Dose conversion coefficients for Chinese reference adult male and female voxel phantoms from idealized neutron exposures

A whole-body voxel phantom called CRAF representing the Chinese adult female was constructed in this study from color photographs. The CRAF has a height of 160 cm and a weight of 54 kg with 80 organs and tissues, including almost all organs required in the ICRP Publication 103. Masses of all organs in CRAF are consistent with the Chinese reference data within 1%. Our research group has previously developed a Chinese reference adult male phantom called CRAM, which has a height of 170 cm and a weight of 60 kg with 80 tissues and organs. A new set of fluence-to-dose conversion coefficients based on the Chinese reference adult voxel phantoms CRAM and CRAF are presented for six idealized external neutron exposures from 10^?8 MeV to 20 MeV. The calculation of organ-absorbed doses and effective doses were performed with the Monte Carlo transport code MCNPX 2.4. The resulting dose conversion coefficients were compared with those published in ICRP Publication 116, which represents the reference Caucasian. It was observed that the two sets of organ-absorbed dose conversion coefficients agreed well for most organs in ROT and ISO geometry; however, considerable deviations were found in some organs in AP, PA, RLAT and LLAT exposures. Organ-absorbed dose coefficients difference showed good agreement with the difference of organ depth distribution between Chinese and ICRP phantoms. For neutrons with energies above 2 MeV, the effective dose conversion coefficients of Chinese reference adult are almost identical to those of ICRP Publication 116 in AP, PA, ROT and ISO geometries, while are generally 15% higher than those of ICRP Publication 116 from lateral geometries. When energies range from 10^?8 MeV to 1 MeV, differences are within 10% in AP (–5%), PA (–8%) and ROT (4%) geometries; however, relatively large discrepancies are shown in lateral and ISO geometries, with differences of 15% for LLAT, 20% for RLAT and 12% for ISO, respectively.     

Evaluation of neutron nuclear data for Technetium-99

Neutron nuclear data of ⁹⁹Tc was evaluated, considering cross-sections and spectra provided from recent experiments. The evaluation was made in the incident neutron energy range from 1 keV to 20 MeV, using the optical model and nuclear reaction models. The optical model calculation based on the coupled-channels method was performed for the interaction of neutrons with ⁹⁹Tc, and potential parameters appropriately chosen reasonably explain the measured data of total cross-section. The cross-section of inelastic scattering, capture, (n, 2n), (n, p), (n, α) and (n, n′α) reactions, and γ-ray emission spectra were calculated on the basis of statistical model with preequilibrium and direct components, and they were compared with available experimental data. It is found that the presently evaluated cross-sections and γ-ray emission spectra well reproduce those experimental values and that there is a large discrepancy among the present result and evaluated data for neutron emission spectra. The obtained capture cross-section increases at the energies below 1 MeV, relative to that in JENDL-4.0. This makes the transmutation efficiency of ⁹⁹Tc into stable ¹⁰⁰Ru by accelerator driven system enhanced. The production cross-section of ⁹⁹Mo important for the medical use of nuclear diagnostics reduces by 5–30% at the energies above 12 MeV, compared with JENDL-4.0.     

Transformation of a system consisting of plane isotropic source and unit sphere detector into a system consisting of point isotropic source and plane detector in Monte Carlo radiation transport calculation

In a nuclear power plant accident, radioactive nuclides may be released which are distributed uniformly on the ground. If estimation of dose rate from such a source by a Monte Carlo calculation is attempted, some difficulty is encountered because the calculation efficiency is very low. To solve this low efficiency problem, we show that a plane isotropic source can be transformed into a point isotropic source by changing the detector shape from a unit sphere to a plane. We verified the validity of this transformation by the numerical comparison of unscattered photon fluence. As an example of this transformation, the ambient dose rate D _i was calculated from the uniform radioactive nuclide distribution on the ground using the EGS5 Monte Carlo code. We also measured the radioactivity and ambient dose rate (M) on the KEK campus within a month after the releases from the Fukushima No. 1 Nuclear Power Plant accident. Using radioactivity data and D _i, we calculated the ambient dose rate (C). The calculated and measured ambient dose rates agreed reasonably well; their ratio (C/M) was 0.62 to 1.28.     

Analysis of core physics experiments on irradiated BWR MOX fuel in the REBUS program

Critical experiments were performed in the REBUS program on a core loaded with a test bundle including 16 irradiated BWR-type MOX rods of average burnup of 61 GWd/t. The experimental data were analyzed using diffusion, transport, and continuous-energy Monte Carlo calculation codes coupled with nuclear data libraries based on JENDL-3.2 or JENDL-3.3. Biases in effective multiplication factors of the critical cores were ?1.0%Δk for the diffusion calculations (JENDL-3.2), ?0.3%Δk for the transport calculations (JENDL-3.3), and 0.2%Δk for the Monte Carlo calculations (JENDL-3.2). The measured core fission rate and co-activation rate distributions were generally well reproduced using the three types of calculations. The burnup reactivity determined using the measured water level reactivity coefficients was ?2.41 ± 0.08%Δk/kk’, which also agreed with the results of the three type of calculations within the measurement and calculation errors. The most probable isotopic inventories in the irradiated MOX rods was tentatively obtained by using the ratios of the calculation to chemical assay data on a pellet sample, and the burnup reactivity was reanalyzed to split the calculation error into those due to the inventory and reactivity calculations. This approach showed that the inventory calculation error compensated the reactivity calculation error.     

10岁中国儿童参考人面元体模的建立

本文基于首都儿科研究所提供的中国儿童真实CT扫描图像建立了10岁中国儿童参考人面元体模。采用了多边形网格模型与NURBS曲面模型相结合的方法,建立了10岁中国儿童的多边形网格面元。体模包括85个器官,经过优化和调整,各组织和器官质量与GBZ/T 200 2-2007和亚洲参考人的参考值差异小于1%,各器官位置和形态符合中国儿童解剖学特征。该面元体模为10岁中国儿童剂量的准确计算提供了重要工具。     

Synthesis and Characterization of a Soft-tissue Substitute for Neutron Dosimetry

To develop a physical phantom for neutron dosimetry, a solid soft-tissue substitute was synthesized. The synthesized tissue substitute, NAN-JAERI, is improved in both hydrogen and oxygen elemental composition in comparison with existing tissue substitutes. To examine the radiation characteristics of the new soft-tissue substitute, absorbed dose distributions in NAN-JAERI were measured using a ²⁵²Cf neutron source. The measured absorbed dose distributions of neutrons and photons agree with those calculated by a Monte Carlo simulation code MCNP. The agreement between the experiment and the simulation verifies this method of evaluating the soft-tissue equivalence of NAN-JAERI for ²⁵²Cf neutrons. Similar simulations for some mono-energetic neutron sources showed that the newly developed tissue substitute has soft-tissue equivalent characteristics in the neutron energy range from 1 MeV up to 14 MeV, in terms of the absorbed dose distributions in a slab phantom.     

Feedback on neutron capture cross sections of 238Pu and 241Am from analyses of measured isotopic compositions of irradiated LWR fuels and MOX core physics experiments

The atomic fractions of ²³⁸Pu and ²⁴¹Am in MOX fuels recycled in light water reactors are 1% to 2% and not significant compared with those of major Pu isotopes. On the other hand, recent evaluated nuclear data libraries, such as JENDL-4.0 and JEFF-3.2, give noticeably different thermal and epithermal neutron capture cross sections for ²³⁸Pu and ²⁴¹Am. The thermal neutron capture cross sections of ²³⁸Pu and ²⁴¹Am in JEFF-3.2 are 31% and 9% larger than those of JENDL-4.0, respectively. This paper shows the effect of the differences in the neutron cross sections on analysis results of two different integral experiments. The first is the isotopic compositions of ²³⁸Pu on UO₂ and MOX fuels irradiated in BWR and PWR, and the second is the critical experiments of the water moderated cores fully loaded with MOX fuels. The former was analyzed by using the continuous energy Monte Carlo burnup calculation code MVP-BURN and the latter by the continuous energy Monte Carlo calculation code MVP. The comparisons between the calculated and measured results indicate that the most likely thermal and epithermal neutron capture cross sections of ²³⁸Pu and ²⁴¹Am should be around at the middle between those of JEFF-3.2 and JENDL-4.0.     

Calculation of response matrix of CaSO4:Dy based neutron dosimeter using Monte Carlo code FLUKA and measurement of 241Am–Be spectra

Response matrix for CaSO₄:Dy based neutron dosimeter was generated using Monte Carlo code FLUKA in the energy range thermal to 20 MeV for a set of eight Bonner spheres of diameter 3–12″ including the bare one. Response of the neutron dosimeter was measured for the above set of spheres for ²⁴¹Am–Be neutron source covered with 2 mm lead. An analytical expression for the response function was devised as a function of sphere mass. Using Frascati Unfolding Iteration Tool (FRUIT) unfolding code, the neutron spectrum of ²⁴¹Am–Be was unfolded and compared with standard IAEA spectrum for the same.     

Intergranular stress corrosion cracking initiation model of austenite stainless steels used in BWRs for optimized water chemistry control

A calculation model on intergranular stress corrosion cracking (IGSCC) initiation time of materials used in boiling water reactors (BWRs) has been developed to evaluate effectiveness of water chemistry control for mitigation of the IGSCC. The model was composed of four terms which determine passive film break time: (1) a chemical term based on electrochemical corrosion potential (ECP) and impurity concentration; (2) a mechanical term based on strain rate; (3) a material term based on sensitization; and (4) an irradiation term based on acceleration of corrosion by γ-rays and neutron irradiation. The contribution of the chemical term in the passive film break was calculated based on a deterministic local corrosion model. Then, the local corrosion model was modified by adding mechanical acceleration of the film rupture to treat the IGSCC phenomenon. The model could reproduce the behavioral tendency seen in the slow strain rate tensile test on high carbon contents with sensitization heat treatment (for example, 620°C × 24 h). Under BWR operating conditions, IGSCC initiation time could be extended by a factor of 5 by lowering the electric conductivity from 1.0 to 0.06 μS/cm. If the ECP was reduced below the critical potential by a mitigation method, the IGSCC initiation time was predicted to become sufficiently long for pipings and components.     

Neutronics Experimental Study on Tritium Production in Solid Breeder Blanket Mockup with Neutron Reflector

Neutronics experiments have been performed for the solid breeder blanket using a DT neutron source at the FNS facility in JAEA. We have applied the blanket mockup composed of two enriched Li₂TiO₃ and three beryllium layers, and measured the detailed spatial distribution of the tritium production rate (TPR) using enriched Li₂CO₃ pellets. TPRs in the pellets have been measured by a liquid scintillation counter. Experiments have been done under a condition with a neutron reflector surrounding the DT neutron source. Numerical simulations have been performed using the MCNP-4C with the FENDL-2.0 and JENDL-3.3. The ranges of ratios of calculation results to experimental ones (C/Es) are 0.97–1.17 concerning with local TPR, and 1.04–1.09 for the integrated tritium production. It is found that the total integrated tritium production, which corresponds to tritium breeding ratio, can be predicted within uncertainty of 10% using the Monte Carlo calculation code and latest nuclear data libraries.     

Voxel-Based Frog Phantom for Internal Dose Evaluation

A voxel-based frog phantom has been developed for radiation protection of the environment. The voxel-based frog phantom was applied to evaluating self-absorbed fractions (self-AFs), which are defined as the fraction of energy emitted by a radiation source that is absorbed within the source organ. The self-AFs were evaluated for both photons and electrons in the spleen, kidneys, and liver using Monte Carlo simulations. Furthermore, self-S values (mGy/MBq.s) for ¹⁸F and ⁹⁰Y in the organs were calculated using the results of the self-AFs. Consequently, the voxel-based frog phantom was found to be useful for the organ dose evaluations, which have not been proposed by the International Commission on Radiological Protection (ICRP). It was also confirmed that the self-AFs and self-S values are largely dependent on the mass of the source organ.     

Measurement and Analysis of Reactivity Worth of 241Am Sample in Water-Moderated Low-Enriched UO2 Fuel Lattices at TCA

The reactivity worths of 22.82 grams of ²⁴¹Am oxide sample were measured and theoretically analyzed in water-moderated UO2 fuel lattices in seven cores of the Tank-Type Critical Assembly (TCA) at the Japan Atomic Energy Agency for an integral test of ²⁴¹Am nuclear data. These cores provided a systematic variation in the neutron spectrum between the thermal and resonance energy regions. The sample reactivity worth was measured with an uncertainty of 2.1% or less. The theoretical analysis was performed using the JENDL-3.3 nuclear data by a Monte Carlo calculation method. Ratios of calculation to experiment (C/Es) of the reactivity worth were between 0.91 and 0.97, and showed no apparent dependence on the neutron spectrum. In addition, sensitivity analysis based on the deterministic calculation method was carried out to obtain the impact of changing the ²⁴¹Am capture cross section on the sample reactivity worth. The result of this analysis showed that the C/E could be significantly improved by almost uniformly increasing the ²⁴¹Am capture cross section of JENDL-3.3 by 25–30%.     

Neutron Self-Shielding Factor of Scattering-Predominant Resonance Foils

The neutron self-shielding factor of ⁵⁹Co resonance foil as an example of foils whose scattering cross section predominate over their absorption cross sections was obtained by both Monte Carlo method (analog) and the collision probability method for various thicknesses of the foil. Also, the transmission and reflection probabilities of neutrons which have various energies near the resonance energy were obtained, and the effects of multiple scattering of neutrons on the neutron self-shielding factor are discussed.

The neutron self-shielding factors obtained by the Monte Carlo method and by the collision probability method agreed well with each other in the cases Σ t ～ 4.0, in which the Monte Carlo method requires considerably longer machine time. Although for the cases of large Σ t (～4.0) the agreement is not always good because of the flat flux approximation in the collision probability method, the calculation time by Monte Carlo is conveniently short. A combination of both methods is useful in obtaining the neutron self-shielding factor of resonance foils.     

Detection of Delayed Neutrons from Fissionable Samples: Monte Carlo Modelling and Physical Assumptions for a Design of the DET-12 Device

An activation of fissionable materials with neutrons has been considered as a possible neutron diagnostic of D–D and D–T fusion plasma. Fission reaction caused by fusion neutrons leads up to emission of secondary neutrons: prompt and delayed. Physical assumptions have been outlined to design a new device (DET-12) for measurements of delayed neutrons emitted from samples of fissionable materials activated with neutrons at big fusion-plasma devices. The aim is to support a classic neutron activation method used as one of plasma diagnostics at tokamaks or stellarators. An interpretation of the time decay of delayed neutrons enables an assessment of the primary neutron flux which induced fission reaction. Monte Carlo calculations have been carried out in order to elaborate the method considered. Nuclides like: pure ²³⁵U, ²³⁸U and ²³²Th, have been selected as possible materials to be irradiated. Physical fundamentals of generation of the delayed neutrons are mentioned and a resulting concept of the DET-12 device, built in the Institute of Nuclear Physics, Poland, is presented. A general size and dimensions of particular constituent material layers, and a number and placement of neutron detectors are optimized by means of Monte Carlo modelling. Recommendations for a technical design of the measuring chamber were formulated. Detection efficiency of DET-12 has been also estimated.     

Measurement and Analysis of Reactivity Worth of 237Np Sample in Cores of TCA and FCA

The reactivity worth of 22.87 grams of ²³⁷Np oxide sample was measured and analyzed in seven uranium cores in the Tank-Type Critical Assembly (TCA) and two uranium cores in the Fast Critical Assembly (FCA) at the Japan Atomic Energy Agency. The TCA cores provided a systematic variation in the neutron spectrum between the thermal and resonance energy regions. The FCA cores, XXI and XXV, provided a hard neutron spectrum of the fast reactor and a soft one of the resonance energy region, respectively. Analyses were carried out using the JENDL-3.3 nuclear data library with a Monte Carlo method for the TCA cores and a deterministic method for the FCA cores. The ratios of calculated to experimental (C/E) reactivity worth were between 0.97 and 0.91, and showed no apparent dependence on the neutron spectrum.     

BNCT人头体模内剂量分布计算

用修正的Synder人头体模几何模型和ICRU-46中的材料数据，用MCNP-4B程序对0．0253ev、1kev、2keV、10keV、100keV、1MeV单能中子束，0．2、0．5、1、2、5、10MeV单能光子束，以及与当前硼中子俘获治疗(BNCT)临床中使用的超热中子相似的超热中子束，计算了在人头体模中的剂量分布，计算结果与有关文献报道的结果一致，初步校验了我们正在编制的BNCT治疗计划软件。     

Fluence and dose measurements for an accelerator neutron beam

The 3 MV Van de Graaff accelerator at McMaster University accelerator laboratory is extended to a neutron irradiation facility for low-dose bystander effects research. A long counter and an Anderson-Braun type neutron monitor have been used as monitors for the determination of the total fluence. Activation foils were used to determine the thermal neutron fluence rate (around 10⁶ neutrons s⁻¹). Meanwhile, the interactions of neutrons with the monitors have been simulated using a Monte Carlo N Particle (MCNP) code. Bystander effects, i.e. damage occurring in cells that were not traversed by radiation but were in the same radiation environment, have been well observed following both alpha and gamma irradiation of many cell lines. Since neutron radiation involves mixed field (including gamma and neutron radiations), we need to differentiate the doses for the bystander effects from the two radiations. A tissue equivalent proportional counter (TEPC) filled with propane based tissue equivalent gas simulating a 2 μm diameter tissue sphere has been investigated to estimate the neutron and gamma absorbed doses. A photon dose contamination of the neutron beam is less than 3%. The axial dose distribution follows the inverse square law and lateral and vertical dose distributions are relatively uniform over the irradiation area required by the biological study.     

Thermal neutron capture cross sections for the Sm(n,γ)Sm and Sm(n,γ)Sm reactions at 0.0536 eV energy

The neutron capture cross sections for the ¹⁵²Sm(n,γ)¹⁵³Sm and ¹⁵⁴Sm(n,γ)¹⁵⁵Sm reactions at 0.0536 eV neutron energy were measured using an activation technique based on the TRIGA Mark-II research reactor, relative to the reference reaction ¹⁹⁷Au(n,γ)¹⁹⁸Au. The activity was measured nondestructively using gamma-ray spectroscopy. Our measured values at this neutron energy are the first ones and are compared with 1/v based evaluated cross sections reported in the ENDF/B-VII and JENDL-3.3 libraries. The measured value for the ¹⁵²Sm(n,γ)¹⁵³Sm reaction is 0.28% lower than JENDL-3.3 and 0.48% higher than ENDF/B-VII. Our value for the production of ¹⁵⁵Sm is about 3% and 2.3% higher than the evaluated value with ENDF/B-VII and JENDL-3.3 at 0.0536 eV, respectively.     

TRIPOLI-4® Monte Carlo code ITER A-lite neutronic model validation

3D Monte Carlo transport codes are extensively used in neutronic analysis, especially in radiation protection and shielding analyses for fission and fusion reactors. TRIPOLI-4^® is a Monte Carlo code developed by CEA. The aim of this paper is to show its capability to model a large-scale fusion reactor with complex neutron source and geometry. A benchmark between MCNP5 and TRIPOLI-4^®, on the ITER A-lite model was carried out; neutron flux, nuclear heating in the blankets and tritium production rate in the European TBMs were evaluated and compared. The methodology to build the TRIPOLI-4^® A-lite model is based on MCAM and the MCNP A-lite model. Simplified TBMs, from KIT, were integrated in the equatorial-port. A good agreement between MCNP and TRIPOLI-4^® is shown; discrepancies are mainly included in the statistical error.