Conceptual design of a small nuclear reactor for large-diameter NTD-Si using short PWR fuel assemblies

A design of a small nuclear reactor for a large-diameter NTD-Si using a conventional Pressurized Water Reactors (PWR) full-length assembly was proposed in previous works. The height of the full-length assembly was 400 cm, and the overall size of the reactor and reflector around the core became large. In addition, the irradiation channel became very long, making handling of the Si ingots in the channel more difficult. The use of a short PWR fuel assembly, with a height of 100 cm, was considered in the current work. With the shorter assembly, the design of the reactor became compact and more practical. Gd₂O₃ and control rods were used to suppress excess reactivity. Criticality, neutron transport, and core burn-up calculations were performed using the MVP/GMVP II code and MVP-BURN code. Steady-state single-channel thermal hydraulic analyses were also performed. The calculation results showed that the reactor could be critical over 1200 days, and that heat removal from core was possible under 1 atm operating pressure. Large-diameter ingot up to 20 cm in height could be doped with sufficient uniformity. The reactor semiconductor production rate was estimated, and varied between 48 tons/year and 70 tons/year for the 50 Ω cm target resistivity depending on the position of the control rod.     

Neutronic calculations of the U3Si LEU as a dispersed fuel for Low-Power Research Reactors

Usability of the LEU U₃Si dispersed fuel together with the actual UAl₄–Al HEU fuel (mixed core) in Low-Power Research Reactors (LPRRs) (～30 kW) was assessed in this paper. The use of both fuels together (33% HEU and 67% LEU) in LPRRs seems to be achievable from the neutronic point of view. High Initial Excess Reactivity (IER) can be achieved. To maintain the reactor performance in terms of neutron flux value in the internal and external irradiation sites the reactor power needs to be increased to about 32 kW. However the safety margin of the mixed core is smaller in both normal and accidental operation conditions.     

Design simplification of a small nuclear reactor for large-diameter neutron transmutation doping silicon using control rods

A design concept for a small nuclear reactor for neutron transmutation doping silicon (NTD-Si) using a Pressurized Water Reactor (PWR) full-length fuel assembly was proposed in our previous work. The excess reactivity was suppressed by a combination of Gd₂O₃ and soluble boron, which results in a flatter flux profile over the core than with control rod insertion; however, the soluble boron system for reactivity control is quite complex and expensive. The removal of this system would make the design much simpler. In the current work, the removal of soluble boron is considered. Criticality, neutron transportation and core burn-up calculations were performed using the MVP/GMVP II code and MVP-BURN code. The calculation results show that the insertion of control rods in five of the nine assemblies is enough to suppress reactivity. The thermal hydraulic analysis showed that heat removal from the core was possible under 1 atm operating pressure. Silicon ingots up to 30 cm in diameter could be irradiated with sufficient uniformity in the irradiation channels.     

Characterization of neutron field in the experimental fast reactor JOYO for fuel and structural material irradiation test

The experimental fast reactor JOYO has been operated as an irradiation test facility for fast reactor fuel and structural material since 1983 with its MK-II core. During this time, an extensive study was conducted to characterize the neutron field in order to assure the accuracy and reliability of neutron fluence. Neutron flux for a given irradiation test was calculated using a core management code system based on three-dimensional diffusion theory. It was then corrected with the adjusted neutron spectrum by means of the multiple foil activation method. The neutron fluence calculation accuracy in the fuel region was evaluated within a 5% error by comparing the burn-up of spent fuel with the measured values, which had been obtained from their post-irradiation examination. At positions away from the fuel region, the neutron flux distribution was calculated using a two-dimensional transport code. A Monte Carlo code was also used to analyze the detailed neutron flux distribution within an irradiation test subassembly that had a heterogeneous internal structure. With the neutron flux results various irradiation parameters, such as displacement per atom (dpa) and helium production, could be evaluated. A helium accumulation fluence monitor has been developed to measure not only neutron fluence but also helium production. Neutron flux and fluence obtained from the core management calculations were compiled as a database for users’ convenience together with related irradiation information and fuel subassembly material compositions. These data are expected to be widely used in the post-irradiation analysis of fuel and structural material.     

Neutronic analysis for core conversion (HEU–LEU) of Pakistan research reactor-2 (PARR-2)

Neutronic analyses for the core conversion of Pakistan research reactor-2 (PARR-2) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel has been performed. Neutronic model has been verified for 90.2% enriched HEU fuel (UAl₄–Al). For core conversion, UO₂ fuel was chosen as an appropriate fuel option because of higher uranium density. Clad has been changed from aluminum to zircalloy-4. Uranium enrichment of 12.6% has been optimized based on the design basis criterion of excess reactivity 4 mk in miniature neutron source reactor (MNSR). Lattice calculations for cross-section generation have been performed utilizing WIMS while core modeling was carried out employing three dimensions option of CITATION. Calculated neutronic parameters were compared for HEU and LEU fuels. Comparison shows that to get same thermal neutron flux at inner irradiation sites, reactor power has to be increased from 30 to 33 kW for LEU fuel. Reactivity coefficients calculations show that doppler and void coefficient values of LEU fuel are higher while moderator coefficient of HEU fuel is higher. It is concluded that from neutronic point of view LEU fuel UO₂ of 12.6% enrichment with zircalloy-4 clad is suitable to replace the existing HEU fuel provided that dimensions of fuel pin and total number of fuel pins are kept same as for HEU fuel.     

A feasibility study of LEU enrichment uranium fuels for MNSR conversion using MCNP

A neutronics feasibility study has been performed to determine the enrichment that would be required to convert a commercial Miniature Neutron Source Reactor (MNSR) from HEU (90.2%) to LEU (<20%) fuel. Two LEU cores with uranium oxide fuel pins of different dimensions were studied. The one has the same dimensions as the current HEU fuel while the other has the dimensions as the special MNSR, the In-Hospital Neutron Irradiator (INHI), which is a variant of the MNSR. The LEU cores that were studied are of identical core configuration as the current HEU core, except for potential changes in the design of the fuel pins. The following reactor core physics parameters were computed for the two LEU fuel options; clean cold core excess reactivity (ρ_ex), control rod (CR) worth, shut down margin (SDM), neutron flux distributions in the irradiation channels and kinetics data (i.e. effective delayed neutron fraction, β_eff and prompt neutron lifetime, l_f). Results obtained are compared with current HEU core and indicate that it would be feasible to use any of the LEU options for the conversion of NIRR-1 in particular from HEU to LEU.     

Preliminary scoping study of some neutronic aspects of new shim safety rods for a typical 5 MW research reactor by Monte Carlo simulation

A Monte Carlo simulation of a typical 5 MW research reactor (TRR) was carried out using MCNP4C code. The geometry of the reactor core was modeled including the details of all fuel elements, control rods, all irradiation channels, graphite reflectors, reactor pool and thermal column. The model predicted neutron flux distributions within the core, control rod (CR) worth, core reactivity (ρ), shutdown margin, and some kinetic parameters when the control rod insert or withdraw. This study was carried out to reduce blockage probability of shim safety rod (SSR)s of the TRR. Two introduced more blackness SSRs were chosen and made thinner in a way adequate blackness, in comparison to the present rods, achieved.     

Monte Carlo simulation of Feynman-α and Rossi-α techniques for calculation of kinetic parameters of Tehran Research Reactor

Noise analysis techniques including Feynman-α (variance-to-mean) and Rossi-α (correlation) have been simulated by MCNP computer code to calculate the prompt neutron decay constant (α₀), effective delayed neutron fraction (β_eff) and neutron generation time (Λ) in a subcritical condition for the first operating core configuration of Tehran Research Reactor (TRR). The reactor core is considered to be in zero power (reactor power is less than 1 W) in the entire simulation process. The effect of some key parameters such as detector efficiency, detector position and its dead time on the results of simulation has been discussed as well. The results of proposed method in the current study are validated against both the experimental data and the results of MTR_PC computer code.     

Power up-grading study for the first Egyptian research reactor

In the present work, power up-grading study is performed, for the first Egyptian Research Reactor (ET-RR-1), using the present fuel basket with 4×4 fuel rods, (17.5 mm pitch), and a proposed fuel basket with 5×5 fuel rods, (14.0 mm pitch), without violating the thermal hydraulic safety criteria. These safety criteria are; fuel centerline temperature (fuel melting), clad surface temperature (surface boiling), outlet coolant temperature, and maximum heat flux (critical heat flux ratio). Different thermal reactor powers (2–10 MW) and different core coolant flow rates (450, 900, 1350 m³ h⁻¹) are considered. The thermal hydraulic analysis was performed using the subchannel code COBRA-IIIC for the estimation of temperatures, coolant velocities and critical heat flux. The neutronic calculations were performed using WIMS-D4 code with 5 — group neutron cross section library. These cross sections were adapted to use in the two-dimensional (2-D) diffusion code DIXY for core calculations. The study concluded that ET-RR-1 power can be upgraded safely up to 4 MW with the present 4×4-fuel basket and with the proposed 5×5-fuel basket up to 5 MW with the present coolant flow rate (900 m³ h⁻¹). With the two fuel arrays, the reactor power can be upgraded to 6 MW with coolant flow rate of 1350 m³ h⁻¹ without violating the safety criterion. It is also concluded that, loading the ET-RR-1 core with the proposed fuel basket (5×5) increases the excess reactivity of the reactor core than the present 4×4 fuel matrix with equal U-235 mass load and gave better fuel economy of fuel utilization.     

Monte Carlo calculation of the core of the heavy water zero power reactor (HWZPR) using MCNP4c

The MCNP4c code, based on the probabilistic approach, was used to simulate 3D configuration of the core of the heavy water zero power reactor (HWZPR). In present work, first, all of the constituents of the core such as fuel pellets, fuel element, moderator (D₂O) and annular graphite reflector were modeled using MCNP4c code. Then calculations of axial and radial neutron fluxes were performed in three energy groups such as thermal (0-0.625 eV), epithermal (0.625-550 eV), and fast (0.550-20 MeV). The cadmium ratio was calculated as well and the neutron flux parameters such as extrapolated height (H_e), extrapolated radius (R_e) and physical center of the core (z₀) were computed using cadmium ratio. Comparison of the neutron flux parameters with the experimental data showed that the MCNP4c model of the HWZPR was validated.     

Main results of the first phase of the upgrade of the SM core

The first phase of the work on checking the main assumptions of the concept for upgrading the core of the SM reactor has been completed. A full-scale reactor experiment has been performed for the purpose of creating in the reactor the conditions necessary for accelerated high-dose irradiation of materials, meeting the requirements for the fast-neutron flux density, water temperature, pressure, and composition, and making it possible to install large-diameter experimental channels and apparatus for regulating the temperature and neutron spectrum. The decrease of the fuel volume and excess reactivity are compensated by a 20% increase of the uranium content in a fuel element and by replacing the corrosion-resistant steel fuel-assembly jackets with zirconium-alloy jackets. The results of the calculations and experiments performed during the first phase have shown that the objective has been achieved-the reactor can be operated efficiently with the new arrangement of the core. The objective and problems of the second phase are formulated: increase of the neutron flux density in the experimental channels by a factor of 1.5 by increasing the power density in the core, using a fuel element with a low harmful absorption of neutrons, and equalizing the power-release distribution by using a consumable absorber. __________ Translated from Atomnaya énergiya, Vol. 102, No. 2, pp. 86–92, February, 2007.     

Development of Hydride Absorber for Fast Reactor—Evaluation of Design Method for Hydride Absorber Rod—

A hydride control rod is being developed to improve the economy of fast reactor plants because it has a longer lifetime than the currently used B₄C control rod. A hydride burnable poison rod is also under development to reduce the number of control rods by decreasing core excess reactivity. Hydrogen in the hydride control rod causes neutron spectrum interference between the fuel and control rod regions. Thus, the study on core design was performed with the continuous-energy Monte Carlo code MVP using the nuclear data library JENDL-3.3 to deal with this phenomenon precisely. To evaluate the applicability of MVP to hydride absorber rod design, two benchmark calculations were carried out. One of them is a hydrogen-contained metal fuel fast core constructed in Fast Critical Assembly (FCA) and the other is the Nuclear Safety Research Reactor (NSRR) core where zirconium-hydride fuel (U-ZrH_1.6) rods are loaded. These benchmark calculations and the design study on a fast reactor core with hafnium-hydride control rods have revealed that MVP is a reliable tool for hydride absorber rod design.     

The effect of silicon on the interaction between metallic uranium and aluminum: A 50 year long diffusion experiment

The core of the BR1 research reactor at SCK•CEN, Mol (Belgium) has a graphite matrix loaded with fuel rods consisting of a natural uranium slug in aluminum cladding. The BR1 reactor has been in operation since 1956 and still contains its original fuel rods. After more than 50 years irradiation at low temperature, some of the fuel rods have been examined. Fabrication reports indicate that a so-called AlSi bonding layer and an U(Al,Si)₃ anti-diffusion layer on the natural uranium fuel slug were applied to limit the interaction between the uranium fuel and aluminum cladding. The microstructure of the fuel, bonding and anti-diffusion layer and cladding were analysed using optical microscopy, scanning electron microscopy and electron microprobe analysis. It was found that the AlSi bonding layer does provide a tight bond between fuel and cladding but that it is a thin USi layer that acts as effective anti-diffusion layer and not the intended U(Al,Si)₃ layer.     

Finite element spherical harmonics (PN) solutions of the three-dimensional Takeda benchmark problems

A set of multi-group eigenvalue (K_eff) benchmark problems in three-dimensional homogenised reactor core configurations have been solved using the deterministic finite element transport theory code EVENT and the Monte Carlo code MCNP4C. The principal aim of this work is to qualify numerical methods and algorithms implemented in EVENT. The benchmark problems were compiled and published by the Nuclear Data Agency (OECD/NEACRP) and represent three-dimensional realistic reactor cores which provide a framework in which computer codes employing different numerical methods can be tested. This is an important step that ought to be taken (in our view) before any code system can be confidently applied to sensitive problems in nuclear criticality and reactor core calculations. This paper presents EVENT diffusion theory (P₁) approximation to the neutron transport equation and spherical harmonics transport theory solutions (P₃–P₉) to three benchmark problems with comparison against the widely used and accepted Monte Carlo code MCNP4C. In most cases, discrete ordinates transport theory (S_N) solutions which are already available and published have also been presented. The effective multiplication factors (K_eff) obtained from transport theory EVENT calculations using an adequate spatial mesh and spherical harmonics approximation to represent the angular flux for all benchmark problems have been estimated within 0.1% (100 pcm) of the MCNP4C predictions. All EVENT predictions were within the three standard deviation uncertainty of the MCNP4C predictions. Regionwise and pointwise multi-group neutron scalar fluxes have also been calculated using the EVENT code and compared against MCNP4C predictions with satisfactory agreements. As a result of this study, it is shown that multi-group reactor core/criticality problems can be accurately solved using the three-dimensional deterministic finite element spherical harmonics code EVENT.     

Validation of the Monte Carlo model supporting core conversion of the Portuguese Research Reactor (RPI) for neutron fluence rate determinations

The Monte Carlo code MCNP was used to calculate absolute values of thermal, epithermal and fast neutron fluence rates in the new RPI core using fresh LEU fuel. Discrepancies smaller than 20% were obtained between calculated results and activation foil measurements. A previous knowledge of general characteristics of the neutron energy spectra, provided by the MCNP reactor model itself, has been fundamental to determine the conditions yielding a proper comparison of simulated and measured results. An excellent agreement (6%) was also obtained for the relative neutron fluence rate profiles along the fuel height. The MCNP model of the reactor core was therefore validated for a tri-dimensional determination of neutron fluence rates in the fuel assemblies and neighbouring irradiation positions.     

Validation of a new continuous Monte Carlo burnup code using a Mox fuel assembly

The reactivity of nuclear fuel decreases with irradiation (or burnup) due to the transformation of heavy nuclides and the formation of fission products. Burnup credit studies aim at accounting for fuel irradiation in criticality studies of the nuclear fuel cycle (transport, storage, etc…). The principal objective of this study is to evaluate the potential capabilities of a newly developed burnup code called “BUCAL1”. BUCAL1 differs in comparison with other burnup codes as it does not use the calculated neutron flux as input to other computer codes to generate the nuclide inventory for the next time step. Instead, BUCAL1 directly uses the neutron reaction tally information generated by MCNP for each nuclide of interest to determine the new nuclides inventory. This allows the full capabilities of MCNP to be incorporated into the calculation and a more accurate and robust analysis to be performed. Validation of BUCAL1 was processed by code-to-code comparisons using predictions of several codes from the NEA/OCED. Infinite multiplication factors (k_∞) and important fission product and actinide concentrations were compared for a MOX core benchmark exercise. Results of calculations are analysed and discussed.     

PECITIS-II, a computer program to predict the performance of collapsible clad UO2 fuel elements

The codes devised and used in India for the design of fuel for their Pressurized Heavy Water Reactor (PHWR) programme are described. The scheme includes the use of collapsible fuel cladding for improved neutron economy.This code is made with reference to collapsible clad UO₂ fuel elements. This evaluates sheath strain and fission gas pressure. The fuel expansion is calculated by a two zone model which assumes that above a certain temperature the UO₂ deforms plastically and below that temperature it cracks radially and behaves as an elastic solid; the plastic core is under compression. The pellet clad gap conductance is calculated by using a modified Ross and Stoute model considering the effects of fuel and clad thermal expansion, fission gas release, dilution of filler gas and irradiation swelling. Stress relaxation of the sheath and its effect on fuel sheath contact pressure is also considered for arriving at the end result.     

Monte Carlo simulation of the Greek Research Reactor neutron irradiation facilities

A Monte Carlo simulation of the Greek Research Reactor was carried out using MCNP-4C2 code and continuous energy cross-section data from ENDF/B-VI library. A detailed model of the reactor core was employed including standard and control fuel assemblies, reflectors and irradiation devices. The model predicted neutron flux distributions within the core in good agreement with calculations performed using the deterministic code CITATION and measurements using activation foils. The model is used for the prediction of the neutron field characteristics at the reactor irradiation devices and enables the design and evaluation of experiments involving material irradiations.     

基于R2S方法的反应堆结构材料活化剂量计算研究

反应堆结构材料在堆芯中子辐照下由于中子活化反应而产生大量的放射性核素,其衰变光子是反应堆停堆检修、换料、退役过程中工作人员职业照射剂量的重要来源。本文基于严格两步法（R2S）,研究了反应堆结构材料栅元活化计算方法,并基于蒙卡粒子输运程序（MCNP）与点活化计算程序（ORIGEN）建立了反应堆结构材料活化剂量计算软件（MOCA）。通过开发功能接口与数据接口程序实现输运程序与活化计算程序的自动耦合,进而实现“中子输运-活化分析-剂量计算”全自动耦合分析。利用M5包壳活化计算模型、不锈钢活化计算模型和NUREG/CR-6115压水堆模型对MOCA进行基准验证,证明了MOCA的正确性与可靠性。