Evaluation of pipe whip impacts on neighboring piping and walls of the Ignalina Nuclear Power Plant

Stress corrosion cracks have been discovered in Group Distribution Headers (GDH) at the Ignalina and Chernobyl Nuclear Power Plants. This increases the probability that a guillotine pipe break can occur that creates a whipping pipe (GDH) with the potential to damage surrounding structures—i.e. adjacent GDH and its attached piping or adjacent reinforced concrete compartment wall. The GDH is the most important component for reactor safety in case of an accident. Emergency Core Cooling System (ECCS) piping is connected to the GDH piping such that, during an accident, coolant passes from the ECSS into the GDH.Presented in this paper is the transient analysis of a Group Distribution Header following a guillotine break at the blind end of the header. Using a very conservative force loading function, the transient response of a whipping RBMK-1500 GDH along with neighboring concrete walls and pipelines is obtained using finite element methodology.The results of the study, assuming that the impacted GDH does not suffer stress corrosion cracking, indicate that the structural integrity of the compartment should be maintained and failure should not propagate from GDH to GDH.     

Strength evaluation of a steam distribution device in the Ignalina NPP accident localisation system

The Ignalina NPP has a pressure suppression type of confinement, which is referred to as the accident localization system (ALS). The ALS prevents the release of the radioactive material from the NPP to the environment during a loss-of-coolant accident (LOCA). Ten water pools are located in the two ALS towers (five pools in each tower), which separate the dry well from the wet well. These water pools condense the accident-generated steam and prevent high overpressures in the compartments.The steam distribution device (SDD), with the vertical vent pipes (nozzles) that are inserted under the water of the condensing pools, connects the dry well and the wet well. In case of an accident, these components must be capable of withstanding the dynamic loads generated by a LOCA for successful pressure suppression function.This paper presents the transient analysis of the SDD and their connections to the vertical steam corridors following a LOCA. A thermo-hydraulic analysis of the SDD was performed using the state-of-the-art COCOSYS code to determine pressure and temperature histories resulting from a LOCA. The finite element code NEPTUNE was used to evaluate the structural integrity of the SDD and its supporting reinforced concrete wall. Results show that, although portions of the SDD undergo plastic response and the outside surface of the vertical steam corridor reinforced concrete wall cracks, the structural integrity of the SDD and wall are maintained during a LOCA.     

Probabilistic safety analyses developments resurgence in Bulgaria—Kozloduy nuclear power plant example

This paper presents some of the main technical features and insights of the Kozloduy nuclear power plant (NPP) units 5 and 6 probabilistic safety analysis (PSA) level 1. Probabilistic analyses and their applications in Bulgaria were given further impetus in recent years. More than 17 years after the first PSA study in Bulgaria in 1992 today probabilistic analyses receive increasing attention and application than ever before. The Bulgarian regulatory body (BNRA) is also interested in expanding their capability of reviewing and using PSA in plant safety assessments. In November 2008 within the framework of the program financed by European Union (PHARE), a project for assisting the BNRA in establishing the regulatory requirements on the base of PSA was completed. One of the objectives of this project was performance of the independent review of Kozloduy NPP units 5 and 6 PSA. This review was a new impulse for the authors to present in more details of Kozloduy NPP probabilistic assessment studies in the present paper.     

On the speed of response of an FPGA-based shutdown system in CANDU nuclear power plants

Several issues in an FPGA based implementation of shutdown systems in CANDU nuclear power plants have been investigated in this paper. A particular attention is on the response time of an FPGA implementation of safety shutdown systems in comparison with operating system based software solutions as in existing CANDU plants. The trip decision logic under “steam generator (SG) level low” condition has been examined in detail. The design and implementation of this logic on an FPGA platform have been carried out. The functionality tests are performed in a hardware-in-the-loop (HIL) environment by connecting the FPGA based system to an NPP simulator, and replacing one channel of Shutdown System Number 1 (SDS1) in the simulator by the FPGA implementation. The response time of the designed system is also measured through multiple tests under different conditions, and statistical data analysis has been performed. The results of the response time tests are compared against those of a software-based implementation of the same trip logic.     

Developing architecture for upgrading I&C systems of an operating nuclear power plant using a quality attribute-driven design method

This paper presents the architecture for upgrading the instrumentation and control (I&C) systems of a Korean standard nuclear power plant (KSNP) as an operating nuclear power plant. This paper uses the analysis results of KSNP's I&C systems performed in a previous study. This paper proposes a Preparation–Decision–Design–Assessment (PDDA) process that focuses on quality oriented development, as a cyclical process to develop the architecture. The PDDA was motivated from the practice of architecture-based development used in software engineering fields. In the preparation step of the PDDA, the architecture of digital-based I&C systems was setup for an architectural goal. Single failure criterion and determinism were setup for architectural drivers. In the decision step, defense-in-depth, diversity, redundancy, and independence were determined as architectural tactics to satisfy the single failure criterion, and sequential execution was determined as a tactic to satisfy the determinism. After determining the tactics, the primitive digital-based I&C architecture was determined. In the design step, 17 systems were selected from the KSNP's I&C systems for the upgrade and functionally grouped based on the primitive architecture. The overall architecture was developed to show the deployment of the systems. The detailed architecture of the safety systems was developed by applying a 2-out-of-3 voting logic, and the detailed architecture of the non-safety systems was developed by hot-standby redundancy. While developing the detailed architecture, three ways of signal transmission were determined with proper rationales: hardwire, datalink, and network. In the assessment step, the required network performance, considering the worst-case of data transmission was calculated: the datalink was required by 120 kbps, the safety network by 5 Mbps, and the non-safety network by 60 Mbps. The architecture covered 17 systems out of 22 KSNP's I&C systems. The architecture is implementable with the equipment developed in South Korea. The architecture can be used as a model to upgrade the existing I&C systems in a planned, large-scale, and one-shot manner. A more detailed architecture down to software level will be developed in the future.     

Approach to accident management in RBMK-1500

In order to ensure the safe operation of the nuclear power plants accident management programs are being developed around the world. These accident management programs cover the whole spectrum of accidents, including severe accidents. A lot of work is done to investigate the severe accident phenomena and implement severe accident management in NPPs with vessel-type reactors, while less attention is paid to channel-type reactors CANDU and RBMK.Ignalina NPP with RBMK-1500 reactor has implemented symptom based emergency operation procedures, which cover management of accidents until the core damage and do not extend to core damage region. In order to ensure coverage of the whole spectrum of accidents and meet the requirements of IAEA the severe accident management guidelines have to be developed.This paper presents the basic principles and approach to management of beyond design basis accidents at Ignalina NPP. In general, this approach could be applied to NPPs with RBMK-1000 reactors that are available in Russia, but the design differences should be taken into account.     

Safety analysis of LBLOCA in BDBA scenarios for the Bushehr's VVER-1000 nuclear power plant

This paper presents the results of thermal-hydraulic calculations of a large break loss of coolant accident (LBLOCA) analysis for a VVER-1000/V446 unit at Bushehr nuclear power plant (BNPP). LBLOCA is analysis in two different beyond design basis accident (BDBA) scenarios using the RELAP5/MOD3.2 best estimate code. The scenarios are LBLOCA with station blackout (SBO) and LBLOCA with pump re-circulation blockage which have been evaluated in the final safety analysis report (FSAR) of BNPP. A model of VVER-1000 reactor based on Unit 1 of BNPP has been developed for the RELAP5/MOD3.2 thermal-hydraulics code consists of 4-loop primary and secondary systems with all their relevant sub-systems important to safety analysis. The analysis is performed without regard for operator's actions on accident management. The safety analysis is carried out and the results are checked against the acceptance criteria which are the possibility of using water inventory in the emergency core cooling system (ECCS) accumulators and the KWU tanks for core cooling and the available time to operators before the maximum design limit of fuel rod cladding damage is reached. These kinds of analyses are performed to provide the response of monitored plant parameters to identify symptoms available to the operators, timing of the loss of critical safety functions and timing of operator actions to avoid the loss of critical safety functions of core damage. The results of performed analyses show that the operators have 2.9 and 3.1 h for LBLOCA with SBO and LBLOCA with pump re-circulation blockage scenarios, respectively, before the fuel rod cladding rupture. The results are also compared with the BNPP FSAR data.     

RELAP5/MOD3.2 sensitivity calculations of loss-of-feed water (LOFW) transient at Unit 6 of Kozloduy NPP

This paper provides a comparison between the real plant data obtained by Unit 6 of Kozloduy nuclear power plant (NPP) during the loss-of-feed water (LOFW) transient and the calculation results received by RELAP5/MOD3.2 computer model of the same NPP unit.RELAP5/MOD3.2 computer model of the VVER-1000 has been developed at the Institute for Nuclear Research and Nuclear Energy-Bulgarian Academy of Sciences (INRNE-BAS) based on Unit 6 of Kozloduy NPP. This model has been used for simulation the behavior of the real VVER-1000 NPP during the LOFW transient. Several calculations have been provided to describe how the different boundary conditions reflect on the prediction of real plant parameters.This paper discusses the results of the thermal–hydraulic sensitivity calculations of loss-of-feed water transient for VVER-1000 reactor design. The report also contains a brief summary of the main NPP systems included in the RELAP5 VVER model and the LOFW transient sequences.This report was possible through the participation of leading specialists from Kozloduy NPP and with the assistance of Argonne National Laboratory (ANL) for the United States Department of Energy (US DOE), International Nuclear Safety Program (INSP).