A pinboard by
Ishita Trivedi

PhD Student, Nuclear Engineering, North Carolina State University


Nuclear power plants produce a lot of heat which is cooled using a coolant such as water and is then used to generate electricity. Majority of commercially operating reactors around the world are cooled by water which is the older generation (Generation II/III) of Nuclear Reactors. However, my research involves development of newest generation of nuclear reactors (Generation IV) cooled by liquid metal (Lead) instead of water cooled. Generation III water cooled reactors have an extensive plant design and are constricted to being built near a water body. However, these issues can be successfully tackled by using liquid lead as coolant. Liquid lead as reactor coolant has several advantages over water such as the compact reactor design, reduced capital cost, more economical and reliable. Lead-cooled reactors also supersede the current water-cooled reactors in safety aspect as lead has a much higher melting a boiling point, thus can operate at high temperatures without the risk of boiling which is considered a safety issue in nuclear reactors. In addition, construction site has no restriction based on the coolant. It can be constructed in the dry Sahara desert or the frozen Gobi dessert. Development of Lead-Cooled Fast Reactors is a unique and innovative step in the field of Nuclear Engineering. Westinghouse Electric Co. is developing an advanced lead-cooled fast reactor with the potential to achieve best in class safety and economics performances. Commercialization of the technology will require demonstration, as a preliminary step, in an appropriately scaled demonstration reactor that will ensure best scalability and flexibility towards the commercial reactor. My research includes performing a sensitivity study on the reactor size, fuel cost and cycle length. This is will help provide the base platform form towards development of a Lead-cooled fast reactor.


Validation of the DYN3D-Serpent code system for SFR cores using selected BFS experiments. Part I: Serpent calculations

Abstract: Publication date: April 2017 Source:Annals of Nuclear Energy, Volume 102 Author(s): Reuven Rachamin, Sören Kliem A comparative study has been performed to evaluate the prediction capability of the DYN3D-Serpent code system for sodium fast reactor (SFR) cores. In this study, the calculation system was tested against the BFS-73-1 and BFS-62-3A experiments conducted at the Russian Institute of Physics and Power Engineering (IPPE). These experiments were designed for full-scale modeling of SFR cores, and for validation of codes and nuclear data. The study was performed in two parts. The first part is aimed at developing and validating a 3D full-core heterogeneous model of each of the experiments using the Serpent Monte-Carlo code (MC) code. This part meant as a first step towards the use of the Serpent MC code as a tool for preparation of homogenized group constants, and as a reference solution for code-to-code comparison with the DYN3D code. The second part is devoted to the homogenized group constants generation procedure and the DYN3D steady-state calculations. This paper covers the first part of the study. The experiments were simulated using the Serpent MC code, and the basic neutronic characteristics were evaluated and compared against experimental values. The calculated results agreed well with the measured values on most of the neutronic characteristics. It suggests that the Serpent MC code can be used for the preparation of homogenized group constants and as a reference solution for code-to-code verification with the DYN3D code.

Pub.: 05 Jan '17, Pinned: 22 Aug '17

Thermal hydraulic parametric investigation of decay heat removal from degraded core of a sodium cooled fast Breeder reactor

Abstract: Publication date: March 2017 Source:Nuclear Engineering and Design, Volume 313 Author(s): Lokesh Verma, Anil Kumar Sharma, K. Velusamy Ensuring post accident decay heat removal with high degree of reliability following a Core Disruptive Accident (CDA) is very important in the design of sodium cooled fast reactors (SFR). In the recent past, a lot of research has been done towards the design of an in-vessel core catcher below the grid plate to prevent the core debris reaching the main vessel in a pool type SFR. However, during an energetic CDA, the entire core debris is unlikely to reach the core catcher. A significant part of the debris is likely to settle in core periphery between radial shielding subassemblies and the inner vessel. Failure of inner vessel due to the decay heat can lead to core debris reaching the main vessel and threatening its integrity. On the other hand, retention of a part of debris in core periphery can reduce the load on main core catcher. Towards achieving an optimum design of SFR and safety evaluation, it is essential to quantify the amount of heat generating core debris that can be retained safely within the primary vessel. This has been performed by a mathematical simulation comprising solution of 2-D transient form of the governing equations of turbulent sodium flow and heat transfer with Boussinesq approximations. The conjugate conduction-convection model adopted for this purpose is validated against in-house experimental data. Transient evolutions of natural convection in the pools and structural temperatures in critical components have been predicted. It is found that 50% of the core debris can be safely accommodated in the gap between radial shielding subassemblies and inner vessel without exceeding structural temperature limit. It is also established that a single plate core catcher can safely accommodate decay heat arising due to ∼70% of the core debris by establishing natural circulation in the lower sodium pool. The influence of heat removal rate by natural circulation on availability of the number of decay heat exchangers (DHX) dipped in the upper pool is also analyzed. It is seen that the temperatures in the inner vessel, source plate and the maximum debris temperature do not increase significantly even when the DHXs are deployed 5h after the accident, demonstrating the benefit of large thermal inertia of the pool.

Pub.: 12 Jan '17, Pinned: 22 Aug '17

Radiation damage in structural materials for fast reactors

Abstract: Radiation damage in structural materials for fast reactors is caused by the action of many different mechanisms, depending on the irradiation conditions, the composition and state of the material, and external factors. This damage affects considerably the physicomechanical and operating characteristics of the material and, thus, the economy of fast reactors. Experimental data and theoretical models of radiation damage make it possible to predict the basic factors which limit the efficiency of the structural material for high burnup values and also single out the basic problems in improving the core elements and materials. This applies in the first place to problems of accommodation of structural material swelling and problems of securing sufficiently high mechanical characteristics for large fluence values. Experimental data on swelling and long-term mechanical characteristics have a rather large scatter. It is necessary to understand the causes of this scatter in order to eliminate indeterminacies in design calculations and determine the conditions ensuring the greatest resistance of materials to radiation damage.The lower values of swelling and plasticity loss encourage optimism and show the potentialities of the materials already in use. However, it must be considered that most reactor data pertain to doses of 50–70 d/α, while it is necessary to know the behavior of the material characteristics for doses of up to 100–120 d/α. Therefore, in substantiating the choice of materials for fuel-element jackets and assembly casings for projected high-power reactors, it is necessary to obtain the characteristics of materials with different compositions for such doses under conditions close to the actual operating conditions.

Pub.: 01 Jul '77, Pinned: 30 Jun '17

Modeling astatine production in liquid lead-bismuth spallation targets

Abstract: Astatine isotopes can be produced in liquid lead-bismuth eutectic targets through proton-induced double charge exchange reactions on bismuth or in secondary helium-induced interactions. Models implemented into the most common high-energy transport codes generally have difficulties to correctly estimate their production yields as was shown recently by the ISOLDE Collaboration, which measured release rates from a lead-bismuth target irradiated by 1.4 and 1 GeV protons. In this paper, we first study the capability of the new version of the Liège intranuclear cascade model, INCL4.6, coupled to the deexcitation code ABLA07 to predict the different elementary reactions involved in the production of such isotopes through a detailed comparison of the model with the available experimental data from the literature. Although a few remaining deficiencies are identified, very satisfactory results are found, thanks in particular to improvements brought recently on the treatment of low-energy helium-induced reactions. The implementation of the models into MCNPX allows identifying the respective contributions of the different possible reaction channels in the ISOLDE case. Finally, the full simulation of the ISOLDE experiment is performed, taking into account the likely rather long diffusion time from the target, and compared with the measured diffusion rates for the different astatine isotopes, at the two studied energies, 1.4 and 1 GeV. The shape of the isotopic distribution is perfectly reproduced as well as the absolute release rates, assuming in the calculation a diffusion time between 5 and 10hours. This work finally shows that our model, thanks to the attention paid to the emission of high-energy clusters and to low-energy cluster induced reactions, can be safely used within MCNPX to predict isotopes with a charge larger than that of the target by two units in spallation targets, and, probably, more generally to isotopes created in secondary reactions induced by composite particles.

Pub.: 04 Mar '13, Pinned: 30 Jun '17