73 research outputs found
Hazard Assessment of NPP Krško for Republic of Croatia
While Croatia does not have nuclear power plant on its territory, NPP Krško in Slovenia is just
10 km from the Croatian border. It is important for Croatia to include NPP Krško in comprehensive
hazard assessment. This article will present hazard assessment based on calculations using RODOS.
Real-time weather prepared by Croatian National Weather Service and collected by the State Office
for Radiological and Nuclear Safety over the years will be used. Scenario resulting in the large release
from the NPP will be analysed. Results from hundreds of calculations will be statistically analysed
and compared to the current protection zones in Croatia around the NPP Kršk
Operation and Performance Analysis of Steam Generators in Nuclear Power Plants
Steam generators are components in which heat produced in the reactor core is transferred to the secondary side, the steam supply system, of the nuclear power plant (NPP). Steam generators (SGs) have to fulfil special nuclear regulatory requirements regarding their size, selection of materials, pressure loads, impact on the NPP safety, etc. The primary-side fluid is liquid water at the high pressure, and the fluid on the secondary side is saturated water-steam mixture at the pressure twice as low. A special attention is given to preserving the boundary between the contaminated water in the primary reactor coolant system and the water-steam mixture in the secondary system. A brief overview of the SG design, its operation and the mathematical correlations used to quantify heat transfer is given in the chapter. Results of the SG transient behaviour obtained by the simulation with the best-estimate computer code RELAP5, developed for safety analyses of NPPs, are also presented. Two types of steam generators are analyzed: the inverted U-tube SG, which is commonly found in the present-day pressurized water reactors and the helical-coil SG that is part of the new-generation reactor designs
INTRODUCTION
Journal of Energy special issue: Papers from 11th International Conference
of the Croatian Nuclear Society “Nuclear Option in Countries
with Small and Medium Electricity Grids”
Welcome to this special issue, which is based on selected papers
presented at the 11th International Conference of the Croatian Nuclear
Society “Nuclear Option in Countries with Small and Medium Electricity
Grids”, held in Zadar, Croatia, on June 5th–8th, 2016.
This International Conference was organized by the Croatian Nuclear Society
in cooperation with International Atomic Energy Agency (IAEA), Croatian
State Office for Nuclear Safety and University of Zagreb, Faculty of
Electrical Engineering and Computing. The goal of the Conference was to
address the various aspects of the implementation of nuclear energy for
electricity production in the countries with small and medium electricity
grids and in power system in general. The conference also focuses on the
exchange of experience and co-operation in the fields of the plant operation,
nuclear fuel cycle, nuclear safety, radioactive waste management,
regulatory practice and environment protection.
The conference was organized in eight main topics covered in ten oral
sessions and one poster session. In three Conference days authors presented
49 papers orally and 23 papers in poster session. 102 participants
came from 16 countries representing equipment manufacturers and utilities,
universities and research centres, and international and government
institutions. Eight invited lectures were held and 72 papers were accepted
by international programme committee.
The importance of international cooperation for the assessment of the
nuclear option has been recognized by everybody planning to introduce
nuclear power plant to the grid. That is even more important for small and
medium countries having limited resources and specific requirements due
to limited grid size. The Conference topics reflect some current emphasis,
such as country energy needs, new reactor technologies (especially small
reactors), operation and safety of the current nuclear power plants, move
of the focus in nuclear safety toward severe accidents and accident management
strategies, improvement in nuclear safety, reactor physics and
radiation shielding calculation tools and ever increasing requirements for
minimization of environmental impact.
From 72 papers presented at the Conference, 16 papers were accepted
for publication in this number of Journal of Energy after having undergone
the additional peer-review process. We would like to thank the authors for
their contributions and the reviewers who dedicated their valuable time
in selecting and reviewing these papers, both during the Conference and
during the preparation of this special issue of Journal of Energy. It was
very challenging to collect a balanced overview of the entire Conference.
We decided to select 16 papers for this issue and additional 14 for the
next one. We believe that the papers which were selected for this number
represent some of the best research related to nuclear plant operation,
energy planning, development of new reactors and technologies, reactor
physics and radiation shielding. We hope this special issue will provide a
valuable insight into different aspects of nuclear and electrical engineering
and reactor physics, as well as a pleasant and inspiring reading
NPP KRŠKO CONTAINMENT MODELLING WITH THE ASTEC CODE
ASTEC is an integral computer code jointly developed by Institut de
Radioprotection et de Sûreté Nucléaire (IRSN, France) and Gesellschaft für Anlagenund
Reaktorsicherheit (GRS, Germany) to assess nuclear power plant behaviour
during a severe accident (SA). The ASTEC code was used to model and to simulate
NPP behaviour during a postulated station blackout accident in the NPP Krško.
The accident analysis was focused on containment behaviour; however the complete
integral NPP analysis was carried out in order to provide correct boundary
conditions for the containment calculation. During the accident, the containment
integrity was challenged by release of reactor system coolant through degraded
coolant pump seals, molten corium concrete interaction and direct containment
heating mechanisms. Impact of those processes on relevant containment
parameters, such as compartments pressures and temperatures, is going to be
discussed in the paper
NPP Krško Station Blackout Analysis after Safety Upgrade Using MELCOR Code
The analysis of a Station blackout (SBO) accident in the NPP Krško including thermalhydraulic
behaviour of the primary system and the containment, as well as the simulation of the
core degradation process, release of molten materials and production of hydrogen and other
incondensable gases will be presented in the paper. The calculation model includes the latest plant
safety upgrade with addition of Passive Autocatalytic Recombiners (PAR) and the Passive
Containment Filter Venting (PCFV) system. The code used is MELCOR, version 1.8.6. MELCOR
is an integral severe accident code which means that it can simulate both the primary reactor
system, including the core, and the containment. The code is being developed at Sandia National
Laboratories for the U.S. Nuclear Regulatory Commission.
The analysis is conducted in two steps. First, the steady state calculation is performed in order
to confirm the applicability of the plant model and to obtain correct initial conditions for the
accident analysis. The second step is the calculation of the SBO accident with the leakage of the
coolant through the damaged reactor coolant pump seals. Without any active safety systems, the
reactor pressure vessel will fail after few hours. The mass and energy releases from the primary
system cause the containment pressurization and rise of the temperature. The newly added safety
systems, PAR and PCFV, prevent the damage of the containment building by keeping the thermalhydraulic
conditions below the design limits. The analysis results confirm the capability of the
safety systems to effectively control the containment conditions.
Results of the analysis are given with respect to the results of the MAAP 4.0.7 analysis for the
same accident scenario. The MAAP and MELCOR codes are the most popular severe accident
codes and, therefore, it is reasonable to compare their results. In addition, sensitivity calculations
performed by varying most influential parameters, such as the hot leg creep failure, blockage of a
pipe connecting the cavity and the sump, inclusion of a radionuclide package in the MELCOR, etc.
are done in order to demonstrate correct physical behaviour and the accuracy of the developed NPP
Krško MELCOR model
Spent Fuel Pool Dose Rate Calculations Using Point Kernel and Hybrid Deterministic-Stochastic Shielding Methods
This paper presents comparison of the Krško Power Plant simplified Spent Fuel Pool (SFP)
dose rates using different computational shielding methodologies. The analysis was performed to
estimate limiting gamma dose rates on wall mounted level instrumentation in case of significant
loss of cooling water. The SFP was represented with simple homogenized cylinders (point kernel
and Monte Carlo (MC)) or cuboids (MC) using uranium, iron, water, and dry-air as a bulk region
materials. The pool is divided on the old and new section where the old one has three additional
subsections representing fuel assemblies (FAs) with different burnup/cooling time (60 days, 1 year
and 5 years). The new section represents the FAs with the cooling time of 10 years. The time
dependent fuel assembly isotopic composition was calculated using ORIGEN2 code applied to the
depletion of one of the fuel assemblies present in the pool (AC-29). The source used in Microshield
calculation is based on imported isotopic activities. The time dependent photon spectrum with total
source intensity from Microshield multigroup point kernel calculations was then prepared for two
hybrid deterministic-stochastic sequences. One is based on SCALE6.2b3/MAVRIC (Monaco and
Denovo) methodology and another uses Monte Carlo code MCNP6.1.1b and ADVANTG3.0.1.
code. Even though this model is a fairly simple one, the layers of shielding materials are thick
enough to pose a significant shielding problem for MC method without the use of effective variance
reduction (VR) technique. For that purpose the ADVANTG code was used to generate VR
parameters for the MCNP fixed-source calculation using continuous energy transport. ADVATNG
employs a deterministic forward-adjoint transport solver Denovo which implements CADIS/FWCADIS
methodology. Denovo uses a structured, Cartesian-grid SN solver based on the Koch-
Baker-Alcouffe parallel transport sweep algorithm across x-y domain blocks. This was our first
application of ANDVANTG/MCNP hybrid sequence for this type of calculation and the results
where compared to SCALE/MAVRIC sequence which we regularly use for similar calculations.
The comparison of gamma dose rates on different point detector locations (central above pool and
at the top of pool periphery) showed a good agreement between Microshield (point-kernel) and
deterministic-stochastic shielding methodologies for the cylindrical approximation of the pool
geometry. More complicated cases for model with multi-source option and for cuboids showed very
good agreement between SCALE/MAVRIC and ANDVANTG/MCNP calculations
NPP Krško Post-UFC Transient Response during MSLB
UpFlow Conversion (UFC) was implemented in NPP Krško during the last outage in order to
reduce the pressure differential across baffle plates and the possibility of the fuel damage caused by
flow induced vibration. The paper describes the coupled code calculation (RELAP5 and PARCS) of
MSLB accident at power for pre and post-UFC configuration of reactor vessel. In the calculation,
the split model of the reactor vessel was used to better describe asymmetric conditions in loops. It
has been demonstrated that the basic parameters (pressure, temperatures) stayed unchanged and
there was little change in the flow rates except in baffle-barrel region of the vessel where both flow
direction and amount of flow were changed
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