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

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

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

ANALYSIS OF SPENT FUEL POOL LOSS OF COOLANT INVENTORY ACCIDENT PROGRESSION

The Spent Fuel Pool (SFP) in a nuclear power plant is a robust structure designed to withstand large seismic loads. After the terrorist attacks on September 11 and the accident in the Fukushima nuclear power plant, special attention was focused on safety assessments and taking measures to mitigate possible accidents related to the spent fuel pool. This paper will provide an insight into spent fuel pool loss of coolant phenomenology and consequence mitigation strategies. Model of NPP Krško SFP was presented in MELCOR code which has been used as a case-study for evaluating accident propagation. The calculations were carried out using the latest version of MELCOR code which was updated for the analysis of severe accidents in nuclear spent fuel pools

Influence of Spacer Grids Homogenization on Core Reactivity and Axial Power Distribution

The paper presents the influence of spacer grid homogenization during cross section generation on core reactivity and axial power distribution. Homogenization calculation was performed at fuel assembly level using FA2D code. The first approach is to smear uniformly all centrally located spacer grids along 120 inches of fuel assembly and carry out 2D transport calculation. The second approach is to smear spacer grid within 6 inches of fuel assembly and perform homogenization calculation. That composition is then assigned to closest 6 in axial subdivision of the core calculation. The last analysed option is to do additional localization of spacer grids and carry out homogenization within 2 inches of fuel assembly height. The additional subdivision is afterward performed of the closest regular axial core subdivision in nodal core calculation. The core calculation was performed using modified PARCS 2.5 code for NPP Krško cycle 29. The normalized axial power distributions obtained by PARCS for three different ways of spacer grid homogenization are then compared to quantify the influence of modelling. Similar comparison was performed for critical boron concentration. As expected larger influence is present for axial power distribution (more details for fine localization), with some influence on axial power offset and global reactivity

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