Radiation Shielding Analysis for Pressurized Heavy Water Reactors (CANDU) Using MCNPX Code

MCNPX(Monte Carlo N-Particle Transport Code)computer code is used to design a model to CANDU(Canada Deuterium Uranium)reactor core and its shielding system.It is assumed that reactor core is fueled with natural uranium.The core radiation sources are calculated which consider prompt neutrons,neutron induced gamma and prompt gamma radiations.The total neutron flux and dose rate are calculated along the shield and at outer shield points.The results indicated that the major dose rate component at outer shield points is due to neutron induced gamma dose rate(μSv/h).

Calculation of photon attenuation coefficient and dose rate in concrete with the addition of SiO_2 and MnFe_2O_4 nanoparticles using MCNPX code and comparison with experimental results

One of the most important safety features of nuclear facilities is the shielding material used to protect the operating personnel from radiation exposure. The most common materials used in radiation shielding are concretes. In this study, a Monte Carlo N-Particle eXtended code is used to calculate the gamma-ray attenuation coefficients and dose rates for a new concrete material composed of MnFe_2O_4 nanoparticles, which is then compared with the theoretical and experimental results obtained for a SiO_2 nanoparticle concrete material. According to the results, the average relative differences between the simulations and the theoretical and experimental results for the linear attenuation coefficient(l) in the SiO_2 nanoparticle materials are 6.4% and 5.5%, respectively. By increasing the SiO_2 content up to 1.5% and the temperature of MnFe_2O_4 up to 673 K, l is increased for all energies. In addition, the photon dose rate decreases up to 9.2% and3.7% for MnFe_2O_4 and SiO_2 for gamma-ray energies of0.511 and 1.274 MeV, respectively. Therefore, it was concluded that the addition of SiO_2 and MnFe_2O_4 nanoparticles to concrete improves its nuclear properties and could lead to it being more useful in radiation shielding.

Neutronic design investigation of a liquid injection-based second shutdown system for a typical research reactor using MCNPX

Safety systems, built on state-of-the-art technology, are essential for achieving acceptable levels of plant safety to minimize hazards to the reactor and the general public. The second shutdown system(SSS) as an engineered safety feature and a part of the reactor protection system(RPS) is a means for rapidly shutting down a nuclear reactor, keeping it in a subcritical state and serving as a backup to the first shutdown system(FSS). In this research, one SSS with two types of optimum chamber designs is proposed that take into account the main current characteristic features of the Tehran research reactor with improvements over earlier designs. They are based on a liquid neutron absorber injection that is preferably different, diverse, and independent from the FSS based on the rod drop mechanism. The major design characteristics of this SSS with two different chambers were investigated using MCNPX 2.6.0 code. The performed calculations showed that the designed SSS is a reliable shutdown system, assuring an appropriate shutdown margin and injection time, with no significant effects on the effective delayed neutron fraction while causing minimal variations to the core structure. Further, the reasonable financial cost and the prolongation of the operation cycle are additional advantages of this design.

基于示范快堆的ADS次临界快堆堆芯研究

选取中国示范快堆作为次临界快堆参考堆芯,研究次临界快堆作为嬗变PWR(U)乏燃料中次锕系元素的可行性。中国示范快堆堆芯设计是参考目前正在建设的俄罗斯示范快堆BN-800。次临界快堆堆芯在示范快堆堆芯基础上去掉中间7盒组件放置铅靶组件,控制棒组件用含贫铀和次锕系元素(MA)的组件代替,转换区组件用反射层组件代替。采用MCNPX和ORIGEN2程序作为计算软件。计算结果表明:次临界快堆中加入MA后能够保持一定的次临界度且具有较好的嬗变效果,因此,选取示范快堆堆芯作为ADS次临界快堆的参考堆芯研究是可行的。

ADS次临界堆脉冲中子源实验动态特性数值模拟研究

强外源驱动与深次临界度使得ADS次临界反应堆在中子学特性上与传统临界堆有较大差异,确定论中子学计算方法难以直接应用于ADS次临界堆.本文采用MCNPX程序对"快热"耦合ADS装置YALINA-Booster的PNS实验进行了模拟,并将模拟与实验结果进行比较.结果表明:在不同的堆芯布置方案和不同脉冲中子源特性下,模拟结果与实验结果具有良好的一致性,验证采用MCNPX程序研究ADS次临界堆中子学动态特性的可行性.

Analysis of CANDU Reactor Performance Using Thorium Fuel:Comparison with Natural UO2 Case

The purpose of the paper is to study the performance of the CANDU(Canada Deuterium Uranium)reactor when the reactor core is loaded with thorium fuel mixed with plutonium isotopes with ratio 3 and 5%.A three dimensional model is designed for the core of CANDU reactor.The computer code MCNPX(Monte Carlo N–Particle Transport)is used to calculate the processes in its core.The results are compared with natural UO2 case which is the typical fuel of the reactor.The results show that the multiplication factor of the reactor is higher even in the case of thorium fuel mixed with 3%plutonium isotopes,which indicates longer neutron life cycle length and more economic utilization of the reactor.

Review the Behavior of Thorium Based Fuel (U,Th) and (Pu,Th)

Study on the behavior of thorium based fuel in a fuel bundle is the aim of this Simulation.check the spectrum flux in theoretical sample Shown that(Th,U)and(Th,Pu)cycle can work in one fuel bundle.

Measurement of mass attenuation coefficients,effective atomic numbers,and electron densities for different parts of medicinal aromatic plants in low-energy region

The mass attenuation coefficients(l/q) for different parts(root, flower, stem, and leaf) of three medicinal aromatic plants(Teucrium chamaedrys L. subsp. sinuatum,Rheum ribes, and Chrysophthalmum montanum) were measured using an ^(241)Am photon source in a stable geometry and calculated using the Monte Carlo N-Particle Transport Code System-extended(MCNPX) code and the WinXCOM program. The experimental and theoretical MCNPX and WinXCOM values exhibited good agreement.The measured mass attenuation coefficient values were then used to compute the effective atomic number(Z_(eff))and electron density(N_E) of the samples. The results reveal that S1-S(stem of Teucrium chamaedrys L. subsp. sinuatum) has the highest values of l/q and Zeff.

Determination of water equivalent ratio for some dosimetric materials in proton therapy using MNCPX simulation tool

The water equivalent ratio(WER) was calculated for polypropylene(PP), paraffin, polyethylene(PE), polystyrene(PS), polymethyl methacrylate(PMMA), and polycarbonate materials with potential applications in dosimetry and medical physics. This was performed using the Monte Carlo simulation code, MCNPX, at different proton energies. The calculated WER values were compared with National Institute of Standards and Technology(NIST) data, available experimental and analytical results,as well as the FLUKA, SRIM, and SEICS codes. PP and PMMA were associated with the minimum and maximum WER values, respectively. Good agreement was observed between the MCNPX and NIST data. The biggest difference was 0.71% for PS at 150 MeV proton energy. In addition, a relatively large positive correlation between the WER values and the electron density of the dosimetric materials was observed. Finally, it was noted that PE presented the most analogous Depth Dose Characteristics to liquid water.

Investigation of neutronic parameters of ^(nat)U spallation target irradiated by low-energy protons

Accelerator-based neutron sources could outstandingly compete with the reactor-based ones, which are widely used for research aims and radioisotope production.Spallation neutron sources are used by many research centers. In this work, the potential of natural uranium spallation target irradiated by low-energy protons for production of an external neutron source was investigated.MCNPX code was used to model the spallation target. The results showed using 30-Me V protons of 100 μA current a neutron flux in order of 10~7n/s cm^2 leaks from an optimized-dimension target. Different physical models available in the computational code do not result in significant relative discrepancies for neutron yield and deposited heat calculations. Water with a velocity of 0.6 m/s can be used as coolant for the spallation target to keep the surface temperature under 100 °C at atmospheric pressure.

Designing a nuclear battery based on the Mo-99 radioactive source soluble in water and aqua regia in order to use in early tests

Today, millions of electrocommunication, electric, medical, and industrial devices use battery. Batteries with long life and high energy density seem to be essential in medical, military, oil and mining, aerospace areas as well as conditions in which access is difficult and in situations where replacement or recharging of battery is costly.In this regard, the use of radiation energy resulting from radioactive materials and its conversion to electric energy can be effective in making batteries. In the present study,various Mo-99 radioisotope values with a half-life of 65.98 h were used as a soluble radioactive source in two materials of water and aqua regia. Then, by comparing the results of the Monte Carlo simulations program MCNPX for these two solutions, it was found that when the water is used instead of aqua regia(for idealization), the values of the superficial current of electrons, the volumetric flux of electrons, and the deposited energy in the volume containing the radioactive solution increased by 10.80, 4.10,and 13.80%, respectively. Also, the short-circuit current and energy conversion efficiency of this battery with a concentration of 0.01 molar, Mo-99 dissolved in the aqua regia are 0.79μA and 16.47%, respectively.

Photoneutron dose and flux determination of a typical LINAC by MCNP simulation

Purpose High-energy electron linear accelerators(LINACs)have a wide use in radiotherapy.The photoneutron production in medical linear accelerators,when operating at energies greater than 10 MeV,is an important subject to consider in patient’s treatment procedures.In this work,we simulate a typical LINAC to calculate photoneutron dose and flux in radiotherapy room.Methods The latest version of MCNPX Monte Carlo code along with MCNP visual editor has been used to simulate the LINAC and its components.Photoneutron production has been successfully simulated by selecting appropriate physics card and parameters.Dose and flux of produced photoneutrons have been calculated by using mesh tally function.Results Production of photoneutrons in LINAC head and its components have been successfully simulated by using MCNPX code.MCNP visual editor has been used to track the particles.Photoneutron dose and flux have been calculated using mesh tally function,with good results of statistical tests.Conclusion The photoneutron production has been successfully simulated and benchmarked.The proposed simulation code is able to calculate photoneutron dose and flux.According to photoneutron production cross sections,the appropriate parameters have been selected to reduce the run-time of simulation code.