604 research outputs found
RBMK Fuel Channel Blockage Analysis by MCNP5, DRAGON and RELAP5-3D© codes
No full textThe aim of this work was to perform precise criticality analyses by Monte-Carlo code
MCNP5 for a Fuel Channel (FC) flow blockage accident, considering as calculation domain a single FC and a 3x3 lattice of RBMK cells. Boundary conditions for MCNP5 input were derived by a previous transient calculation by state-of-the-art codes HELIOS/RELAP5-3D©.
In a preliminary phase, suitable MCNP5 models of a single cell and of a small lattice of RBMK cells were set-up; criticality analyses were performed at reference conditions for 2.0%
and 2.4% enriched fuel. These analyses were compared with results obtained by University of Pisa (UNIPI) using deterministic transport code DRAGON and with results obtained by NIKIET Institute using MCNP4C. Then, the changes of the main physical parameters (e.g. fuel and water/steam temperature,
water density, graphite temperature) at different time intervals of the FC blockage transient were evaluated by a RELAP5-3D© calculation. This information was used to set up further
MCNP5 inputs. Criticality analyses were performed for different systems (single channel and lattice) at those transient’ states, obtaining global criticality versus transient time. Finally the
weight of each parameter’s change (fuel overheating and channel voiding) on global
criticality was assessed. The results showed that reactivity of a blocked FC is always negative; nevertheless, when considering the effect of neighboring channels, the global reactivity trend reverts, becoming slightly positive or not changing at all, depending in inverse
relation to the fuel enrichment
RBMK Fuel Channel Blockage Reactivity Analysis by MCNP5 and RELAP5-3D© codes
No full textThe aim of this work was to perform precise
criticality analyses by Monte-Carlo code MCNP5 for a
Fuel Channel (FC) flow blockage accident,
considering as calculation domain a single FC and a
3x3 lattice of RBMK cells. Boundary conditions for
MCNP5 input were derived by a previous transient
calculation by state-of-the-art codes
HELIOS/RELAP5-3D©.
In a preliminary phase, suitable MCNP5 models
of a single cell and of a small lattice of RBMK cells
were set-up; criticality analyses were performed at
reference conditions, for 2.0% and 2.4% enriched
fuel.
Then, the changes of the main physical parameters (e.g. fuel and water/steam temperature, water density, graphite temperature) at different time intervals of the FC blockage transient were evaluated by a RELAP5-3D© calculation. This information was used to set up further MCNP5 inputs. Criticality
analyses were performed for different systems (single
channel and lattice) at those transient’ states,
obtaining global reactivity versus transient time.
Finally the weight of each parameter’s change (fuel
overheating and channel voiding) on global criticality
was assessed. The results showed that reactivity of a
blocked FC is always negative; nevertheless, when
considering the effect of neighboring channels, the
global reactivity trend reverts, becoming slightly
positive or not changing at all, depending in inverse
relation to the fuel enrichment
The University of Pisa Calculations for the Phase I of the OECD/NEA UAM Benchmark
No full textIn this paper we present the University of Pisa preliminary results for the first exercise of the
Phase I of the OECD/NEA Benchmark on the Uncertainty in Analysis and Modeling. The scope of
exercise one is to address the uncertainties due to the basic nuclear data as well as the impact of
processing the nuclear and covariance data, selection of multi-group structure and self-shielding
treatment. DRAGON code and TSUNAMI code were employed, using the available covariance
data matrix. The execution of DRAGON calculations required the use of ANGELO and LAMBDA
codes for the extension of the covariance matrix from the original SCALE 44 group structure to
DRAGON 69 group structure. The uncertainties for the main cross sections were evaluated and
are presented here
Analysis of MSLB accident in VVER-1000 with coupling between 3-dimensional neutron kinetics and thermal-hydraulics
No full textAnalysis of flow blockage of a single RBMK channel
No full textThe aim of the this study is to perform an evaluation of the behaviour of a single RBMK reactor core
fuel channel focusing on the fuel clad and the Pressure Tube, in case of a flow blockage accident. The
tools used were the RELAP5 and FRAP code. The RBMK channel, the graphite stack and the He-N2
gap were modelled with the RELAP5 code; the thermo-mechanical fuel rod behaviour was studied by
FRAP code. Some cases were analyzed considering different values of coolant flow reduction at
different values of the channel power in order to demonstrate if the break of the pressure tube and the
failure of fuel clad occur. A failure map was drawn identifying safe operational zones for the Pressure
Tube and the fuel clad
Boron dilution and boron transport after SBLOCA in PWR and VVER-1000 nuclear reactors, Book ‘Boron dilutation in pressurizer water reactors’
No full textThree-dimensional Neutron Kinetics-Thermal-Hydraulics VVER1000 Main.Steam Line Break analysis by RELAP5-3D© code
No full textAfter the development and the assessment of Three-Dimensional (3D) Neutron Kinetics (NK) – 1D
Thermal-Hydraulics (TH) coupled codes analyses
methods, deterministic nuclear safety technology is
nowadays producing noticeable efforts for the validation of 3D NK – 3D TH coupled codes analyses methods too.
Thus, the purpose of this work was to address the
capability of the RELAP5-3D© 3D NK – 3D TH code to
reproduce VVER1000 Nuclear Power Plant (NPP) core
dynamic in simulating the mixing effects that could
happen in the vessel downcomer and lower plenum
during some scenarios.
The work was developed in three steps. The first
step dealt with the 3D TH modeling of the Kozloduy-6
VVER1000 reactor pressure vessel. Then this model was validated following a Steam
Generator Isolation transient.
The second step has been the development of a 3D NK nodalization for the reactor core region. Then the 3D NK model was directly coupled with the previously developed 3D TH model. The third step was the calculation of a Main Steam Line Break (MSLB) transient. The 3D NK global
nuclear parameters were then compared with the 0-D
results showing a good agreement; nevertheless only
the 3D NK- 3D TH model allowed the calculation of
each single assembly power trend for this strong NK-
TH asymmetric transient
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