P. Selvaraj

Kinematic Analysis of Carrier Mechanism for In-service Inspection of Sodium-Cooled Fast Reactor Internals

In-service inspection requirements of reactor internals of sodium-cooled fast reactors offer challenges in the design and successful deployment of carrier mechanisms. A single-carrier mechanism consisting of inter-connected rigid links having multiple ultrasonic transducers is envisaged for carrying out all the necessary inspections. The kinematic analysis of the chain assembly of the proposed carrier mechanism is the main focus of the present paper. Initially, a base design of the chain assembly was proposed and the same is analysed with respect to the linearity and smooth motion transmission. The kinematic chain of the base design was modified to improve the motion transmission characteristics. Parametric studies were done to fine-tune the design parameters of modified design. The analysis has given valuable insights into the performance of chain assembly and provided necessary data for the smooth operation of chain assembly.

Sodium-Free Level Fluctuations: Concern on Structural Integrity for the Pool-Type Sodium-Cooled Fast Breeder Reactor

In the case of sodium-cooled pool-type fast breeder reactor, the weight of the entire reactor assembly along with sodium coolant is carried by a large-sized cylindrical vessel called the main vessel. Since main vessel is one of the primary boundaries from the radiation shielding point of view, the structural integrity of the main vessel is an essential safety feature to be ensured under all operating conditions. The inner surface of the main vessel is partly filled with high-temperature sodium coolant, and the remaining part is in contact with relatively low-temperature argon cover gas. Due to large surface area of the main vessel pool, the free surface level of the sodium oscillates. This environment creates drastic temperature cycling in the main vessel wall at the sodium-free level interface. This transient temperature cycling imposed on the main vessel with negligible attenuation can lead to high-cycle thermal fatigue damage in the vicinity of sodium-free level. This high-cycle thermal fatigue can be detrimental in ensuring the structural integrity of the main vessel considering the number of cycles (approximately 9.5 × 108 cycles) applied in the plant’s lifetime. This paper presents the numerical studies carried out towards assessing the structural integrity of the main vessel by considering the effect of sodium-free level fluctuations. The magnitudes of level fluctuations of 100 and 30 mm with a frequency range of 0.1–10 Hz with different increment of frequencies are considered as a parametric study. In this study, the critical portion due to level fluctuation has been identified from the thermal stress cycling point of view, and the effect of level fluctuation frequency on the structural integrity of the main vessel has been quantified.

Thermal Hydraulic Analysis of Severe Accident in PFBR

During a core disruptive accident (CDA), the amount of primary sodium that can be released to Reactor Containment Building (RCB) in Prototype Fast Breeder Reactor (PFBR) is estimated to be 350 kg/s, by a transient fluid dynamic calculation. The pressure and temperature evolutions inside RCB, due to consequent sodium fire have been estimated by a constant burning rate model, accounting for heat absorption by RCB wall, assuming RCB isolation based on area gamma monitors. The maximum pressure developed is 7000 Pa. In case RCB isolation is delayed, then the final pressure inside RCB reduces below atmospheric pressure due to cooling of RCB air. The negative pressure that can be developed is estimated by dynamic thermal hydraulic modeling of RCB air / wall to be −3500 Pa. These investigations were useful to arrive at the RCB design pressure. Following CDA, RCB is isolated for 40 days. During this period, the heat added to RCB is dissipated to atmosphere only by natural convection. Considering all the possible routes of heat addition to RCB, evolution of RCB wall temperature has been predicted using HEATING5 code. It is established that the maximum temperature in RCB wall is less than the permissible value.© 2010 ASME