International Journal of Nuclear Energy

Volume 2014, Article ID 167426, 8 pages

http://dx.doi.org/10.1155/2014/167426

## Modeling of SPERT IV Reactivity Initiated Transient Tests in EUREKA-2/RR Code

Institute of Nuclear Science & Technology (INST), Atomic Energy Research Establishment (AERE), Ganakbari, Savar, P.O. Box 3787, Dhaka 1000, Bangladesh

Received 28 August 2014; Revised 18 November 2014; Accepted 18 November 2014; Published 9 December 2014

Academic Editor: Arkady Serikov

Copyright © 2014 N. H. Badrun et al. This is an open access article distributed under the Creative Commons Attribution License, which permits unrestricted use, distribution, and reproduction in any medium, provided the original work is properly cited.

#### Abstract

EUREKA-2/RR code has been used for SPERT IV reactor benchmark calculations against the experimental results provided by IAEA (International Atomic Energy Agency) obtained for a series of transient tests initiated by step insertion of different magnitudes of positive reactivity with varying degrees of different controlled parameters such as reactor initial power, coolant temperature and coolant flow condition. 20 out of 39 tests that fall under forced convection mode have been considered for the present simulation provided the reactor scram system is disabled. Peak power and peak clad temperature due to transient have been calculated and it was found that although peak clad temperature values agreed, the peak power values seem to underestimate the experimental values. Further study appears to be needed to identify the limitations in modeling or examining the effect of input parameters during modeling to obtain the better simulation results.

#### 1. Introduction

Since SPERT program initiated in 1954, a variety of configurations under a broad range of conditions have been taken into consideration to study transient behavior of nuclear reactors. With the growing need to validate the experimental results, numerical code has got much attention and PARET is one such computer code originally developed for the analysis of SPERT-III experiments in 1969 [1]. Later, this code has been modified incorporating many issues such as selection of flow instability, departure from nucleate boiling, single- and two-phase heat transfer correlations, and a properties library considered more applicable to the low pressures, temperature, and flow rates that enhances the acceptability of the code for its subsequent use in the analysis of transient behavior of research reactor. However, in modeling the research reactor transient condition, the analyst has to consider some conservative assumptions due to lack of experimental data which may lead to false prediction of the real physical phenomenon that takes place in the reactor. One common point that came out from reactor analysis is the importance of proper selection of input parameters such as DNB correlations to obtain the real characteristic of the transient parameters [2, 3]. In this context, reactivity initiated transient analysis code EUREKA-2/RR [4] has been used whether it could be able to model the SPERT IV transient experiment. This work has been done under the framework of the International Atomic Energy Agency Coordinated Research Project (CRP) 1496 “Innovative Methods in Research Reactor Analysis: Benchmark Against Experimental Data on Nuetronics* [sic]* and Thermal Hydraulic Computational Methods and Tools for Operation and Safety Analysis of Research Reactors,” the aim of which is to develop the innovative research reactor modeling methods and to assess and qualify the computational codes for application in the safety analysis of various research reactors. Being the member of this CRP project, the authors have been delivered the design specifications with tests results of SPERT IV transient experiment for analysis with the computer code they have.

#### 2. SPERT IV Transient Tests Description

The SPERT IV D-12/25 core was a light water-cooled and moderated, pool-type reactor with provisions for both upward forced and natural convection cooling. The core is composed of 25 type-D standard fuel assemblies in a square five by five section of the nine by nine supporting grid. The core configuration is provided in Figure 1. Each standard fuel assembly contained 12 removable fuel plates. Four gang-operated boron-alloy double-blade control rods and one transient rod of the same style located in the center of the core were accommodated in the standard fuel assembly replacing its six fuel plates. For complete specification of the reactor components, Day [5] can be referred to. The test sequence is illustrated in Figure 2 where five transient phases are noticed. There are 39 transient tests reported for SPERT IV reactor benchmark analysis [5]. The published results for peak power, peak cladding temperature, and energy released are provided in the document [5] with respect to a number of control variables such as initial reactor power, reactivity insertion, bulk moderator-coolant temperature, hydrostatic head, and coolant circulation rate. Table 1 shows these tests are divided into five groups in terms of coolant flow rate. The first group corresponds to flow rate zero with coolant velocity zero, the second group corresponds to 500 gpm with coolant velocity 0.365 m/s, the third group corresponds to 1000 gpm with coolant velocity 0.731 m/s, the fourth group corresponds to 2500 gpm with coolant velocity 1.828 m/s, and the fifth group corresponds to 5000 gpm with coolant velocity 3.657 m/s. Also, each test of each group is seen to differ with respect to initial ambient reactor temperature, power, reactivity insertion, and corresponding period. It is also apparent that the tests in each group were performed in a progressive manner by increasing the amount of positive insertion of reactivity so the corresponding transient period is found shorter. Most of the tests in the series were initiated with the reactor critical at a power level of approximately 1 W which is related to the initial transient period >15 ms. The remaining tests were performed for <15 ms for which the reactor was initially subcritical at a power in the milliwatt range that allows the complete reactivity insertion before power rises to a significant level.