Monte Carlo N-Particle Radiation Transport Research Articles

Calculations of backpatch factors and conversion coefficients for the personal dose equivalent for X-ray beams used in radiodiagnostic

The purpose of this study is to use the Monte Carlo N-Particle radiation transport code (MCNP) to calculate the conversion coefficients of air kerma for personal dose equivalent (Hp [Sv]) and Back Scattering Factors (BSF) for X-ray beams used in diagnostic radiography at potentials of 40, 50, 60, 70, 80, 90, 100, 120, and 150 kVp. Additionally, effective dose is determined using male and female anthropomorphic voxel-based simulators (REX and REGINA) as recommended by the International Commission on Radiological Protection (ICRP) Publication 110, along with tissue weight factors from ICRP Publication 103 to simulate whole-body irradiation. The results indicate that the relationship between operational metrics such as Hp and risk limitation metrics like effective dose varies from 0.14 to 0.71 for Radiation Qualities in Radiation (RQR) beams from the X-ray source assembly and up to 0.64 for Radiation Qualities based on a phantom with an added aluminium filter (RQA). When comparing the obtained values for air kerma conversion coefficients for personal dose equivalent with the value adopted by the National Commission for Nuclear Energy (CNEN) of 1.14 Sv/Gy for individual dose determination, it is observed that at low energies, the values are overestimated by up to 44%, while at high energies, the values are underestimated by up to 67%.

Validation of the single-event method for low-energy electron transport via stopping power calculations with MCNP

Monte Carlo simulations of low-energy (<50 keV) electron transport in matter are essential for a broad range of application fields. Several Monte Carlo codes have developed specialized treatments for this case, but a comprehensive validation of low-energy electron transport for general-purpose simulations remains lacking in the literature. One approach to accomplish this validation is calculation of stopping power using low-energy electron transport physics, as stopping power is a fundamental radiation transport quantity which must be simulated accurately for nearly any application. In this work, we use the Monte Carlo N-Particle (MCNP) radiation transport code with the single-event method for electron transport to calculate stopping powers of low-energy electrons (50 eV to 30 keV) in 41 elemental solids, 14 compound solids, and five rare gas solids, comparing simulation results to published semi-empirical stopping power calculations from optical measurements. In general, the simulations give good agreement (typically within ±10%) with semi-empirical stopping power values at higher energies: 300 eV and above for most elemental solids, 1 keV and above for compound solids, and 400 eV and above for rare gas solids. Agreement between MCNP and semi-empirical values is worse below these energies. The most significant source of error is the EPRDATA14 cross section data, which does not account for changes in electronic structure due to solid-state bonding, particularly in compound materials. The simplistic model of atomic excitation used to generate the EPRDATA14 cross sections is another key source of error. Additionally, the breakdown of the continuous slowing-down approximation introduces significant uncertainty at low energies, although this is a limitation of the calculation method and not of the simulation procedure. Accounting for these and other uncertainty sources, the single event method in MCNP is robust and able to give good accuracy for a variety of low-energy electron transport problems through diverse kinds of materials.

Photon detector response function methodology using MCNP and shift hybrid radiation transport code for wide-area contamination assay applications

Radiation transport modeling using the Monte Carlo N-Particle (MCNP) radiation transport code and Monte Carlo code, Shift, were employed to model detector responses for a variety of wide-area photon contamination scenarios. In this study, 2′′ × 2′′ and 3′′ × 3′′ cylindrical NaI(Tl) scintillation detector configurations at source detector-distances of 0.5 cm, 1 cm, 2.54 cm, 10 cm, and 30 cm were modeled. Media of soil, concrete, and steel were evaluated for contamination depths ranging from surface to a depth of an infinite thickness in each medium for photon energies ranging from 20 keV to 3 MeV, which correspond to the energies that current detectors can discern. Monoenergetic photon surface contamination detector responses for each of the media, source–detector distances, and detectors were estimated using MCNP v6.2. Shift was harnessed for improved variance reduction of particle transport in highly attenuating media to obtain average cell fluxes in the two MCNP NaI(Tl) scintillation detector configurations. Average cell flux values in Shift were coupled with detector responses from MCNP to convert average cell flux in a void to energy distribution of pulses in the NaI(Tl) scintillation detector crystal of interest. An optimized detector response function methodology was developed by coupling the MCNP radiation transport method with the Consistent Adjoint Driven Importance Sampling (CADIS) hybrid radiation transport method built into Shift to significantly decrease the runtime of thousands of MCNP pulse height simulations. The methodology may be utilized to quickly and accurately facilitate the assessment of a broad range of wide-area environmental contamination assay and decommissioning cleanup applications.

Analysis of Loss of Water Inventory at the Philippine Research Reactor-1 Fuel Storage Facility

The Philippine Research Reactor-1 (PRR-1) fuel storage facility is a wet storage for irradiated nuclear fuel. The behavior of the facility in case of accidental loss of water is currently unknown. Safety analysis of a repository of fissile material involves criticality safety assessment and radiation dose estimation for normal and accident scenarios. To determine the radiological consequences of loss of water inventory in the fuel storage facility, we calculated the effective multiplication factor (keff) and gamma dose rate distribution in the system using Monte Carlo N-particle (MCNP) radiation transport code. Results show that keff will decrease from a maximum of 0.61424 ± 0.00013 when the water level is decreased from its fully moderated condition. In contrast, keff will increase to a maximum of 0.7468 ± 0.0002 when voids are introduced in the water. These results indicate that the system will remain subcritical (keff&lt;1) and a runaway supercritical fission chain reaction cannot result from any moderating conditions. Furthermore, the dose assessment for the case of complete loss of water revealed a maximum dose rate of 5.52 ± 0.045 mSv/h and 288 ± 4.65 𝛍Sv/h at the tank surface and tank platform, respectively. These values are significantly above background radiation levels; however, the calculated spatial distribution is highly asymmetric and the dose rate falls off rapidly with increasing distance. These results suggest that a facility-level emergency response is sufficient to address the accident. Our findings can augment the existing facility emergency procedures, thereby contributing to the improvement of radiological safety.

Hybrid halide perovskite neutron detectors

Interest in fast and easy detection of high-energy radiation (x-, γ-rays and neutrons) is closely related to numerous practical applications ranging from biomedicine and industry to homeland security issues. In this regard, crystals of hybrid halide perovskite have proven to be excellent detectors of x- and γ-rays, offering exceptionally high sensitivities in parallel to the ease of design and handling. Here, we demonstrate that by assembling a methylammonium lead tri-bromide perovskite single crystal (CH3NH3PbBr3 SC) with a Gadolinium (Gd) foil, one can very efficiently detect a flux of thermal neutrons. The neutrons absorbed by the Gd foil turn into γ-rays, which photo-generate charge carriers in the CH3NH3PbBr3 SC. The induced photo-carriers contribute to the electric current, which can easily be measured, providing information on the radiation intensity of thermal neutrons. The dependence on the beam size, bias voltage and the converting distance is investigated. To ensure stable and efficient charge extraction, the perovskite SCs were equipped with carbon electrodes. Furthermore, other types of conversion layers were also tested, including borated polyethylene sheets as well as Gd grains and Gd2O3 pellets directly engulfed into the SCs. Monte Carlo N-Particle (MCNP) radiation transport code calculations quantitatively confirmed the detection mechanism herein proposed.

Computational evaluation of a novel beta radiation probe design using integrated circuits

Researchers at Texas A&M University (TAMU) have designed the radiation integrated circuit (RIC) for deployment as a new radiation detection system. Most integrated circuits are susceptible to radiation-induced failures, and decades of research have gone into solving this problem. Research at TAMU has led to a novel integrated circuit design that utilizes both radiation-hardened areas (RHAs) and radiation-sensitive areas (RSAs) to take advantage of these failures. The RSAs are susceptible to charged particle interactions, allowing the RIC to detect alpha and beta particles. However, beta particles are more penetrating compared to alpha particles, resulting in a lower interaction probability for beta particles incident on a bare RIC. In any material, the higher the beta energy, the deeper the beta particle can penetrate; therefore, the use of a wedge-shaped attenuator for beta particle detection not only increases interaction probability, but also provides the capability to perform maximum beta energy discrimination in the field. The objective of this research was to optimize the design of the RIC. Monte Carlo N-particle radiation transport code (MCNP) simulations assessed the beta particle detection and maximum energy discrimination performance of plate glass, borosilicate (Pyrex®) glass, acrylic (Lucite®), and natural rubber attenuators. In this proof-of-concept analysis, natural rubber was observed to be the optimal attenuating material for the beta probe with respect to maximum energy discrimination capability and weight, but all materials considered proved to be good candidates. The results of this study are promising and indicate the potential to achieve maximum beta particle energy discrimination of 50 keV using a wedged, natural rubber attenuator on the RIC.

Comparison of Neutron Organ and Effective Dose Coefficients for PIMAL Stylized Phantoms in Bent Postures in Cranial and Caudal Irradiation Geometries.

Radiation dose estimations in the human body are performed using computational reference phantoms, which are anatomical representations of the human body. In previous studies, dose reconstructions have been performed focusing primarily on phantoms in an upright posture, which limits the accuracy of the dose estimations for postures observed in realistic work settings. In this work, the International Commission on Radiological Protection (ICRP) Publication 103 recommendations for monoenergetic neutron plane sources directed downward from above the head (cranial) and upward from below the feet (caudal) for adult female and male reference phantoms were used to calculate organ absorbed and effective dose coefficients. The Phantom with Moving Arms and Legs (PIMAL) and the Monte Carlo N-Particle (MCNP) radiation transport code were used to compute organ-absorbed dose and effective dose coefficients for the upright, half-bent (45°), and full-bent (90°) phantom postures. The doses calculated for each of the articulated positions were compared to those calculated for the upright posture by computing the ratios of the coefficients (45°/upright and 90°/upright). These ratios were used to assess the effectiveness of upright phantoms in providing a comparable estimate when conducting dose estimations and dose reconstructions for articulated positions. This work compiling neutron cranial and caudal posture-specific dose coefficients completes the series of dose coefficients computed for posture-specific ICRP Publication 116 irradiation geometries for monoenergetic photons and neutrons, in addition to cranial and caudal monoenergetic photons. Results reported demonstrated that organ-absorbed dose coefficients for most of the organs in the CRA and CAU irradiation geometries were significantly higher for the bent phantoms than for the upright phantom. Since the upright phantom underestimates the organ-absorbed dose, this demonstrates the impact of posture while performing dose calculations. Organ doses reported in past neutron dose coefficient data were found to omit effects from neutron resonances at energies of 0.435, 1.0, and 3.21 MeV from 16O in tissue. Reported data notes as high as 60% underestimation for neutron organ-absorbed doses, specifically at the neutron resonance energy region omitted by smoothing. Ongoing studies are examining the effect of resonances on reported neutron organ-absorbed dose coefficients in ICRP 116 geometries.

A new methodology to estimate stochastic uncertainty of MCNP-predicted isotope concentrations in nuclear fuel burnup simulations

The Monte Carlo N-particle radiation transport code (MCNP) is used widely in nuclear fuel burnup simulation (FBS). In the FBS, MCNP predicts various neutron reaction rates and the associated stochastic relative error (SRE). The SRE is due to the nature of the statistical methods used in MCNP. These SREs for neutron reaction rates and fluxes are reported in the MCNP output for each FBS time-step. However, MCNP does not report SREs for its predictions of isotope concentrations. A new methodology to compute the SRE for the MCNP-predicted isotope concentrations is developed, which uses MCNP-computed neutron reaction rates and the associated SREs. The effectiveness of the methodology is verified for a pressurized water reactor FBS and is found to be satisfactory. The SRE computation for isotope concentrations in irradiated fuel has applications in nuclear science and engineering.

Design of a Monte Carlo model based on dual-source computed tomography (DSCT) scanners for dose and image quality assessment using the Monte Carlo N-Particle (MCNP5) code

Abstract The purpose of this work was to develop and validate a Monte Carlo model for a Dual Source Computed Tomography (DSCT) scanner based on the Monte Carlo N-particle radiation transport computer code (MCNP5). The geometry of the Siemens Somatom Definition CT scanner was modeled, taking into consideration the x-ray spectrum, bowtie filter, collimator, and detector system. The accuracy of the simulation from the dosimetry point of view was tested by calculating the Computed Tomography Dose Index (CTDI) values. Furthermore, typical quality assurance phantoms were modeled in order to assess the imaging aspects of the simulation. Simulated projection data were processed, using the MATLAB software, in order to reconstruct slices, using a Filtered Back Projection algorithm. CTDI, image noise, CT-number linearity, spatial and low contrast resolution were calculated using the simulated test phantoms. The results were compared using several published values including IMPACT, NIST and actual measurements. Bowtie filter shapes are in agreement with those theoretically expected. Results show that low contrast and spatial resolution are comparable with expected ones, taking into consideration the relatively limited number of events used for the simulation. The differences between simulated and nominal CT-number values were small. The present attempt to simulate a DSCT scanner could provide a powerful tool for dose assessment and support the training of clinical scientists in the imaging performance characteristics of Computed Tomography scanners.

Dose distribution in boron neutron capture therapy for the treatment of brain cancer

This work presents a study of the influence of the field size of an epithermal neutron beam on the dose distribution in treatments with boron neutron capture therapy using an anthropomorphic voxel-based simulator. To calculate the doses in the tissues and organs of the head, the Monte Carlo N-Particle radiation transport code is used. The results suggest the possibility of using a beam of 6 cm in diameter for the treatment of tumours in the frontal lobe and parietal regions. With a beam of 10 cm in diameter, we could treat tumours located in the frontal lobe, parietal lobe and thalamus. The work also shows that the contribution of the secondary components (gamma rays, fast and thermal neutrons) in the calculation of the total dose can result in up to 15% of the dose in the tumour tissue, 68% of the dose in the healthy brain tissue and 87% of the dose in the non-cephalic regions.

A Method to Estimate Fission Product Concentration Uncertainty in a Multi-Time-Step MCNP6 Code Nuclear Fuel Burnup Calculation

The Monte Carlo N-Particle (MCNP6) radiation transport code is widely used to perform material transmutation and depletion calculations using the embedded module CINDER90. CINDER90 is capable of obtaining fission product and transuranic nuclide concentrations with a high level of accuracy in irradiated nuclear fuel. This information is very useful for many nuclear applications including reactor design and analysis, nuclear safeguards, nuclear security, and nuclear forensics, to name a few. However, at present the MCNP6 code does not estimate the overall statistical uncertainty in the nuclide concentrations reported at the end of a depletion calculation. We report our approach using a random sampling method to estimate stochastic uncertainty in fission product nuclide concentration using various parameters reported in MCNP6 output and how these uncertainties are affected by the calculation parameters.

Investigations on Detecting Potential Nuclear Material Diversion from a Pyroprocessing Facility

Under normal operating conditions, a pyroprocessing facility removes highly radioactive and nonradioactive fission product waste from used nuclear reactor fuel to recycle the remaining uranium (U), plutonium (Pu), and other actinides contained in it. The products from this facility are separate ingots of U and mixed transuranic elements (TRUs)–uranium (TRU-U). Uranium in both ingots will be depleted U with 235U enrichment less than 1%. The TRU-U ingot will contain neptunium, Pu, americium (Am), and curium (Cm) mixed with U with an approximate TRU:U ratio of 1:1. Four scenarios of nuclear material diversion by potential misuse of the pyroprocessing facility operations are analyzed and compared with the scenario of normal operating condition when the electrowinning process or the TRU-U ingot manufacturing process is misused. These diversion scenario analyses are carried out to understand the proliferation potential and to recommend safeguards measures. The four scenarios of nuclear material diversion analyzed are (1) 50 g Pu, (2) 100 g Pu, (3) 200 g Pu, and (4) all Pu, i.e., 452 g in the 1-kg TRU-U ingot. Plutonium cannot be diverted by itself because other TRUs (Am and Cm) will be simultaneously extracted with Pu. This is because the reduction potentials of those actinides are not distinguishably different from that of Pu on a liquid cadmium cathode of the electrowinning step of the pyroprocess. Hence, in addition to Pu, simultaneous diversion of respective amounts of Am and Cm for the four diversion scenarios are considered. The diversion scenario analysis also considered the concealment of Pu and Cm removal from the TRU-U ingot by adding an equivalent amount of 252Cf to replenish the neutron source emissions. These five scenarios (four nuclear material diversion scenarios and one normal operation scenario) are modeled and simulated using the Monte Carlo N-Particle (MCNP6) radiation transport computer code by incorporating the model of a NaI gamma radiation detection system. The results show that the presence and absence of Pu in the TRU-U ingot can be confirmed by the NaI gamma radiation detection system. However, identifying the presence of U in the TRU-U ingot is difficult using the NaI gamma radiation detection system due to interference from TRU gamma radiation. To identify the U presence in the TRU-U ingot, an application of nuclear magnetic resonance (NMR) is studied. The NMR technology employs a numerical calculation approach based on density functional theory (DFT) simulation. The DFT calculation results show that the detection of U in a pyroprocess is feasible by NMR technology. In addition, these four nuclear material diversion scenarios are analyzed through MCNP6 simulations by incorporating the model of a coincidence neutron detection system. To conceal the nuclear material diversion, the simulations are performed by replacing the diverted Pu and Cm by an appropriate mass of 252Cf neutron source that is equivalent to the neutron source strengths of the diverted mass. Simulation results show that this concealment (misuse) results in a deceived Pu mass estimate in the TRU-U ingot if the Pu-to-244Cm–ratio method (proposed method in the literature) is used.

PIECEWISE POLYNOMIAL APPROXIMATIONS TO THE ICRP 116 EFFECTIVE DOSE COEFFICIENTS: PHOTONS AND NEUTRONS.

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

Analysis of Radiation Transport due to Activated Coolant in the ITER Neutral Beam Injection Cell

Detailed spatial distributions of the biological dose rate due to a variety of sources are required for the design of the ITER tokamak facility to ensure that all radiological zoning limits are met. During operation, water in the Integrated loop of Blanket, Edge-localized mode and vertical stabilization coils, and Divertor (IBED) cooling system will be activated by plasma neutrons and will flow out of the bioshield through a complex system of pipes and heat exchangers. This paper discusses the methods used to characterize the biological dose rate outside the tokamak complex due to 16N gamma radiation emitted by the activated coolant in the Neutral Beam Injection (NBI) cell of the tokamak building.Activated coolant will enter the NBI cell through the IBED Primary Heat Transfer System (PHTS), and the NBI PHTS will also become activated due to radiation streaming through the NBI system. To properly characterize these gamma sources, the production of 16N, the decay of 16N, and the flow of activated water through the coolant loops were modeled. The impact of conservative approximations on the solution was also examined. Once the source due to activated coolant was calculated, the resulting biological dose rate outside the north wall of the NBI cell was determined through the use of sophisticated variance reduction techniques. The AutomateD VAriaNce reducTion Generator (ADVANTG) software implements methods developed specifically to provide highly effective variance reduction for complex radiation transport simulations such as those encountered with ITER. Using ADVANTG with the Monte Carlo N-particle (MCNP) radiation transport code, radiation responses were calculated on a fine spatial mesh with a high degree of statistical accuracy. Advanced visualization tools were also developed and used to determine pipe cell connectivity, to facilitate model checking, and to post-process the transport simulation results.

A new response matrix for a [formula omitted]LiI scintillator BSS system

A new response matrix was calculated for a Bonner Sphere Spectrometer (BSS) with a 6LiI(Eu) scintillator, using the Monte Carlo N-Particle radiation transport code MCNPX. Responses were calculated for 6 spheres and the bare detector, for energies varying from 1.059E(−9) MeV to 105.9 MeV, with 20 equal-log(E)-width bins per energy decade, totalizing 221 energy groups. A comparison was done among the responses obtained in this work and other published elsewhere, for the same detector model. The calculated response functions were inserted in the response input file of the MAXED code and used to unfold the total and direct neutron spectra generated by the 241Am-Be source of the Universidad Politécnica de Madrid (UPM). These spectra were compared with those obtained using the same unfolding code with the Mares and Schraube matrix response.

Comparison of photon organ and effective dose coefficients for PIMAL stylized phantom in bent positions in standard irradiation geometries

Computational phantoms with articulated arms and legs have been constructed to enable the estimation of radiation dose in different postures. Through a graphical user interface, the Phantom wIth Moving Arms and Legs (PIMAL) version 4.1.0 software can be employed to articulate the posture of a phantom and generate a corresponding input deck for the Monte Carlo N-Particle (MCNP) radiation transport code. In this work, photon fluence-to-dose coefficients were computed using PIMAL to compare organ and effective doses for a stylized phantom in the standard upright position with those for phantoms in realistic work postures. The articulated phantoms represent working positions including fully and half bent torsos with extended arms for both the male and female reference adults. Dose coefficients are compared for both the upright and bent positions across monoenergetic photon energies: 0.05, 0.1, 0.5, 1.0, and 5.0MeV. Additionally, the organ doses are compared across the International Commission on Radiological Protection's standard external radiation exposure geometries: antero-posterior, postero-anterior, left and right lateral, and isotropic (AP, PA, LLAT, RLAT, and ISO). For the AP and PA irradiation geometries, differences in organ doses compared to the upright phantom become more profound with increasing bending angles and have doses largely overestimated for all organs except the brain in AP and bladder in PA. In LLAT and RLAT irradiation geometries, energy deposition for organs is more likely to be underestimated compared to the upright phantom, with no overall change despite increased bending angle. The ISO source geometry did not cause a significant difference in absorbed organ dose between the different phantoms, regardless of position. Organ and effective fluence-to-dose coefficients are tabulated. In the AP geometry, the effective dose at the 45° bent position is overestimated compared to the upright phantom below 1MeV by as much as 27% and 82% in the 90° position. The effective dose in the 45° bent position was comparable to that in the 90° bent position for the LLAT and RLAT irradiation geometries. However, the upright phantom underestimates the effective dose to PIMAL in the LLAT and RLAT geometries by as much as 30% at 50keV.

A neutron transport and thermal hydraulics coupling scheme to study xenon induced power oscillations in a nuclear reactor

A multi-physics computational methodology to analyze xenon induced power oscillations in a nuclear reactor is developed and presented. The methodology development takes into account both neutron transport and thermal hydraulics behaviors and their inter-dependent feedbacks in a nuclear power reactor as a function of fuel burn-up. The methodology uses the Monte Carlo N-Particle radiation transport computational code (MCNP6) along with its fuel transmutation module (CINDER90), a semi-analytic single channel analysis tool for thermal hydraulics and the SIGACE code suite for temperature dependent neutron interaction cross section processing. A Python script developed couples the physics codes for full automation. The accuracy of the multi-physics computational methodology developed was verified through a benchmark calculation for the published core parameters of a startup test performed at Yonggwang Power Reactor Unit No. 3. The power axial offset and xenon axial offset parameters were calculated for this benchmark case and used to quantify the oscillatory behavior observed, the results of this benchmark study is also presented here. The results showed that the developed methodology was able to capture the underlying phenomena governing xenon induced power oscillations in a nuclear reactor.

Benchmarking of Monte Carlo model of PWR (CNPP-II) core against theoretical and experimental results

The purpose of this study is to develop a complete Monte Carlo model of Chashma Nuclear Power Plant Unit-2 (CNPP-II) core and benchmark against the theoretical (Final Safety Analysis Report FSAR) and experimental results. For this purpose, a comprehensive model of CNPP-II core is developed using Monte Carlo N-Particle (MCNP5) radiation transport code which is equipped with temperature dependent ENDF/B-VII.1 cross-section library. The model is validated against the reference results of CNPP-II at seven different operating conditions of the initial core. For Cold Zero Power (CZP) condition, the overall temperature is taken as 40 °C, while for Hot Zero Power (HZP) condition, the temperature is taken as 280 °C. In Hot Full Power (HFP), the fuel, cladding and coolant temperatures are taken as 873 °C, 340 °C and 302 °C respectively. To enhance the model validation from global level (keff) to local level (flux or power distribution), the simulated pin & assembly wise neutron flux and power distribution is compared with the available experimental results of CNPP-II. The comparison demonstrates a reasonable agreement between computational and reference results. This study infers that the criticality value and excess reactivity are within design limits of CNPP-II and the reactor has a symmetrical flat power distribution, ensuring the safe reactor operation. The deviations between the calculated and experimental results have been explained in this paper. This model is interesting in the sense that it presents a new full core benchmark with experimental validation. This research also reveals that CNPP-II, which is in SMR category, is a safe and reliable reactor.

Integrated doses calculation in evacuation scenarios of the neutron generator facility at Missouri S&T

Any source of ionizing radiations could lead to considerable dose acquisition to individuals in a nuclear facility. Evacuation may be required when elevated levels of radiation is detected within a facility. In this situation, individuals are more likely to take the closest exit. This may not be the most expedient decision as it may lead to higher dose acquisition. The strategy followed in preventing large dose acquisitions should be predicated on the path that offers least dose acquisition. In this work, the neutron generator facility at Missouri University of Science and Technology was analyzed. The Monte Carlo N-Particle (MCNP) radiation transport code was used to model the entire floor of the generator's building. The simulated dose rates in the hallways were used to estimate the integrated doses for different paths leading to exits. It was shown that shortest path did not always lead to minimum dose acquisition and the approach was successful in predicting the expedient path as opposed to the approach of taking the nearest exit.

Rapid MCNP simulation of DNA double strand break (DSB) relative biological effectiveness (RBE) for photons, neutrons, and light ions

To account for particle interactions in the extracellular (physical) environment, information from the cell-level Monte Carlo damage simulation (MCDS) for DNA double strand break (DSB) induction has been integrated into the general purpose Monte Carlo N-particle (MCNP) radiation transport code system. The effort to integrate these models is motivated by the need for a computationally efficient model to accurately predict particle relative biological effectiveness (RBE) in cell cultures and in vivo. To illustrate the approach and highlight the impact of the larger scale physical environment (e.g. establishing charged particle equilibrium), we examined the RBE for DSB induction (RBEDSB) of x-rays, 137Cs γ-rays, neutrons and light ions relative to γ-rays from 60Co in monolayer cell cultures at various depths in water. Under normoxic conditions, we found that 137Cs γ-rays are about 1.7% more effective at creating DSB than γ-rays from 60Co (RBEDSB = 1.017) whereas 60–250 kV x-rays are 1.1 to 1.25 times more efficient at creating DSB than 60Co. Under anoxic conditions, kV x-rays may have an RBEDSB up to 1.51 times as large as 60Co γ-rays. Fission neutrons passing through monolayer cell cultures have an RBEDSB that ranges from 2.6 to 3.0 in normoxic cells, but may be as large as 9.93 for anoxic cells. For proton pencil beams, Monte Carlo simulations suggest an RBEDSB of about 1.2 at the tip of the Bragg peak and up to 1.6 a few mm beyond the Bragg peak. Bragg peak RBEDSB increases with decreasing oxygen concentration, which may create opportunities to apply proton dose painting to help address tumor hypoxia. Modeling of the particle RBE for DSB induction across multiple physical and biological scales has the potential to aid in the interpretation of laboratory experiments and provide useful information to advance the safety and effectiveness of hadron therapy in the treatment of cancer.