Analysis of safety limits of the Moroccan TRIGA MARK II research reactor
Creators
- 1. Groupe de Physique des Reacteurs, Faculte des Sciences, Rabat (Morocco)
Description
A 2 MW TRIGA MARK II research reactor has been designed by General Atomics (GA) for the National Centre for Energy and Nuclear Sciences and Techniques (CNESTEN) in Morocco. This TRIGA reactor has the particularity of being the only TRIGA reactor designed to operate at the power level of 2 MW with the use of natural convection cooling. The main objective of this study is to check the ability of the reactor to operate at its nominal power with sufficient safety margins. The analysis of the reactor core starts from the basic reactor cells calculations which were performed for all the reactor cells using the LEOPARD code. The zone averaged group constants provided by cell calculations are used to compute the multiplication factor keff of the cold and clean core using the diffusion theory code Mcrac which is a recent version of the earlier code EXTERMINATOR-2. The main objective of the core calculations is to predict the core excess reactivity in cold zero power conditions and the power peaking factors which are very important data for the thermal hydraulic analysis of the core. For the maximum power peaking factors, our results agree with the values given by the reactor designer. Concerning the core excess reactivity, our results from both XY and RZ core calculations models lead to higher values than the results given by GA (about +2000 pcm). However, we should mention that GA results correspond probably to the minimum core excess reactivity which is guaranteed. The thermal hydraulic analysis of the TRIGA core was performed using the French code FLICA developed in CEA/Saclay for transient and study state thermal hydraulic analysis of a large variety of reactor cores. The objective of this analysis is to evaluate the main safety related parameters of the core and to ensure that they are within the safety limits in any operating conditions. The parameters considered in our study are: maximum fuel temperature, minimum DNBR and maximum void fraction. The obtained results confirm the GA conclusions except for the void fraction (the calculated values are beyond the usual accepted limits). (author)
Additional details
Identifiers
Publishing Information
- Publisher
- Faculty of Nuclear Sciences and Physical Engineering
- Imprint Place
- Prague (Czech Republic)
- ISBN
- 80-01-02180-7
- Imprint Title
- ISRP-8. 8th international symposium on radiation physics. Abstracts
- Imprint Pagination
- 340 p.
- Journal Page Range
- p. 123
Conference
- Title
- 8. international symposium on radiation physics (ISRP-8)
- Dates
- 5-9 Jun 2000
- Place
- Prague (Czech Republic)
INIS
- Country of Publication
- Czech Republic
- Country of Input or Organization
- Czech Republic
- INIS RN
- 32009837
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- COMPUTER CALCULATIONS; E CODES; F CODES; L CODES; M CODES; MULTIPLICATION FACTORS; REACTIVITY; REACTOR CORES; REACTOR SAFETY; SAFETY ANALYSIS; THERMAL ANALYSIS; TRIGA TYPE REACTORS
- Descriptors DEC
- COMPUTER CODES; ENRICHED URANIUM REACTORS; HOMOGENEOUS REACTORS; HYDRIDE MODERATED REACTORS; REACTOR COMPONENTS; REACTORS; RESEARCH AND TEST REACTORS; SAFETY; SOLID HOMOGENEOUS REACTORS; WATER COOLED REACTORS; WATER MODERATED REACTORS
Optional Information
- Notes
- The abstract in the publication is identical with that reproduced below