[en] Full text: Static power coefficient is calculated for the reactor PFBR using ABBN perturbation data. It is calculated as the change in reactivity for a unit change in power from one steady state to the other steady state, keeping all the other reactor parameters like coolant flow and inlet temperature constant. Power coefficient are calculated using the code PREDIS, for all the representative pin using ABBN perturbation data with the corresponding axial and radial core boundary movement worths. Isothermal temperature coefficients are calculated for constant thermal expansion coefficients of fuel and steel. Analysis of unprotected uncontrolled withdrawal of a CSR is done with the ABBN perturbation data and new radial and axial boundary movement worths, with inlet coolant temperature as a varying function of time. From uncontrolled withdrawal of CSR, external reactivity of 1.36 $ is given in 200 s, with the negative feedback reactivity, the resultant net reactivity becomes 0.002 $ at 250s. The peak melt fraction is 28% in the central subassembly, over 40 cm length in core middle, with 0.8% over all fraction of core melt

[en] Highlights: • Possibilities of enhancing safety under ULOF and UTOP accidents. • CSRDM expansion feedbacks under unprotected transients. • CSRDM expansion feedback enhances the safety of fast reactors. • CSRDM expansion feedbacks ensuring enough time for initiating safety actions. - Abstract: Possibilities of enhancing core safety under unprotected loss of flow (ULOF) and unprotected transient over power (UTOP) accidents with control and safety rod drive mechanism (CSRDM) expansion feedbacks are explored in a medium sized oxide fuelled fast breeder reactor. This feedback is expected to take the reactor to a safe shutdown under ULOF and to an another steady state under UTOP where there is no significant fuel melting. Under ULOF, with CSRDM feedback net reactivity was maintained negative throughout the transient (up to 2000 s) and the power dropped to a level of heat removal capacity of decay heat removal system based on natural circulation. Similarly, under UTOP with the above feedback reactor power goes to a lower peak value. The fuel temperature is just touching the melting temperature and the melt fraction does not cross 5%. With CSRDM expansion feedbacks both ULOF and UTOP transients prolong beyond 2000 s. It ensures, availability of time for initiating any safety actions against the transients, and thus it helps to preclude core disruptive accidents (CDA) in a medium sized oxide fuelled reactors.Classification: L. safety and risk analysis

[en] Accident analysis for unprotected loss of flow (ULOF) has been done for the BN-800 reactor with non-zero sodium void coefficient of reactivity. The calculations are performed in three phases, steady state fuel pin characterisation, transient calculations up to onset of sodium boiling and post sodium boiling up to first fuel pin failure. The results of the calculations are compared with those obtained by the standard computer codes used by west European countries. The agreement of our results with other codes are fairly good and our code predicts all the generic features of ULOF. (author)

[en] Conclusions: • Metal fuel shows superior inherent passive safety features for both ULOFA and UTOPA. • Negative Doppler may be enhanced by addition of moderating material in metal fuel reactors to protect the reactor against UTOPA. • Increase thermal conductivity of fuel by suitable alloy. • Better expansion co-efficient of metal fuel to enhance negative reactivity. • Decrease gap between spacer pads to enhance the flowering effect. • Enhancing negative reactivity through axial and radial boundary movement by adjusting the enrichment

[en] Conclusions: • Clad relocation model and in-pin fuel motion models are implemented in the safety analysis code PREDIS and the effect of those phenomena in the severe accident characteristics of fast reactors are studied. • A benchmark analysis for PREDIS code is carried against the FFTF LOFWOS test 13. • The models on clad relocation, in-pin fuel motion and the fuel expansion helped in simulating fast reactor severe accidents in more detail.

[en] Unprotected loss of flow (ULOF) analyses was carried out for an extended flow coast down on a 1000 MWe metal core to verify the passive shutdown capability of the reactor and its inherent safety parameters. Based on earlier analysis nominal behaviors of 1000 MWe under ULOFA with 8 s flow halving time is found to be benign and the over all reactivity remains negative till the power reduces below SGDHRS capacity. It is also found that, with considering uncertainties on sensitive parameters such as the core radial expansion feedback and sodium void reactivity the transient heads for CDA. From the studies it was concluded that the sodium void worth is required to be reduced by 20 % by design. In the present work sensitivity analyses are carried out with enhanced flow coast down (15 s) without making any design changes. Sensitivity analyses ensure the safe shutdown of 1000 MWe with the consideration of uncertainties on these parameters. (author)

[en] Methodology is derived to determine the power reactivity decrement between any two asymptotic states. Vested reactivity feedback parameters from the fuel side (A) and the coolant side (B) are determined separately to examine the inherent safety character of the reactor. Power reactivity decrement is the combination of vested reactivity feedback parameters from the fuel side (A) and the coolant side (B). This is the amount of reactivity which has to be compensated with the addition of excess reactivity while taking the reactor from zero power to nominal power, or this much of reactivity comes back as a positive reactivity when the reactor go to zero power from nominal power. Inherent safety of medium sized 500 MWe oxide fuel core (PFBR) and metal fuel core are examined based on these values, it is found power reactivity decrement of metal fuel reactor is found to be small as compared to oxide fuel. (author)

[en] Generalised methodology is derived to determine the power coefficients between any two asymptotic states. Methodology is derived by simplifying heat transfer equations into lumped parameter model. If the reactor coolant flow and thermo physical properties are assumed to be constant, change in reactivity at a given reactor location is purely a function of linear power, and the net change in reactivity is a function of net change in total power. Power coefficient is almost a constant value, when it is calculated between any two asymptotic states, except the Doppler contribution, a non linear reactivity component. It is verified between zero power and nominal power, and between zero power and 50 % of nominal power. (author)

[en] Point kinetics equations are stiff differential equations, and their solution by the conventional explicit methods will give a stable consistent result only for very small time steps. Since the neutron lifetime in a LMFBR is very short, the point kinetics equations for LMFBRs become even stiffer. In this study the power series solution (PWS) method is applied for solving the point kinetics equations for a typical LMFBR. A Fortran program is developed for accident analysis of LMFBRs with the PWS method for solving the point kinetics and a lumped model for solving the heat transfer equations. A new technique is developed with fixing factor to find out the average temperature at the peak power node (PPN) without performing temperature calculations at all axial nodes in a reactor fuel pin. The temperature at PPN also decides whether the reactor is within the design safety limit (DSL) or it has entered a serious transient that may lead to an accident. The coupled heat transfer and point kinetics models for a peak power node give the average fuel, clad and coolant temperatures. For the transient over power accidents (TOPA), this is the best way for calculating the temperature, with minimum amount of computations. TOPA analyses are carried out with PWS method. It is found that the PWS methodology uses a small number of numerical operations, while the computational time and the accuracy are comparable with the available fast computational tools. This methodology can be used in nuclear reactor simulation studies and accident analysis

[en] Point kinetics equations (P. K. E) are system of differential equations, which is solved simultaneously to get the neutron density as a function of time for a given reactivity input. P. K. E are stiff differential equations, computational solution through the conventional explicit method will give a stable consistent result only for smaller time steps. Analytical solutions are available either with step or ramp reactivity insertion without considering the source power contribution. When a reactor operates at low power, the neutron source gives a considerable contribution to the net reactor power. Similarly, when the reactor is brought to delayed critical with the presence of external source, the sub critical reactor kinetics studies with source power are important to understand the power behavior as a function of reactivity insertion rate with respect to the initial reactivity. In the present work, P.K.E with one group delayed neutron are solved analytically to determine the reactor power as a function of reactivity insertion rate in the presence of neutron source. The analytical solution is a combination of converging two infinite series. Truncated infinite series is the analytical solution of P.K E. A general formulation is made by Combining both the ramp reactivity and step reactivity solution. So that the analytical solution could be useful in analyzing either step and ramp reactivity insertion exclusively or the combination of both. This general formulation could be useful in analyzing many reactor operations, like the air bubble passing through the core, stuck rod conditions, uncontrolled withdrawal of controlled rod, discontinuous lifting of control rod, lowering of rod and etc. Results of analytical solutions are compared against the results of numerical solution which is developed based on Cohen's method. The comparisons are found to be good for all kind of positive and negative ramp reactivity insertions, with or without the combination of step reactivity. The methodology is found to be a promising tool for analyzing low power reactor with the inclusion of external source. (author)