Volume 81, Issue 1

The Korea Atomic Energy Research Institute (KAERI) has been developing the 150 MWe Prototype Generation-IV Sodium-cooled Fast Reactor (PGSFR). The design concept is highly based on passive safety mechanisms, minimizing the need for engineered safety systems. Presently, it is of primary importance to assure the reactor dynamics and stability against small reactivity disturbances under power operating conditions. KAERI has therefore developed the NuSTAB code for stability analysis of the PGSFR. In NuSTAB, the neutron-kinetic and thermal-hydraulic coupling equations are linearized to form the characteristic equation, which is solved as a generalized eigenvalue problem for determining the decay ratio, an indicator of the system stability. In this paper, the stability of the PGSFR was analyzed by applying the point kinetic and spatial kinetic options in the NuSTAB code. System responses to temperature feedbacks including the Doppler effect, thermal expansion, coolant density change, and overall feedback were studied. The results indicate that the initial U and final TRU cores of the PGSFR are both inherently stable thanks to the temperature feedbacks.

In an attempt to understand the built-up of natural circulation in MTR pool type upward flow research reactors after loss of power, an experimental test rig was built to simulate the loop of natural circulation in MTR reactors. The test rig consisting of two vertically oriented branches, in one of them the core is simulated by two rectangular, electrically heated, parallel channels. The other branch simulates the part of the return pipe that participates in the development of core natural circulation. In the first phase of the work, many experimental runs at different conditions of channel's power and branch's initial temperatures are performed. The channel's coolant and surface temperatures were measured. The measurements and their interpretation were published by the first three authors. In the present work the thermal hydraulic behavior of the test rig is complemented by theoretical analysis using RELAP5 Mod 3.3 system code. The analysis consisting of two parts; in the first part RELAP5 model is validated against the measured values and in the second part some of the other not measured hydraulic parameters are predicted and analyzed. The test rig is typically nodalized and an input dick is prepared. In spite of the low pressure of the test rig, the results show that RELAP5 qualitatively predicts the thermal hydraulic behaviour and the accompanied phenomenon of flow inversion of such facilities. Quantitatively, there is a difference between the predicted and measured values especially the channel's surface temperature. This difference may be return to the uncertainties in initial conditions of experimental runs, the position of the thermocouples which buried inside the heat structure, and the heat transfer package in RELAP5.

Developing a reliable thermal-hydraulic model of a nuclear reactor is an essential process in the steady state and transient analyses. This paper provides the results of best estimate calculation carried out with reference to Iranian Miniature Neutron Source Reactor (MNSR) using the RELAP5 code. Applying the qualified nodalization and the cross-flow effects are some of the advantages in the present model. Here, various transients including step and ramp reactivity insertions were inspected for safety analysis. The obtained results from the code showed a reasonable agreement with the MNSR Safety Analysis Report (SAR) and existing experimental and reference data.

Main purpose of the study is justifying doping of protactinium-231 into fuel compositions of advanced nuclear reactors with the ultimate aim to improve their operation safety and economic efficiency. Protactinium-231 could be generated in thorium blankets of hybrid thermonuclear facilities. The following results were obtained: 1. Protactinium-231 has some favorable features for its doping into nuclear fuel; 2. Protactinium containing fuel compositions can be characterized by the higher values of fuel burn-up, the longer values of fuel lifetime and the better proliferation resistance; 3. as protactinium-231 is the stronger neutron absorber than uranium-238, remarkably lower amounts of protactinium-231 may be doped into fuel compositions. The free space could be occupied by materials which are able to improve heat conductivity and refractoriness of fuel. As a consequence, operation safety of nuclear reactors could be upgraded.

The aim of the present paper is to assess the calculations of pin-by-pin group integrated fission rates within MOX/UO 2 Fuel assemblies using the Monte Carlo code MCNPX2.7c with two sets of the available latest nuclear data libraries used for calculating MOX-fueled systems. The data that are used in this paper are based on the benchmark by the NEA Nuclear Science Committee (NSC). The k ∞ and absorption/fission reaction rates per isotope, k eff and pin-by-pin group integrated fission rates on 1/8 fraction of the geometry are determined. To assess the overall pin-by-pin fission rate distribution, the collective per cent error measures were investigated. The results of AVG, MRE and RMS error measures were less than 1 % error. The present results are compared with other participants using other Monte Carlo codes and with CEA results that were taken in the benchmark as reference. The results with ENDF/B-VI.6 are close to the results received by MVP (JENDL3.2) and SCALE 4.2 (JEF2.2). The results with ENDF/BVII.1 give higher values of k ∞ reflecting the changes in the newer evaluations. In almost all results presented here, the MCNP calculated results with ENDF/B VII.1 should be considered more than those obtained by using other Monte Carlo codes and nuclear data libraries. The present calculations may be consider a reference for evaluating the numerical schemes in production code systems, as well as the global performance including cross-section data reduction methods as the calculations used continuous energy and no geometrical approximations.

Numerical investigation of the rewetting of single sector fuel assembly of Advanced Heavy Water Reactor (AHWR) has been carried out to exhibit the effect of coolant jet diameters (2, 3 and 4 mm) and jet directions (Model: M, X and X2). The rewetting phenomena with various jet models are compared on the basis of rewetting temperature and wetting delay. Temperature-time curve have been evaluated from rods surfaces at different circumference, radial and axial locations of rod bundle. The cooling curve indicated the presence of vapor in respected location, where it prevents the contact between the firm and fluid phases. The peak wall temperature represents as rewetting temperature. The time period observed between initial to rewetting temperature point is wetting delay. It was noted that as improved in various jet models, rewetting temperature and wetting delay reduced, which referred the coolant stipulation in the rod bundle dominant vapor formation.

An experimental study has been undertaken concerning natural convection heat transfer of nanofluids over a cylindrical heater with a constant wall heat flux condition. The cylindrical heater having 7.1 mm O. D. and 0.5 mm thickness with 100 mm heated length was used as heater surface. The heat flux was varied from 0 to 50 000 W/m 2 and the corresponding Rayleigh number range is varied from 3 × 10 4 to 1.65 × 10 5 . Different Nanofluids were used i. e. Aluminium oxide, Titanium dioxide, Silicon dioxide and Copper oxide with concentration varying from 0.0005 % to 0.05 % by weight. Results show that there was a reduction in natural convection heat transfer coefficient of nanofluids as compared to water. Experimental results were compared with existing models for similar geometry. However, the available correlation was found to be unable to predict experimental data as it does not take into account the effect of particle concentration. A new empirical model was developed based on the experimental data including the effect of nanoparticles concentration which predicts the experimental data satisfactorily.

Natural circulation systems find extensive applications in industrial engineering systems. One of the applications is in nuclear reactor where the decay heat is removed by natural circulation of the fluid under off-normal conditions. The upcoming reactor designs make use of natural circulation in order to remove the heat from core under normal operating conditions also. These reactors employ multiple vertical fuel channels with provision of on-power refueling/defueling. Natural circulation systems are relatively simple, safe and reliable when compared to forced circulation systems. However, natural circulation systems are prone to encounter flow instabilities which are highly undesirable for various reasons. Presence of parallel channels under natural circulation makes the system more complicated. To examine the behavior of parallel channel system, studies were carried out for single-phase natural circulation flow in a multiple vertical channel system. The objective of the present work is to study the flow behavior of the parallel heated channel system under natural circulation for different operating conditions. Steady state and transient studies have been carried out in a parallel channel natural circulation system with three heated channels. The paper brings out the details of the system considered, different cases analyzed and preliminary results of studies carried out on a single-phase parallel channel system.

In this paper, the PDS-XADS LBE-cooled core open-loop transfer function was calculated by considering the source importance in point-kinetic equations. For this purpose, the overall-feedback transfer function was calculated considering the lumped-model for 14-steps of subcritical levels. Following effects were considered in three steps: 1. Doppler broadening, fuel expansion, coolant density and structure expansion, 2. Delayed-reactivity and void-worth inserted to prior step, 3. Severe-accident condition, inserted to prior steps. The linear stability analysis was modeled by using the Bode diagrams, Nyquist stability criterion and Nichols chart in MATLAB for each subcritical level and six groups of delayed neutrons. For optimized subcritical level determination, a conservative severe accident was considered. According to calculation results and analysis, the PDS-XADS core is stable and in optimized subcritical level, has the higher safety margin. The results are in good agreement with SIMMER-III code and main neutronic results. The optimized subcritical level by using the lumped-model is 0.97687.

In this paper, the generalisation of the 4th-order Adams-Bashforth-Moulton predictor-corrector method is proposed to numerically solve the point kinetic equations of the nuclear reactivity calculations without using the nuclear power history. Due to the nature of the point kinetic equations, different predictor modifiers are used in order improve the precision of the approximations obtained. The results obtained with the prediction formulas and generalised corrections improve the precision when compared with previous methods and are valid for various forms of nuclear power and different time steps.