NEUTRONIC MODELING STRATEGIES FOR A LIQUID FUEL TRANSIENT CALCULATION

Framework • A detailed and highly flexible numerical tool to study criticality accidents has been developed • The tool implements a Multi-Physics coupling using neutronics, thermal-hydraulics and thermal-mechanics models based on Open FOAM and SERPENT codes • Two neutronics models: Quasi-Static Monte Carlo and SPN Objective: In this work a system composed by a 2D square liquid fuel cavity filled with a fuel molten salt has been used to: • Investigate the performance of the tool’s thermal-hydraulics and neutronics solvers coupling numerical scheme • Evaluate possible strategies for the implementation of the Quasi-Static (QS) method with the Monte Carlo (MC) neutronics code • Compare the QS-MC approach precision and computational cost against the Simplified P3 (SP3) method

A Monte Carlo Method for Simulating Stochastic Neutron Fields in Criticality Transients

In presence of a weak neutron source, the initial growth of neutron population in a supercritical system exhibits a significant stochastic feature, both initiation and burst waiting times are uncertain. As a result, the energy released during criticality excursions is stochastic, obeying a probability distribution. When criticality accidents and pulsed reactor experiments are analyzed, it is important to estimate this kind of stochastic feature, including assessing the initiation probability and then the fission energy probability distribution. Thus a Monte Carlo direct simulation method has been proposed and the corresponding code MES has been developed. By taking random factors during criticality excursions into account in dynamic Monte Carlo simulations, this method is capable of simulating the whole process from source injection to exponential growth of the neutron population, and finally to extinction of the neutron pulse. A set of static initiation probability problems and a figurative criticality excursion problem have been applied to validate this method and MES. Results demonstrate that with the proposed method MES is able to simulate stochastic transient neutron fields in multiplying systems during criticality excursions.

ANALISIS KESELAMATAN TERMOHIDROLIK BULK SHIELDING REAKTOR KARTINI

ABSTRAK ANALISIS KESELAMATAN TERMOHIDROLIK BULK SHIELDING REAKTOR KARTINI. Bulk shielding merupakan fasilitas yang terintegrasi dengan reaktor Kartini yang berfungsi sebagai penyimpanan sementara bahan bakar bekas. Fasilitas ini merupakan fasilitas yang termasuk dalam struktur, sistem dan komponen (SSK) yang penting bagi keselamatan. Salah satu fungsi keselamatan dari sistem penanganan dan penyimpanan bahan bakar adalah mencegah kecelakaan kekritisan yang tak terkendali dan membatasi naiknya temperatur bahan bakar. Analisis keselamatan paling kurang harus mencakup analisis keselamatan dari sisi neutronik dan termo hidrolik Bulk shielding. Analisis termo hidrolik ditujukan untuk memastikan perpindahan panas dan proses pendinginan bahan bakar bekas berjalan baik dan tidak terjadi akumulasi panas yang mengancam integritas bahan bakar. Code tervalidasi PARET/ANL digunakan untuk analisis pendinginan dengan mode konveksi alam. Hasil perhitungan menunjukkan bahwa mode pendinginan konvekasi alam cukup memadai dalam mendinginkan panas sisa tanpa mengakibatkan kenaikan temperatur bahan bakar yang signifikan. Kata kunci: Bulk shielding, bahan bakar bekas, konveksi alam, PARET. ABSTRACT THERMAL HYDRAULIC SAFETY ANALYSIS OF BULK SHIELDING KARTINI REACTOR. Bulk shielding is an integrated facility to Kartini reactor which is used for temporary spent fuels storage. The facility is one of the structures, systems and components (SSCs) important to safety. Among the safety functions of fuel handling and storage are to prevent any uncontrolable criticality accidents and to limit the fuel temperature increase. Safety analyses should, at least, cover neutronic and thermal hydraulic calculations of the bulk shielding. Thermal hydraulic analyses were intended to ensure that heat removal and the process of the spent fuels cooling takes place adequately and no heat accumulation that challenges the fuel integrity. Validated code, PARET/ANL was used for analysing the spent fuels cooling with natural convection mode. The calculations results concluded that natural convection cooling mode can adequately cools down the decay heat without significant increase in fuel temperatur. Keywords: bulk shielding, spent fuels, natural convection, PARET.