety requirements, floor load limits, and cost constraints. To verify the design a prototype was assembled consisting of 120 mm thick Pb(96%)Sb(4%) walls resting on a concrete block. There was good agreement between experimental measurements and calculated dose rates for bulk shield regions.

Staff dose calculation during a cobalt-60 source stuck

One of the most postulated accidents in 60Co radiotherapy units is the source getting stuck, where one or more of the staff should enter the treatment room to deal with the problem. For such an accident, an emergency plan is important. A three-dimensional model of a 60Co therapy room has been done using the Monte Carlo code MCNP4B. The radiation safety measures taken and the drawings of the device are given together with suggestions for future use of the source for irradiation purposes. Moreover, the calculated results were compared with those of an experimental study dealing with this problem and were found to be in very good agreement.

Design of a single moderator-type neutron spectrometer with enhanced energy resolution in the energy range from a few to 100 keV

The moderator structure for a neutron spectrometer was optimized with the Monte Carlo code MCNP-4B. The spectrometer consists of a cylindrical moderator and a position-sensitive thermal neutron detector and obtains energy spectra from thermal neutron distribution along its cylindrical axis. The structure of the moderator was improved by using a low hydrogen density material on one end and a high hydrogen density on the other, and inserting a neutron absorber that eliminates thermal neutron diffusion. This design improves the energy resolution of the spectrometer, especially for low-energy neutrons from a few to 100 keV. The designed spectrometer can be applied to the measurement of energy spectra over a neutron energy range from a few keV to 20 MeV.

Monte carlo parametric study of stent impact on dose for catheter-based intravascular brachytherapy with 90Sr/90Y.

The radiation treatment of catheter-based beta-emitter sources is being used to prevent restenosis following interventional coronary procedures. We present the results of a Monte Carlo calculation study to assess the dosimetric impact in the vessel tissue due to the presence of the stent. A catheter-based beta-emitter system is modeled using the Monte-Carlo code MCNP4B. Dose distributions are calculated in annular voxels (0.050 x 0.025 mm2 section) along the axis of a 40 mm. 90Sr/90Y source with and without the stent (at a distance of 1.5-3.0 mm from the longitudinal axis of the source). The main results include: (a) a clear difference between the local perturbation just behind the strut and a more general perturbation seen deeper into the vessel tissue; (b) the local perturbations disappears at a depth of 300-400 microm while the more general perturbation affects the tissue in its full thickness including the prescription point; (c) in the local perturbation the maximum impact is determined mainly by the material and the thickness of the strut while the spatial attenuation of this impact is defined mainly by the strut width; (d) in the general perturbation, the most important magnitude is the free-area ratio for the path of the electrons, being the material characteristics and strut thickness of secondary importance; (e) analytical expressions are presented to estimate the magnitude of this perturbation according to the complete characteristics of the expanded strut, i.e., thickness, free-area ratio, and material; and (f) a simple algorithm is presented for estimating the free-area ratio when this information is not available.

Radiation Shielding Analysis for Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors (DUPIC)

As a part of the compatibility analysis of DUPIC fuel in Canada deuterium uranium (CANDU) reactors, the radiation physics calculations have been performed for the CANDU primary shielding system, which was originally designed for natural uranium core. At first, the conventional CANDU primary shield analysis method was validated using the Monte Carlo code MCNP-4B in order to assess the current analysis code system and the cross-section data. The computational benchmark calculation was performed for the CANDU end shield system, which has shown that the conventional method produces results consistent with the reference calculations as far as the total dose rate and total heat deposition rate are concerned. Second, the primary shield system analysis was performed for the DUPIC fuel core based on the power distribution obtained from the time-average core model, and the results have shown that the dose rates and heat deposition rates through the primary shield of the DUPIC fuel core are not much different from those of the natural uranium core because the power levels on the core periphery are similar for both cores. This study has shown that the current primary shield system is adaptable for the DUPIC fuel CANDU core without design modification.

Benchmarking MCNP and WIMS/RFSP against Measurement Data—I: Deuterium Critical Assembly

Benchmark calculations have been performed for the conventional Canadian deuterium uranium (CANDU) core analysis code RFSP and the Monte Carlo code MCNP-4B using experimental data from the deuterium critical assembly. The benchmark calculation was carried out for the effective multiplication factor (keff), void reactivity, local power peaking factor (LPPF), and power distribution of a uniform core with 1.2 wt% UO2 and two-region cores with PuO2-UO2 fuels. The RFSP calculation was performed with two energy groups, using lattice parameters generated by WIMS-AECL with the ENDF/B-V cross-section library. The RFSP calculation has shown that the root-mean-square (rms) errors of the keff and the void reactivity are within 0.6% δk and 0.3% δ(1/k), respectively. The MCNP simulation was performed using a fully heterogeneous core model that explicitly describes the individual fuel rod and channel. The simulation showed an excellent agreement for the keff against the measurement, while the rms error of the void reactivity was 0.4% δ(1/k). The LPPF and core power distribution estimated by both codes matched those of the measurements within 4 and 9%, respectively. Conclusively, the physics analysis by the RFSP code in conjunction with the WIMS-AECL produces credible results for the light water–cooled and heavy water–moderated system. In addition, the MCNP-4B code has proved its potential as a computational benchmarking tool for the heavy water–moderated system.

Neutronics experiments for DEMO blanket at JAERI/FNS

In order to verify the accuracy of the tritium production rate (TPR), neutron irradiation experiments have been performed with a mockup relevant to the fusion DEMO blanket consisting of F82H blocks, Li2TiO3 blocks with a 6Li enrichment of 40% and 95%, and beryllium blocks. Sample pellets of Li2TiO3 were irradiated and the TPR was measured by a liquid scintillation counter. The TPR was also calculated using the Monte Carlo code MCNP-4B with the nuclear data library JENDL-3.2 and ENDF-B/VI. The results agreed with experimental values within the statistical error (10%) of the experiment. Accordingly, it was clarified that the TPR could be evaluated within 10% uncertainty by the calculation code and the nuclear data. In order to estimate the induced activity caused by sequential reactions in cooling water pipes in the DEMO blanket, neutron irradiation experiments have been performed using test specimens simulating the pipes. Sample metals of Fe, W, Ti, Pb, Cu, V and reduced activation ferritic steel F82H were irradiated as typical fusion materials. The effective cross-sections needed to calculate the formation of the radioactive nuclei (56Co, 184Re, 48V, 206Bi, 65Zn and 51Cr) due to sequential reactions were measured. From the experimental results, it was found that the effective cross-sections increased remarkably while coming closer to polyethylene board, which was a substitute for water. As a result of this present study, it has become clear that the sequential reaction rates are important factors in the accurate evaluation of induced activity in fusion reactor design.

Measurement of radiation skyshine with D–T neutron source

The D–T neutron skyshine experiments have been carried out at the Fusion Neutronics Source (FNS) of JAERI with the neutron yield of ∼1.7×10 11 n/s. The concrete thickness of the roof and the wall of a FNS target room are 1.15 and 2 m, respectively. The FNS skyshine port with a size of 0.9×0.9 m 2 was open during the experimental period. The radiation dose rate outside the target room was measured a maximum distance of 550 m from the D–T target point with a spherical rem-counter. Secondary gamma-rays were measured with high purity Ge detectors and NaI scintillation counters. The highest neutron dose was about 9×10 −22 Sv/(source neutron) at a distance of 30 m from the D–T target point and the dose rate was attenuated to 4×10 −24 Sv/(source neutron) at a distance of 550 m. The measured neutron dose distribution was analyzed with Monte Carlo code MCNP-4B and a simple line source model. The MCNP calculation overestimates the neutron dose in the distance range larger than 230 m. The line source model agrees well with the experimental results within the distance of 350 m.