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1 Nu clear Sci ence and Tech nol ogy Re search In sti tute, Atomic En ergy Or ga ni za tion of Iran, Teh ran, Iran 2 De part ment of Phys ics, Imam Hussein Uni ver sity, Teh ran, Iran 3 De part ment of Nu clear En gi neer ing, Sci ence and Re search Branch, Is lamic Azad Uni ver sity, Teh ran, Iran Sci en tific pa perAn ac cu rate anal y sis of the flow tran sient is very im por tant in safety eval u a tion of a nu clear power plant. In this study, anal y sis of a WWER-1000 re ac tor is in ves ti gated. In or der to perform this anal y sis, a model is de vel oped to sim u late the cou pled ki net ics and ther mal-hy draulics of the re ac tor with a sim ple and ac cu rate nu mer i cal al go rithm. For ther mal-hy drau lic calcu la tions, the four-equa tion drift-flux model is ap plied. Based on a multi-chan nel ap proach, core is di vided into some re gions. Each re gion has dif fer ent char ac ter is tics as rep re sented in a sin gle fuel pin with its as so ci ated cool ant chan nel. To ob tain the core power dis tri bu tion, point ki netic equa tions with dif fer ent feed back ef fects are uti lized. The ap pro pri ate ini tial and bound ary con di tions are con sid ered and two sit u a tions of de creas ing the cool ant flow rate in a pro tected and un pro tected core are an a lyzed. In ad di tion to anal y sis of nor mal op er a tion con di tion, a full range of ther mal-hy drau lic pa ram e ters is ob tained for tran sients too. Fi nally, the data ob tained from the model are com pared with the cal cu la tions con ducted us ing RELAP5/MOD3 code and Bushehr nu clear power plant data. It is shown that the model can pro vide ac cu rate pre dic tions for both steady-state and tran sient con di tions.

In view of the importance of studying coolant transient behavior in a nuclear reactor, this work is devoted to the thermal-hydraulic analysis of protected and unprotected loss of flow transients in a WWER-1000 reactor. A series of corresponding mathematical and physical models based on the four-equation Drift-Flux model has been applied. Based on a multi-channel approach, the core has been divided into different regions. Each region has different characteristics as represented in a single fuel pin with its associated coolant channel. Appropriate initial and boundary conditions have been considered and two situations of tripping four and two primary pumps in a protected core in addition to situation of tripping all four pumps in an unprotected core have been analyzed. For each transient, a full range of thermal-hydraulic parameters has been obtained. For verification of the proposed model, the results have been compared with those of the RELAP5/MOD3 and Bushehr nuclear power plant Final Safety Analysis Report (FSAR). A good agreement between results has been attained for the aforementioned transients.

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