here - Atomic Energy Regulatory Board

CHAPTER 9
SAFETY RESEARCH INSTITUTE
Research activities in the following areas are being
pursued at AERB’s Safety Research Institute (SRI),
Kalpakkam.
l Nuclear Safety Studies.
l Reactor Physics Studies.
l Radiological Safety Studies.
l Environmental Safety Studies.
9.1 NUCLEAR SAFETY STUDIES
9.1.1 Seismic Re-evaluation of FBTR
For seismic re-evaluation of FBTR, a criteria
document was developed consistent with the
internationally accepted practices. This document also
describes the methodology to be adopted for carrying out
seismic re-evaluation and the different tasks involved in the
exercise and interfaces with the different activities. Work
was also undertaken to identify the frontline systems and
support systems that perform safety functions in case of a
seismic event. This will involve identification of (a) the safe
shutdown path, (b) systems for maintaining the plant in
the safe shutdown condition and decay heat removal and
(c) systems to maintain containment integrity. To meet this
objective, a list of 19 Initiating Events (IEs) for the
functions found to be most important for seismic induced
core melt was identified. The frontline and support systems
that perform safety functions for each of the initiating event
were also identified and the corresponding fault trees were
developed based on the system functions. A total of about
90 fault trees for 19 event trees of IEs were developed as
part of this activity. Further work is in progress.
9.1.2 Functional Reliability Analysis of Safety
Grade Decay Heat Removal System
(SGDHRS) of PFBR
As part of the PFBR level-1 PSA activity,
evaluation of functional reliability analysis of SGDHRS has
been carried out. A list of critical parameters with reference
to initiating event groups that will have significant impact
on the mission success was prepared. Sensitivity analysis
was carried out for some of the parameters and these
parameters were ranked according to their importance. Each
of the selected parameters was assigned suitable
probability distribution and ranges of variation and a
functional reliability analysis of SGDHRS was carried out.
Uncertainty in the parameters was assessed using 1D plant
dynamics computer code, DHDYN. From a set of 50 runs
of the computer code, a multi response surface for 3
important responses was constructed. A large number of
Monte Carlo simulations were conducted using the response
surface model to estimate the functional failure probability
of SGDHRS. The probability of functional failure of
SGDHRS (on natural convection) is found to be
dependent on the number and duration of sodium loop
availability during the initial few hours of mission of
operation of SGDHRS.
The SGDHRS fails a) due to component failures
in the system OR b) due to failure of forced convection
AND functional failure. That is,
λ SGDHR-TOT =
∑ i
f i
* ( P Comp,i
+ P FUN-F,i
* P FC
) ,
where f i
is number of demands per year of type i for DHR,
P Comp,i
is the probability of failure of SGDHR due to
component failures for initiating event i, and P FUN-F,i
is the
functional failure probability for initiating event of type i,
P FC
is the failure probability of forced convection. The
integrated failure frequency of SGDHRS with both
functional and component failures is found to be 2.1E-7/y.
9.1.3 Development of Database on Fast Reactor
Components
The development of database on failure rates for
fast reactor components was continued. So far failure data
have been collected for about 3000 components and stored
in the database. The data is stored and retrieved from a
relational database system. The application can be used to
obtain a suitable failure rate for a given component
specification, viz., category, group. The module to
combine the operating experience with the stored data and
estimation of the posterior failure data using Bayesian
technique has been completed. While the security of the
database is ensured, user interface is made available to
retrieve component reliability information over the IGCAR
intranet.
9.2 REACTOR PHYSICS STUDIES
9.2.1 PWR Physics Analysis
In continuation of the work initiated last year on
the development of expertise in PWR physics analysis and
fuel management strategy (taken up in collaboration with
Reactor Physics Design Division (RPDD) /BARC and
NPCIL), the computer code system, EXCEL &
TRIHEX-FA along with 172-energy groups IAEAGX/
ENDFB6GX cross section libraries in WIMS-D format,
developed at Light Water Reactor Physics Section (LWRPS),
RPDD, BARC, have been acquired and commissioned at
SRI. Few groups lattice database has been generated using
EXCEL code for all types of fuel assemblies of KKNPP.
Core physics analysis for the proposed 8 fuel cycles of
KK-NPP was carried out with TRIHEX-FA code using the
aforementioned lattice database. The results of the
analysis include critical soluble boron concentration, 3D
distribution of power, burn-up, fuel/coolant temperature,
coolant density, reactivity coefficients and kinetics
parameters of core average delayed neutron fraction ‘β’,
prompt neutron mean life time ‘l’ and the material
52