MCNP-4A Code Research Articles

Optimization Studies of Elemental Composition of Composite Polymer for Neutron Shielding Using Factorial Design Analysis

The present study is envisaged with an aim to highlight a novel approach of applying the statistical factorial design analysis (FDA) technique in radiation shielding design. In FDA, the estimated total dose rate (TDR) and concentration of elements in shielding material are termed as “Response” and “Factors.” The impact on the response due to the change in the level of factors is defined as “Effects.” Monte Carlo simulation (MCS) is performed using the MCNP4A code to compute the surface TDRs due to the coupled neutron-gamma field arising from the 740-GBq 241Am-Be source housed inside a shielding container made of composite polymer (CP). The composition of CP is hydrogen, carbon, and oxygen with lead and natural boron as fillers. In the present work, the weight percent of hydrogen, carbon, and lead is optimized in the CP by minimizing the surface TDR at the exit of the shield, and a proposed CP (PCP) is obtained. For the first time, a detailed regression analysis is performed to develop a model linking TDR and the three factors, namely, hydrogen weight percent, carbon weight percent, and lead weight percent. Three levels of each factor are considered, and the impact due to the linear, quadratic, and interaction effects of the factors that influences the TDR is investigated using Student’s t-test analysis. The results from the statistical analysis indicate that the weight percent of hydrogen and lead have a greater influence on TDR. The interaction effects arising out of the combination of hydrogen, carbon, and lead are observed to be negligible. Hence, the regression model is modified by dropping the statistically insignificant terms from the equation, and the new model has shown excellent correlation within ±1% of the estimated TDRs using MCS. The R2 and R2 Adj values are found to be 0.99970 and 0.99966, which explains the computation power of the model. The model can be applied to compute the TDRs for any combinations of factors within the range of variability as studied in the present work. The shielding container made of PCP obtained from the present study provides a minimum of 20% reduction in volume and mass compared to the conventional high-density polyethylene and other polymer-based materials.

Deterministic Analysis of the Low Enriched Uranium SLOWPOKE-2 Research Reactor Using DRAGON-5 and DONJON-5 Codes System

This paper discusses the capability of the deterministic code system DRAGON-5 and DONJON-5 to reliably simulate the SLOWPOKE-2 research reactor. The characteristics of complex geometry, highly heterogeneity and large leakage of the SLOWPOKE-2 core, poses a challenge for classical neutronics calculation tools and schemes for power reactors. The simulation model is presented (geometry, cross section library, assumption, etc.), and the results obtained by DRAGON-5/DONJON-5 codes and the evaluated nuclear data from WLUP project were compared to the calculations performed with Monte Carlo code MCNP 4A using a detailed geometrical model of the reactor and the experimental data. Criticality calculations have been performed to verify and validate the model based on the multiplication factor comparisons for 4 energy groups with two different leakage models. The results indicate that the DRAGON-5 and DONJON-5 codes system agrees well with the MCNP 4A continuous-energy Monte Carlo simulation and experiment. Since created model properly describes the reactor core it can be used for calculations of reactor core parameters and for optimization of research reactor application.

Neutron flux spectra investigations in breeding blanket assembly containing lithium titanate and Lead irradiated with DT neutrons

A benchmark experiment was performed to obtain the neutron spectra in a breeding blanket assembly consisting of lithium titanate and Lead using multiple foil activation technique with unfolding code SAND-II-SNL. The purpose of the experiment was to validate the calculated neutron spectra and neutron induced reaction rates with MCNP-4A code and FENDL-2.1 & FENDL-3.0 transport libraries in the layered geometry of lithium titanate and Lead. Measured reaction rates in different activation foils were compared with MCNP-IRDFF-1.05 and FISPACT-EAF2007 calculated values. In general, FENDL-3.0 calculated reaction rates are closer to the measured reaction rates than FENDL-2.1 calculated rates. Calculated Lead multiplication is improved from FENDL-2.1 to FENDL-3.0 up to 2%. Unfolded neutron spectra in the multiplier zones are slightly underestimated by calculations in the secondary neutron emission range of Pb(n, xn) reactions indicating the need of improvement in the (n, 2n) cross-section of Lead isotopes. The observed underestimation in the Lead multiplication puts the tritium production calculations of breeding blankets using Lead as multiplier on conservative side. The inelastic scattering range of Lead isotopes is overestimated by calculations as large as 80% in one of the breeder zone. This observation requires improvement in the inelastic scattering cross-sections of Lead isotopes in FENDL evaluations.

Use of a Si(Li) detector as β spectrometer

The aim of this work is to demonstrate the capability of a Si(Li) detector for the measurement of β spectra, despite the energy absorption in air and in the Be window. A simple source holder fixes the source on the symmetry axis at 3mm from the detector window. The β-sources are produced by evaporation on a plastic backing plate. Absorbing materials between the source and the sensitive volume of the detector are 3mm of air, a Be window, 0.1μm Si and 20nm of gold. A model of the detector was created for β spectra simulation using the MCNP 4A code. Experimental spectra of 14C, 147Pm, 204Tl, 90Sr/90Y were compared with simulated spectra.

Determination of Pu-239 content in Pu–Be neutron sources using Gamma-ray spectrometry

The need to determine the content of plutonium in Pu–Be neutron sources of Russian provenience arises from the administrative regulations applying to nuclear materials. The determination of a plutonium (all isotopes) amount was based on the measurement of an activity Pu-239. Gamma-ray spectrometry with semiconductor detector HPGe was applied. The determination of plutonium (all isotopes) amount was based on the measurement of Pu-239 activity by means of Gamma-ray spectrometry with semiconductor detector HPGe. Gamma lines of radionuclides Am-241, Pu-238, Pu-239 and Pu-241 were detected in spectra. Detection efficiencies were calculated by Monte Carlo method using MCNP4A code. Computations were done for several hypothetical plutonium amounts. Activities determined from peak areas of 129 and 413 keV photons emitted from Pu-239 were confronted. The right amount was established under the condition of equality of both activities.

Optimized concept design of the target station of Chinese spallation neutron source

CSNS (Chinese spallation neutron source) target station, with proton beam power of 100 kW, consists of Tungsten rectangular target surrounded by a beryllium/steel reflector, three wing-moderators and the shield having 18 beam tubes. The leakage neutron intensity from the target (with reflector) and heat deposition on the target, reflector and shield were calculated using Monte Carlo code NMTC/JAM respectively. It is reported that the target having rectangular section will produce more leakage neutron intensity than a square one for the same proton power. The temperature and thermal stress distribution in the target disks were calculated by the finite element method. The performances of moderators were calculated using MCNP-4A code.

Calculations of neutron and gamma transport in ITER H-alpha and CXRS diagnostic channels

Calculations of neutron and gamma transport were carried out with the MCNP-4A code for some optical diagnostic systems located in the upper and equatorial ports of the reactor ITER. The main emphasis was placed on gamma-neutron fluxes and energy deposition in mirrors, quartz windows and optical fibres. Neutron and gamma spectra at primary vacuum windows were calculated for the purpose of determining accurately their possible transparency degradation in ITER on the basis of irradiation test data. The effectiveness of neutron shielding was considered from a viewpoint of radiation safety. A similar analysis was done for the CXRS diagnostic system: the nuclear energy deposition in mirrors and the dose rate responses at toroidal field coils were calculated.

Near surface photon energy spectra outside a 6 MV field edge

The purpose of this study was to investigate the difference between a 6 MV linear accelerator x-ray energy spectrum outside the field edge near a phantom surface, and the corresponding spectrum on the central axis. The Monte Carlo code MCNP-4A was used to calculate the spectra on the central axis and at 1, 2, 5 and 10 cm from the edge of a 4 × 4 cm2, 10 × 10 cm2 and 15 × 15 cm2 field. Compared to the spectrum on the central axis, the spectra outside the field edge showed two distinct regions: a broad peak below about 0.5 MeV, and a lower amplitude, less rapidly changing region at higher energies from 0.5 to 6 MeV. The lower energy peak was due to scattered photons, and the higher energy component was due mainly to primary photons transmitted through the jaws of the secondary collimator. The potential impact of these spectral differences on critical organ photon dosimetry was determined by calculating the ratio of the sensitivity of a Scanditronix EDD-5 diode and of a LiF:Mg:Ti thermoluminescent dosimeter (TLD) outside the field edge to their respective sensitivity at the calibration position on the central axis. The lower energy peak combined with the non-uniform energy sensitivity of each detector produced up to a two-thirds overestimate of x-ray dose outside the field by the diode, whereas the response ratio of the TLD was about unity. These results indicated that a similar evaluation was required for profile measurements of a dynamic wedged field and measurements in an intensity modulated beam with either type of detector.

Sensitivity analysis for a 14MeV neutron benchmark using Monte Carlo and deterministic computational methods

A sensitivity analysis has been performed for a 14 MeV neutron benchmark on an iron assembly, typical for a fusion neutronic integral experiment. Probabilistic and deterministic computational methods have been used in the sensitivity calculations with the main objective to check and validate the novel Monte Carlo technique for calculating point detector sensitivities. Good agreement has been achieved between the Monte Carlo and the deterministic approaches for the individual calculated sensitivity profiles, the uncertainties and the neutron flux spectra. It is thus concluded that the Monte Carlo technique for calculating point detector sensitivities and related uncertainties as being implemented in MCSEN, a local version of the MCNP4A code with the capability to calculate point detector sensitivities, is well qualified for sensitivity and uncertainty analyses of integral experiments.

Assessment of IAEA Exemption Quantities on the Basis of Effective Dose

The exemption quantities of the 20 most common radionuclides in use were evaluated according to the methodology applied in establishing exemption quantities by the IAEA in Basic Safety Series 115. The parts of the external dose conversion coefficients of the IAEA for various exposure modes, which affect exemption level significantly, were not based on effective dose of ICRP 60, but on an effective dose equivalent of ICRP 26. The internal effective dose conversion factors of the IAEA were somewhat different from those of ICRP 68. This paper calculated the effective dose conversion coefficients for external exposure, using MCNP4A code to replace those effective dose equivalent conversion coefficients of the IAEA and applied the internal dose conversion of ICRP 68, then combined them into the re-assessment of the exemption quantities.Energy Fluences in an exposed body for 20 mono photon energies ranging from 0.01 to 2 MeV(29 mono photon energies from 1.2keV to 2MeV for skin dose equivalent) were calculated using MCNP4A code and the response functions of effective doses for the energies were obtained by applying the effective dose coefficients of ICRP 74. Finally, effective dose coefficients of each radionuclide were calculated, using the response functions above and the nuclear transformation data provided in ICRP 38.When the exemption quantities were calculated by IAEA methodology as it is, the results matched the current IAEA value, admitting 10% of relative error except for P-32. Meanwhile, being evaluated with effective doses assessed in this study, the values of S-35 and Tl-201 were greatly different from IAEA’s to a magnitude of an order in addition to the deviation of P-32.The exemption quantities of radionuclides emitting low energy photon, beta, including Tl-201, S-35 will vary with the present IAEA values, if assessed in terms of effective dose coefficients, and it is necessary to revaluate the current IAEA exemption quantity of P-32.

The effect of the imaging geometry and the impact of neutron scatter on the detection of small features in accelerator-based fast neutron radiography

In fast neutron radiography, the neutron energy spectrum, the imaging geometry, neutron scattering and the position-sensitive detector influence the feature contrast, resolution and the signal-to-noise ratio in the image. The effect of the imaging geometry can be explored by using a ray-tracing method. This requires following the path of neutrons through the imaging field, which includes the sample of interest. A relationship between imaging geometry and feature detectability can be developed. There are several important physical processes that govern image formation in fast neutron radiography. These are often examined individually in order to develop an understanding of how they affect the detectability of a feature in a radiograph; however, their combined effect is required to develop a model that can be regarded as a benchmark for future applications. In order to accomplish this, diverse techniques and tools are used. In this study it was expected that the ray-tracing experiments would show that changes in the imaging geometry would greatly affect the detectability of small features as a result of a loss in the resolution and indirectly the contrast. The effect of neutron scatter on the detectability of small features was taken into consideration by using Monte Carlo code MCNP-4A.

Measurement of Burnup in FBR MOX Fuel Irradiated to High Burnup

The burnup of fuel pins in the subassemblies irradiated at the range from 0.003 to 13.28%FIMA in the JOYO MK-II core were measured by the isotope dilution analysis. For the measurement, 75 and 51 specimens were taken from the fuel pins of driver fuel and irradiation test subassemblies, respectively. The data of burnup could be obtained within an experimental error of 4%, and were compared with the ones calculated by 3-dimensional neutron diffusion codes MAGI and ESPRIT-J, which are used for JOYO core management system. Both data of burnup almost agree with each other within an error of 5%. For the fuel pins loaded at the outer region of the subassembly in the 4th row, which was adjacent to reflectors, however, some of the calculation results were 15% less at most than the measured values. It is suggested from the calculation by a Monte Carlo code MCNP-4A that this difference between the calculated and the measured data attribute from the softening of neutron flux in the region adjacent to the reflector.

Measurement of Leakage Neutron Spectra from a Spherical Pile of Zirconium Irradiated with 14MeV Neutrons and Validation of Its Nuclear Data

In order to make a benchmark validation of the nuclear data for Zr, the leakage neutron spectrum from a Zr sphere of a 61-cm diameter was measured between 0.1 and 16MeV using a time-of-flight technique with a 14MeV neutron source facility, OKTAVIAN. The result was compared with the calculation using the Monte Carlo code MCNP-4A. To investigate the spectrum dependence on the individual neutron reactions, test calculations were carried out with the MCNP-4A code using the JENDL-3.2-based libraries, in which partial cross section values were reduced from the original values. From the comparison between the measured and the calculated spectra, it was found that each of the results could predict well the experiment in general. However, in detail, both ENDF/B-VI and EFF-2.4 gave considerable overestimation above 1 MeV. The JENDL-3.2 predicts the spectrum almost satisfactorily except below 0.8 MeV and around 10 MeV. The discrepancy found in JENDL-3.2 calculation is considered due to the cross section values of the (n, 2n) reaction and its secondary energy distributions (SED). The modified JENDL-3.2 library with the reduced (n, 2n) reaction values and the lower SED below 1 MeV reproduced the experiment with better agreement over the whole energy range.

Measurement of Burnup in FBR MOX Fuel Irradiated to High Burnup

The burnup of fuel pins in the subassemblies irradiated at the range from 0.003 to 13.28%FIMA in the JOYO MK-II core were measured by the isotope dilution analysis. For the measurement, 75 and 51 specimens were taken from the fuel pins of driver fuel and irradiation test subassemblies, respectively. The data of burnup could be obtained within an experimental error of 4%, and were compared with the ones calculated by 3-dimensional neutron diffusion codes MAGI and ESPRIT-J, which are used for JOYO core management system. Both data of burnup almost agree with each other within an error of 5%. For the fuel pins loaded at the outer region of the subassembly in the 4th row, which was adjacent to reflectors, however, some of the calculation results were 15% less at most than the measured values. It is suggested from the calculation by a Monte Carlo code MCNP-4A that this difference between the calculated and the measured data attribute from the softening of neutron flux in the region adjacent to the reflector.

Benchmark and MC Sensitivity/Uncertainty Analyses of FNS Beryllium and Iron TOF Experiments and IPPE Nickel TOF Experiment Using EFF-3 Data

In order to test new EFF-3 evaluations for 9Be, 56Fe, 58Ni and 60Ni, MCNP based analyses of the JAERI/FNS beryllium and iron TOF experiments and the IPPE Obninsk nickel TOF experiment were carried out. The analyses showed the followings results; EFF-3.0 56Fe (the previous version of EFF-3 56Fe) reproduces the measured neutron fluxes best. EFF-3.156Fe (the newest EFF-3 data of 56Fe) tends to overestimate the neutron flux from 5 to 10 MeV by 20 - 30 % and to underestimate it from 0.05 to 0.1 MeV by 20 %. For the newest 9Be, 58Ni and 60Ni EFF-3 data, the achieved accuracy is comparable to FENDL-1 and -2. Monte Carlo sensitivity and uncertainty calculations were also performed with a local version of the MCNP4A code. In general the uncertainties of the neutron flux calculated with EFF-3 data were smaller than those with FENDL-1.

359. 大きな超過倍率決定のための修正法の適用

The fuel addition method or the neutron absorption substitution method have been used for determination of large excess multiplication factor of large sized reactors. It has been pointed out, however, that all the experimental methods are possibly not free from the substantially large systematic error up to 20%, when the value of the excess multiplication factor exceeds about 15%Δk. Then, a basic idea of a revised procedure was proposed to cope with the problem, which converts the increase of multiplication factor in an actual core to that in a virtual core by calculation, because its value is in principle defined not for the former but the latter core. This paper proves that the revised procedure is able to be applicable for large sized research and test reactors through the theoretical analyses on the measurements undertaken at the JMTRC and JMTR cores. The values of excess multiplication factor are accurately determined utilizing the whole core calculation by the Monte Carlo code MCNP4A.

The Shielding Effect of Iron Louvers for Neutrons and Gamma Rays

The shielding effect for neutrons and gamma rays provided by a louver-type steel structure inserted into the second leg of a concrete duct was measured and analyzed. The louver is an assembly of steel plates that are stacked at the same interval to reduce the radiation streaming while keeping air flowing through the duct. The experiment was carried out at the Japan Research Reactor-4 (JRR4) using a large concrete duct that was temporarily installed for this experiment.Experimental data for the shielding effect of the louver were obtained through the use of thermoluminescent dosimeters (CaSO4 and BeO) and an ionization chamber for the gamma dose. A rem counter was used to obtain neutron dose, while a solid-state track detector was used for fast neutron dose. Finally, indium activation foil was used to obtain the thermal neutron flux. A NaI(Tl) scintillation spectrometer was used for the measurement of gamma rays from activated foil.The measured data were compared to that derived from numerical analyses. Numerical analyses included the use of the conventional Sn transport code DOT3.5, the Monte Carlo code MCNP4A, and calculations with empirical formulas.MCNP4A provided satisfactory estimates for all cases. If proper calculations were carried out, then DOT3.5 provided acceptable estimates except for the thermal neutrons in spite of the limitations of the code’s two-dimensional geometrical modeling. Calculations by hand using simple empirical formulas with modifications, like that for the angular flux correction, also could provide fairly accurate estimates.

Calculated energy dependence of CaSO 4:Dy TL phosphor and phosphor embedded Teflon for X and gamma rays

The energy dependence of a CaSO 4 : Dy TL phosphor for 21 monoenergetic photons of energy ranging from 0.01 to 3 MeV was calculated by using MCNP4A code. The calculation results show good agreement with those of other authors within ± 5% relative error. Calculations and experiments are also performed to determine the energy dependence of CaSO 4 : Dy phosphor with Dy concentrations from 0.01 to 1.5 mol %, and with the compositions of TL–Teflon from 10 to 90 wt %. The calculated energy responses show good agreement with the experiment results within ± 20% relative error except for a concentration lower than 0.1 mol % Dy and the low energy regions of M30 (Eave.=20 keV) and M60 (Eave.=34 keV). For the TL–Teflon mixture, the energy dependence shows an decreasing trend with the increase of Teflon concentration in MCNP calculation but shows no dependence for TL–Teflon mixture proportions in experiment. These differences are due to the non-homogeneous distribution of Dy concentration and the effect of large grain size in the phosphors. The energy dependence of the 30 wt/o CaSO 4 (0.1 mol%;Dy) and 70 wt/o Teflon that is placed behind the filters of plastic, aluminum, copper, tin and lead with a thickness from 0.1 to 2 mm were calculated, respectively.

Neutron energy spectra inside a PET cyclotron vault room

Neutron energy spectra were measured at three locations inside the vault room of a PET cyclotron. Measured neutron fields were produced during 17.2 MeV protons and 8.6 MeV deuterons colliding with Faraday cup, and 8.6 MeV deuterons colliding with 14N. Measurements were performed using a Bonner sphere spectrometer whose active 6LiI(Eu) scintillator was replaced by two pairs of thermoluminscent dosimeters. Modified neutron spectrometer was calibrated using bare 252Cf and D 2O moderated 252Cf neutron sources. Thermoluminiscent dosimeters were calibrated using γ-rays from 137Cs and 60Co sources. From this calibration, a single factor was derived that allow us to obtain the thermal neutron net signal used to unfold neutron spectra. MCNP 4A code was used to calculate neutron spectra at those sites were measurements were taken. For MCNP calculations a source term for protons colliding with graphite was calculated using a semiempirical model. Due the lack of source term for neutrons produced during 8.6 MeV deuterons nuclear reactions, with carbon and 14N, several models were tried. From experimental and calculated results a systematic behavior of neutron spectra was observed regardless the primary neutron source. This behavior was a peak indicating the presence of evaporation neutrons due to the large amount of iron in cyclotron.

Investigations on neutron-induced prompt gamma ray analysis of bulk samples

A systematic investigation was carried out for the improvement of the prompt gamma interrogation method used for contraband detection by the pulsed fast/thermal neutron analysis (PFTNA) technique. Optimizations of source–detector shielding and geometry, role of the type and dimension of the gamma detector, attenuation of neutrons and gamma rays in bulky samples were also studied. Results obtained for both the shielding materials and elemental content of cocaine simulants have been compared with the values calculated by the MCNP-4A code.