MCNP5 modeling of the IPR-R1 TRIGA reactor for criticality calculation and reactivity determination

The IPR-R1 TRIGA is a research nuclear reactor managed and located at the Nuclear Technology Development Center (CDTN) a research institute of the Brazilian Nuclear Energy Commission (CNEN). It is mainly used to radioisotopes production, scientific experiments, training of nuclear engineers for research and nuclear power plant reactor operation, experiments with materials and minerals and neutron activation analysis. In this work, criticality calculation and reactivity changes are presented and discussed using two modelings of the IPR-R1 TRIGA in the MCNP5 code. The first model (Model 1) analyzes the criticality over the reactor. On the other hand, the second model (Model 2) includes the possibility of radial and axial neutron flux evaluation with different operation conditions. The calculated results are compared with experimental data in different situations. For the two models, the standard deviation and relative error presented values of around 4.9 × 10^?4. Both models present good agreement with respect to the experimental data. The goal is to validate the models that could be used to determine the neutron flux profiles to optimize the irradiation conditions, as well as to study reactivity insertion experiments and also to evaluate the fuel composition.     

Optimization study of some neutron imaging parameters in a reactor based neutron radiography facility

Neutron imaging technique can be used as a means of material Non-Destructive testing. One of common neutron sources for neutron radiography is nuclear research reactor. In this work, several neutron imaging parameters such as aperture distance and the radiography plane location from the neutron source as well as the aperture diameter have been computationally optimized to deliver a proposed neutron beam. According to the results, the aperture diameter of 3.5–4 cm which was located at 55–85 cm from the outer layer of reactor core and the position of image plane at 300–400 cm fulfills delivering of the suitable neutron flux and other required conditions. W, Fe and Pb walls with an identified length formed the convergent-divergent collimator and shielded the neutron and gamma out of beam path. Bi and Fluental filters with an optimal dimension were used to efficiently improve the neutron beam profile at a sample position.     

Multi purpose research reactor

At present Dhruva and Cirus reactors provide the majority of research reactor based facilities to cater to the various needs of a vast pool of researchers in the field of material sciences, physics, chemistry, bio sciences, research & development work for nuclear power plants and production of radio isotopes. With a view to further consolidate and expand the scope of research and development in nuclear and allied sciences, a new 20 MWt multi purpose research reactor is being designed. This paper describes some of the design features and safety aspects of this reactor.     

Refurbishment and safety upgradation of research reactor Cirus

Cirus, a 40 MW t, vertical tank type research reactor, having wide range of research facilities, was commissioned in the year 1960. This research reactor, situated at Mumbai, India has been operated and utilized extensively for isotope production, material testing and neutron beam research for nearly four decades. With a view to assess the residual life of the reactor, detailed ageing studies were carried out during the early 1990s. Based on these studies, refurbishment of Cirus for its life extension was taken up. During refurbishment, additional safety features were incorporated in various systems to qualify them for the current safety standards. This paper gives the details of the operating experiences, utilization of the reactor along with methodologies followed for carrying out detailed ageing studies, refurbishment and safety upgradation for its life extension.     

An innovative research reactor design

A new and innovative core design for a research reactor is presented. It is shown that while using the standard, low enriched uranium as fuel, the maximum thermal flux per MW of power for the core design suggested and analyzed here is greater than those found in existing state of the art facilities without detrimentally affecting the other design specs. A design optimization is also carried out to achieve the following characteristics of a pool type research reactor of 10 MW power: high thermal neutron fluxes; sufficient space to locate facilities in the reflector; and an acceptable life cycle. In addition, the design is limited to standard fuel material of low enriched uranium. More specifically, the goal is to maximize the maximum thermal flux to power ratio in a moderate power reactor design maintaining, or even enhancing, other design aspects that are desired in a modern state of the art multi-purpose facility. The multi-purpose reactor design should allow most of the applications generally carried out in existing multi-purpose research reactors. Starting from the design of the German research reactor, FRM-II, which delivers high thermal neutron fluxes, an azimuthally asymmetric cylindrical core design with an inner and outer reflector, is developed. More specifically, one half of the annular core (0 < θ < π) is thicker than the other half. Two variations of the design are analyzed using MCNP, ORIGEN2 and MONTEBURNS codes. Both lead to a high thermal flux zone, a moderate thermal flux zone, and a low thermal flux zone in the outer reflector. Moreover, it is shown that the inner reflector is suitable for fast flux irradiation positions. The first design leads to a life cycle of 41 days and high, moderate and low (non-perturbed) thermal neutron fluxes of 4.2 × 10¹⁴ n cm⁻² s⁻¹, 3.0 × 10¹⁴ n cm⁻² s⁻¹, and 2.0 × 10¹⁴ n cm⁻² s⁻¹, respectively. Heat deposition in the cladding, coolant and fuel material is also calculated to determine coolant flow rate, coolant outlet temperature and maximum fuel temperature under steady-state operating conditions. Finally, a more compact version of the asymmetric core is developed where a maximum (non-perturbed) thermal flux of 5.0 × 10¹⁴ n cm⁻² s⁻¹ is achieved. The core life of this more compact version is estimated to be about 23 days.     

Research of Radioisotope Production with Fast Neutrons, (VII)

Production of radioisotopes of high specific activity was studied in the JRR-1 reactor using several (n,p) and (n,α) reactions, such as ²⁴Mg(n,p)²⁴Na, ²⁷A1(n,α)²⁴N_a, ³⁵Cl(n,p)³⁵S, ³⁵C1(n,α)³²P, ⁵⁸Ni(n,p)⁵⁸Co, ⁶⁴Z_n (n,p) ⁶⁴C_u and ⁶⁷Z_n(n,p) ⁶⁷Cu. The target materials for these reactions were irradiated in several experimental holes of JRR-1 and the radioisotopes formed in the target materials were separated. The amount of the radioisotopes produced and the specific activity were determined, and the possibility of producing high specific activity radioisotopes by these reactions was investigated. The specific activity of the radioisotopes produced by the (n,p) and (n,γ) reactions was more than several hundreds times higher than when produced by the corresponding (n,γ)reactions. Although the yield of the radioisotopes by the former two reactions was fairly small, practical production of high specific activity radioisotopes by this method was thought to be possible, at least for elements of lower atomic number such as those studied in the present work.

For each experimental hole, the thermal and the fast neutron fluxes were determined respectively by the reactions ¹⁹⁷Au(n,γ)¹⁹⁸Au and ⁵⁸Ni(n,p)⁵⁸Co. In order to apply these (n,p) and (n,α) reactions effectively to radioisotope production, such basic informations as the dependence of the reactions on neutron energy and the effect of irradiation position on the reaction yield were studied on the basis of the neutron flux distribution, and the cross section of the reactions for fast neutrons in JRR-1 was estimated.     

中国先进研究堆(CARR)应用设计及其规划

中国先进研究堆(CARR)是一座多用途、高性能指标的研究堆,CARR采用反中子阱型堆芯结构,便于提供更多的空间进行水平中子束流孔道和垂直辐照孔道的布置,满足多用途的需要。CARR综合应用研究平台建成后,具备开展中子敢射实验研究、燃料和材料性能研究、在线中子活化分析、中子照相、硼中子俘获治疗癌症、放射性同位素及单晶硅中子掺杂研发以及核能和核技术人才教育培训等的能力。本文就CARR可能开展的主要应用以及与应用有关的物理设计、结构设计及配套系统设计进行了简要介绍。     

Evaluation of irradiation facility options for fusion materials research and development

Successful development of fusion energy will require the design of high-performance structural materials that exhibit dimensional stability and good resistance to fusion neutron degradation of mechanical and physical properties. The high levels of gaseous (H, He) transmutation products associated with deuterium–tritium (D–T) fusion neutron transmutation reactions, along with displacement damage dose requirements up to 50–200 displacements per atom (dpa) for a fusion demonstration reactor (DEMO), pose an extraordinary challenge. One or more intense neutron source(s) are needed to address two complementary missions: (1) scientific investigations of radiation degradation phenomena and microstructural evolution under fusion-relevant irradiation conditions (to provide the foundation for designing improved radiation resistant materials), and (2) engineering database development for design and licensing of next-step fusion energy machines such as a fusion DEMO.A wide variety of irradiation facilities have been proposed to investigate materials science phenomena and to test and qualify materials for a DEMO reactor. Some of the key technical considerations for selecting the most appropriate fusion materials irradiation source are summarized. Currently available and proposed facilities include fission reactors (including isotopic and spectral tailoring techniques to modify the rate of H and He production per dpa), dual- and triple-ion accelerator irradiation facilities that enable greatly accelerated irradiation studies with fusion-relevant H and He production rates per dpa within microscopic volumes, D–Li stripping reaction and spallation neutron sources, and plasma-based sources.The advantages and limitations of the main proposed fusion materials irradiation facility options are reviewed. Evaluation parameters include irradiation volume, potential for performing accelerated irradiation studies, capital and operating costs, similarity of neutron irradiation spectrum to fusion reactor conditions, temperature and irradiation flux stability/control, ability to perform multiple-effect tests (e.g., irradiation in the presence of a flowing coolant, or in the presence of complex applied stress fields), and technical maturity/risk of the concept. Ultimately, it is anticipated that heavy utilization of ion beam and fission neutron irradiation facilities along with sophisticated materials models, in addition to a dedicated fusion-relevant neutron irradiation facility, will be necessary to provide a comprehensive and cost-effective understanding of anticipated materials evolution in a fusion DEMO and to therefore provide a timely and robust materials database.     

Feasibility of an epithermal neutron source for BNCT based on RFQ accelerator

The use of boron neutron capture therapy for the treatment of deep-seated tumours, such as glioblastoma multiforme, requires neutron beams of suitable energy and intensity. The analysis of the therapeutic gain shows that a high tumour control probability with sublethal dose at healthy tissues can be achieved, in most cases, by using neutron beams of a few keV energy, with a flux of about 10⁹ neutrons/cm² s. Therapeutic neutron beams with high-spectral purity in this energy range could be produced by accelerator-based neutron sources through a suitable neutron-producing reaction. We investigate the feasibility of a solution based on a small radio frequency quadrupole for a proton beam current of 30 mA and an energy of 2 MeV. An appropriate choice of the function parameters of the RFQ (modulation, efficiency of acceleration, phase shift, etc., …) allows one to design relatively compact accelerators, which could eventually lead to setup hospital-based BNCT facilities.     

Design of moderator and multiplier systems for D–T neutron source in BNCT

Boron Neutron Capture Therapy (BNCT) is an outstanding way to treat Glioblastoma Multiforme. Epithermal neutrons with energy from 1 eV to 10 keV represent the most effective range for brain tumor therapy. In this research we have focused on ³H(d, n)⁴He reaction as a neutron source using Cock Craft Walton accelerator. High neutron yield with 14.1 MeV energy can be generated via accelerating a deuteron beam with 110 keV energy.A Monte Carlo simulation code (MCNP4C) was used to design the D–T source. Pb and ²³⁸U are suggested as neutron multipliers; AlF₃ and BeO as a moderator and reflector, respectively. An Al layer is used for decreasing the ratio of fast to total neutron fluxes. Epithermal neutron flux in the suggested system is 10⁸ n/cm² s and is a suitable flux for BNCT applications. Finally the suggested configuration is compared to the most recent works and it is shown that the proposed configuration works better.     

Design calculation of an epithermal neutronic beam for BNCT at the Syrian MNSR using the MCNP4C code

This article describes the design calculation of an epithermal neutronic beam for the boron neutron capture therapy at the Syrian MNSR by using the MCNP4C code and ENDF/B-V cross-section library. To produce a high flux of epithermal neutrons at the beam exit, the moderator/filter from Al, Cd, Fluental and Bi was used with Pb as reflector for neutrons along the beam. In addition, the Bi lined collimator with Li₂CO₃-PE and Pb at the end. The calculated beam parameters under 30.0 kW of reactor power at the beam exit are Ф_epi = 2.83 × 10⁸ n/cm² s, D_f/Ф_epi = 7.98 × 10⁻¹¹ cGy cm²/n, D_γ/Ф_epi = 1.70 × 10⁻¹¹ cGy cm²/n, Φ_epi/Φ_the = 0.05 and J_n⁺/Ф_n = 0.77. As well as, the calculated values of the advantage depth and advantage ratio are 7.51 cm and 3.49, respectively. If such beam was built into the Syrian MNSR the scientific applications of the reactor would increase.     

Design of the thermal neutron beam for neutron radiography at the Syrian MNSR

Neutron beam design was studied at the Syrian reactor (MNSR, 30 kW) with a view to generating thermal neutron beam in the vertical irradiation sites for neutron radiography. The design of the neutron collimator was performed using MCNP4C and the ENDF/B-V cross-section library. Thermal, epithermal and fast neutron energy ranges were selected as <0.4 eV, 0.4 eV–10 keV, >10 keV, respectively. To produce a good neutron beam quality, bismuth was used as photon filter. In this design, the L/D ratio of this facility had the value of 125. The thermal neutron flux at the beam exit was about 2.548 × 10⁵ n/cm² s. If such neutron beam were built into the Syrian MNSR many scientific applications would be available using the neutron radiography.     

Calculational tools to conduct experimental optimization in Tehran research reactor

Utilization of irradiation facilities in Tehran research reactor (TRR) requires proper computational tools to deliver accurate and precise results. In this paper validity of different schemes are checked against experimental measurements. A reference core with a neutron flux trap in the middle is chosen as the reference model for this purpose. ©     

Research reactor RBT-6

Conclusions Reactor RBT-6 is simple in construction and is easily accessible for conducting experiments. The values of neutron flux in it are high for small thermal power; this together with the large duration of continuous operation ensures the possibility of conducting a wide range of experimental investigationa. Such a reactor may be recommended as a research reactor for irradiation of samples of materials up to moderate flux values (10¹⁹–10²¹ neutrons/cm²) and for conducting experiments for studying the change in the properties of materials during irradiation, which are becoming increasingly more important. Estimates show that the number of used fuel assemblies of the SM-2 reactor are sufficient for the operation of several such reactors.If necessary, the number of experimental channels in the active zone of such a reactor can be increased by increasing the number of fuel assemblies and the thermal power. Beryllium can be used as the lateral reflector. This results in a decrease of the volume of the active zone and the thermal power of the reactor, but increases its cost.Translated from Atomnaya Énergiya, Vol. 43, No. 1, pp. 3–7, July, 1977.     

Activation analysis of the water cooling system of the LIPAc beam dump

LIPAc stands for Linear IFMIF Prototype Accelerator. LIPAc generates a 9 MeV deuteron beam, which is stopped at a beam dump, depositing over 1 MW of thermal power. A water cooling system has been devised for extracting this energy while keeping operational temperatures within range. The existing high neutron fluxes in the beam dump during operation produce activation of both coolant and beam stopper, which also suffers from corrosion into the coolant. The presence of radioisotopes in the cooling water leads to a radiological hazard.Water purification systems are located outside the accelerator vault and accumulate activated products during filtration, requiring a specific radiological shield to comply with target dose rates. Also devices containing large volume of activated cooling water, like N-16 decay pipes, require specific radioprotection analysis and design. This work identifies the most relevant radiation sources due to the activated cooling fluid, which may result in radiation doses to workers, and propose radioprotection measures into the design to mitigate their effect.     

Monte Carlo simulation of the thermal column and beam tube of the TRIGA Mark II research reactor

The Monet Carlo simulation of the TRIGA Mark II research reactor core has been performed employing the radiation transport computer code MCNP5. The model has been confirmed experimentally in the PhD research work at the Atominstitute (ATI) of the Vienna University of Technology. The MCNP model has been extended to complete biological shielding of the reactor including the thermal column, radiographic collimator and four beam tubes. This paper presents the MCNP simulated results in the thermal column and one of the beam tubes (beam tube A) of the reactor. To validate these theoretical results, thermal neutron flux density measurements using the gold foil activation method have been performed in the thermal column and beam tube A (BT-A). In the thermal column, the theoretical and experimental results are in fairly good agreement i.e. maximum thermal flux density in the centre decreases in radial direction. Further, it is also agreed that thermal flux densities in the lower part is greater than the upper part of the thermal column. In the BT-A experiment, the thermal flux density distribution is measured using gold foil. The experimental and theoretical diffusion lengths have been determined as 10.77 cm and 9.36 cm respectively with only 13% difference, reflecting good agreement between the experimental and simulated results. To save the computational cost and to incorporate the accurate and complete information of each individual Monte Carlo MC particle tracks, the surface source writing capability of MCNP has been utilized to the TRIGA shielding model. The variance reduction techniques have been applied to improve the statistics of the problem and to save computational efforts.     

Evaluation of gamma-ray and neutron energy for area monitoring system in the IFMIF/EVEDA accelerator building

The engineering validation of the IFMIF/EVEDA prototype accelerator, up to 9 MeV by supplying the deuteron beam of 125 mA, will be performed at the BA site in Rokkasho. A design of this area monitoring system, comprising of Si semiconductors and ionization chambers for covering wide energy spectrum of gamma-rays and ³He counters for neutrons, is now in progress. To establish an applicability of this monitoring system, photon and neutron energies have to be suppressed to the detector ranges of 1.5 MeV and 15 MeV, respectively. For this purpose, the reduction of neutron and photon energies throughout shield of water in a beam dump and concrete layer is evaluated by PHITS code, using the experimental data of neutron source spectra. In this article, a similar model using the beam dump structure and the position with a degree of leaning for concrete wall in the accelerator vault is used, and their energy reduction including the air is evaluated. It is found that the neutron and photon flux are decreased by 10⁴-order by employing the local shields using concrete and polyethylene around beam dump, and the photon energy can be suppressed in the low energy.     

Monte Carlo simulation of the NaI(Tl) detector response to measure gold activated foils

This work deals with the implementation of a NaI(Tl) detector for the assessment of the specific saturation activities of pure gold foils after neutron irradiation. These gold foils can be placed in the centre of a set of polyethylene spheres with different diameters. This configuration, known as a passive Bonner sphere system, is suitable to measure neutron spectra normally extended over a wide energy range containing up to 11 decades (from thermal to a few MeV), at places where the neutron field is very intense, high frequency pulsed or where it is mixed with an important high-energy photon component. The MCNPX code was used to evaluate the NaI(Tl) responses to different incident photon energies in terms of pulse-height distributions. An experimental validation of the calculated NaI(Tl) responses, using certified standard sources at a given measurement arrangement, indicates that MCNPX is a valid tool for routine calibration and benchmarking studies of this detector. A good agreement is found between the measured pulse-height distributions of the certified standard sources and those obtained from MCNPX simulations. As a preliminary application, a bare disc Au foil was directly exposed to a Bremsstrahlung photon beam at the isocentre of an 18 MV medical LINAC, in order to test the suitability of this activation material to measure the photo-neutrons generated in such facility. Two differentiated main photo-peaks, arising from ¹⁹⁶Au and ¹⁹⁸Au predominant γ-ray emissions, were observed. The two isotopes are produced mainly by the photonuclear, ¹⁹⁷Au(γ, n)¹⁹⁶Au, and radiative capture, ¹⁹⁷Au(n, γ)¹⁹⁸Au, reactions of, respectively, high-energy photons and thermal neutrons on the gold foil. From the measured ¹⁹⁸Au saturation activity, a rough estimation of (378 ± 68) × 10⁴ cm⁻² Gy⁻¹ was derived for the thermal neutron flux within the LINAC treatment room. This value, although being very approximate, is comparable to those reported by other authors for similar LINAC facilities but with different treatment room configurations, nominal acceleration potentials and Bremsstrahlung photon irradiation areas.