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7.7.6 Development of In-house Code for Transient materials are classified into three categories with respect
Analysis of Sodium Cooled Fast Reactor (SCFR) to reactivity loads: best suitable (with small reactivity
load 0 to -5 mk), may be acceptable (with medium
An in-house computer code has been developed reactivity load -5 mk to -10 mk) and not suitable (with
for analysing flow and power transients in fast breeder large reactivity load more than -10 mk). Full power
reactors. Capability of the code has been improved by days penalty due to the ATF material coating are also
incorporating slug-expulsion model for analysing void assessed.
progression and flow inside the core with significant
voiding. Comparison of pre-disassembly phase of an 7.7.9 Effect of Americium Build-up on Core Physics
un-protected transient over-power accident (UTOPA) Parameters of Fast Reactor
scenario was carried out with FBR accident analysis
code, SAS-1A. Pressures, temperature, void fraction, In order to evaluate effect of Americium build up on
241
vapour quality, saturation temperature of coolant and core neutronics parameters, Am build-up due to decay
temperature distributions of fuel, clad and structures of 241 Pu isotope in fresh fuel subassemblies has been
were estimated at different times and compared with estimated for residence periods of two and five years.
SAS-1A results. By using the evaluated number densities of the isotopes
after considering the residence period, neutronics
7.7.7 Independent Verification Analysis of PHWR- calculations have been performed to evaluate core
700 Stability excess reactivity in case of PFBR. It is found that the
delay of fuel loading in core, could lead to considerable
As a part of independent verification, stability reduction in core excess reactivity due to decay of Pu
analysis of the total power control loop of 700 MWe isotope.
PHWR was carried out. Analysis was carried out in
discrete-time domain by linearizing the system around 7.7.10 Investigation of Attenuation Characteristics
its equilibrium points and identifying Eigen values of of Advanced Shield Materials
the closed-loop system. The dynamics of the system
vary widely depending on operating power levels, Investigation of neutron attenuation characteristics of
core–fuelling states and cycle time of reactor regulating advanced shielding materials like Portland cement mixed
system. The input to the stability analysis are dynamics with 10wt% Samarium Oxide (Sm O ) and ordinary
3
2
governing system parameters and the output is a stability concrete admixture with 10wt% Borated Polyethylene
characterization in terms of gain of total power control was carried out for typical Light Water Reactor (LWR)
loop at any power level. The results independently neutron spectrum. Neutron spectra simulated with and
verified the stability analysis and respective gain values without shielding materials of interests given in Fig. 7.23
of total power control loop. clearly demonstrate the effect of each shielding material
on neutron spectrum.
7.7.8 ATF Coating Material for PHWR based on
Neutronics Evaluation

Coating around Zircalloy based cladding material
enhances accident tolerance capability due to its various
favourable properties such as high melting point, low
oxidation rate, low hydrogen generation and remain
stable even in nuclear accident scenarios. However
loading of coating material in the reactor will impact on
neutronic properties. Based on the exhaustive literature
survey, in total 24 materials have been identified and
considered for their reactivity load assessment for
PHWR-700 reactor. The reactivity load for both side
coating thickness of 10 μm, 20 μm, 30 μm and 40 μm
are used for estimating the reactivity load for potential
24 number of Zircaloy based ATF clad coating materials Fig. 7.23: Comparison of Neutron Dose Rate Attenuation
for PHWR700. Based on the analysis, the coating Behaviour

92 AERB Annual Report 2019